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One of the US NRC’s tricks for ignoring risks seems to be saying that reactor pressure vessel (RPV) fracture and the subsequent catastrophic release of radioactive materials into the environment is a “beyond design basis accident”. They then sneak these changes through in a laundry list of what they call “Non-Significant Hazards”. This change for Brown’s Ferry Nuclear Power Station is found below Cybersecurity changes on the same page, suggesting that it is intentionally hidden.
TVA map Brown's Ferry
Tennessee Valley Authority (TVA) sites owned by the US Federal government

Brown’s Ferry Nuclear Power station was the site of what was considered a major fire and near miss for nuclear disaster, which was widely reported, in 1975. It seems that cable fires have become fairly routine in the aging “fleets” of nuclear reactors world-wide.

It was a manager at Brown’s Ferry who took a bribe or kickback on behalf of Kris Singh’s company, Holtec, for its dry cask storage, as well: https://miningawareness.wordpress.com/2015/02/05/why-was-holtec-debarred-as-tva-contractor/ A Brown’s Ferry whistleblower, Joni Johnson, has alleged (ca 2013) that her reports were modified: http://youtu.be/IJSfanjuTFM She has been quoted as saying: “I found myself in the position of becoming a whistleblower when TVA management altered root cause reports I authored to subdue their findings… I hope that bringing this story to public light will force TVA to address the safety significance of altering the findings of teams of engineers and experts for the sake of protecting production and their own bonuses.http://nuclear-news.net/2013/07/08/unsafety-at-browns-ferry-nuclear-plant-and-a-whistleblowers-case/ (Emphasis our own) Ah-hah! Small wonder that they’ve had so many problems. The TVA is owned by the Federal government and should be quickly and easily shut down permanently. An accident will cost the government (i.e. taxpayer) far more than money made. Unit 3 was actually offline for a decade due to problems! How many rats chewed on cables in the interim? Rats are actually more resistant to radiation than most species, as is black mold. Rats are resistant but not immune to radiation. Black mold, and most mold, is able to prosper in highly radioactive environments. Rats and mold!
ORNL PT embrittlement curve 1998
Image from the 1998 ONRL document below

In 1998 ONRL (Oak Ridge National [Nuclear] Lab) reported: “The pressure-temperature (P-T) curve controls the upper-bound to the permissible operating envelope for a reactor pressure vessel (RPV) during the normal start-up and cool-down transients. The P-T operating envelope is progressively restricted because of irradiation embrittlement of the RPV material. In recent years, a number of electric utility companies have reported that the plant-specific P-T operating envelope has become so restricted that operation of the reactor during the heat-up and cool-down transients has become very difficult…. the level of conservatism was less than had been anticipated. Examples of unnecessary conservatism include the use in the P-T curve analyses of, (a) the lower-bound (KIJ crack arrest fracture toughness curve, and (b) a conservative inner surface flaw, having a depth corresponding to 25% of the RPV wall thickness (1/4t). Use of the KIR curve reflects an early concern [3] that fracture could originate from local brittle zones in the RPV as a pop-in and propagate in a dynamic manner to the 1/4t depth. Fracture toughness tests on irradiated weld material, however, showed that data from specimens which had pop-ins fell within the scatter band of data from specimens which faiied with no prior pop-ins [4]. Use of the KI, curve to safeguard against failure initiating from a local brittle zone was shown to be unnecessary….http://www.osti.gov/scitech/servlets/purl/1747/ORNL/CP-100102 Evaluation of Margins in the ASME Rules for Defining t P-T Curve for a RPV“, by T. L. Dickson, W. J. McAfee, W. E. Pennell, and P. T. Williaw Oak Ridge National Laboratory

Above we see what may have been the beginning of manipulating the data to get what they wanted. One sees the term “unnecessary conservatism” used as an excuse to weaken safety. That was almost 20 years ago. They have continued to whittle away at safety margins. One can but hope that this hasn’t been going on even longer. The only place this stops is 1) shut-down of reactors, and ideally the arrest of those who are corrupt in the NRC-Nuclear labs or 2) catastrophic nuclear accident, which will have deadly consequences on North America. The many new immigrants working for the NRC-DOE-nuclear labs will return to their respective homes leaving devastation behind; Obama can go to Kenya. However, the average American who has been in the US for generations will have no place else to go.

True science starts with facts and arrives at hypotheses. It is based on inductive logic. Then the hypotheses can then be tested.

