[Federal Register Volume 65, Number 29 (Friday, February 11, 2000)]
[Notices]
[Pages 7072-7074]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 00-3187]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-254 and 50-265]


Commonwealth Edison Company (Quad Cities Nuclear Power 
Station,Units 1 and 2);

    Exemption

I.

    The Commonwealth Edison Company (ComEd, the licensee) is the holder 
of Facility Operating Licenses Nos. DPR-29 and DPR-30 which authorize 
operation of the Quad Cities Nuclear Power Station, Units 1 and 2 (Quad 
Cities). The license provides, among other things, that the facility is 
subject to all rules, regulations, and orders of the U.S. Nuclear 
Regulatory Commission (the Commission) now or hereafter in effect.
    The facility consists of boiling water reactors (Units 1 and 2) 
located on the licensee's Quad Cities site in Rock Island County, 
Illinois. This exemption refers to both units.

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, ``The appropriate requirements on both the 
pressure-temperature limits and the minimum permissible temperature 
must be met for all conditions.'' Appendix G of 10 CFRPart 50 specifies 
that the requirements for these limits are the American Society of 
Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), 
Section XI, Appendix G Limits.
    To address provisions of the proposed amendments to the technical 
specification (TS) P-T limits, the licensee requested in its submittal 
of November 12, 1999, that the staff exempt Quad Cities from 
application of specific requirements of 10 CFR Part 50, Section 
50.60(a) and Appendix G, and substitute use of ASME Code Cases N-588 
and N-640. Code Case N-588 permits the postulation of a 
circumferentially-oriented flaw (in lieu of an axially-oriented flaw) 
for the evaluation of the circumferential welds in RPV P-T limit 
curves. Code Case N-640 permits the use of an alternate reference 
fracture toughness (KIC fracture toughness curve instead of 
KIa fracture toughness curve) for reactor vessel materials 
in determining the P-T limits. Since the pressure stresses on a 
circumferentially-oriented flaw are lower than the pressure stresses on 
an axially-oriented flaw by a factor of 2, using Code Case N-588 for 
establishing the P-T limits would be less conservative than the 
methodology currently endorsed by 10 CFR Part 50, Appendix G and, 
therefore, an exemption to apply the Code Case would be required by 10 
CFR 50.60. Likewise, since the KIC fracture toughness curve 
shown in ASME Section XI, Appendix A, Figure A-2200-1 (the 
KIC fracture toughness curve) provides greater allowable 
fracture toughness than the corresponding KIa fracture 
toughness curve of ASME Section XI, Appendix G, Figure G-2210-1 (the 
KIa fracture toughness curve), using Code Case N-640 for 
establishing the P-T limits would be less conservative than the 
methodology currently endorsed by 10 CFR Part 50, Appendix G and, 
therefore, an exemption to apply the Code Case would also be required 
by 10 CFR 50.60. It should be noted that, although Code Case N-640 was 
incorporated into the ASME Code recently, an exemption is still needed 
because the proposed P-T limits (excluding Code Cases N-588 and N-640) 
are based on the 1989 edition of the ASME Code.

Code Case N-588

    The licensee has proposed an exemption to allow the use of ASME 
Code Case N-588 in conjunction with ASME Section XI, 10 CFR 50.60(a) 
and 10 CFR Part 50, Appendix G, to determine the P-T limits.
    The proposed amendments to revise the P-T limits for Quad Cities 
rely, in part, on the requested exemption. These proposed P-T limits 
have been developed using the postulation of a circumferentially-
oriented reference flaw as the limiting flaw in a RPV circumferential 
weld in lieu of an axially-oriented flaw required by the 1989 Edition 
of ASME Section XI, Appendix G.
    Postulating the Appendix G [axially-oriented flaw] reference flaw 
in a circumferential weld is physically unrealistic and overly 
conservative, because the length of the flaw is 1.5 times the vessel 
thickness, which is much longer than the width of the reactor vessel 
girth weld. Industry experience with the repair of weld indications 
found during preservice inspection, and data taken from destructive 
examination of actual vessel welds, confirms that any remaining flaws 
are small, laminar in nature, and do not transverse the weld bead 
orientation. Therefore, any potential defects introduced during the 
fabrication process, and not detected during subsequent nondestructive 
examinations, would only be expected to be oriented in the direction of 
weld fabrication. For circumferential welds this indicates a postulated 
defect with a circumferential orientation.
    An analysis provided to the ASME Code's Working Group on Operating 
Plant Criteria (WGOPC) (in which Code Case N-588 was developed) 
indicated that if an axial flaw is postulated on a circumferential 
weld, then based on the stress magnification factors (Mm) 
given in the Code Case for the inside diameter circumferential (0.443) 
and axial (0.926) flaw orientations, it is equivalent to applying a 
safety factor of 4.18 on the pressure loading under normal operating 
conditions. Appendix G requires a safety factor of 2 on the 
contribution of the pressure load in the case of an axially-oriented 
flaw in an axial weld, shell plate, or forging. By postulating a 
circumferentially-oriented flaw on a circumferential weld and

