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Comments must be filed by January 2, 2015, at 11.59 PM (23.59) US Eastern Time (DC). However a request for a hearing can be filed until February 2, 2015. http://www.regulations.gov/#!documentDetail;D=NRC-2014-0250-0001
[Comments do not have to be long. One or two sentences is good. It needs to be your own words. If they are the same as others, they don’t post them. Demanding a hearing is a good idea. Below we discuss only two of five nuclear power stations which require comment.]

The US NRC wants to sneak through changes from Maximum Extended Load Line (MELLLA) to Maximum Extended Load Line Plus (MELLLA+), for the largest US reactor, Grand Gulf, and Peach Bottom. While they state that the changes “involve no significant hazards consideration,“: http://www.regulations.gov/#!documentDetail;D=NRC-2014-0250-0001
other NRC documents indicate otherwise: “implementation of MELLLA+ could lead to operation closer to system stability boundaries” (US NRC, 2014). http://pbadupws.nrc.gov/docs/ML1404/ML14041A136.pdf And, Grand Gulf is not only the largest single nuclear reactor in the United States, but also has a Mark III containment which is so prone to hydrogen build up that they have put a system to burn off the hydrogen, which requires a back-up energy source. Peach Bottom, near Baltimore, has the better known Mark 1 containment of Fukushima Daiichi infamy. The new NRC solution for Mark 1 containment problems is to vent unfiltered radioactive air out of the containment.
Fukushima US NRC photo via Greenpeace.org
US NRC photo via Greenpeace.org

This MELLLA+ uprate is such a dangerous prospect for Grand Gulf that even Alan Wang of the NRC, who appears always happy to comply with giving the nuclear industry “relief” from ASME and NRC standards, actually required an “audit” of Grand Gulf. http://pbadupws.nrc.gov/docs/ML1427/ML14279A585.pdf http://pbadupws.nrc.gov/docs/ML1411/ML14119A009.pdf

However, that didn’t stop him-the NRC from agreeing to hide GE-Hitachi information, which could be of critical importance for MELLLA+ safety issues, as will be seen further below.

The US NRC, the US’ Brooklyn lab and others have warned of the dangers of MELLLA+.

You don’t need to be an expert to understand that instability of a nuclear reactor core and the need for emergency depressurization, discussed below, are not good. This is all the more true where the reactors are large and have containment issues. As for ATWS, the NRC states: “An ATWS is one of the ‘worst case’ accidents, consideration of which frequently motivates the NRC to take regulatory action. Such an accident could happen if the scram system (which provides a highly reliable means of shutting down the reactor) fails to work during a reactor event (anticipated transient). The types of events considered are those used for designing the plant.” http://www.nrc.gov/reading-rm/basic-ref/glossary/anticipated-transient-without-scram-atws.html A transient is: “A change in the reactor coolant system temperature, pressure, or both, attributed to a change in the reactor’s power output. Transients can be caused by (1) adding or removing neutron poisons, (2) increasing or decreasing electrical load on the turbine generator, or (3) accident conditions.” http://www.nrc.gov/reading-rm/basic-ref/glossary/transient.html

According to the US Government Brooklyn lab:
“Boiling Water Reactors (BWRs) have in recent years been increasing operating power; sometimes to 120% of their original licensed thermal power (OLTP). This places them in an expanded operating domain and changes how they maneuver in the power-flow operating map. One option being pursued, ‘maximum extended load line limit analysis plus” (MELLLA+) operation [1], raises questions about how the plant will respond to anticipated transients without scram (ATWS)….

