[Federal Register Volume 66, Number 106 (Friday, June 1, 2001)]
[Notices]
[Pages 29846-29848]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-13741]


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NUCLEAR REGULATORY COMMISSION

[Docket Nos. 50-272 and 50-311]


PSEG Nuclear LLC; Salem Nuclear Generating Station, Unit Nos. 1 
and 2; Exemption

1.0 Background

    PSEG Nuclear LLC (PSEG or the licensee) is the holder of Facility 
Operating License Nos. DPR-70 and DPR-75 that authorize operation of 
the Salem Nuclear Generating Station, Unit Nos. 1 and 2. The licenses 
provide, 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 two pressurized water reactors located at 
the licensee's site on the southern end of Artificial Island in Lower 
Alloways Creek Township, Salem County, New Jersey. Salem, New Jersey, 
is located approximately 7.5 miles northeast of the site.

2.0 Purpose

    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 to 10 CFR Part 50 also 
specifies that the requirements for these limits are the American 
Society of Mechanical Engineers (ASME) Code, Section XI, Appendix G 
Limits. In Generic Letter 88-11, the NRC staff advised licensees that 
the staff would use Regulatory Guide (RG) 1.99, Revision 2, to review 
P-T limit curves. RG 1.99, Revision 2, provides guidance for 
implementing 10 CFR Part 50, Appendix G, and contains conservative 
methodologies for determining the increase in transition temperature 
and the decrease in upper-shelf energy (USE) resulting from neutron 
radiation.
    In order to address provisions of amendments to the Technical 
Specifications (TS) P-T limit curves, the licensee requested in its 
application dated November 10, 2000, that the staff exempt, as 
permitted by 10 CFR 50.60(b), Salem, Unit Nos. 1 and 2, from 
application of specific requirements of 10 CFR 50.60(a) and 10 CFR Part 
50, Appendix G, and substitute use of ASME Code Case N-640. Code Case 
N-640 provides an alternate reference fracture toughness methodology 
for reactor vessel materials for use in determining the P-T limits. The 
proposed action is in accordance with PSEG's application for exemption 
contained in its November 10, 2000, letter, as supplemented by letters 
dated March 28 and April 2, 2001. The proposed action is needed to 
support PSEG's license amendment request to increase thermal power 
levels by 1.4% submitted under the same application (the final revision 
of the proposed P-T limit curves was submitted by the licensee by 
letter dated March 28, 2001). The proposed license amendment will, in 
part, revise the P-T limits for heatup, cooldown, core criticality, and 
hydrostatic/leak test limitations for the reactor coolant system (RCS) 
to 32 effective full power years (EFPYs).

Code Case N-640

    The licensee has proposed an exemption to allow the use of Code 
Case N-640, in conjunction with ASME Section XI, Appendix G, 10 CFR 
50.60(a), and 10 CFR Part 50, Appendix G, to determine the P-T limits, 
and stated that this proposed alternative meets the underlying intent 
of the NRC's regulations.
    Standard Review Plan (NUREG-0800) Section 5.3.2 provides an 
acceptable method for determining the P-T limit curves for ferritic 
materials in the beltline of the RPV based on the linear elastic 
fracture mechanics (LEFM) methodology of Appendix G to Section XI of 
the Code. The basic parameter of this methodology is the stress 
intensity factor K<INF>I</INF>, which is a function of the stress state 
and flaw configuration. Appendix G requires a safety factor of 2.0 on 
stress intensities resulting from reactor pressure during normal and 
transient operating conditions, and a safety factor of 1.5 on the same 
stresses for hydrostatic testing curves. The methods of Appendix G 
postulate the existence of a sharp surface flaw in the RPV that is 
normal to the direction of the maximum stress. This flaw is postulated 
to have a depth that is equal to 1/4 of the RPV beltline thickness and 
a length equal to 1.5 times the RPV beltline thickness. The critical 
locations in the RPV beltline region for calculating heatup and 
cooldown P-T curves are the 1/4 thickness (1/4T) and 3/4 thickness (3/
4T) locations, which correspond to the maximum depth of the postulated 
inside surface and outside surface defects, respectively.
    The methodology provided in Appendix G to Section XI of the ASME 
Code requires that licensees determine

