[Federal Register Volume 68, Number 142 (Thursday, July 24, 2003)]
[Notices]
[Pages 43765-43766]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-18842]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-315]
Indiana Michigan Power Company, Donald C. Cook Nuclear Plant,
Unit 1; Exemption
1.0 Background
Indiana Michigan Power Company (the licensee) is the holder of
Facility Operating License No. DPR-58 which authorizes operation of the
Donald C. Cook (D.C. Cook) Nuclear Plant, Unit 1. The licensee
provides, among other things, that the facility is subject to all
rules, regulations, and orders of the U.S. Nuclear Regulatory
Commission (NRC, the Commission) now or hereafter in effect.
The facility consists of a pressurized water reactor located in
Stevensville, Michigan.
2.0 Request/Action
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, Appendix
G to 10 CFR part 50 states that ``[t]he appropriate requirements on * *
* the pressure-temperature limits and minimum permissible temperature
must be met for all conditions.'' Further, Appendix G of 10 CFR part 50
specifies that the requirements for these limits are based on the
application of evaluation procedures given in Appendix G to section XI
of the American Society of Mechanical Engineers (ASME) Boiler and
Pressure Vessel Code (Code). In this exemption, consistent with the
current provisions of 10 CFR 50.55(a), all references to the ASME Code
denote the 1995 Edition through the 1996 Addenda of the ASME Code.
In order to address provisions of amendments to the D.C. Cook, Unit
1, Technical Specification (TS) P-T limit curves, the licensee
requested in its submittal dated December 10, 2002, that the NRC staff
exempt D.C. Cook, Unit 1, from application of specific requirements of
Appendix G to 10 CFR part 50, and substitute the use of ASME Code Case
N-641. ASME Code Case N-641 permits the use of an alternate reference
fracture toughness curve for RPV materials and permits the postulation
of a circumferentially-oriented flaw for the evaluation of
circumferential RPV welds when determining the P-T limits. The proposed
exemption request is consistent with, and is needed to support, the
D.C. Cook, Unit 1, TS amendment that was contained in the same
submittal. The proposed D.C. Cook, Unit 1, TS amendment will revise the
P-T limits for heatup, cooldown, and inservice test limitations for the
reactor coolant system (RCS) through 32 effective full power years of
operation.
Code Case N-641
The licensee has proposed an exemption to allow the use of ASME
Code Case N-641 in conjunction with Appendix G to ASME section XI, 10
CFR 50.60(a) and 10 CFR part 50, Appendix G, to establish the P-T
limits for the D.C. Cook, Unit 1 RPV.
The proposed TS amendment to revise the P-T limits for D.C. Cook,
Unit 1, relies in part, on the requested exemption. These revised P-T
limits have been developed using the lower bound KIC
fracture toughness curve shown in ASME section XI, Appendix A, Figure
A-2200-1, in lieu of the lower bound KIA fracture toughness
curve of ASME section XI, Appendix G, Figure G-2210-1, as the basis
fracture toughness curve for defining the D.C. Cook Unit 1 P-T limits.
In addition, the revised P-T limits have been developed based on the
use of a postulated circumferentially-oriented flaw for the evaluation
of RPV circumferential welds in lieu of the axially-oriented flaw which
would be required by Appendix G to section XI of the ASME Code. 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 as the basis fracture toughness
curve for the development of P-T operating limits is more technically
correct than use of the KIA curve. The KIC curve
appropriately implements the use of a relationship based on static
initiation fracture toughness behavior to evaluate the controlled
heatup and cooldown process of a RPV, whereas the KIA
fracture toughness curve codified into Appendix G to section XI of the
ASME Code was developed from more conservative crack arrest and dynamic
fracture toughness test data. The application of the KIA
fracture toughness curve was initially codified in Appendix G to
section XI of the ASME Code in 1974 to provide a conservative
representation of RPV material fracture toughness. This initial
conservatism was necessary due to the limited knowledge of RPV material
behavior in 1974. However, additional information has been gained about
RPV materials which demonstrates that the lower bound on fracture
toughness provided by the KIA fracture toughness curve is
well beyond the margin of safety required to protect the public health
and safety from potential RPV failure.
Likewise, the use of a postulated circumferentially-oriented flaw
in lieu of an axially-oriented one for the evaluation of a
circumferential RPV weld is more technically correct. The flaw size
required to be postulated for P-T limit determination has a depth of
one-quarter of the RPV wall thickness and a length six times the depth.
Based on the direction of welding during the fabrication process, the
only technically reasonable orientation for such a large flaw is for
the plane of the flaw to be circumferentially-oriented (i.e., parallel
to the direction of welding). Prior to the development of ASME Code
Case N-641 (and the similar ASME Code Case N-588), the required
postulation of an axially-oriented flaw for the evaluation of a
circumferential RPV weld provided an additional, unnecessary level of
conservatism to the overall evaluation.
