Proposed breakaway corrosion testing is Absurd

DG-1261 is absurd.

Some background.
The U. S. Nuclear Regulatory Commission (NRC) is managing a series of public meetings on THE PERFORMANCE-BASED EMERGENCY CORE COOLING
SYSTEMS CLADDING ACCEPTANCE CRITERIA (TITLE 10 OF THE
CODE OF FEDERAL REGULATIONS SECTION 50.46c) PROPOSED
RULE AND ASSOCIATED DRAFT REGULATORY GUIDANCE, 
Tuesday June 24, 2014 through Thursday June 26, 2014.

Westinghouse presented a series of 20 slides and on June 25 I briefly commented that DG-1261 is absurd.  Thus I stated, "I have modified the first line of slide 20 by substituting  absurd for the remainder of the slide."  Following is that slide as modified:

Westinghouse Non-Proprietary Class 3
© 2014 Westinghouse Electric Company LLC. All Rights Reserved.
10 CFR 50.46c Proposed Rule and Associated Draft Regulatory Guides
Westinghouse Electric Power Company Recommendations
David Mitchell, Fellow Engineer, Rockville, MD
June 24, 2014

Conclusion and summary on DG-1261 (Slide 20)
•  Proposed breakaway corrosion testing and reporting is absurd.

On June 26, I briefly promised to submit a formal comment and I added that an appropriate reference is  WCAP-12610, Appendix E, August, 1990.  As readers will read read  several lines below in the DOCKET of December 14, 2002, I had access to a highly censored version of proprietary WCAP-12610 at that time.  However, when I again sought that report on June 27, 2014, I had no success, and NRC's PDR reported as follows;

Subject:	RE: WCAP-12610, Appendix E, August, 1990
Date:	6/27/2014 1:53:03 P.M. Mountain Daylight Time
From:	PDR.Resource@nrc.gov
Reply To: To: Bobleyse@aol.com

Hi Bob,

This document is non-public.  I found its record in ADAMS Legacy, which is below.  You would need to file a FOIA request for it.

------------------------------------------------------------------------------------------------

Document Title: Proprietary WCAP-12610,App E, "ZIRLO High Temp Oxidation Tests." Withheld.

  Document Date: Aug 31, 1990

  Estimated Page Count: 18

  Document Type: "REPORTS-TOPICAL (BY MANUFACTURERS-VENDORS ETC)"; "TEXT-SAFETY REPORT"

  Availability: Non-Publicly Available

  Accession Number: 9009130156

  Document/Report Number: "WCAP-12610-APP"; "WCAP-12610-APP-E"

------------------------------------------------------------------------------------------------

Sincerely,

Mary Mendiola

Technical Librarian

US NRC Public Document Room

http://www.nrc.gov/reading-rm/pdr.html

301-415-4737

800-397-4209

PDR.Resource@nrc.gov

From: Bobleyse@aol.com [mailto:Bobleyse@aol.com]
Sent: Friday, June 27, 2014 2:45 PM
To: PDR Resource
Subject: WCAP-12610, Appendix E, August, 1990

Hello:

10 years ago or more I found the subject documentation, WCAP-12610, Appendix E, August, 1990.  Today I am having no luck.  I believe that PDR experts may find this.

Thank you,

Bob   bobleyse@aol.com

DOCKET
(6 1 F 511 s3 )
December 14, 2002
Petitioner's Responses to Comments by Westinghouse and NEI
DOCKETED
USNRC
December 16, 2002 (4:30PM)
OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF
On page 2 of the attachment to its comments dated October 22, 2002, Westinghouse states, "More recently, Westinghouse conducted tests with pure oxygen instead of steam." With difficulty, the Petitioner located a reference that apparently describes this work, WCAP-12610, Appendix E, August, 1990. Only a limited portion of the report is available to the public and it is classified by Westinghouse as a proprietary report. The high temperature oxidation tests were performed by Nuclear Electric, plc in the United Kingdom. Twenty four ZIRLO alloy and six Zr-4 samples were tested at temperatures ranging from 1832F to 2372F. The cylindrical tubing specimens were approximately 0.6 inches long and were from production grade 17x1 7 tubing.

