[Federal Register Volume 64, Number 149 (Wednesday, August 4, 1999)]
[Notices]
[Pages 42418-42420]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-19986]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-269, 50-270, and 50-287]


Duke Energy Corporation (Oconee Nuclear Station, Units 1, 2, and 
3); Exemption

I

    The Duke Energy Corporation (Duke/the licensee) is the holder of 
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55, that 
authorize operation of the Oconee Nuclear Station, Units 1, 2, and 3 
(Oconee), respectively. The licenses provide, among other things, that 
the facilities are subject to all rules, regulations, and orders of the 
US Nuclear Regulatory Commission (the Commission) now or hereafter in 
effect.
    The facilities consist of pressurized water reactors located on 
Duke's Oconee site in Seneca, Oconee County, South Carolina.

[[Page 42419]]

II

    Title 10 of the Code of Federal Regulations (10 CFR) part 50, 
Appendix G requires that pressure-temperature (P-T) limits be 
established for reactor pressure vessels (RPVs) during normal operating 
and hydrostatic or leak rate testing conditions. Specifically, 10 CFR 
part 50, Appendix G states that ``[t]he appropriate requirements on * * 
* the pressure-temperature limits and minimum permissible temperature 
must be met for all conditions.'' Appendix G of 10 CFR part 50 
specifies that the requirements for these limits are the American 
Society of Mechanical Engineers (ASME) Code, section XI, Appendix G 
limits.
    Pressurized water reactor licensees have installed cold 
overpressure mitigation systems/low temperature overpressure protection 
(LTOP) systems in order to protect the reactor coolant pressure 
boundary (RCPB) from being operated outside of the boundaries 
established by the P-T limit curves and to provide pressure relief of 
the RCPB during low temperature overpressurization events. The licensee 
is required by the Oconee Units 1, 2, and 3 Technical Specifications 
(TS) to update and submit the changes to its LTOP setpoints whenever 
the licensee is requesting approval for amendments to the P-T limit 
curves in the Oconee Units 1, 2, and 3 TS.
    Therefore, in order to address provisions of amendments to the TS 
P-T limits and LTOP curves, the licensee requested in its submittal 
dated May 11, 1999, that the staff exempt Oconee Units 1, 2, and 3 from 
application of specific requirements of 10 CFR part 50, section 
50.60(a) and 10 CFR part 50, Appendix G, and substitute use of three 
ASME Code Cases as follows:
    1. N-514 as an alternate methodology for determining the low 
temperature overpressure protection system enable temperature,
    2. N-588 for determining the reactor vessel P-T limits derived from 
postulating a circumferentially-oriented reference flaw in a 
circumferential weld, and
    3. N-626 as an alternate reference fracture toughness for reactor 
vessel materials for use in determining the P-T limits. (As a result of 
recent ASME code committee action, the designation for Code Case N-626 
was changed to N-640. Therefore, Code Case N-640 will be discussed 
below rather than Code Case N-626, the designation referenced in the 
submittal.)
    The proposed action is in accordance with the licensee's 
application for exemption contained in a submittal dated May 11, 1999, 
and is needed to support the TS amendments that are contained in the 
same submittal and are being processed separately. The proposed 
amendments will revise the P-T limits of TS 3.4.3 for Oconee Units 1, 
2, and 3 related to the heatup, cooldown, and inservice test 
limitations for the Reactor Coolant System of each unit to a maximum of 
33 Effective Full Power Years (EFPY). It will also revise TS 3.4.12, 
Low Temperature Overpressure Protection System, to reflect the revised 
P-T limits of the reactor vessels.

Code Case N-514

    During staff review of this submittal, the staff determined that 
granting of an exemption to use Code Case N-514 to redefine the LTOP 
enable temperature as RTNDT +50  deg.F was not necessary. 
Since the prior definition of the enable temperature as 
RTNDT +90  deg.F is found only in an NRC Branch Technical 
Position, an exemption is not required.