[[Page 7073]]

using the appropriate stress magnification factor, the margin of 2 is 
maintained for the contribution of the pressure load to the integrity 
calculation of the circumferential weld. Consequently, the staff 
determined that the postulation of an axially-oriented flaw on a 
circumferential RPV weld is a level of conservatism that is not 
required to establish P-T limits to protect the RCS pressure boundary 
from failure during hydrostatic testing, heatup, and cooldown.
    The staff noted that ASME Code Case N-588 also includes changes to 
the methodology for determining the thermal stress intensity, 
KIT, which was incorporated into Section XI of the ASME Code 
after the 1989 Edition. However, the licensee still used the 
methodology in the 1989 edition of the ASME Code to calculate 
KIT. The staff already accepted the use of Code Case N-588 
including the modifications made to the KIT methodology for 
exemption requests by other licensees. Hence, the licensee may use the 
methodology in the 1989 Edition of ASME Section XI or the methodology 
contained in Code Case N-588 for determining KIT.
    In summary, the ASME Section XI, Appendix G, procedure was 
developed for axially-oriented flaws, which is physically unrealistic 
and overly conservative for postulating flaws of this orientation to 
exist in circumferential welds. Hence, the NRC staff concurs that 
relaxation of the ASME Section XI, Appendix G, requirements by 
application of ASME Code Case N-588 is acceptable and would maintain, 
pursuant 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.

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

    The licensee has proposed an exemption to allow use of ASME Code 
Case N-640 in conjunction with ASME Section XI, 10 CFR 50.60(a) and 10 
CFR Part 50, Appendix G, to determine P-T limits.
    The proposed amendments to revise the P-T limits for Quad Cities 
rely in part on the requested exemption. These revised P-T limits have 
been developed using the Klc fracture toughness curve, in 
lieu of the Kla fracture toughness curve, as the lower bound 
for fracture toughness.
    Use of the Klc curve in determining the lower bound 
fracture toughness in the development of P-T operating limits curve is 
more technically correct than use of the Kla curve since the 
rate of loading during a heatup or cooldown is slow and is more 
representative of a static condition than a dynamic condition. The 
Klc curve appropriately implements the use of static 
initiation fracture toughness behavior to evaluate the controlled 
heatup and cooldown process of a reactor vessel. The staff has required 
use of the initial conservatism of the Kla curve since 1974 
when the curve was codified. This initial conservatism was necessary 
due to the limited knowledge of RPV materials. Since 1974, additional 
knowledge has been gained about RPV materials, which demonstrates that 
the lower bound on fracture toughness provided by the Kla 
curve is well beyond the margin of safety required to protect the 
public health and safety from potential RPV failure. In addition, P-T 
curves based on the Klc 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.
    Since the RCS P-T operating window is defined by the P-T operating 
and test limit curves developed in accordance with ASME Section XI, 
Appendix G, continued operation of Quad Cities with these P-T curves 
without the relief provided by ASME Code Case N-640 would unnecessarily 
require the RPV to maintain a temperature exceeding 212 degrees 
Fahrenheit in a limited operating window during the pressure test. 
Consequently, steam vapor hazards would continue to be one of the 
safety concerns for personnel conducting inspections in primary 
containment. Implementation of the proposed P-T curves, as allowed by 
ASME Code Case N-640, does not significantly reduce the margin of 
safety and would eliminate steam vapor hazards by allowing inspections 
in primary containment to be conducted at lower coolant temperature. 
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 RPV 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, pursuant 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 the exemption would be required to 
approve the use of Code Cases N-588 and N-640. The staff examined the 
licensee's rationale to support the exemption requests and concurred 
that the use of the code cases would 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 Regulatory Guide 1.99, Revision 
2, the staff concludes that application of the code cases as described 
would provide an adequate margin of safety against brittle failure of 
the RPV. 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 
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-640 may 
be used to revise the P-T limits for Quad Cities Nuclear Power Station, 
Units 1 and 2.

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 Commonwealth 
Edison Company exemption from the requirements of 10 CFR Part 50, 
Section 50.60(a) and 10 CFR Part 50, Appendix G, for Quad Cities 
Nuclear Power Station, Units 1 and 2.
    Pursuant to 10 CFR 51.32, an environmental assessment and finding 
of no significant impact has been prepared and published in the Federal 
Register (65 FR 5702). Accordingly, based upon the environmental 
assessment, the Commission has determined that the granting of this 
exemption will not result in any significant effect on the quality of 
the human environment.

[[Page 7074]]

    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 4th day of February 2000.

    For The Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 00-3187 Filed 2-10-00; 8:45 am]
BILLING CODE 7590-01-P