In a previous report [2] a discussion is given of how MELLLA+ operation affects the power-flow operating map and the impact of this in an ATWS event. If the initiating event is a turbine trip, then after the automatic trip of the recirculation pumps, the reactor evolves to a relatively high power to flow condition and specifically to a region of the power-flow map where unstable power oscillations are likely to occur. The occurrence of these power oscillations, if left unmitigated, may result in fuel damage. Additionally, the severity of the power oscillations may hamper the effectiveness of mitigation strategies. For example, ATWS events are typically mitigated through the injection of dissolved neutron absorber (boron) through the standby liquid control system (SLCS). The occurrence of oscillation induced core inlet flow reversal may reduce the rate at which this soluble absorber is delivered to the active region of the reactor core. The results of studies of these ATWS events with core instability (ATWS-I) are given in [2, 3].

If the initiating event is the closure of main steam isolation valves (MSIVs) the concern is the amount of energy placed into containment during the mitigation period. This thermal load may exhaust available pressure suppression capacity of the containment wetwell, which would prompt the operators to conduct a manual emergency depressurization according to standard emergency operating procedures. The emergency depressurization raises several concerns: (1) the reactor has undergone a beyond-design-basis event, and fuel damage may have occurred, (2) the pressure suppression capacity of the containment has been exhausted, and (3) the reactor coolant pressure boundary has been bypassed by manually opening the automatic depressurization system valves….” “BNL-98984-2012-IR,
BWR ANTICIPATED TRANSIENTS WITHOUT SCRAM IN THE MELLLA+ EXPANDED OPERATING DOMAIN Part 4: Sensitivity Studies for Events Leading to Emergency Depressurization.” https://www.bnl.gov/isd/documents/79650.pdf (2013)

“Operation in the MELLLA+ domain could lead to conditions that are more difficult to manage during ‘special events,’ such as an anticipated transient without scram (ATWS). The higher power could increase the peak reactor vessel pressure. The higher power / lower flow conditions and the closer proximity to stability boundaries could result in more rapid growth of oscillations during ATWS instability events. There are large uncertainties associated with the predictions of plant responses to ATWS, due to the difficulties in modeling such events and the paucity of data available to validate the calculational methods.” He cheerfully adds the following non-sensical statement: “These uncertainties are tempered to a large extent by the low frequency and the resulting low risk significance of these events.” Huh? Low frequency? How many more nuclear accidents does he think the world can afford to have? And, how are they of low risk significance? He later adds: “The Safety Evaluations were very demanding tasks for which the staff should be commended. The staff performed thorough evaluations and carried out convincing confirmatory analyses where tools were available, such as for the reactor physics and fuel related issues. Unfortunately, the staff did not have the thermal-hydraulic code capability that would have been needed to independently confirm some important parts of the evaluation such as ATWS instability. The TRACE thermal-hydraulic system analysis code has the capabilities needed to address such issues.” http://pbadupws.nrc.gov/docs/ML0717/ML071760346.pdf (NRC, 2007)

“The reviewers of PRM-50-84 may note that among the fuel design features that accommodate operation at the higher power / lower flow conditions there is no allowance for the impact of crud deposits.

The reviewers may further note that the so-called thorough evaluations of MELLA+ by the NRC staff did not include any attention to the impact of crud deposits on fuel related issues. Regarding TRACE, even if it becomes operational, it has no specifications that incorporate the impact of crud deposits.

This letter cites the lack of allowances for crud deposits in the NRC’s evaluations of MELLLA+. Crud deposits are ubiquitous among the worldwide fleet of LWRs, and the issues are of very high safety significance.” http://pbadupws.nrc.gov/docs/ML0720/ML072080298.pdf (Leyse, 2007)

The US NRC has even allowed the hiding of information about Grand Gulf, which appears relevant. The excuse is so-called proprietary reasons:
“Revised Response to Items 1 and 2 of the Request for Additional Information Regarding ‘Maximum Extended Load Line Limit Plus’ Amendment Request, dated 5/19/2014. Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29“:
“The revision removes the text above the 50 MWtlMlbm/hr line on the Thermal Power vs. Core Flow map as agreed during a call with the Nuclear Regulatory Commission (NRC) on September 12, 2014.” Oct. 2, 2014, “Revised Response to Items 1 and 2 of the Request for Additional Information Regarding ‘Maximum Extended Load Line Limit Plus” http://pbadupws.nrc.gov/docs/ML1427/ML14275A050.pdf

