Friday, April 18, 2014

Comment Number 1, Performance-Based Emergency Core Cooling Systems Cladding Acceptance Criteria

I'll edit this later,for now, here it is. ADAMS ML14098A491

1
Rulemaking1CEm Resource
From: RulemakingComments Resource
Sent: Tuesday, April 08, 2014 4:41 PM
To: Rulemaking1CEm Resource
Cc: RulemakingComments Resource
Subject: FW: Strange reference to Baker-Just
DOCKETED BY USNRC—OFFICE OF THE SECRETARY
SECY-067
PR#: PR-50 and PR-52
FRN#: 79FR16105
NRC DOCKET#: NRC–2008–0332, NRC–2012–0041, NRC– 2012–0042, NRC–2012–0043
SECY DOCKET DATE: 04/03/14
TITLE: Performance-Based Emergency Core Cooling Systems Cladding
Acceptance Criteria
COMMENT#: 01
From: Bobleyse@aol.com [mailto:Bobleyse@aol.com]
Sent: Wednesday, April 02, 2014 9:10 PM
To: Inverso, Tara
Subject: Strange reference to Baker-Just
A Strange reference to Baker-Just
16116 Federal Register / Vol. 79, No. 56 / Monday, March 24, 2014 / Proposed Rules
For this case, appendix K to 10 CFR part
50 ECCS evaluation models would
continue to use the Baker-Just (BJ)
weight gain correlation for estimating
the rate of energy release and hydrogen
generation from the metal/water
reaction.
It is strange that BJ is deployed, because neither Baker-Just nor BJ is used elsewhere in
the entire document.
Furthermore, it is strange that only Cathcart-Pawel is used in Figure 1; there is no
comparison with Baker-Just:
2
Robert H. Leyse bobleyse@aol.com
Hearing Identifier: Secy_RuleMaking_comments_Public
Email Number: 987
Mail Envelope Properties (377CB97DD54F0F4FAAC7E9FD88BCA6D0016513891473)
Subject: FW: Strange reference to Baker-Just
Sent Date: 4/8/2014 4:40:57 PM
Received Date: 4/8/2014 4:40:59 PM
From: RulemakingComments Resource
Created By: RulemakingComments.Resource@nrc.gov
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Saturday, April 12, 2014

ANL-6548 Baker-Just Equation 5/1962 ML050550198


Saturday, April 5, 2014

An update on Ultrasonic Cleaning of Nuclear Reactor Fuel

Here is a reference to a patent that issued during 2013:

United States Patent 8,372,206
Gross ,   et al. February 12, 2013

High power density ultrasonic fuel cleaning with planar transducers

Abstract
Provided are a range of ultrasonic cleaning assemblies that include radiating surfaces activated by corresponding arrays of planar transducers configured to increase the power applied to a reduced volume of fluid associated with a fuel assembly, thereby increasing that applied power density for improved cleaning. The individual ultrasonic cleaning assemblies may be arranged in a variety of modules that, in turn, may be combined to increase the length of the cleaning zone and provide variations in the power density applied to improve the cleaning uniformity.

Inventors: Gross; David J. (Bethesda, MD), Arguelles; David (Herndon, VA)
Applicant:
Name City State Country Type

Gross; David J.
Arguelles; David

Bethesda
Herndon

MD
VA

US
US

Assignee: Dominion Engineering, Inc. (Reston, VA)
Family ID: 41115275
Appl. No.: 12/353,950
Filed: January 14, 2009

And here is a reference to a patent that issued during 2002:
United States Patent 6,396,892
Frattini ,   et al. May 28, 2002

Apparatus and method for ultrasonically cleaning irradiated nuclear fuel assemblies

Abstract
An apparatus for cleaning an irradiated nuclear fuel assembly includes a housing adapted to engage a nuclear fuel assembly. A set of ultrasonic transducers is positioned on the housing to supply radially emanating omnidirectional ultrasonic energy to remove deposits from the nuclear fuel assembly.

Inventors: Frattini; Paul L. (Los Altos, CA), Varrin; Robert Douglas (Reston, VA), Hunt; Edwin Stephen (Arlington, VA)
Assignee: Electric Power Research Institute, Inc. (Palo Alto, CA)
Family ID: 22435141
Appl. No.: 09/545,354
Filed: April 7, 2000

Here is the link to Mark Leyse's PRM-50-84:
http://pbadupws.nrc.gov/docs/ML0708/ML070871368.pdf

Here is link to Summary of Public Comments on PRM-50-84:
http://pbadupws.nrc.gov/docs/ML0831/ML083110761.pdf

Here is a link to a significant inspection report;
http://www.nrc.gov/NRR/OVERSIGHT/ASSESS/REPORTS/rbs_2005008.pdf
Following is from the above inspection report:
Cycle 8 Fuel Failures: During Operating Cycle 8, the licensee experienced seven fuel
pin failures in high power regions of the core. All the fuel pin failures were located in
first burn fuel bundles. During Refueling Outage 8, the licensee found a significant layer
of crud on the fuel surface. Pictures of the crud indicated that it was primarily composed
of loose iron oxide deposits but the team observed some tenacious crud on the cladding
surface as well. The licensee did not perform a chemical analysis of the crud.
 