Trying to prove what you want to prove, as the nuclear industry-US government does, is not a foundation for science. It is especially dangerous in the computer age when it is far more easy to play around with numbers to prove what they want to prove to justify keeping the nuclear reactors open.

Thus, their “solution” for age and radiation induced embrittlement was and continues to be playing with the data to prove what they wanted. This is not good science. Science already has enough limitations without them making a mockery of scientific method.

The NRC says in this call for comment on Brown’s Ferry 3:
“The adjusted reference temperature calculations were performed in accordance with the requirements of 10 CFR 50 Appendix G, using the guidance contained in Regulatory Guide 1.190, “Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,” to reflect use of the operating limits to no more than 54 Effective Full Power Years (EFPY).

Furthermore, the 1998 conclusions were based on weld material, which is not going to be the same material, nor the same strength, as the rest of the reactor pressure vessel. And, how wide was the “scatter band of data”? Their scatter band of data was so scattered, as to make it meaningless. It is impossible to get a best fit. If this were a basic chemistry lab they would be made to re-do it. To make things worse, they use the arithmetic average or mean. See pp. 9-10. This is all of the crap that we’ve come to expect from the nuclear industry who could never pass a normal chemistry lab so they either bribed the teacher or went to “nuclear science” which, by all appearances, is a special degree for lazy imbeciles who would never be able to get a normal degree. The rules of science do not count for the nuclear industry, rather it is anything goes. And, with such severe repercussions they need to take the most conservative data points, rather than best fit or any sort of average, anyway.

Returning to 2015:
Dangerous NRC Nonsense

P/T curves are to assess safety margins to protect against brittle fracture and yet brittle fracture appears to be excluded from NRC evaluation because it is beyond design basis. This is clearly nonsensical.

The NRC says: “The proposed changes deal exclusively with the reactor vessel P/T limit curves, which define the permissible regions for operation and testing. Failure of the reactor vessel is not considered as a design basis accident… The proposed change replaces the current set valid up to 20 effective full power years (EFPYs) with a new set valid up to 38 EFPYs, and replaces the current set valid up to 28 EFPYs with a new set valid up to 54 EFPYs.

Failure of the pressure vessel of a PWR or a BWR constitutes an accident beyond the design basis for which there is no safety system – inevitably leading to a catastrophic release of radioactive material to the environment.” “Nuclear Reactor HazardsGreenpeace, 2005, cited in:
http://www.greenpeace.de/files/publications/briefing-cracking-rpv-20150217.pdf

Pressure Temperature curves are to include adequate safety margins to protect against brittle fracture of reactor pressure vessels (RPV), especially in the more at risk RPV beltline. The beltline is at high risk due to both extra irradiation and hoop stress.

Dangers of Mark 1 containment at Brown’s Ferry

The warnings were stark and issued repeatedly as far back as 1972: If the cooling systems ever failed at a “Mark 1” nuclear reactor, the primary containment vessel surrounding the reactor would probably burst as the fuel rods inside overheated. Dangerous radiation would spew into the environment.http://www.nytimes.com/2011/03/16/world/asia/16contain.html?_r=0

The Lamentable History of Brown’s Ferry Unit 3
(NB: Unit 1 had a major fire early on.)

Browns Ferry Unit 3, Athens, AL, Owner: Tennessee Valley Authority Outage dates (duration): March 19, 1985 to November 19, 1995 (10.7 years). Reactor type: Boiling water reactor Reactor age when outage began: 8.0 years. Commercial operations began: March 1, 1977 Fleet status: Third oldest of five reactors owned by the company.
Synopsis The Tennessee Valley Authority (TVA) shut down Unit 3 in September 1983 per NRC order to inspect recirculation system piping. Repairs to cracking identified by the inspections kept Unit 3 out of service until November 1984. During restart, the instrumentation monitoring the water level above the reactor core malfunctioned, forcing the reactor to be shut down. TVA restarted the reactor without fixing the root cause of the instrumentation problem. The instrumentation malfunctioned again during a startup in February 1985. TVA manually shut down Unit 3 on March 9, 1985, to investigate the root cause of the water level instrumentation problems. On March 18, 1985, TVA announced that it was suspending operation of all three Browns Ferry reactors until programmatic weaknesses were corrected. Deficiencies caused by the programmatic weaknesses continued to surface. In 1985, cable tray supports were found to be poorly designed, the emergency diesel generators were found to be poorly maintained, and the control room operators were found to be poorly trained. In 1986, the senior managers brought in to lead the restart effort were found to violate ethical standards and recirculation system piping was found to be cracked and in need of replacement. In 1987, 28 percent of key personnel were found to be unqualified for their duties. In 1988 and again in 1989, Browns Ferry was found to be in non-compliance with fire protection regulations adopted after the 1975 Browns Ferry fire. Consequently, Unit 3 did not restart until November 1995 with a price tag estimated to be nearly $1.4 billion ($1.83 billion in 2006 dollars.)
” (Emphasis our own) Read the entire saga here: http://www.ucsusa.org/sites/default/files/legacy/assets/documents/nuclear_power/browns-ferry-3-ii.pdf

