[Federal Register Volume 66, Number 174 (Friday, September 7, 2001)]
[Notices]
[Pages 46846-46848]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-22514]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-289]


AmerGen Energy Company, LLC; Three Mile Island Nuclear Station, 
Unit 1; Exemption

1.0  Background

    The AmerGen Energy Company, LLC (the licensee) is the holder of 
Facility Operating License No. DPR-50 which authorizes operation of the 
Three Mile Island Nuclear Station, Unit 1 (TMI-1). The license 
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 
Dauphin County in Pennsylvania.

2.0  Request/Action

    Title 10 of the Code of Federal Regulations (10 CFR), part 50, 
Appendix G requires, in part, 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 these limits be at least as conservative as those 
obtained by following the methods of analysis and the margins of safety 
of the American Society of Mechanical Engineers (ASME) Code, Section 
XI, Appendix G.
    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 TMI-1 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 TMI-1 
TS.
    By an application dated March 29, 2001, the licensee requested 
amendments to the P/T limit curves in the TS. In the same application, 
the licensee requested an exemption from application of specific 
requirements of 10 CFR part 50, Appendix G, and 10 CFR part 50, Section 
50.61(a)(5), in order to address provisions of amendments to the TS P/T 
limits curves. Specifically, the exemption would instead allow the use 
of ASME Code Cases and an alternative approach as follows:
    1. Code Case N-588, which permits the use of circumferentially-
oriented flaws in circumferential welds for development of P/T limits,
    2. Code Case N-640, which permits application of the lower bound 
static initiation fracture toughness value equation as the basis for 
establishing the P/T curves in lieu of using the lower bound crack 
arrest fracture toughness value equation, and
    3. The master curve approach for determining the initial reference 
temperature value for weld metal WF-70 in the TMI-1 reactor vessel.

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 the public health 
and safety, and are consistent with the common defense and security; 
and (2) when special circumstances are present. The three exemptions 
and their associated special circumstances are discussed below.

3.1  Code Case N-588

    The licensee has proposed an exemption to allow use of ASME Code 
Case N-588 in conjunction with ASME Section XI and 10 CFR Part 50, 
Appendix G, to determine P/T limits for TMI-1. The proposed amendment 
to revise the P/T limits for TMI-1 relies in part on the requested 
exemption. These revised P/T limits have been developed using 
postulated flaws in the circumferential orientation for the 
circumferential weld in the TMI-1 RPV, in lieu of postulating axial 
flaws in the circumferential welds.
    The use of circumferential flaws in circumferential welds is more 
appropriate than the use of axial flaws in circumferential welds. Since 
the flaws postulated in the development of P/T limits have a through-
wall depth of one-quarter of the vessel wall thickness (1.94 in. for 
the TMI-1 RPV), the length of the postulated flaw, six times the depth, 
is more than 11 inches. For the circumferential weld in the TMI-1 RPV, 
an axial flaw of this length centered at the weld would place the tips 
of the postulated flaw within the adjacent base metal above and below 
the weld.

[[Page 46847]]

Therefore, the only way to maintain a flaw within the circumferential 
weld metal is to postulate a circumferential flaw within the weld, as 
accomplished using Code Case N-588. For the base metals adjacent to the 
circumferential welds, axial flaws are and continue to be postulated 
for the development of P/T limits.
    The underlying purpose of ASME Section XI and 10 CFR Part 50, 
Appendix G, is to ensure that (1) the RCPB 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 
curve limits 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, and application of 10 CFR part 50, 
Appendix G in these circumstances is not necessary to achieve that 
purpose.

3.2  Code Case N-640

    The licensee has proposed an exemption to allow use of the ASME 
Code Case N-640 in conjunction with ASME Section XI and 10 CFR part 50, 
Appendix G, to determine P/T limits for TMI-1. The proposed license 
amendment to revise the TS P/T operating limits for TMI-1 relies, in 
part, on the requested exemption. These revised P/T operating limits 
have been developed using the KIC fracture toughness curve 
shown in 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 the P/T operating limits curve 
is more technically correct than using the KIA curve. The 
KIC curve appropriately implements the use of static 
initiation fracture toughness behavior to evaluate the controlled 
heatup 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 
RPV materials. Since 1974, additional knowledge has been gained about 
RPV 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 RPV 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 operating window through which the 
operator heats up and cools down the reactor coolant system (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 (RCP) seals adjusted for 
instrument uncertainties.
    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 TMI-1 with these P/T 
curves without the relief provided by ASME Code Case N-640 may 
unnecessarily restrict the P/T operating window, especially at low 
temperature conditions. The operating window becomes more restrictive 
with continued reactor vessel service. 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, 
and application of 10 CFR Part 50, Appendix G, in these circumstances 
is not necessary to achieve that purpose.
    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.