The crud increased the thermal resistance between the fuel cladding and the coolant
such that cladding surface temperatures were substantially higher than would normally
be expected. Normal cladding surface temperatures are about 560 EF (close to the bulk
coolant temperature). General Electric (the fuel vendor) calculated that the cladding
surface temperatures approached 1200 degrees F in localized areas. The higher temperatures
increased the cladding oxidation rate and, at approximately 1 year into the cycle, the
cladding oxidation layer extended the entire way through the cladding, creating a hole.
 

The team reviewed one technical study that discussed the behavior of crud on the
surface of boiler tubes (“Two-Phase Flow and Heat Transfer,” D. Butterworth and G.F.
Hewitt, Oxford University Press, 1977). The team noted that the thermal resistance of
crud is not normally sufficient to cause cladding temperature increases consistent with
those observed during Cycle 8. In most circumstances, “wick boiling” occurs within the
crud. That is, capillary coolant channels within the crud deliver coolant to the cladding
surface. Steam then escapes from the cladding surface in chimney type plumes. This
is a fairly effective method of heat transfer. However, in some instances the capillary
coolant channels can become clogged, creating a static steam blanket on the cladding
surface. Steam is an exceptionally good thermal insulator. This is the process that
caused the very high cladding surface temperatures and ultimately resulted in fuel
cladding failure.

Following is my blog entry from 2008:

Wednesday, September 3, 2008


Ultrasonic Fuel Cleaning: AREVA, EPRI, Westinghouse, Dominion and more


Go to GOOGLE and enter Ultrasonic Fuel Cleaning. Hunt a bit and you will find a lot: Areva, EPRI, 50.59 game,
Following is the text of Areva's advertisement, minus the photographs.

Ultrasonic Fuel Cleaning

Effective fuel cleaning technology to help assure performance and improve safety. AREVA NP offers patented Electric Power Research Institute (EPRI) Ultrasonic Fuel Cleaning (UFC) to prevent uneven crud deposits that can negatively affect fuel performance. With proven performance in applications at several domestic U.S. utilities, UFC can also reduce dose rates on primary components contaminated by the migration of activation products from core surfaces. Plus, we are an official EPRI licensee authorized to supply UFC equipment and services to nuclear stations worldwide. We can provide UFC for your next outage.

UFC was developed by EPRI to eliminate in-core flux depression by effectively removing deposits from fuel assemblies during refueling outages. Ultrasonic waves cause small particles of crud to release from the fuel assembly. Fuel pool water cools the fuel and transports particles to the filter banks where they are collected for final disposal. The system employs disposable filters to remove radioactive corrosion and activation products. Customers can store the filters in their fuel pool or process them for immediate shipping.

Cleaning Chamber Ensures Even Distribution A special cleaning chamber, similar to a fuel rack, holds ultrasonic transducers positioned on each face of the fuel assembly in an overlapping pattern. This configuration ensures even distribution of the ultrasonic energy into the fuel assembly. Reliable Console Controls the Process An operating console, located on the refuel floor near the edge of the spent fuel pool or reactor vessel, controls the process. The operator can easily observe the cleaning parameters and performance of the filtration unit. Underwater Filters Capture Removed Deposits. The underwater filters contain removed deposits while maintaining radiation to acceptable levels. A variety of filtration system designs are available to provide custom optimization.

BENEFITS
BWR or PWR application
Effectively removes crud
Improves fuel flux distribution
Improves fuel utilization
Reduces radiation source term
Reduces primary system dose rate

And here is the notice of EPRI's R&D award, also on GOOGLE:

EPRI's Patented Nuclear Fuel Cleaning Technology Receives R&D 100 Award; Award Reception Slated for Oct. 20
PALO ALTO, Calif.--(BUSINESS WIRE)--Oct. 5, 2005--The Electric Power Research Institute (EPRI), three member companies, AmerenUE, Exelon Corp., and South Texas Project Nuclear Operating Co., and Dominion Engineering, Inc. (DEI) have earned a prestigious 2005 R&D 100 Award for ultrasonic cleaning of nuclear fuel, a promising new technology that safely removes deposits from irradiated fuel assemblies in nuclear power plants.
The annual awards are given by R&D Magazine for the most outstanding technology developments with commercial potential. The award reception will take place Thursday, Oct. 20 in Chicago; EPRI Senior Vice President and Chief Technology Officer Ted Marston is scheduled to attend.
"The future of the energy industry relies on pursuing innovative technologies that advance efficient, reliable and environmentally sensitive power generation and transmission," said EPRI CEO Steven R. Specker. "I applaud our team and member companies for their contribution towards this end."
The technology awarded delivers a patented process for removing corrosion products deposited on irradiated nuclear fuel pins using a unique form of ultrasonic technology. The technology was first applied at their nuclear power plants by the three EPRI member companies noted above, using equipment supplied by DEI.
"We were pleased to hear that our technology received an R&D Award," said Christopher J. Wood, a technical manager in EPRI's Nuclear Sector. "This breakthrough technology allows the full potential of current nuclear fuel designs to be achieved while maintaining excellent fuel reliability. Availability of a safe, reliable cleaning technology will also now allow utilities to further optimize fuel performance, core design, and reduce radiation fields and electricity generating costs."
This unique technology, developed in EPRI's Fuel Reliability Program, solves a significant emerging problem by removing deposits from nuclear fuel assemblies in nuclear power plants. Enhancing the performance of nuclear fuel is crucial to continue the improvement in electricity production from nuclear units. Over the past decade, nuclear power production has increased by over 20 percent, but this has placed additional demands on the fuel, as fuel temperatures have increased.
Some of the potential problems with fuel reliability result from the buildup of deposits on the surfaces of the fuel elements, which produces an insulating layer that could result in corrosion of the fuel cladding material at increased fuel pin temperatures. Until EPRI's developed technology, there was no effective way of removing these deposits during the working life of the fuel. Including early development demonstrations, this ultrasonic fuel cleaning technology has been used successfully eight times at nuclear power plants in the USA through 2004, and has been licensed worldwide. Seven additional commercial applications have taken place in 2005, including one in Spain. The technology used cleans all the fuel elements in every fuel assembly without any adverse effects. The cleaning process does not extend the schedule of routine refueling outages and is very cost-effective in pressurized water reactors. It is expected to result in a major reduction in radiation fields in boiling water reactors.
About the Electric Power Research Institute
The Electric Power Research Institute (EPRI), with major locations in Palo Alto, California, and Charlotte, North Carolina, was established in 1973 as an independent, nonprofit center for public interest energy and environmental research. EPRI brings together member organizations, the Institute's scientists and engineers, and other leading experts to work collaboratively on solutions to the challenges of electric power. These solutions span nearly every area of power generation, delivery, and use, including health, safety, and environment. EPRI's members represent over 90% of the electricity generated in the United States. International participation represents nearly 15% of EPRI's total R&D program.

