[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