Appendix E candidly discloses: "Since, particularly at high temperatures, the self heating of the specimen results in its being at a higher temperature than its surroundings, any temperature measured will be equal to or lower than that of the test specimen." In other words, in order for the investigators at Nuclear Electric to prevent runaway from the heat of reaction at high temperatures (self heating) it was necessary to maintain the surroundings at a substantially lower temperature than the specimen. In this manner, the heat loss by radiation to the relatively cold surroundings compensated for the heat produced by chemical reaction with the pure oxygen. This then leads to the question: What if Nuclear Electric had conducted the investigation with a 17x 17 arrangement of ZIRLO or Zr-2 tubes captured within a Zircaloy-4 structural grid with ZIRLO thimbles as depicted in FIGURE 2-1 of WCAP-12610? The answer is that the assembly would have rapidly been destroyed in runaway if a sufficient flow of oxygen had been maintained.

The oxidation tests of the ZIRLO alloy and Zr-4 samples were conducted within a very "quiet" oxygen atmosphere. The apparatus was extremely delicate. The investigators reported, "Pure oxygen gas was used as the oxidant rather than steam. It is believed that, if steam were used, condensation on the suspension wire could invalidate the weight gain measurements." From this it may be inferred that the apparatus was certainly insufficiently robust to accommodate the turbulent thermal hydraulic conditions of LOCA. The oxygen supply system and flow rates are not disclosed in Appendix E, but it must have been a very tender application of oxygen to not upset the suspension wire and the weight gain apparatus.

For Whom is this Blog? 65 YEARS IN THE GAME

Silver Nitrate and Berl Saddles
Fish and Type-K
ANL-172
SBLOCA before the acronym
CP-5 Autoclave  
Shipping Bolts in Expansion Joints 105-C
Quenching a Quatrefoil
YTB-268
Arikara Muffler
Oriskany Accumulators
EBWR Scale
Stainless burnout in CP-5
Flocculation
GETR Power Level
Fuel Storage
B3B and loop
COOSHOO
SL-1 at
Aluminum Nitride
RTDs
J
FLECHT 
Wasted 
RADCAL
TMI
Kemeny
UHI ultra high risk
Clerical in Coal (decompose)
 HITBIT at WJ
Wench games there
Proposal to feds
HITBIT at UCLA
Penn State Junk
Vent Sizing
Runaway and PRMs
Fouling and PRMs
Core exit thermocouples

So, how do I cover all of the above fast?  Each game is involved.  Moreover, on the average is it is more of a set of rackets than games although most of the items are games.

Game Silver Nitrate and Berl Saddles.  My first pay as a chemical engineer out of the University of Wisconsin, August 1950, was from Hanford Works, Richland, Washington, run by General Electric Company.  First I was in the stack gas group for three months where we were proof testing silver reactors for iodine removal from dissolver off gases.  As we walked across the test site, the boss, Al Blasewitz asked me where I would place the control thermocouple. I quickly responded that it should be placed at the inlet to the silver reactor.  That was done.  Later, in service, a batch of callrod heaters melted down when there was power but no gas flow.

Game Fish and Type-K.  Next, I spent three months in site survey.  Fish from the Columbia River were placed on type-K X-ray paper.  The skeleton showed up.

Game ANL-172.  This is worth more space than I'll use partly because the experience was helpful about 20 years later when I designed a gamma thermometer with in-core electric calibration that is in use today in some BWRs as well as licensing of the ESBWR. The experiment proved  that thermal conductivity of uranium-zirconium alloys is unchanged with accumulated fissions.   ANL and ORNL sent an experiment to Hanford that I installed and operated in a production reactor.  It was still the early days and I was pretty much left alone.  Also, Rickover's gang was not on top of this job and never has been.  I pressure tested that capsule assembly and found that it leaked like a sieve.  A reasonable approach would have been to send it back to ANL.  Instead, I worked out a game to operate the experiment as a nitrogen pressurized assembly which prevented inleakage to the water-cooled  capsule.  The capsule housed  two cylinders of alloy which were water cooled on one face and the temperature gradient was measured.  A separate assembly measured the power; it was a thin strip of uranium-aluminum alloy wrapped in a wire heater with thermocouples to monitor a temperature difference.  After a week of so of exposure it became apparent that the power meter was deteriorating.  I had gas samples analyzed which showed that organic insulation was outgassing.  So I purged the capsule by alternately increasing and decreasing the nitrogen pressure until the contaminants were removed and I had a valid calibration point. Thus it was demonstrated that accumulated fissions were not changing the thermal conductivity of the alloy.  After a month of so of operation, R. Bergren, a scientist from ORNL, visited and checked my work. He reported that the experiment was being operated in a competent manner.  As an aside, Bergren guessed that the power level of the pile was few megawatts; I responded that it was over a hundred times that and he damn near passed out.