And here is how NRC accepted Ultrasonic Fuel Cleaning under 50.59!


And here we have EPRI, way back in 1999, highlighting its Ultrasonic Fuel Cleaning Process at Callaway as a 1999 payoff:


And, during September 2003, Westinghouse advertised its ultrasonic fuel cleaning service. "As a result, the plant safety review committee granted the application 10 CFR 50.59 approval."
Ultrasonic cleaning means fast, safe removal of fuel-assembly crud buildup
Crud — corrosion products that accumulate on fuel surfaces — can break loose and spread to other parts of the system, causing radioactive buildup. Over time, crud that builds up on fuel surfaces becomes activated by neutrons to form radioactive nuclides, making crud cleanup a high priority.
Ultrasonic fuel cleaning can break up crud deposits during normal refueling, trapping particulates in filters for storage in the fuel pool. Designed by Dominion Engineering, Inc. (DEI), and patented by EPRI, the technique blasts crud with ultrasonic transducers.
Ultrasonic cleaning reduces the risk of fuel damage and takes a fraction of the time required by other methods. Controlling crud and other particulate inventory reduces out-of-core radiation fields and lowers radiation dosage levels.
Eliminating crud also mitigates local in-core flux supression and decreases the likelihood of axial offset anomaly (AOA) caused by lithium and boron concentrations. Ultrasonic cleaning also helps prevent crud-induced power shifts that can reduce output by as much as 20 percent.
Ultrasonic fuel cleaning was first used and verified at the Callaway plant in Missouri in 2001. After a year, no evidence of core-wide AOA was found, and early ex-core dosage was reduced significantly with no impact on critical path time. Measurements of assemblies before and after cleaning, and of particulate discharge at the filters, showed that ultrasound cuts crud deposits by about 80 percent. As a result, the plant safety review committee granted the application 10 CFR 50.59 approval.Ultrasonic cleaning is fast, too. During routine refueling, an assembly scheduled for reuse can be cleaned in as little as seven to ten minutes. Westinghouse is the first vendor to use this technique commercially. Our partnership with DEI gives utilities access to ultrasonic cleaning with minimal incremental costs.
Dominion (DEI), the inventors of Ultrasonic Fuel Cleaning, may have discussed this at a very recent meeting of PWR operators.
Sunday, July 20, 2008
PWR ALARA Association Board Meeting - Board Room
Wednesday, July 23, 2008
General PWR Session – Day 2
10:00 Ultrasonic Fuel Cleaning Process/Success – Dr. Robert Verrin (Dominion Engineering) Tentative

Thursday, April 3, 2014

Mark Leyse and PRM-50-84 at ACRS, December 15, 2011

Here is the reference:

http://pbadupws.nrc.gov/docs/ML1202/ML120200495.pdf

Mark Leyse is referred to as Mark Lacey

The Members wanted to see Statements of Consideration if possible.
There was a petition from Mr. Lacey concerning the effects of crud. 132-133, 161


Official Transcript of Proceedings
NUCLEAR REGULATORY COMMISSION
Title: Advisory Committee on Reactor Safeguards
Materials, Metallurgy and Reactor Fuels
Docket Number: (n/a)
Location: Rockville, Maryland
Date: Thursday, December 15, 2011

MEMBER SHACK: There was a statement that
19 the new rule was going to address Mr. Lacey's
20 petition, and I haven't seen anything that really does
21 that. Is it something in the Statement of
22 Considerations?
23 MR. CLIFFORD: There was an analytical
24 requirement added to the rule itself that said the
25 effects of crud have to be accounted for.