Browns Ferry Nuclear Plant is on the north shore of Wheeler Reservoir in north Alabama. It was TVA’s first nuclear power plant, and the largest in the world when it began operation in 1974. It was the first nuclear plant in the world to generate more than 1 billion watts of power.http://www.tva.gov/sites/brownsferry.htm

There is already a song called “Brown’s Ferry Blues, (starts at 1:08): http://youtu.be/dJv3gQNGYmM It dates from 1930, by the Delmore Brothers, before the nuclear age. It may need to soon be re-written as Brown’s Ferry Nuclear Blues. It also indirectly speaks to the abusive siting of nuclear reactors in areas with economically disadvantaged populations, whether white Appalachians, African Americans, American Indians or Mexican-Americans.

Fire protection issues still not addressed 32 years after the fire

The TVA had not addressed fire safety issues 32 years after the fire. We do not know if they have finally addressed them, but it is doubtful, based on what we have seen of the NRC (e.g. firewalls not up to standard at one reactor; plastic pipes which melt in fire at others): “An investigation by the Nuclear Information and Resource Service (NIRS) has found that the recently-restarted Browns Ferry-1 reactor still does not comply with federal fire protection regulations put into place because of a near-catastrophic fire at the reactor in 1975.
The reactor does not meet the regulations despite the Tennessee Valley Authority having spent $1.8 Billion to revive the long-shuttered reactor.
http://www.nirs.org/press/06-20-2007/1 Rats!

Here is their sneak through amendment. Comment by Thursday, June 4, 11.59 pm NY-DC time (Eastern):
Applications and Amendments to Facility Operating Licenses and Combined Licenses Involving Proposed No Significant Hazards Considerations and Containing Sensitive Unclassified Non-Safeguards Information and Order Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards Information
This Notice document was issued by the Nuclear Regulatory Commission (NRC)
For related information, Open Docket Folder

TENNESSEE VALLEY AUTHORITY (TVA), DOCKET NO. 50-296, BROWNS FERRY NUCLEAR PLANT (BFN), UNIT 3, LIMESTONE COUNTY, ALABAMA
Date of amendment request: January 27, 2015. A publicly-available version is in ADAMS under Accession No. ML15040A698.

Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendment would revise the Technical Specifications (TSs) for Limiting Condition for Operation (LCO) 3.4.9, “RCS [Reactor Coolant System] Pressure and Temperature (P/T) Limits.” The TVA submitted this license amendment request to satisfy a commitment to prepare and submit revised BFN, Unit 3, P/T limits prior to the start of the period of extended operation, as discussed in “Browns Ferry Nuclear Plant (BFN)—Units 1, 2, and 3—Application for Renewed Operating Licenses,” dated December 31, 2003 (ADAMS Accession No. ML040060359).

Specifically, the proposed change affects the current sets of TS Figures 3.4.9-1, “Pressure/Temperature Limits for Mechanical Heat up, Cooldown following Shutdown, and Reactor Critical Operations,” and 3.4.9-2, “Pressure/Temperature Limits for Reactor In-Service Leak and Hydrostatic Testing.” The proposed change replaces the current set valid up to 20 effective full power years (EFPYs) with a new set valid up to 38 EFPYs, and replaces the current set valid up to 28 EFPYs with a new set valid up to 54 EFPYs.

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 any accident previously evaluated?
Response: No.
” [This is false, unless reactor pressure failure wasn’t evaluated. The beyond design event isn’t evaluated?]

The proposed changes are to accept operating parameters that have been approved in previous license amendments. The changes to P/T limit curves were developed based on NRC-approved methodologies. The proposed changes deal exclusively with the reactor vessel P/T limit curves, which define the permissible regions for operation and testing. Failure of the reactor vessel is not considered as a design basis accident. Through the design conservatisms used to calculate the P/T limit curves, reactor vessel failure has a low probability of occurrence and is not considered in the safety analyses. The proposed changes adjust the reference temperature for the limiting material to account for irradiation effects and provide the same level of protection as previously evaluated and approved.