Code Case N-588

    This requested exemption will allow the use of ASME Code Case N-588 
to determine stress intensity factors for postulated defects in 
circumferential welds. Appendix G of 10 CFR part 50 requires, in part, 
that Article G-2120 of ASME section XI, Appendix G, be used to 
determine the maximum postulated defects in reactor pressure vessels 
(RPV) when determining the P-T limits for the vessel. Article G-2120 
specifies that the postulated defect be in the surface of the vessel 
material and normal (perpendicular in the plane of the material) to the 
direction of maximum stress. ASME section XI, Appendix G, also provides 
methodology to determine the stress intensity factors for a maximum 
postulated defect normal to the maximum stress. The purpose of this 
article is to prevent non-ductile failure of the RPV by providing 
procedures to identify the most limiting postulated fractures to be 
considered in the development of P-T limits.
    Per Article G-2120 of ASME section XI, Appendix G, the postulated 
flaw ``normal to the direction of maximum stress'' would be an axially-
oriented flaw for each reactor vessel beltline material. This 
postulated reference flaw is intended to be a conservative, bounding 
defect when compared to those defects that may have gone undetected 
during the fabrication process.
    Engineering experience and non-destructive examinations over the 
course of the last thirty years have shown this to be a valid 
assumption and have shown that no service-induced degradation mechanism 
exists in pressurized water reactors that would cause significant 
growth of preservice flaws.
    However, for a circumferential weld, it is extremely unlikely that 
axial flaws of appreciable size would be introduced perpendicular to 
the weld seam during fabrication since the nature of the welding 
process leads to any extended flaws being introduced parallel to the 
direction of travel of the welding head. In addition, the size of flaw 
required to be postulated by the ASME Code, if oriented axially, would 
extend across the entire nominal width of the circumferential weld and 
into the base material on either side. Given the strict procedure 
controls required during the fabrication of ASME Code Class 1 reactor 
vessels and the extensive amount of preservice and inservice non-
destructive examination to which their welded regions have been 
subjected, it has been confirmed that any remaining defects are small 
and do not cross transverse to the weld bead orientation. Therefore, 
the NRC staff finds that the application of this degree of non-physical 
conservatism is not necessary to achieve the underlying intent of 10 
CFR part 50, Appendix G.
    In summary, the underlying purpose of 10 CFR part 50, Appendix G 
and ASME section XI, Appendix G, is to satisfy the requirement that: 
(1) The reactor coolant pressure boundary be operated in a regime 
having sufficient margin to ensure that when stressed the vessel 
boundary behaves in a non-brittle manner and the probability of a 
rapidly propagating fracture is minimized, and (2) P-T operating and 
test curves provide margin in consideration of uncertainties in 
determining the effects of irradiation on material properties.
    Application of Code Case N-588 to determine P-T operating and test 
limit curves per ASME section XI, Appendix G, provides appropriate, 
conservative procedures to determine limiting maximum postulated 
defects and to consider those defects in the P-T limits. This 
application of the code case maintains the margin of safety for 
circumferential welds equivalent to that originally contemplated for 
plates/forgings and axial welds.
    Therefore, pursuant to 10 CFR 50.12(a)(2)(ii), application of the 
code case would continue to achieve the underlying purpose of the rule.

Code Case N-640 (Formerly Code Case N-626)

    The licensee has proposed an exemption to allow use of ASME Code 
Case N-626 (which is now Code Case N-640) in conjunction with ASME

[[Page 42420]]

