PSEG Nuclear LLC; Notice of Consideration of Issuance of Amendment to Facility Operating License, Proposed No Significant Hazards Consideration Determination, and Opportunity for a Hearing, 24627-24630 [E7-8437]

Agencies

• NUCLEAR REGULATORY COMMISSION

[Federal Register Volume 72, Number 85 (Thursday, May 3, 2007)]
[Notices]
[Pages 24627-24630]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: E7-8437]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-354]


PSEG Nuclear LLC; Notice of Consideration of Issuance of 
Amendment to Facility Operating License, Proposed No Significant 
Hazards Consideration Determination, and Opportunity for a Hearing

    The U.S. Nuclear Regulatory Commission (NRC or the Commission) is 
considering issuance of an amendment to Facility Operating License No. 
NPF-57 issued to PSEG Nuclear (the licensee) for operation of the Hope 
Creek Generating Station (Hope Creek) located in Salem County, New 
Jersey.
    The proposed amendment would increase the authorized maximum power 
level from 3339 megawatts thermal (MWt) to 3840 MWt, an increase of 
approximately 15 percent.
    Before issuance of the proposed license amendment, the Commission 
will have made findings required by the Atomic Energy Act of 1954, as 
amended (the Act), and the Commission's regulations.
    The Commission has made a proposed determination that the amendment 
request involves no significant hazards consideration. Under the 
Commission's regulations in Title 10 of the Code of Federal Regulations 
(10 CFR), Section 50.92, this means that operation of the facility in 
accordance with the proposed amendment would not (1) involve a 
significant increase in the probability or consequences of an accident 
previously evaluated; or (2) create the possibility of a new or 
different kind of accident from any accident previously evaluated; or 
(3) involve a significant reduction in a margin of safety. As required 
by 10 CFR 50.91(a), the licensee has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:
    1. Does the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The CPPU [Constant Pressure Power Uprate] analyses, which were 
performed at or above CPPU power levels, included a review and 
evaluation of the structures, systems, and components (SSCs) that could 
be affected by the proposed change. The proposed amendment does not 
change the design function or operation of the affected SSCs.
    Plant specific analyses were performed in the following areas: 
Reactor Core and Reactor Internals (e.g., steam dryer), Reactor Coolant 
System and associated systems, Containment, Emergency Core Cooling 
Systems, Control and Instrumentation Systems, Electrical Systems, 
Balance of Plant Systems, and Radwaste Systems. The results of the 
analyses, which included evaluating the increase in the likelihood of 
an SSC malfunction, concluded that the SSCs are capable of performing 
their design functions at CPPU conditions.
    Comprehensive evaluations were performed on the steam dryer and 
other reactor internals for both operational and structural 
performance. Predicted steam dryer peak and alternating stress ratios 
remain within allowable levels. The existing margins to steam dryer 
alternating stress limits and the steam dryer monitoring program during 
power ascension provide assurance that steam dryer integrity will be 
maintained.
    Vibration evaluations at CPPU conditions were performed on the 
Reactor Internal components and Reactor Coolant and associated system 
piping. These included the Main Steam, Feedwater and Reactor 
Recirculation systems piping and supports. The results of the vibration 
analyses demonstrate that operation at CPPU conditions will not result 
in any detrimental effects. System values will remain within allowable 
American Society of Mechanical Engineers Boiler and Pressure Vessel 
Code (ASME Code) limits. In addition, the ASME Code and regulatory 
guidelines require vibration test data be taken on high-energy piping 
during initial CPPU startup. The vibration start-up test program will 
validate the vibration analyses that were performed, demonstrating 
adequate performance of the SSCs.
    Engineered Safety Features (ESF) were evaluated at CPPU conditions 
using NRC-approved methods. The Emergency Core Cooling Systems (ECCS) 
were evaluated to ensure they are capable of performing their design 
function during loss-of-coolant-accidents (LOCA). Adequate net positive 
suction head is maintained without reliance on post-accident 
containment pressure. CPPU does not result in an increase or decrease 
in the available water sources, and does not result in any change in 
the maximum nominal reactor operating pressure. The CPPU evaluations 
demonstrate that the ECCS performance satisfy the requirements of 10 
CFR 50.46 and 10 CFR [Part] 50 Appendix K.
    Balance-of-plant (BOP) systems and equipment were also evaluated 
for CPPU operation. The resulting evaluations demonstrate adequate 
performance with limited modifications that were or will be made to BOP 
components.
    These analyses, which included evaluating the increased likelihood 
of an SSC malfunction, confirm acceptable performance of plant SSCs 
under CPPU conditions. On this basis, PSEG concludes that there is no 
significant change in the ability of the SSCs to preclude or mitigate 
the consequences of accidents.
    The probability (frequency of occurrence) of postulated Design 
Basis Accidents (DBA), and other Updated Final Safety Analysis Report 
(UFSAR) evaluated accidents, occurring is not affected by the increased 
power level, and Hope Creek continues to comply with the regulatory and 
design basis criteria established for plant equipment. The changes in 
consequences of hypothetical accidents, which are assumed to occur at 
102% of the CPPU RTP [Rated Thermal Power], compared to those 
previously evaluated, are in all cases insignificant. The CPPU accident 
evaluations do not exceed any of the NRC-approved acceptance limits. 
The spectrum of hypothetical accidents and transients has been 
investigated, and is shown to meet the plant's currently licensed 
regulatory criteria. Consequently, there is no significant increase in 
the probability or consequences of an accident previously evaluated.
    The impact of CPPU on the radiological consequences of postulated 
DBAs, operational transients and other UFSAR accidents was evaluated. 
The magnitude of the potential consequences is dependent upon the 
quantity of fission products released to the environment, the 
atmospheric dispersion factors and the dose exposure pathways. The 
atmospheric dispersion factors and the dose exposure pathways are not 
changed by CPPU operation. The only factor which could influence the 
magnitude of the consequences is the quantity of activity released to 
the environment. For CPPU, the Control Rod Drop Accident (CRDA), Loss-
of-Coolant Accident (LOCA), Fuel Handling Accident (FHA), Main 
Steamline Break Accident (MSLBA) and instrument line break accident 
(ILBA) were reanalyzed.

