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Related documents-info: https://miningawareness.wordpress.com/2015/04/30/us-nrc-sneak-through-alert-comment-deadline-today-at-11-59-et-request-hearing-by-1-june/ Information on requesting a public hearing here: http://www.regulations.gov/#!documentDetail;D=NRC-2015-0073-0001 Anyone living east of the Rockies could be impacted by a nuclear reactor accident anywhere in the US, especially property-owners. One need only look at the serious, long-lasting impacts of Chernobyl upon the UK and Scandinavia.

Failure of a reactor pressure vessel at Grand Gulf Nuclear Reactor in Mississippi (along with many others) has the potential of devastating agriculture (and health) over the entire North American continent with radioactive fallout, based on the huge distances travelled by the Chernobyl plume, which reached as far as the UK. US weapons testing impacted most of North America, as well. Grand Gulf is a Mark III reactor type. https://miningawareness.wordpress.com/2015/01/01/dangerous-maximum-extended-load-line-limit-plus-for-largest-us-nuclear-reactor-urgent-comment-demand-hearing-now/

Mississippi itself is an agricultural and forestry state and arguably the greenest, which would make the impacts all the more devastating. The very worst impacts would fall upon a very poor, majority black, rural population.

There must a public hearing for something so serious as calculating potential reactor pressure vessel failure. Neutron bombardment and hydrogen attack must both be considered, not just neutron bombardment. Plus or minus 20% uncertainty, i.e. 40% total uncertainty, for anything, but especially for the reactor pressure vessel beltline embrittlement is unacceptable. Furthermore, there seems to be a more general side-stepping of statistical method.

There must be a public hearing as the document is not written in an English which is understandable to literate people and because about 20% of the most severely impacted population is estimated to be illiterate. Thus, it should be re-written in proper English and presented orally to the most severely impacted population. Furthermore, it appears to have been made intentionally difficult to find – located toward the bottom of a document, located toward the bottom of the NRC page. It is not clear if all related material is available, either.

The NRC grossly abuses and misleads regarding the ASME (American Society of Mechanical Engineers) code by saying things such as “overpressurization The condition that occurs when pressure exceeds the design pressure of the component of interest by more than 10 percent, in accordance with the ASME Code“, and “Pressure in the reactor coolant and main steam systems should be maintained below 110 percent of the design values in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.” (NUREG-0800) which is false. The ASME standard is not 110 percent of design values. This is also an abuse of math and language. You cannot be 110 percent of the maximum! This is especially true of boiler and pressure vessels!

Plus-minus 20%, which is 40% uncertainty (error-variation) is unacceptable by any standard. 20% uncertainly for beltline is also unacceptable. For something so dangerous they need to be at 98 to 99% certainty with a 50 to 100% contingency. Instead they say: “An extensive benchmarking program has been carried out to qualify the MPM neutron transport methodology. All of the requirements of RG 1.190 have been met. In particular, all C/M results fall within allowable limits (+/- 20% ), and it was determined that no bias need be applied to MPM fluence results. The uncertainty analysis indicates that all fluence results in the beltline region have uncertainty of less than 20%. The results of this analysis are documented in References 1 and 2. This meets the requirement of RP 1.4.1, 1.4.2, and 1.4.3.
The beltline is most at risk of embrittlement-sudden failure.

They are modeling the reactor “as built”, whereas material degradation occurs over time, and the “as built” excluded the nuclear reactor power uprate.

Can “as-built” data for plant structures and material compositions be acceptable when the reactor pressure vessel has undergone material degradation – both neutron bombardment and hydrogen attack- and other parts have been changed? For materials the answer is clearly no. For geometry it appears ok, unless there is bulging-geometrical changes from degraded materials (or component changes). Plus they say “wherever these data are available“, providing a huge loophole.

Trying to understand if the definition of conservative means understating risk, so as to protect the utility (here Entergy), or if their definition means an added a safety margin to protect the public, as it’s supposed to mean, is problematic and impossible without knowing many technical details. Their usage of words appears inconsistent and often Orwellian. Apparently the NRC is populated by functionally illiterate Americans mixed with functionally illiterate immigrants and they have made up a world which is non-sensical from a language, methodological, standards and safety point of view. This mockery of all is a deadly game. The only consolation is that radioactive fallout from an NRC-induced accident will almost certainly land on the NRC headquarters.

Are their safety margins adequate when the repercussions are so serious? They appear not. Based on the little time we have had to examine documents, the safety margin seems to vary from plus 1% to negative 10% with error margins of 20-40%. Treating the old reactor pressure vessel as new means the error is even greater (towards the negative). Thus, their safety margins can be said to be non-existent. They seem to eat into the danger zone. There appears not even 0% safety margin but rather negative!

About embrittlement from hydrogen attack: https://miningawareness.wordpress.com/2015/03/03/nuclear-reactor-cracks-widespread-disease-scourge-warns-nobel-in-chemistry-nominee/

More from some of the related documents:
This change is needed to address a legacy issue in which the current method was determined to have required Nuclear Regulatory Commission (NRC) approval prior to being utilized, and also applies to the Maximum Extended Load Line Limit Plus Analysis Plus (MELLLA+) License Amendment Request (LAR) (Accession No. ML 13269A140).