[[Page 29847]]

the adjusted reference temperature (ART or adjusted RT<INF>NDT</INF>). 
The ART is defined as the sum of the initial (unirradiated) reference 
temperature (initial RT<INF>NDT</INF>), the mean value of the 
adjustment in reference temperature caused by irradiation 
(<greek-D>RT<INF>NDT</INF>), and a margin (M) term by application of RG 
1.99, Revision 2. The <greek-D>RT<INF>NDT</INF> is a product of a 
chemistry factor and a fluence factor. The chemistry factor is 
dependent upon the amount of copper and nickel in the material and may 
be determined from tables in RG 1.99, Revision 2, or from surveillance 
data. The fluence factor is dependent upon the neutron fluence at the 
maximum postulated flaw depth. The margin term is dependent upon 
whether the initial RT<INF>NDT</INF> is a plant-specific or a generic 
value and whether the chemistry factor (CF) was determined using the 
tables in RG 1.99, Revision 2, or surveillance data. The margin term is 
used to account for uncertainties in the values of the initial 
RT<INF>NDT,</INF> the copper and nickel contents, the fluence and the 
calculational procedures. RG 1.99, Revision 2, describes the 
methodology to be used in calculating the margin term.
    The Pressurized Thermal Shock (PTS) rule, 10 CFR 50.61, requires 
that licensees demonstrate that facility RPV materials will continue to 
possess an adequate level of fracture resistance to protect the RPV 
from potential failure as a result of PTS events. Each material's PTS 
reference temperature, RT<INF>PTS,</INF> is determined in a manner like 
that used to determine ART, except that the neutron fluence at the 
clad-to-base metal interface at end of license (EOL) conditions is used 
in lieu of either the 1/4T or 3/4T fluence. Each material's 
RT<INF>PTS</INF> value is then compared to the screening limits given 
in 10 CFR 50.61, 270  deg.F for plates, forging, and axial welds, and 
300  deg.F for circumferential welds. Provided that all RPV materials' 
RT<INF>PTS</INF> values remain below these screening limits, the 
fracture resistance of the RPV is demonstrated to be adequate to meet 
the requirements of 10 CFR 50.61 through end of life.
    The proposed license amendments to revise the P-T limits for Salem, 
Unit Nos. 1 and 2, rely in part on the requested exemption. These 
revised P-T limits have been developed using the K<INF>Ic</INF> 
fracture toughness curve shown in ASME Section XI, Appendix A, Figure 
A-2200-1, in lieu of the K<INF>Ia</INF> 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 for establishing P-T limit curves remain 
unchanged.
    Use of the K<INF>Ic</INF> curve in determining the lower bound 
fracture toughness in the development of P-T operating limit curves is 
more technically correct than the K<INF>Ia</INF> curve. The 
K<INF>Ic</INF> 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 stated 
that the use of the K<INF>Ia</INF> curve, with its initial 
conservatism, was justified when the curve was codified in 1974. This 
initial conservatism was necessary due to the limited knowledge of RPV 
materials. Since 1974, additional knowledge has been gained about RPV 
materials, that demonstrates that the lower bound on fracture toughness 
provided by the K<INF>Ia</INF> 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 K<INF>Ic</INF> 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 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 Salem, Unit Nos. 1 and 2, 
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.
    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 NRC 
regulations to ensure an acceptable margin of safety.

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 are 
present whenever, according to 10 CFR 50.12(a)(2)(ii), ``Application 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.'' The staff accepts the licensee's 
determination that an exemption would be required to approve the use of 
Code Case N-640. The staff examined the licensee's rationale to support 
the exemption request and concurred that the use of the code case 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 ASME 
Code; and RG 1.99, Revision 2, the staff concluded that application of 
the code case as described would provide an acceptable margin of safety 
against brittle failure of the RPV.
    Therefore, since strict compliance with the requirements of 10 CFR 
Part 50, Appendix G, is not necessary to serve the underlying purpose 
of the regulation, the staff concludes that application of Code Case N-
640 to the P-T limit calculations meets the special circumstance 
provisions stated in 10 CFR 50.12(a)(2)(ii), for granting this 
exemption to the regulation, and that the methodology of Code Case N-
640 may be used to revise the P-T limits for Salem, Unit Nos. 1 and 2.

4.0 Conclusion

    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 PSEG Nuclear 
LLC an exemption from the requirements of 10 CFR Part 50, Section 
50.60(a) and 10 CFR Part 50, Appendix G, for Salem, Unit Nos. 1 and 2.

[[Page 29848]]

    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 24410).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 25th day of May 2001.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 01-13741 Filed 5-31-01; 8:45 am]
BILLING CODE 7590-01-P