[[Page 24628]]

    The DBA that has historically been limiting from a radiological 
criterion is the LOCA, for which USNRC Regulatory Guide 1.183, Appendix 
A guidance was applied. Adherence to the guidance in RG 1.183, and the 
use of the specific values/limits contained in the Technical 
Specifications with as-tested post-accident performance of the safety 
grade engineered safety functions (ESF), provide the assurance for 
sufficient safety margin, including a margin to account for analysis 
uncertainties. The CPPU LOCA evaluation results include the 2% power 
uncertainty factor from Regulatory Guide 1.49.
    The results of the CPPU radiological analyses remain below the 
allowable limits of 10 CFR 50.67 and Table 6 in Regulatory Guide 1.183; 
the CPPU impact is minimal and all radiological limits are met at CPPU 
conditions. Therefore, the proposed change does not involve a 
significant increase in the radiological consequences of an accident 
previously evaluated.
    While the proposed CPPU amendment is not being submitted as a risk-
informed licensing action, it was evaluated from a risk perspective 
using the NRC guidelines established in Regulatory Guide 1.174. Level 1 
and Level 2 Probabilistic Risk Assessments (PRAs) were performed for 
the CPPU. When compared to the risk-acceptance guidelines presented in 
Regulatory Guide 1.174, the calculated changes in core damage frequency 
(CDF) and large early release frequency (LERF) are insignificant. Based 
on these results, PSEG concludes that the proposed amendment would not 
involve a significant increase in the probability of an accident 
previously evaluated.
    The impact of CPPU operation on plant operator actions and 
procedures was also evaluated. The operator action response times 
credited in the safety analyses in the UFSAR are not changed by CPPU. 
In addition, there is no change in Emergency Operating Procedure (EOP) 
strategy for CPPU operation.
    Based on the above, PSEG concludes that the proposed amendment 
would not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    As discussed above, the evaluation of the proposed amendment 
included review of the SSCs that could be affected by the proposed 
change. The proposed amendment does not change the design function or 
operation of the affected SSCs. The proposed amendment does not 
introduce any new or different plant safety-related equipment, and only 
involves instrument set-point changes for CPPU conditions, and minimal 
modifications to plant BOP power generation equipment. The proposed 
amendment does not significantly impact the manner in which the plant 
is operated, and does not have any significant impact on the capability 
the SCCs involved to perform their design function.
    No new operating mode, safety-related equipment lineup, accident 
scenario or equipment failure mode was identified. The CPPU evaluations 
also addressed the impact to postulated accidents, accident 
radiological consequences and operator response. No significant impacts 
were identified. The full spectrum of accident considerations has been 
evaluated, and no new, different, or limiting kind of accident has been 
identified. CPPU uses developed technology, and applies it within the 
capabilities of existing plant equipment in accordance with presently 
existing regulatory criteria to include NRC approved codes, standards 
and methods. The CPPU analyses results confirm acceptable performance 
of plant SSCs under CPPU conditions. Consequently, there are no new 
credible failure mechanisms, malfunctions, or accident initiators that 
were not previously evaluated in the plant design and licensing bases.