Entergy has reviewed the U.S. Nuclear Regulatory Commission Regulatory Guide, RG 1.190 (Reference 3). This guide describes the application and qualification of a methodology acceptable to the NRC staff for determining the best-estimate neutron fluence experienced by materials in the beltline region of light water reactor (LWR) pressure vessels, as well as for determining the overall uncertainty associated with those best-estimate values.
….

RP 1.1.1 Modeling Data. The calculation modeling (geometry, materials, etc.) should be based on documented and verified plant-specific data.

GGNS Response: The GGNS proposed methodology is based on documented and verified plant-specific data. Further, the calculations use as-built data for plant structures and material compositions whenever these data are available. The fuel data is specific for each fuel cycle and includes results for power distributions and water densities taken from the fuel depletion analysis. This meets the requirement of RP 1.1.1.

All calculations are performed with the BUGLE-96 library which is collapsed to 47 neutron groups. This library has been thoroughly benchmarked. This meets the requirement of RP 1.1.2.” [See: https://rsicc.ornl.gov/codes/dlc/dlc1/dlc-185.html%5D

They are modeling the reactor “as built”, whereas material degradation occurs over time, and the “as built” excluded the uprate. Can “as-built” data for plant structures and material compositions be acceptable when the reactor pressure vessel has undergone material degradation – both neutron bombardment and hydrogen attack- and other parts have been changed? For materials the answer is clearly no. For geometry it appears ok, unless there is bulging from degraded materials or some other geometric change. Plus they say “wherever these data are available”.
RP 1.1.1 Modeling Data. The calculation modeling (geometry, materials, etc.) should be based on documented and verified plant-specific data.
GGNS Response: The GGNS proposed methodology is based on documented and verified plant-specific data. Further, the calculations use as-built data for plant structures and material compositions whenever these data are available. The fuel data is specific for each fuel cycle and includes results for power distributions and water densities taken from the fuel depletion analysis. This meets the requirement of RP 1.1.1
“. Excerpted from “Grand Gulf, Unit 1 – License Amendment Request Application … – NRC http://pbadupws.nrc.gov/docs/ML1432/ML14325A752.pdf

It seems that they are side-stepping basic statistics needed to evaluate error by saying it doesn’t apply because they are not using the Monte Carlo method, for instance:
RP 1.3.2 Statistical Tests. The Monte Carlo estimated mean and relative error should be tested and satisfy all statistical criteria.
GGNS Response: This requirement only applies to Monte Carlo calculations which are not used here.” “Grand Gulf, Unit 1 – License Amendment Request Application … – NRC

http://pbadupws.nrc.gov/docs/ML1432/ML14325A752.pdf

Why are they using American Nuclear Society Standards? It is a nuclear lobby group.
Incidents of moderate frequency and infrequent events are also known as Condition II and Condition III events, respectively, in the commonly used, oft-cited but unofficial American Nuclear Society (ANS) standards.” “NUREG-0800 U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN
15.0 INTRODUCTION – TRANSIENT AND ACCIDENT ANALYSES

http://pbadupws.nrc.gov/docs/ML0707/ML070710376.pdf

In June 2001, the fuel vendor General Electric reported that licensees that implemented stability detect and suppress trip systems at their plants may be making nonconservative errors in their licensing calculations for reloads, resulting in inadequate MCPR safety limit protection (Part 21 Report 2001-23-0).
Optional stability solutions requiring these calculations are defined as Options I-D, II, and III in the vendor’s document NEDO-32465-A, “Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload Applications,” August 1996. This document gives two generic so-called Delta CPR/Initial CPR Vs. Oscillation Magnitude (DIVOM) curves, one for core-wide mode oscillations and one for regional mode oscillations. The curves relate normalized critical power ratio to hot bundle oscillation magnitude
“. http://www.nrc.gov/reading-rm/doc-collections/gen-comm/info-notices/2001/in01015.html Related: http://www.nrc.gov/reading-rm/doc-collections/event-status/event/2000/20000919en.html

RELATED AND IMPORTANT:
The balance between heat flow and fluid flow can be upset from either side; too much heat flow or too little fluid flow. In a waterwall tube, steam forms as discrete bubbles, nucleate boiling. When the bubble is large enough, the bubble is swept away by the moving fluid, and the cycle repeats. At too high a heat flux or too low a fluid flow, steam-bubble formation is too fast for removal by the moving fluid. Several bubbles join to form a steam blanket, a departure from nucleate boiling, DNB. Heat transfer through the steam blanket is poor, steam is an excellent insulator, and tube-metal temperatures rapidly rise and failure occurs quickly.
http://www.nationalboard.org/index.aspx?pageID=164&ID=186
Short-Term High Temperature Failures“, by David N. French, President, David N. French Inc, Mettalurgists,Category: Operations
Summary: The following article is a part of the National Board Technical Series. This article was originally published in the April 1991 National Board BULLETIN. (4 printed pages)http://www.nationalboard.org/index.aspx?pageID=164&ID=186 See also: http://en.wikipedia.org/wiki/Superheating

In September 2000, the Grand Gulf licensee reported that a main generator partial load rejection can actuate a control circuit that may not always activate a reactor scram or recirculation pump downshift as assumed in the analysis. This condition could adversely affect MCPR limits (Event Notification 37342)“. http://www.nrc.gov/reading-rm/doc-collections/gen-comm/info-notices/2001/in01015.html

http://www.nirs.org/factsheets/bwrfact.htm