[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