MEMBER SHACK: I missed it. I missed it.
2 MR. CLIFFORD: So, any new LOCA model we
3 review they would have to say how are they accounting
4 for crud.
5 MEMBER SHACK: I missed it.
6 CHAIR ARMIJO: Paul, don't most of them
7 already do that?
8 MR. CLIFFORD: A lot of them. I can't say
9 -- there are a lot of LOCA models dating back decades
10 some of them. A lot of them do.
11 CHAIR ARMIJO: Yes.
12 MR. CLIFFORD: I can't say that they all
13 do.
14 CHAIR ARMIJO: Okay.
15 MR. CLIFFORD: But they may not
16 specifically account for it, but the way you measure
17 oxidation layers, sometimes you get the tenacious crud
18 that's mixed in with the oxide when you do your eddy
19 current testing.
20 CHAIR ARMIJO: Yes.
21 MR. CLIFFORD: You get a combination of the
22 two, so when you adjust your oxidation model you're
23 kind of getting the inherent -- some inherent effects
24 of tenacious crud.
25 CHAIR ARMIJO: Yes. Yes. Is that it?

 Slide by  Tara Iverso
Rulemaking Purpose
• Revise ECCS acceptance criteria to
reflect recent research findings
• SECY-02-0057
– Replace prescriptive analytical
requirements with performance-based
requirements
– Expand applicability to all fuel designs
and cladding materials
• Address concerns raised in two
PRMs: PRM-50-71 and PRM-50-84


Here is the link to PRM-50-84:

http://pbadupws.nrc.gov/docs/ML0708/ML070871368.pdf

Here is the biography of ACRS Member William Shack at the time of his appointment to ACRS on October 19,1993:

No. 93-154 FOR IMMEDIATE RELEASE
Tel. 301/504-2240 (Tuesday, October 19, 1993)
 

NRC NAMES DR. WILLIAM J. SHACK
TO ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
 

The Nuclear Regulatory Commission has appointed Dr. William
J. Shack, Associate Director of the Energy Technology Division,
Argonne National Laboratory in Chicago, Illinois, to the
independent Advisory Committee on Reactor Safeguards (ACRS).
The ACRS was established in 1957 by the Congress to advise
the former Atomic Energy Commission, and subsequently the Nuclear
Regulatory Commission, with regard to the safety aspects of
proposed and existing nuclear facilities and the adequacy of
related safety standards. The ACRS also performs such other
special assignments as the Commission may request, including the
review of related research programs.
 

Dr. Shack was born in 1943, in Pittsburgh, Pennsylvania. He
received his B.S. degree in civil engineering from the
Massachusetts Institute of Technology (MIT) in 1964, and his M.S.
and Ph.D. degrees in applied mechanics from the University of
California-Berkeley in 1965 and 1968, respectively.
In 1968, Dr. Shack joined the Mechanical Engineering
Department at MIT as an assistant professor. He taught there
until 1975.
 

He joined Argonne National Laboratory in 1975. His work has
included measurement and modeling of residual stresses, fracture
mechanics analyses of stress corrosion crack growth, assessment
of leak-before-break behavior in piping systems, and fatigue of
reactor materials.
 

Dr. Shack is the author or coauthor of more than 75
publications on a variety of topics in applied mechanics and
materials behavior. He has served on the NRC Piping Review
Committee and various ad hoc NRC committees to assess the impact
of environmentally enhanced material degradation on reactor
safety and operation. He has also been involved in research on
corrosion and stress corrosion cracking of candidate materials
for the Yucca Mountain, Nevada, high-level waste repository
project.

Friday, March 28, 2014

Regulatory Analysis For reference 10 CFR 50.46c


No: 14-020 March 24, 2014
CONTACT: Scott Burnell, 301-415-8200
NRC Seeks Comment on Proposed Revision to Acceptance Criteria
For Emergency Cooling Systems at U.S. Reactors

The Nuclear Regulatory Commission is seeking comment on a proposed change to agency regulations regarding the acceptance criteria for emergency systems to cool the reactor core if an accident occurred at a U.S. nuclear power plant.

The proposed rule reflects recent research findings that identified new damage mechanisms for zirconium alloy-covered fuel rods during a loss-of-coolant accident. The proposed rule would also apply to all fuel types and cladding materials, as well as address a petition for rulemaking regarding crud, oxide deposits and hydrogen content in zirconium-alloy fuel cladding. The proposed rule would ensure an acceptable level of fuel rod performance following a loss-of-coolant accident, providing adequate protection of public health and safety. The proposed rule would also provide licensees the option to use risk-informed methods to address the effects of debris during long-term cooling following a loss-of-coolant accident.

The proposed rule is not part of the NRC’s response to the 2011 events at Fukushima, but an outcome of a Nuclear Energy Institute petition for rulemaking in 2000, direction given to the staff by -+_the Commission in 2003, and findings of a 10-year research program ending in 2008. Thus, the development of the proposed rule pre-dates the Japan events by several years.