On October 23rd, 2014, Alan Wang approved the hiding of information from public disclosure regarding “Maximum Extended Load Line Limit Plus“:
“SUBJECT: GRAND GULF NUCLEAR STATION, UNIT 1 -REQUEST FOR WITHHOLDING INFORMATION FROM PUBLIC DISCLOSURE (TAC NO. MF2798)…
By letter dated August 26, 2014, Entergy Operations, Inc. (Entergy, the licensee), submitted an affidavit dated June 19, 2014, executed by Mr. Peter M. Yandow, GE-Hitachi Nuclear Energy Americas LLC (GEH), requesting that the information contained in the following document be withheld from public disclosure pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR), Part 2, Section 2.390:
Attachment 1 to GNR0-2014100045, “Responses to Request for Additional Information Dated 5/19/2014 Pertaining To License Amendment Request-Maximum Extended Load Line Limit Plus” http://pbadupws.nrc.gov/docs/ML1426/ML14265A384.pdf

However, the following Westinghouse patent makes clear both the dangers of MELLA Plus and that methods related to MELLA Plus can be patented and so there appears no grounds for GE-Hitachi and the NRC hiding information, other than tricking the public!
“(19) United States (12) Patent Application Publication (10) Pub. No.: US 2013/0266107 A1 BLAISDELL et al. (43) Pub. Date: Oct. 10, 2013
METHODS FOR PROTECTION OF NUCLEAR REACTORS FROM THERMAL HYDRAULIC/NEUTRONIC CORE INSTABILITY”

Since the Japanese government approved “Convention on Supplementary Compensation for Nuclear Damage“, already signed by the USA, GE Hitachi and other suppliers are no longer liable-cannot be sued, so they appear ready now to blow the world to smithereens.

The largest US Nuclear Reactor, Grand Gulf, is located along the Mississippi River near Port Gibson, Mississippi, famous for its church steeple, rather than for its nuclear reactor. The finger was NOT meant to point to a radioactive plume!
"First Presbyterian Church, circa 1859, Gold hand atop the steeple pointing heavenward, Port Gibson, Mississippi, USA", photo by Gsmith, CC-BY-SA-3.0
“First Presbyterian Church, circa 1859, Gold hand atop the steeple pointing heavenward, Port Gibson, Mississippi, USA”, photo by Gsmith, CC-BY-SA-3.0 http://commons.wikimedia.org/wiki/File:Port_Gibson_steeple2.JPG

Entergy itself states: “The MELLLA+ operating domain affects only design and operating margins. Challenges to the fuel, reactor coolant pressure boundary, and containment were evaluated for MELLLA+ operating domain conditions.” “Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi” http://www.regulations.gov/#!documentDetail;D=NRC-2014-0250-0001 ONLY?!

Regarding issues with (Grand Gulf) Mark III design problems:
“The staff extended the issue to include BWRs with Mark III containments because their relatively low free volume and strength …
BWR Mark III containment designs have the potential to fail as a result of large hydrogen detonations. These types of severe core damage events have a very low probability of occurrence, but they may produce large quantities of hydrogen through a reaction between hot metal fuel cladding and steam. Plants with these containment designs use hydrogen igniters to control the buildup of hydrogen. The existing hydrogen igniters rely on the plant AC power distribution system for power, and AC power may not be available for certain events with the potential to result in severe core damage. A detonation of a large buildup of hydrogen has the potential to fail these containment types because of their low design pressure and low free internal volume. Therefore, for the hydrogen igniters to be effective in preserving containment integrity, the igniters must have power before a buildup of hydrogen has developed (i.e., before severe core damage has occurred).” http://www.nrc.gov/reading-rm/doc-collections/insp-manual/temp-instructions/ti_2515_174.doc There is actually a several hundred pages long document dealing with the problems of Mark III containment, using Grand Gulf as the model: Schroeder, J.A., Pafford, D.J., Kelly, D.L., Jones, K.R., & Dallman, F.J. (EG and G Idaho. (1991). An assessment of BWR (boiling water reactor) Mark III containment challenges, failure modes, and potential improvements in performance. doi:10.2172/6051208
Grand Gulf Containment
Grand Gulf Containment, as it appears in Schroeder, et. al.