Game SBLOCA before the acronym.  This is very condensed.  It was late 1953 or early 1954.  I had been working at ANL since November 1952.  Again, Rickover's gang and others had little interest in the pressurized water loop that I was installing in the Material Testing Reactor at the National Reactor Testing Station (S1W had been in operation and things worked).  Anyway, I was at the loop during initial runs when a small break LOCA opened up and it could not be isolated.  I led the recovery with extreme skill.  I valved the low-capacity piston pump to feed the lower end of the pressurizer level standpipe.  As the pressurizer level became low, I turned off the 100-A canned rotor pump.  Fed McMillan and I valved off the in-pile tube and valved in standby cooling and that was that.  There was no failure of the fuel under test.  Of course this became a very involved situation. Back at Argonne several months later I overheard a remark that Leyse was nuts to think that the low-capacity piston pump would impact cooling.  

Game CP-5 Autoclave.  It's torture to be brief.  Zinn included this in his Geneva talk around 1954. EBWR fuel, a zirconium clad uranium alloy, was supposed to be corrosion resistant if the cladding leaked.  I designed and mocked-up a pressurized water capsule for irradiating bare fuel pins.  The air-cooled finned capsule was to operate with subcoled nucleate boiling in relatively cold water. An external expansion chamber was pressurized with nitrogen and this pressure established  the saturation temperature and thus the approximate temperature of the 0.10 diameter by 1 inch specimen.  Enrichment was 10% so there was self-shielding which would be lost if the specimen disintegrated (power would then increase).  End of story.

Game Quenching a Quatrefoil.  During the summer of 1954 ANL cancelled my draft deferment and I chose not to be an Army private working on some kind of PWR for the Antarctic at Ft. Belvoir.  So I went to DuPont Savannah River  Plant near Aiken, South Carolina.  One task was extending the time allowed in fuel transfer through air from the water in the reactor to the water cooling in the spent fuel pool.  Construction work was slowing and men and  equipment  were available.  I found access to a rejected aluminum quatrefoil (four joined tubes) loaded with natural uranium slugs.  A heat treating furnace was available. Nearby the construction gang dug a 2 foot diameter hole about 20 feet deep and installed a steel liner that was filled with water.  So, we heated the quatrefoil in the furnace, quenched it in the pool, the slugs remained intact, and we increased the allotted time before emergency cooling would be required in the event of a hang-up during fuel transfer.

Racket Shipping Bolts in Expansion Joints 105-C.  The reactor was in cold shakedown and we had assignments.  Curious, I looked at an expansion joint in one (of maybe four) primary loops.  I could not figure out how  that worked.  It turns out that the shipping bolts had never been removed although the loop assembly was complete and  in the process of hydraulic shakedown.  So, I checked around elsewhere.  I walked through several pipe tunnels with 2 foot (or more) diameter cooling water supply and return lines.  I found several more shipping bolts in place, one was stretched.  My young boss, bright but not too wise, told be to write it up, so I issued Shipping Bolts in Expansion Joints - 105-C.  That was not a smart move and after about a year on the job I got a 3% raise and decided that I better move on.  

Game YTB-268. I left DuPont;  the draft board was still there so I went into OCS Newport, confident that I'd wind up in Navy Nuke work.  Luckily I was wrong, the San Francisco Naval Shipyard was far better territory for almost three years.  YTB-268 was a tug boat with direct diesel drive.  As I learned decades later, YTB means Yard Tug Big, however today I found out that it means

YTB - Harbor Tugs, Big

     Harbor tugs with more than 800 h.p

RED CLOUD

YT-268

YTB-268

 The boat wiped its crankshaft, was repaired and wiped its crankshaft again during dock trials when nobody kept track of a temporary filter that clogged and blocked oil flow.  So, Ship Superintendent  Robert H. Leyse, was assigned full time to the tasks.  Quickly, a new crankshaft arrived from Enterprise Diesel (across the bay), Shop 31 (Inside Machinists) chocked it up and decided that the end connecting flange was not perpendicular to the axis of the shaft.  Somehow I got wind of this and naively mentioned this to Carl Fixman who ran a design gang.  Carl flipped and sent his engineer, Mr. Barry to accompany me to Shop 31.  Barry had substantial experience in the game, having worked for Enterprise Diesel.  He locked horns with the Shop Master, Mr. Stiver, who insisted that he was Master of Shop 31.  In the end, Shop 31  checked its work and the flange was again machined.  Jumping ahead, I closely monitored the dock trials and bay trials came next.  It was pleasant summer day as we circled Angel Island and elsewhere.  In the pilot house I asked the Chief why the stack gas temperature was higher that each of the cylinder exhaust temperatures.  He replied that the sensor tips kept burning off so they installed shorter ones.
  