    Based on the preceding, PSEG concludes that the proposed change 
would not introduce any new or different kind of accident, or failure 
mode, not previously analyzed.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    Safety margins are applied to plant parameters to account for 
various uncertainties and to avoid exceeding regulatory and licensing 
limits. The proposed change does not involve a significant reduction in 
any margin of safety. First, due to continuing improvements in the 
analytical techniques (computer codes and data) based on several 
decades of BWR safety technology, plant performance feedback, and 
improved fuel and core designs, a significant increase has resulted in 
the design and operating margins between calculated safety analysis 
results and the licensing limits. These available safety analyses 
differences, combined with the excess as-designed equipment, system and 
component capabilities, provide BWR plants the capability to achieve an 
increase in their thermal power ratings within the existing design and 
licensing basis. The proposed CPPU will reduce some of the existing 
design and operational margins. However, safety margins are considered 
to not be significantly reduced if: (1) Applicable regulatory 
requirements, codes and standards or their alternatives approved for 
use by the NRC, are met, and (2) if safety analysis acceptance criteria 
in the licensing basis are met, or if proposed revisions to the 
licensing basis provide sufficient margin to account for analysis and 
data uncertainty. This is the case for the proposed CPPU amendment.
    Safety margin is related to the ability of the fission product 
barriers to limit the level of radiation dose to the public. The impact 
of the proposed CPPU amendment on the: (1) Fuel cladding barrier, (2) 
reactor coolant pressure boundary (RCPB) barrier, and (3) containment 
fission product barrier is discussed below.
    To assure that fuel cladding damage limits are not exceeded, the 
impact of the proposed amendment on fuel system design, nuclear system 
design, thermal and hydraulic design, accident and transient analyses, 
and fuel design limits was evaluated. No new fuel design, or change in 
the specified fuel design limits, is required for CPPU. The current 
fuel and core design limits will continue to be met; both the Safety 
Limit Minimum Critical Power Ratio (SLMCPR) and other applicable 
Specified Acceptable Fuel Design Limits (SAFDLs) are still met. 
Analyses for each fuel reload will continue to meet the criteria 
accepted by the NRC. Continued compliance with the SLMCPR and other 
SAFDLs will be confirmed on a cycle specific basis consistent with the 
criteria accepted by the NRC as specified in NEDO-24011, ``General 
Electric Standard Application for Reactor Fuel, GESTAR II.'' The ECCS 
evaluation for CPPU demonstrates the continued conformance to the 
acceptance criteria of 10 CFR 50.46, for peak cladding temperature 
(PCT) and the other 10 CFR 50.46 parameters. The increased PCT 
consequences for CPPU are insignificant and remain substantially below 
the regulatory criteria. Therefore, the ECCS safety margin and fuel 
cladding margin (PCT) are not significantly impacted by CPPU.
    Challenges to the Reactor Coolant Pressure Boundary were evaluated 
at CPPU conditions (pressure, temperature, flow, and radiation) and 
were found to meet their acceptance criteria for allowable stresses and 
overpressure margin. These evaluations included (1) overpressure 
protection, (2) structural integrity of the RCPB piping, components, 
and supports, and (3) structural integrity of the reactor vessel.

[[Page 24629]]