The following is about Monticello Nuclear Generating Plant which is less than one half of the size of Grand Gulf. Note that this says “because MNGP has a small core with low power density,…” Grand Gulf is not small, but rather the largest!
“The MELLLA+ expanded operating domain increases operating flexibility by allowing control of reactivity at maximum power by changing flow rather than control rod insertion and withdrawal. …implementation of MELLLA+ could lead to operation closer to system stability boundaries. As a result, enhanced protection against instabilities must be instituted to enable such operation.
…
Because MNGP has a small core with low power density, ATWS events with timely operator actions are predicted to cause cladding temperatures well below the regulatory limit. The models used in TRACG04 are acceptable for this application. MELLLA+ applications with larger cores and higher power densities may result in instabilities that require the use of heat transfer models in TRACG04 for conditions that are still under NRC review.
….
Broadening of the MNGP operating domain by allowing operation at lower flow without requiring additional compensating measures could reduce the plant’s safety margin. However, limitations and conditions adopted by the licensee maintain acceptable safety margins and satisfy regulatory criteria under MELLLA+ operation. As discussed below, these limitations and conditions affect fuel and nuclear design, thermal and hydraulic design, emergency systems, and transient and accident analyses.
…
Restrictions that no safety relief valves (SRVs) be out-of-service when operating in the MELLLA+ domain
….
The licensee also completed evaluations to assess the impact of MELLLA+ operation on the radiological consequences of design basis accidents and other special events, such as station blackout. The evaluation concluded that MELLLA+ operation is bounded by existing analyses because the accident is not impacted if it occurs in the MELLLA+ domain or the accident is bounded by events in the MELLLA domain.” http://pbadupws.nrc.gov/docs/ML1404/ML14041A136.pdf
“MONTICELLO NUCLEAR GENERATING PLANT MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS PLUS (MELLLA+) LICENSE AMENDMENT REQUEST” The last sentence seems to be a BS way of saying that a nuclear accident is a nuclear accident whether it occurs in MELLLA or MELLLA Plus!

Comparing Monticello and Grand Gulf:
Units operational 1 x 671 MW http://en.wikipedia.org/wiki/Monticello_Nuclear_Generating_Plant
Units operational 1 x 1,500 MWe http://en.wikipedia.org/wiki/Grand_Gulf_Nuclear_Generating_Station

By now it should be clear that when the following statement by the US Nuclear Regulatory Commission is patently false (aka BS):
“The Commission has made a proposed determination that the following amendment requests involve no significant hazards consideration. Under the Commission’s regulations in 10 CFR 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. The basis for this proposed determination for each amendment request is shown below.
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 a notice of issuance in theFederal Register.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.” http://www.regulations.gov/#!documentDetail;D=NRC-2014-0250-0001