For more information on the proposed rule, contact NRC staff members Tara Inverso by phone at 301-415-1024 or via e-mail tara.inverso@nrc.gov; or Paul M. Clifford by phone 301-415-4043, via e-mail paul.clifford@nrc.gov
.
Comments on the changes will be accepted until June 9, following publication of the proposed rule in the Federal Register . The Federal Register notice also opens a comment period for three related draft regulatory guides. The notice includes instructions and the relevant Docket IDs for submitting comments on both the rule and on the guides. Comments may be submitted on the Regulations.gov website; by e-mail to Rulemaking.Comments@nrc.gov ; hand delivered to: 11555 Rockville Pike, Rockville, Md., between 7:30 a.m. and 4:15 p.m. (EST) federal workdays; telephone: 301-415-1677; mailed to
Secretary
U.S. Nuclear Regulatory Commission
Washington, DC 20555-0001
ATTN: Rulemakings and Adjudications Staff



http://adamswebsearch2.nrc.gov/webSearch2/doccontent.jsp?doc={56DE33D1-6742-42F2-8AFD-AF53BDA12DA6}




Regulatory Analysis for Proposed Rulemaking 10 CFR 50.46c:
“Emergency Core Cooling System Performance during Loss-of-Coolant Accidents”
March 24, 2014  


From page 20

The proposed rule would require licensees to evaluate the thermal effects of crud and
oxide layers that accumulate on the fuel cladding during plant operation. Because licensees are
required to account for various thermal parameters under the current regulation, the NRC’s
position is that the proposed requirement to evaluate crud is a clarification of the current
requirement. As such, there would be no additional cost incurred as a result of the rule.

Blogger (Leyse) comment on the above paragraph.  It certainly is not obvious that there would be no additional cost.  If the rule includes: During or immediately
after plant operation, if actual crud
layers on reactor fuel are implicitly
determined or visually observed after
shutdown to be greater than the levels
predicted by or assumed in the ECCS
evaluation model, licensees would be
required to determine the effects of the
increased crud on the calculated results, 
that could be very expensive.


http://www.gpo.gov/fdsys/pkg/FR-2014-03-24/pdf/2014-05562.pdf

10 CFR Parts 50 and 52
[NRC–2008–0332, NRC–2012–0041, NRC–
2012–0042, NRC–2012–0043]
RIN 3150–AH42
Performance-Based Emergency Core
Cooling Systems Cladding Acceptance
Criteria




Following is from page 16140
(iv) To the extent practicable,
predictions of the ECCS evaluation
model, or portions thereof, must be
compared with applicable experimental
information.




 from page 16122
Technical Issues in PRM–50–84
Licensees use approved fuel
performance models to determine fuel
conditions at the start of a LOCA, and
the impact of crud and oxidation on fuel
temperatures and pressures may be
determined explicitly or implicitly by
the system of models used. With the
addition of an unambiguous regulatory
requirement to address the
accumulation of crud and oxide during
plant operation, the NRC believes that
fuel performance and LOCA evaluation
models must include the thermal effects
of both crud and oxidation whenever
their accumulation would affect the
calculated results. The NRC notes that
licensees are required to operate their
facilities within the boundary
conditions of the calculated ECCS
performance. During or immediately
after plant operation, if actual crud
layers on reactor fuel are implicitly
determined or visually observed after
shutdown to be greater than the levels
predicted by or assumed in the ECCS
evaluation model, licensees would be
required to determine the effects of the
increased crud on the calculated results.

In many cases, engineering judgment or
simple calculations could be used to
evaluate the effects of increased crud
levels; therefore, detailed LOCA
reanalysis may not be required. In other
cases, engineering judgment is used to
determine that new analyses would be
performed to determine the effect the
new crud conditions have on the final
calculated results. If unanticipated or
unanalyzed levels of crud are
discovered, then the licensee must
determine if correct consideration of
crud levels would result in a reportable
condition as provided in the relevant
reporting paragraphs. Should this
proposed rule be adopted in final form,
the NRC believes this regulatory
approach to address crud and oxide
accumulation during plant operation
would satisfactorily address the issues
raised by the petitioner’s first request.
The formation of cladding crud and
oxide layers is an expected condition at
nuclear power plants. Although the
thickness of these layers is usually
limited, the amount of accumulated
crud and oxidation varies from plant to
plant and from one fuel cycle to
another. Intended or inadvertent
changes to plant operational practices
may result in unanticipated levels of 

crud deposition. The NRC agrees with
the petitioner (the petitioner’s second
request) that crud and/or oxide layers
may directly increase the stored energy
in reactor fuel by increasing the thermal
resistance of cladding-to-coolant heat
transfer, and may also indirectly
increase the stored energy through an
increase in the fuel rod internal
pressure. As such, to ensure that
licensee ECCS models properly account
for the thermal effects of crud and/or
oxide layers that have accumulated
during operations at power, the
proposed rule would add a requirement
to evaluate the thermal effects of crud
and oxide layers that may have
accumulated on the fuel cladding
during plant operation.
If the NRC
adopts the proposed rule in final form,
then the second request of PRM–50–84
would be resolved.
The petitioner’s third request is for
the NRC to establish a maximum
allowable percentage of hydrogen
content in fuel rod cladding. The
purpose of this request is to prevent
embrittlement of fuel cladding during a
LOCA. Although the NRC has decided
not to propose the specific rule language
recommended by the petitioner, the
proposed new zirconium-specific
requirements, if adopted in final form,
would address the petitioner’s third
request by considering cladding
hydrogen content in the development of
analytical limits on integral time at
temperature.
The NRC believes that this proposed
rule addresses each of the three issues
raised in PRM–50–84. If the NRC adopts
the proposed rule in final form, PRM–
50–84 would be granted in part and
resolved.
 



http://www.inel.gov/relap5/rius/yellowstone/leyse.pdf 

Unmet Challenges for SCDAP/RELAP5-3D
Analysis of Severe Accidents
for Light Water Nuclear Reactors
with Heavily Fouled Cores

Wednesday, March 26, 2014

Things Take Time: Leyse Gamma Thermometer in ESBWR

Regarding Things Take Time:

Leyse Gamma Thermometer in ESBWR


NEDO-33197-A Revision 3
Page 157
11. REFERENCES
1. R.H. Leyse, R.D. Smith: “Gamma Thermometer Developments for Light Water
Reactors,” IEEE Transactions on Nuclear Science, Vol.N5.26, No. 1, February 1979, pp.
934–943.