Notice what they just said: “FAILURE OF THE REACTOR VESSEL IS NOT CONSIDERED AS A DESIGN BASIS ACCIDENT. THROUGH DESIGN CONSERVATISMS USED TO CALCULATE THE P/T LIMIT CURVES, REACTOR VESSEL FAILURE HAS A LOW PROBABLY OF OCCURRENCE AND IS NOT CONSIDERED IN THE SAFETY ANALYSIS” BUT THEY STARTED REMOVING CONSERVATISMS AS EARLY AS 1998! HOW “LOW PROBABLY” IS LOW? NO PROBABILITY IS ACCEPTABLE.

The adjusted reference temperature calculations were performed in accordance with the requirements of 10 CFR 50 Appendix G, using the guidance contained in Regulatory Guide 1.190, “Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,” to reflect use of the operating limits to no more than 54 Effective Full Power Years (EFPY).

These changes do not alter or prevent the operation of equipment required to mitigate any accident analyzed in the BFN Final Safety Analysis Report.” AND SO
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.” HUH? HUH? HUH? HELLO? WHAT KILLED THEIR BRAIN CELLS? DRUGS OR RADIATION? THIS MAKES NO SENSE. MITIGATE MEANS TO MAKE AN ACCIDENT LESS SEVERE. WE DON’T WANT THE ACCIDENTS IN THE FIRST PLACE! THERE IS NO SUCH THING AS A “LESS SEVERE” FRACTURED PRESSURE VESSEL IN A NUCLEAR REACTOR! OH, IT WASN’T EVALUATED? BUT THE WHOLE POINT OF THE PRESSURE TEMP LIMIT EVALUATION IS TO PREVENT REACTOR PRESSURE VESSEL FAILURE.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes are accepted operating parameters that have been approved in previous license amendments. The changes to the P/T limit curves were developed based on NRC-approved methodologies. The proposed changes to the reactor vessel P/T limit curves do not involve a modification to plant equipment. No new failure modes are introduced. There is no effect on the function of any plant system, and no new system interactions are introduced by this change.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
” YES IT DOES IF THEY DIDN’T EVALUATE THE RISK OF PRESSURE VESSEL FAILURE. OR IT WAS PREVIOUSLY EVALUATED. OR IT WAS NOT. EITHER STATEMENT ONE OR STATEMENT TWO IS INACCURATE.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
” YES IT DOES – LIARS!
The proposed changes are accepted operating parameters that have been approved in previous license amendments. The changes to P/T curves were developed based on NRC-approved methodologies. The proposed curves conform to the guidance contained in Regulatory Guide 1.190, “Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence,” and maintain the safety margins specified in 10 CFR 50 Appendix G.
Therefore, the proposed change 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.
Attorney for licensee: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., WT 6A-K, Knoxville, Tennessee 37902. NRC Branch Chief: Shana R. Helton
“. http://www.regulations.gov/#!documentDetail;D=NRC-2015-0092-0001