section XI, 10 CFR 50.60(a) and 10 CFR part 50, Appendix G, to 
determine that the P-T limits meet the underlying intent of the NRC 
regulations.
    The proposed amendment to revise the P-T limits for Oconee Units 1, 
2, and 3 rely in part on the requested exemption. These revised P-T 
limits have been developed using the KIc fracture toughness 
curve shown on ASME section XI, Appendix A, Figure A-2200-1, in lieu of 
the KIa fracture toughness curve of ASME section XI, 
Appendix G, Figure G-2210-1, as the lower bound for fracture toughness. 
The other margins involved with the ASME section XI, Appendix G process 
of determining P-T limit curves remain unchanged.
    Use of the KIc curve in determining the lower bound 
fracture toughness in the development of P-T operating limits curve is 
more technically correct than the KIa curve. The 
KIc curve appropriately implements the use of static 
initiation fracture toughness behavior to evaluate the controlled heat-
up and cooldown process of a reactor vessel. The licensee has 
determined that the use of the initial conservatism of the 
KIa curve when the curve was codified in 1974 was justified. 
This initial conservatism was necessary due to the limited knowledge of 
reactor pressure vessel materials. Since 1974, additional knowledge has 
been gained about reactor pressure vessel materials, which demonstrates 
that the lower bound on fracture toughness provided by the 
KIa curve is well beyond the margin of safety required to 
protect the public health and safety from potential reactor pressure 
vessel failure. In addition, P-T curves based on the KIc 
curve will enhance overall plant safety by opening the P-T operating 
window with the greatest safety benefit in the region of low 
temperature operations. The two primary safety benefits in opening the 
low temperature operating window are a reduction in the challenges to 
RCS power operated relief valves and elimination of RCP impeller 
cavitation wear.
    Since the RCS P-T operating window is defined by the P-T operating 
and test limit curves developed in accordance with the ASME section XI, 
Appendix G procedure, continued operation of Oconee with these P-T 
curves without the relief provided by ASME Code Case N-640 would 
unnecessarily restrict the P-T operating window. This restriction 
requires, under certain low temperature conditions, that only one 
reactor coolant pump in a reactor coolant loop be operated. The 
licensee has found from experience that the effect of this restriction 
is undesirable degradation of reactor coolant pump impellers that 
results from cavitation sustained when either one pump or one pump in 
each loop is operating. Implementation of the proposed P-T curves as 
allowed by ASME Code Case N-640 does not significantly reduce the 
margin of safety. Thus, pursuant to 10 CFR 50.12(a)(2)(ii), the 
underlying purpose of the regulation will continue to be served.
    In summary, the ASME section XI, Appendix G procedure was 
conservatively developed based on the level of knowledge existing in 
1974 concerning reactor pressure vessel materials and the estimated 
effects of operation. Since 1974, the level of knowledge about these 
topics has been greatly expanded. The NRC staff concurs that this 
increased knowledge permits relaxation of the ASME section XI, Appendix 
G requirements by application of ASME Code Case N-640, while 
maintaining, pursuit to 10 CFR 50.12(a)(2)(ii), the underlying purpose 
of the ASME Code and the NRC regulations to ensure an acceptable margin 
of safety.

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. The staff accepts the 
licensee's determination that an exemption would be required to approve 
the use of Code Cases N-588 and N-626 (now Code Case N-640). The staff 
examined the licensee's rationale to support the exemption request and 
concurred that the use of the code cases would also meet the underlying 
intent of these regulations. Based upon a consideration of the 
conservatism that is explicitly incorporated into the methodologies of 
10 CFR Part 50, Appendix G; Appendix G of the Code; and RG 1.99, 
Revision 2, the staff concluded that application of the code cases as 
described would provide an adequate margin of safety against brittle 
failure of the RPVs. This is also consistent with the determination 
that the staff has reached for other licensees under similar conditions 
based on the same considerations. Therefore, the staff concludes that 
requesting the exemption under the special circumstances of 10 CFR 
50.12(a)(2)(ii) is appropriate and that the methodology of Code Cases 
N-588 and N-626 may be used to revise the LTOP setpoints and P-T limits 
for the Oconee Units 1, 2, and 3 reactor coolant system.

IV

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), the 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 Duke an 
exemption from the requirements of 10 CFR part 50, section 50.60(a) and 
10 CFR part 50, Appendix G, for the Oconee Nuclear Station, Units 1, 2, 
and 3.
    Pursuant to 10 CFR 51.32, the Commission has determined that the 
granting of this exemption will not result in any significant effect on 
the quality of the human environment (64 FR 40901).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 29th day of July 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-19986 Filed 8-3-99; 8:45 am]
BILLING CODE 7590-01-P