Game Oriskany Accumulators.   Of course there is a lot going on in the conversion of an aircraft carrier.  The Oriskany got an angled deck, a mirror landing system, steam cats and a lot more.  Right now is submerged off the Florida coast somewhere a site for recreational diving.

USS Oriskany Dive Site, 25 miles south of Pensacola FL

The yard's boilermakers fabricated the accumulators for the steam cats.  Those had to be heat treated, but they were too long for the furnace so they were angled from the lower left front corner to the right top corner.  We had thermocouples at several spots.  Of course, with my quatrefoil experience at DuPont I was an expert in flame management to get an acceptable uniform temperature distribution.  It was a night shift operation for some reason and the crew followed my instructions just as if I had that authority.  No Shop Master was on the scene.  

Another Game Oriskany Accumulators was a trivial but important correction.  The accumulators were hung from the top and expanded downward.  The lower drain line, maybe a one inch line, extended from the bottom, then angled horizontally for perhaps 2 feet and then again angled vertically to a fixed penetration downward through the hangar deck.  Anyway it was a bad scene with excessive pipe stress at the fixed penetration.  It was corrected by Shop 56 with added bends and length to add flexibility.  I insisted that the plans should be corrected, however the piping designers insisted that corrections were unnecessary.

Game Arikara Muffler.   The Arikara's muffler was totally shot.  I didn't give it a second thought when I told the boilermakers to replace it with stainless steel.  Decades later I have looked  a bit, a very little bit, into marine exhaust system design.  Maybe unspecified stainless steel was not smart.  I'll never know how well the Arikara's muffler fared since 1957.  
ARIKARA   ATF-98
ATF  -  Auxiliary - Ocean Tugs, Fleet
Fleet tugs for combat operation - having large radius of  action, good fire fighting, salvage and all around facilities. 

Game EBWR Scale.  So I left the Navy and returned to the Argonne National Laboratory during November 1958.  Immediately I initiated deep investigations related to fouling of fuel elements in the Experimental Boiling Water Reactor.  I did not label it crud; I called it scale which it was.
I issued several memoranda;

Effect of Scale Deposits on Fuel Element Temperatures with EBWR at 100 Megawatts,June 17, 1959,

Thermal Conductivity of Scale on EBWR Fuel Elements, December 24, 1959,

Scale on Fuel Elements in EBWR, June 22, 1959,

Post –Irradiation Heating of EBWR Element H-18, December 18, 1959.

 
I was free to initiate and execute the above work, and never considered that it could be otherwise.  In today's world that would not be allowed, the masters inside the beltway would not allow it and if a proposal was written, the lobbyists would have it killed and the initialtor would be fired.  As it turned out, I managed to get the work done before the beltway knew about  it.  My leader was deadwood and that helped.  Also, hot labs and  money was there and EBWR was in operation.


 http://www.pnl.gov/main/publications/external/technical_reports/PNNL-17644.pdf
Predictive Bias and Sensitivity in
NRC Fuel Performance Codes
Manuscript Completed: April 2009
Date Published: October 2009
Prepared by
K.J. Geelhood, W.G. Luscher, C.E. Beyer, D.J. Senor, and
M.E. Cunningham, D.D. Lanning, H.E. Adkins
Pacific Northwest National Laboratory
P.O. Box 999
Richland, WA 99352
NRC Job Code N6326
Office of
A search was performed for published CRUD thermal conductivity data. There was very little
information found. Leyse (2003) stated a value of 0.8 W/m-K for hydrated alumina in the
Experimental Boiling Water Reactor (EBWR). This is close to the value used in FRAPCON-3.3.


108. Leyse, R.H. 2003. “Unmet Challenges for SCDAP/RELAP5-3D. Analysis of Severe
Accidents for Light Water Nuclear Reactors with Heavily Fouled Cores,” 2003
RELAP5 International Users Seminar, West Yellowstone, Montana, 27-29 August
2003

Proposed 10 CFR 50.46c and PRM-50-93 (In preparation)

Here is the link to PRM-50-93. 
http://pbadupws.nrc.gov/docs/ML0932/ML093290250.pdf

And, here is a link to the Proposed 10 CFR 50.46c

• ML12283A174 - Proposed Rule: Performance-Based Emergency Core Cooling Systems(ECCS) Cladding Acceptance Criteria. (151 page(s), 3/6/2014)

Proposed 10 CFR 50.46c and Rigged Research (in preparation).