NEDO-33197-A Revision 3
Page 117
7.3.3 Conclusions
The comparison with the gamma scan established that core monitoring based on GTs is nearly equivalent in accuracy to core monitoring with neutron TIPs. In addition, it was shown that the thermal limits, MCPR and MLHGR, evaluated by the two core monitoring systems were very similar throughout the cycle.
The overall conclusion was that the GT system is “practical as a substitute” for the TIP system.



United States Patent 4,393,025
Leyse July 12, 1983

Method of and apparatus for measuring the power distribution in nuclear reactor cores

Abstract
The invention disclosed is the method of exact calibration of gamma ray detectors called gamma thermometers prior to acceptance for installation into a nuclear reactor core. This exact calibration increases the accuracy of determining the power distribution in the nuclear reactor core. The calibration by electric resistance heating of the gamma thermometer consists of applying an electric current along the controlled heat path of the gamma thermometer and then measuring the temperature difference along this controlled heat path as a function of the amount of power generated by the electric resistance heating. Then, after the gamma thermometer is installed into the nuclear reactor core and the reactor core is operating at power producing conditions, the gamma ray heating of the detector produces a temperature difference along the controlled heat path. With the knowledge of this temperature difference, the calibration characteristic determined by the prior electric resistance heating is employed to accurately determine the local rate of gamma ray heating. The accurate measurement of the gamma heating rate at each location of a set of locations throughout the nuclear reactor core is the basis for accurately determining the power distribution within the nuclear reactor core.

Inventors: Leyse; Robert H. (Rockville, MD)
Family ID: 26901622
Appl. No.: 06/206,741
Filed: November 14, 1980

Here is some text:

The method of calibration which has been described may be performed after the gamma thermometer assembly has been fabricated, but prior to installation into the nuclear reactor core. The calibration may also be determined after the apparatus is installed into the nuclear reactor core, for example, prior to first power operation of the nuclear reactor core or during shutdown of the nuclear reactor core after a period of extended operation. In addition, the calibration may be checked while the nuclear core is at power operation.

In this latter case, one procedure would be the following:

a. With no electrical power input, measure the temperature difference that results from gamma heating with the core at power and then utilizing the calibration curve of FIG. 4, determine the corresponding value of the power per unit length of thermocouple tube.

b. Next, apply an increment of electric power to the thermocouple tube. Add this value of electric power to the gamma heating power determined in step a. Measure the temperature difference of the gamma thermometer detector. This temperature difference and the total of the gamma heating power and the electrical heating power may then be plotted on the original calibration curve as a check on the retention of the original calibration.

c. Step b may be performed for several increments of electric power heating.


Here are some CLAIMS:

17. The method of monitoring elongated fuel elements, which emit gamma rays, of a nuclear reactor core, comprising:

(a) providing a flow path for the flow of a cooling fluid to be used for calibration purposes, and passing said cooling fluid along said flow path for calibration purposes,

(b) providing an elongated instrument element including electrical conducting material having first and second zones,

(c) locating said instrument in said flow path and exposing it to said fluid so that the temperature of the second zone depends on said temperature and rate of flow of said cooling fluid more than the temperature of the first zone depends on the temperature and rate of flow of said cooling fluid,

(d) passing an electrical current, for calibration purposes, through said electrical conducting material to supply heat to both of said zones with the first zone rising in temperature more than the second zone due to cooling effect of said cooling fluid on said second zone,

(e) measuring the temperature difference between said first and second zones to calibrate the instrument,

(f) placing the instrument parallel to and adjacent said elongated fuel elements,

(g) passing a cooling fluid past the instrument while it is adjacent said elongated fuel elements,

(h) the step of passing a cooling fluid past the instrument for calibration purposes as aforesaid involving fluid cooling conditions substantially identical to those characterizing the cooling fluid that is passed by the instrument while it is adjacent to the elongated fuel elements, and

(i) measuring the temperature difference between said two zones while the instrument is adjacent the elongated fuel elements with cooling fluid flowing past the same and without said electrical current flowing, whereby in view of the previous calibration of the instrument with said flow of current the output of the elongated fuel elements may be determined.

18. The method of claim 17 in which during step (i), the first zone rises in temperature above the second zone by an amount related to the output of the elongated fuel elements, and in which water is selected as the cooling fluid.

19. The method of claim 18 in which the cooling fluid is in such good thermal contact with the second zone that the second zone remains at a temperature substantially the same as that of the cooling fluid with the first zone rising to a higher temperature both during calibration as well as during operation adjacent the elongated fuel elements.