3.3  Master Curve Approach

    The licensee has proposed an exemption from 10 CFR Part 50.61(a)(5) 
to allow the use of the master curve approach as an alternative to 
Paragraph NB-2331 of the ASME Code to determine the initial reference 
temperature (RTNDT) value for weld metal WF-70 in the TMI-1 
reactor vessel. The evaluation was part of a pressurized thermal shock 
(PTS) reevaluation for the TMI-1 RPV.
    The current Charpy V-notch and drop weight-based methodology 
described in NB-2331 establishes an RTNDT value and then 
relies on surveillance data from the testing of Charpy specimens and/or 
general material embrittlement models incorporated into Regulatory 
Guide 1.99, Revision 2 to predict the amount this value will shift due 
to a given level of neutron radiation exposure. This ``initial plus 
shift'' methodology has been consistently used to assess RPV 
embrittlement in the U.S. The master curve approach, however, proposes 
that ``direct measurement'' of fracture toughness can be made on 
unirradiated specimens.
    The unirradiated RTNDT for WF-70 weld metal was 
determined from drop weight tests and fracture toughness tests from 
welds fabricated with WF-70 and WF-209-1 weld metal. Since WF-70 and 
WF-209-1 welds were fabricated using the same heat number of weld wire 
and the same type of flux, their material properties are considered 
equivalent. Charpy V-notch impact and drop weight tests (the current 
methodology) were applied to the WF-70 weld metal by the licensees for 
Zion Nuclear Power Station, Units 1 and 2, and Oconee Nuclear Station, 
Units 1, 2, and 3, in the early 1990s for a PTS evaluation. The tests 
resulted in wide variability in RTNDT values. The staff 
concluded that the large uncertainty in RTNDT values for WF-
70 weld metal is due to the low upper-shelf behavior of the material. 
Therefore, the definition of RTNDT in the ASME Code is not 
applicable for WF-70 weld metal due to the large variability in 
RTNDT values. In lieu of using Charpy V-notch and drop 
weight data, the licensee proposed to determine the initial reference 
temperature value using the test results from the master curve 
methodology. Since the licensee did not follow the method in Section 
III of the ASME Code, the methodology for determining the 
RTNDT of WF-70 does not meet the requirements of 10 CFR 
50.61 and requires an exemption.

[[Page 46848]]

    By letter dated February 22, 1994, the NRC approved the use of the 
master curve approach for the Zion Nuclear Power Station, Units 1 and 
2, and the RTNDT value is -26  deg.F for WF-70 weld metal. 
The exemption approval for the Zion station also stated that other 
procedures for determination of RTNDT may serve as 
acceptable alternatives to NB-2331 contingent on staff review and 
approval. The staff acceptance of the alternative procedure in that 
evaluation was based, in part, on the analysis of a significant amount 
of fracture toughness data for the WF-70 weld metal. Therefore, since 
TMI-1 used the same weld metal as Zion and the data considered for the 
Zion exemption resulted in a more representative RTNDT 
value, the TMI-1 use of the master curve approach for WF-70 weld metal 
is acceptable.
    In summary, the underlying purpose of 10 CFR 50.61 is to ensure 
that the RPV is adequately protected from PTS. Application of the 
master curve approach to determine the unirradiated RTNDT 
value for weld metal WF-70 is acceptable because the master curve 
approach is more appropriate for material with low upper-shelf behavior 
like WF-70 weld metal.
    Therefore, pursuant to 10 CFR 50.12(a)(2)(ii), application of the 
master curve approach to determine the unirradiated RTNDT 
value for weld metal WF-70 would continue to achieve the underlying 
purpose of the rule, and application of the definition of 
RTNDT(U) in 10 CFR 50.61(a)(5) in these circumstances is not 
necessary to achieve that purpose.

4.0  Conclusion

    Accordingly, the Commission has determined that, pursuant to 10 CFR 
50.12(a), the exemptions are authorized by law, will not endanger life 
or property or common defense and security, and are, otherwise, in the 
public interest. Also, special circumstances are present. Therefore, 
the Commission hereby grants AmerGen Energy Company, LLC exemptions 
from the requirements of 10 CFR part 50, Appendix G, and 10 CFR part 
50, Sec. 50.61(a)(5), for TMI-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 (66 FR 45874).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 30th day of August 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-22514 Filed 9-6-01; 8:45 am]
BILLING CODE 7590-01-P