[Federal Register Volume 62, Number 193 (Monday, October 6, 1997)]
[Notices]
[Pages 52164-52165]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-26405]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-390]
Tennessee Valley Authority, Watts Bar Nuclear Plant; Exemption
I
On February 7, 1996, the Nuclear Regulatory Commission issued
Facility Operating License No. NPF-90 to Tennessee Valley Authority
(TVA or the Licensee) for the Watts Bar Nuclear 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 20, 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 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 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
[[Page 52165]]
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 bound of static, dynamic, and crack
arrest fracture toughness tests on material similar to the Watts Bar
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 50630).
This exemption is effective upon issuance.
Dated at Rockville, Maryland, this 29th day of September 1997.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 97-26405 Filed 10-3-97; 8:45 am]
BILLING CODE 7590-01-P