20. The method of monitoring elongated fuel elements as defined in claim 17 in which steps (a) to (e) inclusive are performed with said instrument positioned in a remote location with reference to said elongated fuel elements so that those elements do not supply substantial gamma rays to the instrument and so that the instrument is calibrated while the only heat supplied to the instrument during calibration results from said electrical current, and performing steps (f), (g) and (i) after the instrument has been calibrated in said remote location.

21. The method of monitoring elongated fuel elements as defined in claim 20 in which the instrument is calibrated as set forth in said steps (a) to (e) using a first flow path for the cooling fluid, and the elongated fuel elements are monitored as set forth in steps (f), (g) and (i) using a second flow path for the cooling fluid which second path is adjacent said elongated fuel elements and is remote from the first flow path.

22. The method of monitoring elongated fuel elements as defined in claim 17 in which step (f) is performed before the instrument is calibrated, and in which:

the nuclear reactor core is shut down before the instrument is calibrated and in which the instrument is calibrated as called for by said steps (a) to (e) while the instrument is adjacent the elongated fuel elements and the nuclear reactor core is shut down.

23. The method of monitoring elongated fuel elements as recited in claim 22 in which the same flow path for the flow of the cooling fluid is used during said calibration steps (a) to (e) inclusive as is used for the monitoring steps (g) and (i).

24. The method of monitoring elongated fuel elements as recited in claim 17 in which the calibration steps (a) to (e) inclusive are performed while said instrument is adjacent said elongated fuel elements and while the nuclear reactor core is in operation,

said calibration and monitoring steps comprising comparing the temperature differences between said zones under two conditions one of which conditions occurs while said electrical current is off and the other of which conditions occurs while said electrical current is on.

25. The method of monitoring elongated fuel elements as recited in claim 24 in which the calibration steps (a) to (e) inclusive are performed using several increments of electric power heating.

26. The method of monitoring elongated fuel elements as defined in claim 17 in which said measuring step (e) includes measuring the temperature difference between the "hot" and "cold" junction of a thermocouple, comprising:

spacing said "hot" junction from all nearby liquid and solid matter while exposing said "hot" junction to said gamma rays.

27. The method of monitoring elongated fuel elements as defined in claim 26, comprising:

positioning said "cold" junction in a bed of solid material and exposing said solid material to said cooling fluid,

whereby said "hot" junction is heated to a temperature above said cold junction by reason of the direct impingement of said gamma rays on said "hot" junction with said cooling fluid having only a secondary effect on the temperature of said "hot" junction.

28. In apparatus for monitoring fuel elements, a nuclear reactor core:

a measuring instrument comprising a thermocouple having a "hot" junction and a "cold" junction,

said measuring instrument having a body, said body having an outer wall,

said measuring instrument including means for mounting said "hot" junction inside said body and spaced from any and all liquid and solid material,

said measuring instrument including solid material surrounding said "cold" junction and providing a heat conduction path from said cold junction to said outer wall,

means for passing a cooling fluid along the outer wall of said body, and

means for positioning said body in the path of said gamma rays to thus directly heat said hot junction, whereby the heat from said gamma rays elevates the temperature of the "hot" junction above that of the "cold" junction due to the better thermal contact between the "cold" junction and the cooling fluid than between the "hot" junction and the cooling fluid.  


Here is another link to ESBWR Design Summary:
http://pbadupws.nrc.gov/docs/ML0217/ML021770054.pdf
See page 56 for a Cross Section of Gamma Thermometer
 
 


 

Sunday, March 23, 2014

Spent Fuel Pool LOCA cover-up via NRC International Programs

During the past several years I have been unsuccessful in tracking NRC managed work at Sandia in the arena of heat and mass transfer in spent fuel pools under accident conditions.  Two reports, released to the public during 2013, reveal that the work has been going on for over 10 years.  Those reports also reveal that the work is unsatisfactory.

I emailed the following to the ACRS on January 3, 3014:



For ACRS from Robert H. Leyse

NRC Safety Research Program item NUREG/CR-7143 (ML13072A056)

Characterization of Thermal-Hydraulic and Ignition Phenomena in Prototypic, Full-Length Boiling Water Reactor Spent Fuel Pool Assemblies After a Postulated Complete Loss-of-Coolant Accident, published March 2013.

The ACRS should declare that this is unsatisfactory work. 

On page 5 of 247 of ML13072A056 :
The close coupling of the experimental and numerical programs allowed for rapid validation and improvement of the MELCOR whole pool calculations. Because of the success of this approach, this project will be used as a model for subsequent studies.  

The work does not approach being prototypic for at least the following reasons:

Diameter of the heater rods is 0.375 inches.  The heaters were made, “in a process whereby the 0.440-in. tubing was drawn through a die that reduced the diameter to 0.375 in. and compressed the magnesium oxide powder considerably.”  The 0.440 inch tubing is BWR fuel tubing.  However, the heater rods, with compressed (swaged) magnesium oxide and a diameter of 0.375 inches, are not anywhere close to being prototypic of the BWR case.

BWR fuel rods will swell (balloon) and burst from internal gas pressure during a Spent Fuel Pool Complete Loss-of-Coolant Accident (SFPLOCA). 