Here are the BS specifics for Grand Gulf, that they are trying to sneak through during the Holiday Season. The details of the other cases are at the link above and below:
“ENTERGY OPERATIONS, INC., SYSTEM ENERGY RESOURCES, INC., SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION, AND ENTERGY MISSISSIPPI, INC., DOCKET NO. 50-416, GRAND GULF NUCLEAR STATION, UNIT 1, CLAIBORNE COUNTY, MISSISSIPPI
Date of amendment request: September 25, 2013, as supplemented by letters dated December 30, 2013, March 10, 2014, and April 11, 2014. Publicly-available versions are in ADAMS under Accession Nos. ML13269A140, ML13364A286, ML14069A103, and ML14104A144, respectively.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The proposed license amendment would allow Grand Gulf Nuclear Station, Unit 1 (GGNS) to operate in the expanded Maximum Extended Load Line Limit Analysis Plus (MELLLA+) domain. Specifically, the amendment would change the Technical Specifications (TSs) including the operating power/flow map and a number instrument allowable values and setpoints, and the current core stability solution.
Basis for proposed no significant hazards consideration determination: 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 amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The probability (frequency of occurrence) of design basis accidents occurring is not affected by the MELLLA+ operating domain because GGNS continues to comply with the regulatory and design basis criteria established for plant equipment. Furthermore, a probabilistic risk assessment demonstrates that the calculated core damage frequencies do not significantly change due to the MELLLA+.
There is no change in consequences of postulated accidents when operating in the MELLLA+ operating domain compared to the operating domain previously evaluated. The results of accident evaluations remain within the NRC-approved acceptance limits. The spectrum of postulated transients has been investigated and shown to meet the plant’s currently licensed regulatory criteria. In the area of fuel and core design, for example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) is still met. Continued compliance with the SLMCPR is confirmed on a cycle-specific basis consistent with the criteria accepted by the NRC.
Challenges to the reactor coolant pressure boundary were evaluated for the MELLLA+ operating domain conditions (pressure, temperature, flow, and radiation) and were found to meet their acceptance criteria for allowable stresses and overpressure margin.
Challenges to the containment were evaluated and the containment and its associated cooling systems continue to meet the current licensing basis. The calculated post LOCA [loss-of-coolant accident] suppression pool temperature remains acceptable.
Based on the above, operating in the MELLLA+ domain does not increase the probability or consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
Equipment that could be affected by the MELLLA+ operating domain has been evaluated. No new operating mode, safety-related equipment lineup, accident scenario, or equipment failure mode was identified. The full spectrum of accident considerations has been evaluated and no new or different kind of accident has been identified. The MELLLA+ operating domain uses developed technology, which is applied within the capabilities of existing plant safety-related equipment in accordance with the regulatory criteria (including NRC-approved codes, standards and methods). No new accident or event precursor has been identified. In addition, the changes have been assessed and determined not to introduce a different accident than that previously evaluated.
Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The MELLLA+ operating domain affects only design and operating margins. Challenges to the fuel, reactor coolant pressure boundary, and containment were evaluated for MELLLA+ operating domain conditions. Fuel integrity is maintained by meeting existing design and regulatory limits. The calculated loads on affected structures, systems, and components, including the reactor coolant pressure boundary, will remain within their design allowables for design basis event categories. No NRC acceptance criterion is exceeded.
Because the GGNS configuration and responses to transients and postulated accidents do not exceed the NRC-approved acceptance limits, the proposed changes do 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.
Attorney for licensee: Joseph A. Aluise, Associate General Counsel—Nuclear, Entergy Services, Inc., 639 Loyola Avenue, New Orleans, Louisiana 70113.
NRC Branch Chief: Douglas A. Broaddus.” http://www.regulations.gov/#!documentDetail;D=NRC-2014-0250-0001

Request for hearing- to intervene specifics: “Each contention must consist of a specific statement of the issue of law or fact to be raised or controverted. In addition, the requestor/petitioner 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 requestor/petitioner intends to rely in proving the contention at the hearing. The requestor/petitioner must also provide references to those specific sources and documents of which the petitioner is aware and on which the requestor/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. The petition must demonstrate that the matters raised are within the scope of the proceeding. The issues raised must be material to the finding the NRC must make to support the action involved in the proceeding. A requestor/petitioner who fails to satisfy these requirements with respect to at least one contention will not be permitted to participate as a party.” http://www.regulations.gov/#!documentDetail;D=NRC-2014-0250-0001