Included on the same page is a request, which appears to be testimony to the dangers of operating an old Mark 1 “Fukushima style” containment. While a move in the right direction, is this sufficient? Probably not. It’s probably based on the lowest number for when it is new and other data manipulation. These reactors are clearly too old to operate and have been defective from the beginning. The TVA is owned by the US Federal government so it is easier for these reactors to be promptly shut down!
TENNESSEE VALLEY AUTHORITY, DOCKET NOS. 50-259, 50-260, AND 50-296, BROWNS FERRY NUCLEAR PLANT, UNITS 1, 2, AND 3, LIMESTONE COUNTY, ALABAMA
Date of amendment request: December 11, 2014. A publicly-available version is in ADAMS under Accession No. ML14363A158.
Description of amendment request: This amendment request contains sensitive unclassified non-safeguards information (SUNSI). The amendments would revise Section 2.1.1, “Reactor Core SLs [Safety Limits],” of the Technical Specifications (TSs) for all three units, to lower the value of the reactor steam dome pressure safety limit from the current 785 pounds per square inch gauge (psig) to 585 psig. The proposed lowering of this safety limit will effectively expand the validity range for the units’ critical power correlations and the calculation of the minimum critical power ratio. Specifically, the revised value of 585 psig is consistent with the lower range of the critical power correlations currently in use at the units. The revised value will also adequately bound a pressure regulator failure open transient event. No hardware, design, or operational change is involved with this proposed amendment.
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. The NRC staff performed its own analysis, 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 proposed change to the safety limit in TS Section 2.1.1 will continue to support the validity of the existing critical power correlations applied at the units. The proposed TS revision involves no change to the operation of any system or component during normal, accident, or transient operating conditions. The proposed amendment does not involve any modification to plant hardware, design, or operation.
Therefore, the proposed amendment does not involve a significant increase in 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.
The proposed reduction in the reactor dome pressure safety limit from 785 psig to 585 psig is an administrative change and does not involve changes to the plant hardware or its operating characteristics. As a result, no new failure modes are being introduced.
Therefore, the proposed amendment does not introduce a new or different kind of accident from those previously evaluated.
3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No.
The margin of safety is established through the design of plant structures, systems, and components, and through the parameters for safe operation and setpoints of equipment relied upon to respond to transients and design basis accidents. The proposed change in reactor dome pressure does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. The change does not alter the behavior of the plant equipment, which remains unchanged.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee’s analysis and, based on its own analysis, determines 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: General Counsel, Tennessee Valley Authority, 400 West Summit Hill Dr., WT 6A-K, Knoxville, Tennessee 37902.
NRC Branch Chief: Shana R. Helton.
Order Imposing Procedures for Access to Sensitive Unclassified Non-Safeguards Information for Contention Preparation
Dominion Energy Kewaunee, Inc., Docket No. 50-305, Kewaunee Power Station, Kewaunee County, Wisconsin

http://www.regulations.gov/#!documentDetail;D=NRC-2015-0092-0001

Nuclear Engineering is clearly B.S. and not a Bachelor’s of Science Degree!

What then is M.S.E? Master of Satanic Energy?
Shana Helton Ms. Shana Helton has served as the Chief of the Rulemaking Branch in the Division of Policy and Rulemaking in NRR for the past 4 years. She joined the NRC in 2002 through the Nuclear Safety Professional Development Program, as a reactor safety engineer in RES. Ms. Helton joined NMSS in 2004, holding progressively more responsible positions associated with licensing, security, and technical reviews of radioactive materials transportation packages and spent nuclear fuel storage cask designs. She has also completed rotational assignments to RIII and OEDO. Ms. Helton holds a B.S. in Nuclear Engineering from the University of Illinois, and an M.S.E. from the University of Michigan in Nuclear Engineering and Radiological Sciences.
Page Last Reviewed/Updated Friday, May 23, 2014

http://www.nrc.gov/public-involve/conference-symposia/ric/past/2014/docs/bios/bio111.html

Brown’s Ferry Study Commissioned by Concerned Citizens-Watchdogs:

RADIOACTIVE EMISSIONS AND HEALTH HAZARDS SURROUNDING BROWNS FERRY NUCLEAR POWER PLANT IN ALABAMA RADIATION AND PUBLIC HEALTH PROJECT“, by JOSEPH MANGANO, MPH, MBA AND BEST/MATRR GRETEL JOHNSTON, JUNE 4, 2013, “REPORT COMMISSIONED BY BEST/MATRR A CHAPTER OF THE BLUE RIDGE ENVIRONMENTAL DEFENSE LEAGUE; BELLEFONTE EFFICIENCY & SUSTAINABILITY TEAM; MOTHERS AGAINST TENNESSEE RIVER RADIATION; MATRR.ORGhttp://best-matrr.org/pdfs/AL_BFN_Report_2013-final-dig2.pdf

Joni Johnson at NRC hearing: http://youtu.be/IJSfanjuTFM

Historical info: http://en.wikipedia.org/wiki/Battle_of_Brown%27s_Ferry
[It is supposed to be Brown’s Ferry. Due to apparent laziness it seems to have become Browns Ferry for the nuclear reactor.

Update Comment Sample. This is to show how easy it is. Unless you upload a pdf there are limitations on length. A comment needs to be in your own words so that they have no excuse for hiding the comments. The EPA and possibly the NRC do not upload comments which are identical.
Brown's Ferry 3 PT comment
Point of clarification. In the context of the Belgian nuclear reactor cracks, Dr. MacDonald explained that the computer programs did not predict as expected. What this most likely means is that something was wrong in the input assumptions. He has invested a lot of time and energy in this topic and it must surely have been difficult for him to admit that. There are many input variables and unknowns about radiation impacts on materials.