BWR fuel rods will be in intimate contact with inconel grids and will fuse with inconel grids at the elevated temperatures of the SFPLOCA.  Furthermore the intimate contact will be augmented by ballooning. 

The swelled cladding of BWR fuel rods has a significant impact on the ignition phenomena of a SFPLOCA.  (The ratio of surface area to heat capacity of separated cladding is far greater than that of swaged heater rods.)

Ballooned cladding of BWR fuel rods is not prototyped in extensive thermal-hydraulic testing with Incoloy heaters.  It is reported:  The diameter of the Incoloy heaters was slightly smaller than prototypic pins, 1.09x10-2versus1.12x10-2m.  That is 0.429 versus 0.440 inches.  However, that difference is trivial compared with the impact of ballooning.

The work dates back to 2004-2005, however, it is held under-cover until March 2013.  Other work remains under wraps.  Apparently PWR work will be reported sometime and this likely means that more trash will be exposed in defense of MELCOR and more.

Apparently, ACRS did not review this work in progress during 2004 and 2005. 
 

Moving on, I have continued my pursuit of the facts. 


Here is a revealing paragraph from the second of the following exchange of emails:


Further information on the PWR spent fuel pool experiments is not publically available at this time.  The work was conducted under an agreement with the Organisation for Economic Co-operation and Development (OECD).  Reports documenting this work will be released to the public in accordance with the terms of the international agreement, which is typically three years after completion of the program.  







Subject: Re: Responses to your Emails
Date: 3/21/2014 10:21:07 A.M. Mountain Daylight Time
From: Bobleyse@aol.com
To: Robert.Beaton@nrc.gov
 
Robert:
 
Thank you, I have seen the NUREGs and I've recently viewed the poster.
 
I would really like to know how NRC got into this way of doing business, but that is likely beyond your scope.  Clearly, the American public is essentially uninformed, and ACRS does not evaluate the worth of the activity until all of the money is spent, and even then there is a chance that they will never look at it.  For example, it was not until February 5, 2014, that RES suggested that the following are candidates, although not required matters, for ACRS review:
 
 The candidate projects are listed here for your consideration:
• NUREG/CR-7143: Characterization of Thermal-Hydraulic and Ignition Phenomena in
Prototypic, Full-Length Boiling Water Reactor Spent Fuel Pool Assemblies After a
Postulated Complete Loss-of-Coolant Accident, March 2013 (ML13072A056)
• NUREG/IA-0216, Vol. 3: International HRA Empirical Study – Phase 3 Report, January
2013 (ML12349A075)
• NUREG/CR-7149: Effects of Degradation on the Severe Accident Consequences for a
PWR Plant with a Reinforced Concrete Containment Vessel, June 2013 (ML13172A089)
• NUREG/CR-7144: Laminar Hydraulic Analysis of a Commercial Pressurized Water
Reactor Fuel Assembly, January 2013 (ML13028A415)
• NUREG/CR-7148: Confirmatory Battery Testing: The Use of Float Current Monitoring to
Determine Battery State-of-Charge, November 2012 (ML12313A413)
• NUREG/CR-7171: A Review of the Effects of Radiation on Microstructures and
Properties of Concrete Used in Nuclear Power Plants, November 2013 (ML13325B077) 
 
The above list is copied from
 
On March 1, 2012, Borchardt (then EDO) told my Senator Risch about the openness of the NRC:
 
Robert H. Leyse     bobleyse@aol.com
 
In a message dated 3/21/2014 5:48:45 A.M. Mountain Daylight Time, Robert.Beaton@nrc.gov writes:
Mr. Leyse,

The following are responses to your emails to NRC as follows:

From Email to Robert Beaton on March 3, 2014 with subject “PWR Spent Fuel Pool LOCA at RIC 2014”

The publically available information on the spent fuel pool experiments conducted at Sandia National Laboratories is in the following NUREGs.

NUREG/CR-7143, "Characterization of Thermal-Hydraulic and Ignition Phenomena in Prototypic, Full-Length Boiling Water Reactor Spent Fuel Pool Assemblies After a Postulated Complete Loss-of-Coolant Accident", (ADAMS Accession Number ML13072A056)

NUREG/CR-7144, “Laminar Hydraulic Analysis of a Commercial Pressurized Water Reactor Fuel Assembly”, (ADAMS Accession Number ML13028A415)

Further information on the PWR spent fuel pool experiments is not publically available at this time.  The work was conducted under an agreement with the Organisation for Economic Co-operation and Development (OECD).  Reports documenting this work will be released to the public in accordance with the terms of the international agreement, which is typically three years after completion of the program.

From Email to Robert Beaton on March 11, 2014 with subject “Please”

All the posters and presentations from the RIC are available on the USNRC website.  From the main RIC website (http://www.nrc.gov/public-involve/conference-symposia/ric/), near the bottom, click on “Technical Poster and Tabletop Presentations”, then click on “Investigation of a Pressurized Water Reactor Spent Fuel Assembly under Complete Loss of Coolant Accident Conditions,” then finally on “View Presentation.”  Alternatively, you can go directly to an electronic copy of the specific poster you requested at: https://ric.nrc-gateway.gov/docs/posters/56_res-investigation-of-a-pressurized-water-reactor-spent-fuel-assembly-under.pdf

Regards,
Robert Beaton

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