Some References-Information which may be useful:

“Boiling water reactors generally have negative void coefficients, and in normal operation the negative void coefficient allows reactor power to be adjusted by changing the rate of water flow through the core. However, the negative void coefficient can cause an unplanned reactor power increase in events (such as sudden closure of a steamline valve) where the reactor pressure is suddenly increased. In addition, the negative void coefficient can result in power oscillations in the event of a sudden reduction in core flow, such as might be caused by a recirculation pump failure. Boiling water reactors are designed to ensure that the rate of pressure rise from a sudden steamline valve closure is limited to acceptable values, and they include multiple safety systems designed to ensure that any sudden reactor power increases or unstable power oscillations are terminated before fuel or piping damage can occur.” http://en.wikipedia.org/wiki/Void_coefficient

“BNL-98984-2012-IR
BWR ANTICIPATED TRANSIENTS WITHOUT SCRAM IN THE MELLLA+ EXPANDED OPERATING DOMAIN ” “Part 4: Sensitivity Studies for Events Leading to Emergency Depressurization” https://www.bnl.gov/isd/documents/79650.pdf

“(19) United States (12) Patent Application Publication (10) Pub. No.: US 2013/0266107 A1 BLAISDELL et al. (43) Pub. Date: Oct. 10, 2013
METHODS FOR PROTECTION OF NUCLEAR REACTORS FROM THERMAL HYDRAULIC/NEUTRONIC CORE INSTABILITY”

“GRAND GULF NUCLEAR STATION, UNIT 1 – VERBAL AUTHORIZATION OF RELIEF REQUEST ISI-17, PROPOSED ALTERNATIVE REGARDING STRUCTURAL WELD OVERLAY (SWOL) ON THE N06-KB RESIDUAL HEAT REMOVAL SYSTEM/ LOW-PRESSURE COOLANT INJECTION “C” NOZZLE TO SAFE-END DISSIMILAR METAL WELD (TAC NO. ME8525)” http://pbadupws.nrc.gov/docs/ML1213/ML121380483.pdf

Schroeder, J.A., Pafford, D.J., Kelly, D.L., Jones, K.R., & Dallman, F.J. (EG and G Idaho. (1991). “An assessment of BWR (boiling water reactor) Mark III containment challenges, failure modes, and potential improvements in performance“. doi:10.2172/6051208 Found online if you look.

Regarding comments: “The U.S. Nuclear Regulatory Commission (NRC) received and is considering approval of 4 amendmentrequests. The amendment requests are for Grand Gulf Nuclear Station, Unit 1; Peach Bottom Atomic Power Station, Units 2 and 3; Vogtle Electric Generating Plant, Units 1 and 2 and Joseph M. Farley Nuclear Plant, Units 1 and 2; and South Texas Project, Units 1 and 2. The NRC proposes to determine that each amendment request involves no significant hazards consideration. In addition, each amendment request contains sensitive unclassified non-safeguards information (SUNSI).
Dates
Comments must be filed by January 2, 2015. A request for a hearing must be filed by February 2, 2015. Any potential party as defined in § 2.4 of Title 10 of the Code of Federal Regulations (10 CFR), who believes access to SUNSI is necessary to respond to this notice must request document access by December 12, 2014. http://www.regulations.gov/#!documentDetail;D=NRC-2014-0250-0001

Reactors involved in comment period:
“Entergy Operations, Inc., System Energy Resources, Inc., South Mississippi Electric Power Association, and Entergy Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne County, Mississippi
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-277 and 50-78, Peach Bottom Atomic Power Station, Units 2 and 3, York and Lancaster Counties, Pennsylvania
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia and Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda County, Texas
Order Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards Information for Contention Preparation” http://www.regulations.gov/#!documentDetail;D=NRC-2014-0250-0001

NB: EMPHASIS-BOLD ADDED BY US THROUGHOUT THIS POST.