Along with the proposed 10 CFR 50.46c, the NRC has published  three proposed Regulatory Guides.  Following is excerpted from one of the recent draft regulatory guides as evidence of the rigged research that has characterized the continuous maneuvers that have delayed reform of 10 CFR 50.46.  This has been going on for decades, the recent draft regulatory guide refers to year 1996.

DRAFT REGULATORY GUIDE DG-1261  ML12284A324

(Proposed New Regulatory Guide)

CONDUCTING PERIODIC TESTING FOR BREAKAWAY

OXIDATION BEHAVIOR


Background

In 1996, the NRC initiated a fuel-cladding research program intended to investigate the behavior of high-exposure fuel cladding under accident conditions. This research program included an extensive LOCA research and testing program at Argonne National Laboratory (ANL), as well as jointly funded programs at the Kurchatov Institute (Ref. 2) and the Halden Reactor Project (Ref. 3), to develop the body of technical information needed to evaluate LOCA regulations for high-exposure fuel. 

Note: ACRS reviewed the NRC fuel-cladding research program during 2002,  following are Blogger Leyse's notes from then.

Runaway Discussions at the ACRS (2002).

The USNRC is currently working on revisions to rule 10 CFR 50.46 concerning emergency core cooling systems for reactors.  The process is called  risk-informing the regulation.  The ACRS discussions of Friday, May 31, 2002, are revealing in that several aspects of the revisions were discussed, however, the ubiquitous fouling of today’s LWRs was not considered.  This was a combined meeting of three of the most influential subcommittees of the ACRS: Materials and Metallurgy; Thermal Hydraulic Phenomena & Reliability and Probabilistic Risk Assessment

        
 Member Graham B. Wallis was especially enraged by the limited approaches to fuel integrity under LOCA conditions.  In response to detailed descriptions of fracture of corroded specimens of cladding from irradiated power reactor fuel he asserted:  “It seems to be that both these coursing tests and hitting tests, impact tests and the squeezing tests are not really typical of the loads imposed on the real cladding.. I keep wondering what the relevance of all these tests are to the real truth.”  He also reacted to the discussions of  runaway,  “I think when you come back and talk about run-away to this committee you better have a criterion for run-away and not this sort of vagueness about heat transfer.” 

Returning to the excerpts from DG-1261  
Establishing the Onset of Breakaway Oxidation

The experimental procedure provided in Appendix A to this regulatory guide defines a procedure acceptable to the NRC staff to measure the onset of breakaway oxidation (Appendices B through E are provided to expand on critical aspects of the testing procedure in Appendix A). This experimental procedure may be used to characterize the onset of breakaway oxidation as a function of temperature for a zirconium cladding alloy. 

A-7.2 Steamflow Rate
The average steamflow rate used in breakaway oxidation studies should be determined (and reported) from the mass of condensed water collected during these long-time tests or by the mass of water
that is input to the steam chamber divided by the test time and normalized to the net cross-sectional area of the steam chamber. The average steamflow rate should be in the range of 0.5 to 30 mg/square
centimeter per second (cm2 · s).

A-7.3 Steam Pressure
Breakaway oxidation tests should be conducted at a steam pressure at or slightly above atmospheric pressure. 

At this point Blogger Leyse again interrupts the excerpts to convert 0.5 to 30 mg/square centimeter per second to meaningful units, like feet per second of steam flow.

1. First, convert 1 mg of steam at 100 degrees Centigrade and one atmosphere to cubic centimeters of steam:
So (0.001/18) x (373/273) x 22,400 = 1.7 cubic centimeters

2. Then 1 mg/square centimeter per second becomes 1.7 centimeters per second or 1.7/2.54 = 0.67 inches/second

3. So, an average steamflow rate that is in the range of 0.5 to 30 mg/square centimeter per second (cm2 · s) is 0.33 to 20.1 inches per second, or 0.03 to 1.67 feet per second.

The steaming conditions of this regulatory guide are thus far removed of the realities of the mixed thermal hydraulic conditions of severe LOCAs.  The NRC uses the data in "improving" its series of codes such as TRACE.  In this way, licensing of nuclear power plants proceeds including life extensions and power level increases.

The nuclear power industry and its lobbyists do not mind arguing about various aspects of proposed 50.46c as well as the assortment of proposed regulatory guides.  In the meantime, aspects including Baker-Just, Cathcart-Powell, and 2200 degrees Fahrenheit are blessed in the proposed 50.46c, and the assertions of PRM-50-93 (ML093290250) are not addressed.

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 Figure 1. E110 cladding test specimen
Source: NUREG/CR-6967