For the most limiting pressurization event, the peak calculated 
pressure remains below the ASME Code allowable peak pressure. The 
structural integrity of the RCPB piping, components, and supports was 
evaluated using NRC-approved methodology. The changes in flow, pressure 
and temperature associated with CPPU do not result in load limits being 
exceeded. Sufficient margin remains between the calculated stresses and 
ASME Code limits. In addition, the ASME Code and regulatory guidelines 
require vibration test data be taken on high-energy piping during 
initial CPPU startup. The vibration start-up test program will validate 
the vibration analyses that were performed, demonstrating adequate 
performance.
    The structural integrity of the reactor vessel was evaluated. The 
neutron fluence was re-analyzed in accordance with the requirements of 
10 CFR [Part] 50 Appendix G. The existing Pressure-Temperature (P-T) 
limit curves have been revised for CPPU conditions (a previous 
amendment to the Hope Creek license changed the P-T curves and included 
CPPU conditions). The reactor vessel materials surveillance program is 
unchanged by CPPU. The maximum normal operating reactor dome pressure 
for CPPU is unchanged and the vessel remains in compliance with 
regulatory requirements. Consequently, CPPU operation does not have an 
adverse effect on the reactor vessel fracture toughness. The structural 
evaluation of the vessel demonstrates that ASME Code requirements are 
met for normal, upset, emergency and accident conditions.
    Based on the preceding, PSEG concludes that the RCPB structural 
integrity will be maintained and the licensing basis requirements will 
continue to be met following implementation of the proposed CPPU.
    The impact of the proposed CPPU on the Containment was evaluated. 
The effect of CPPU on the peak values for containment pressure and 
temperature confirms the suitability of the plant for operation at CPPU 
RTP. Also, the effects of CPPU on the conditions that affect the 
containment dynamic loads were determined to be satisfactory for CPPU 
operation. Where plant conditions with CPPU are within the range of 
conditions used to define the current dynamic loads, current safety 
criteria are met and no further structural analysis was required. The 
change in short-term containment response is negligible. Because there 
will be more residual heat with CPPU, the containment long-term 
response slightly increases. However, containment pressures and 
temperatures remain below their design limits following any design 
basis accident, and thus, the containment and its cooling systems are 
satisfactory for CPPU operation. The small increase in the calculated 
post LOCA suppression pool temperature above the currently assumed peak 
temperature was evaluated and determined to be acceptable. Based on the 
use of conservative assumptions in these evaluations, PSEG concludes 
that containment structural integrity will be maintained under the 
proposed CPPU conditions, and the containment parameters will remain 
below design limits. Therefore there is no significant reduction in 
safety margin.
    In summary, challenges to the fuel, RCPB, and containment were 
evaluated for CPPU conditions. The structural integrity of the fission 
product barriers will be maintained under CPPU conditions. As such, the 
proposed amendment would not degrade confidence in the ability of the 
barriers to limit the level of radiation dose to the public. Fuel 
integrity is maintained by meeting existing design and regulatory 
limits. The calculated loads on all affected structures, systems and 
components, including the reactor coolant pressure boundary, will 
remain within their design allowables for all design basis event 
categories. The containment parameters remain below design limits. No 
NRC acceptance criterion will be exceeded. Because the Hope Creek 
configuration and responses to transients and hypothetical accidents do 
not result in exceeding the presently approved NRC acceptance limits, 
CPPU does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example, in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place after 
issuance. The Commission expects that the need to take this action will 
occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rulemaking, 
Directives and Editing Branch, Division of Administrative Services, 
Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D59, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Documents may be examined, and/or copied for a fee, at the NRC's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
O1 F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
    The filing of requests for hearing and petitions for leave to 
intervene is discussed below.
    Within 60 days after the date of publication of this notice, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.309, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, https://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition 
for

[[Page 24630]]

leave to intervene is filed by the above date, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the requestors/petitioner's interest. The petition must 
also identify the specific contentions which the petitioner/requestor 
seeks to have litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner intends to rely to establish those facts or expert opinion. 
The petition must include sufficient information to show that a genuine 
dispute exists with the applicant on a material issue of law or fact. 
Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner/requestor 
who fails to satisfy these requirements with respect to at least one 
contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held. If 
the final determination is that the amendment request involves no 
significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment. If the final determination is that the amendment 
request involves a significant hazards consideration, any hearing held 
would take place before the issuance of any amendment.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(c)(1)(i)-(viii).
    A request for a hearing or a petition for leave to intervene must 
be filed by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) e-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, HEARINGDOCKET@NRC.GOV; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. A copy of the request for hearing and 
petition for leave to intervene should also be sent to the Office of 
the General Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by e-mail to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to Jeffrie J. Keenan, 
Esquire, Nuclear Business Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 
08038, attorney for the licensee.
    For further details with respect to this action, see the 
application for amendment dated September 18, 2006, as supplemented by 
letters dated October 10, 2006, October 20, 2006, February 14, February 
16, February 28, March 13, and April 18, 2007 which is available for 
public inspection at the Commission's PDR, located at One White Flint 
North, File Public Area O1 F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the ADAMS Public Electronic Reading Room on the Internet at the NRC Web 
site, https://www.nrc.gov/reading-rm/adams.html. Persons who do not have 
access to ADAMS or who encounter problems in accessing the documents 
located in ADAMS, should contact the NRC PDR Reference staff by 
telephone at 1-800-397-4209, 301-415-4737, or by e-mail to pdr@nrc.gov.

    Dated at Rockville, Maryland, this 27th day of April 2007.

    For the Nuclear Regulatory Commission.
James J. Shea,
Project Manager, Plant Licensing Branch I-2, Division of Operating 
Reactor Licensing, Office of Nuclear Reactor Regulation.
[FR Doc. E7-8437 Filed 5-2-07; 8:45 am]
BILLING CODE 7590-01-P