In addition, P-T limit curves based on the KIC fracture
toughness curve and postulation of a circumferentially-oriented flaw
for the evaluation of RPV circumferential welds 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
operating window through which the operator heats up and cools down the
RCS, is determined by the difference between the maximum allowable
pressure determined by Appendix G of ASME section XI, and the minimum
required pressure for the reactor coolant pump seals adjusted for
instrument uncertainties. A narrow operating window could potentially
have an adverse safety impact by increasing the possibility of
inadvertent overpressure protection system actuation due to pressure
surges associated with normal
[[Page 43766]]
plant evolutions such as RCS pump starts and swapping operating
charging pumps with the RCS in a water-solid condition.
Since application of ASME Code Case N-641 provides appropriate
procedures to establish maximum postulated defects and to evaluate
those defects in the context of establishing RPV P-T limits, this
application of the Code Case maintains an adequate margin of safety for
protecting RPV materials from brittle failure. Therefore, the licensee
concluded that these considerations were special circumstances pursuant
to 10 CFR 50.12(a)(2)(ii), ``[a]pplication 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.''
In summary, the ASME section XI, Appendix G, procedure was
conservatively developed based on the level of knowledge existing in
1974 concerning reactor coolant pressure boundary materials and the
estimated effects of operation. Since 1974, the level of knowledge
about the fracture mechanics behavior of RCS materials has been greatly
expanded, especially regarding the effects of radiation embrittlement
and the understanding of fracture toughness properties under static and
dynamic loading conditions. 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-641, 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
against brittle failure of the RPV.
The NRC staff has reviewed the exemption request submitted by the
licensee and has concluded that an exemption should be granted to
permit the licensee to utilize the provisions of ASME Code Case N-641
for the purpose of developing D.C. Cook, Unit 1, RPV P-T limit curves.
3.0 Discussion
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, pursuant to 10 CFR 50.12(a)(2)(ii), are
present in that continued operation of D.C. Cook, Unit 1, with the P-T
curves developed in accordance with ASME section XI, Appendix G,
without the relief provided by ASME Code Case N-641 is not necessary to
achieve the underlying purpose of Appendix G to 10 CFR part 50. The
underlying purpose of the regulations in Appendix G to 10 CFR part 50
is to provide an acceptable margin of safety against brittle failure of
the RCS during any condition of normal operation to which the pressure
boundary may be subjected over its service lifetime. Application of
ASME Code Case N-641 in lieu of the requirements of ASME Code section
XI, Appendix G provides an acceptable alternative methodology which
will continue to meet the underlying purpose of Appendix G to 10 CFR
part 50.
The NRC staff examined the licensee's rationale to support the
exemption request, and agrees within the licensee's determination that
an exemption would be required to approve the use of Code Case N-641.
The NRC staff agrees that the use of ASME Code Case N-641 would meet
the underlying intent of Appendix G to 10 CFR part 50. The NRC staff
concludes that the application of the technical provisions of ASME Code
Case N-641 provides sufficient margin in the development of RPV P-T
limit curves such that the underlying purpose of the regulations
(Appendix G to 10 CFR part 50) continue to be met so that the use of
all provisions in Appendix G to section XI of the ASME Code are not
necessary. Therefore, the NRC staff concludes that the exemption
requested by the licensee is justified based on the special
circumstances of 10 CFR part 50(a)(2)(ii), ``[a]pplication 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.''
Based upon a consideration of the conservatism that is explicitly
incorporated into the methodologies of Appendix G to 10 CFR part 50;
Appendix G to section XI of the ASME Code; and Regulatory Guide 1.99,
Revision 2; the staff concludes that application of ASME Code Case N-
641 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 NRC
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 Case N-641 may be used to revise the P-T limits for
the D.C. Cook, Unit 1, RPV.
4.0 Conclusion
Accordingly, the Commission has determined that, pursuant to 10 CFR
50.12(a), the 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. Also, special circumstances are present.
Therefore, the Commission hereby grants the licensee an exemption from
the requirements of 10 CFR 50.60 and 10 CFR part 50, Appendix G, to
allow application of ASME Code Case N-641 in establishing TS
requirements for the reactor vessel pressure limits at low temperatures
for D.C. Cook, Unit 1.
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 (68 FR 42137).
This exemption is effective upon issuance.
Dated at Rockville, Maryland, this 18th day of July 2003.
For The Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-18842 Filed 7-23-03; 8:45 am]
BILLING CODE 7590-01-P