FOIA Appeal NRC-2020-000097 Submitted

Subject: FOIA Appeal NRC-2020-000097 Submitted
Date: 12/26/2019 6:14:58 AM Mountain Standard Time
From: admin@foiaonline.gov
To: bobleyse@aol.com
Sent from the Internet (Details)

This message is to notify you of a new appeal submission to the FOIAonline application. Appeal information is as follows:

• Appeal Tracking Number: NRC-2020-000097
• Request Tracking Number: NRC-2020-000007
• Requester Name: Robert Leyse
• Date Submitted: 12/26/2019
• Appeal Status: Submitted
• Description: Appeal the administrative closure due to failure to pay fees associated with NRC-2020-000007

Subject: FOIA Appeal NRC-2020-000097 Modified
Date: 12/26/2019 7:21:12 AM Mountain Standard Time
From: admin@foiaonline.gov
To: bobleyse@aol.com

The FOIA appeal - NRC-2020-000097 description has been modified. Additional details for this item are as follows:

• Tracking Number: NRC-2020-000097
• Requester: Robert Leyse
• Submitted Date: 12/26/2019
• Description: Appeal the administrative closure due to failure to pay fees associated with NRC-2020-000007

Subject: NRC-2020-000097 Admin Close
Date: 12/26/2019 7:34:53 AM Mountain Standard Time
From: TINA.ENNIS@nrc.gov
To: bobleyse@aol.com

Dear Mr. Leyse,

Please find attached a copy of NRC’s final response to your FOIA appeal, NRC-2020-000097.

Thank you,

Tina Ennis

FOIA Analyst (QualX Contractor)

OCIO/GEMS/ISB

U.S. Nuclear Regulatory Commissi

UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON, D.C. 20555-0001
December 26, 2019
NRC-2020-000097
(NRC-2020-000007)
Mr. Robert Leyse
P.O. Box 2850
Sun Valley, ID 83353
Dear Mr. Leyse:
This in response to your letter dated December 18, 2019, in which you sought to appeal our November 5, 2019, administrative closure of your request for various AEOD and/or NSAC documents that were referenced in a December 13, 1984 letter.1
Our October 21, 2019 email notified you of an advanced payment of $1,089.45, as the NRC's FOIA regulations direct us to obtain an advance payment from a requester whenever the estimated fees exceed $250.00, before we continue with the processing of such a request. We also informed you that if we did not receive payment by November 4, 2019, we would administratively close our file on your request. You responded by justifying that you “should not be classified as ‘non-excepted because this is an investigation of corrupt practices between NRC and NSAC.” After sending you a copy of the Office of Management and Budget (OMB)’s government-wide Guidelines for FOIA Fees, and acknowledging your stated reason for submitting your request, that reason does not fall within any of the favored fee categories (i.e., representative of the news media, an educational institution, or non-commercial scientific institution) under the FOIA and the OMB guidelines.
Since we did not receive the payment of estimated fees associated with processing your request by the given deadline, your request was not perfected under the statute. The agency’s administrative closing of our file on your request is not an adverse determination from which you may appeal. See 10 CFR 9.29.
If you choose to re-submit your request, please keep in mind that the fee estimate associated with processing your request, as originally described, is expected to remain the same and you will be required to pay the estimated fee before your request will be processed. However, you may choose to narrow the scope of your request to reduce the cost.
Sincerely,
Stephanie Blaney /s/
Stephanie A. Blaney
FOIA Officer
Office of the Chief Information Officer
1 In your original request, you identified four documents, as well as all records reflecting the technical relationship between NSAC and AEOD. You then supplemented/clarified that you were interested in obtaining copies of any AEOD documents that reflect upon its relationship with NSAC (even if not a document between the two entities).

NRC-2020-000097 closed

Subject: Final Disposition, Request NRC-2020-000097
Date: 12/26/2019 7:38:29 AM Mountain Standard Time
From: DoNotReply-NRCfoia@regulations.gov
To: bobleyse@aol.com

NRC-2020-000097 has been processed with the following final disposition: Closed for Other Reasons.

Leyse FOIA appeal UPS delivery to NRC

UPS package #1Z2Y292X1543895128

www.ups.com

1. Print
2. Feedback

Delivered: Mon, Dec 23, 10:37 AM

This is an FOIA request,  based on attached letter, Lang to Layman
10/5/2019 5:29:09 PM Mountain Standard Time
bobleyse@aol.com
foia@nrc.gov

This FOIA consists of a request for documents that are discussed in the attached letter, Layman December 13, 1984.  This letter documents a telephone conversation between Wayne Lanning of the NRC  and Gary Vine of EPRI's Nuclear Safety Analysis Center.  This telephone conversation was initiated by the NRC.

In paragraph 1. there is a reference to an AEOD memorandum dated February 28, 1984.  Under this  FOIA I am demanding a copy of that AEOD memorandum that is dated February 28,1984.

In paragraph 3, there is a reference to an AEOD evaluation.  I  want copy of that AEOD evaluation.

In paragraph 4.a. there is reference to a memorandum dated October 3, 1984, "UHI-Ultra High Risk" from which the author had been obscured.  I want a copy of that memorandum that has the obscured features.

In paragraph 4.c. there is reference to a joint AEOD/NRR response to Keppler.  I want a copy  of that response.

In paragraph 5. there is discussion of the valued technical relationship between NSAC and AEOD.  I want a copy of all of the records of that technical relationship.

Robert H. Leyse

December 6, 2019

Robert H. Leyse

P. O. Box 2850

Sun Valley, ID 83353

FOIA Officer

U. S. Nuclear Regulatory Commission

Mail Stop T-2 F-43

Washington, D. C. 20555-0001

FOIA Appeal

Enclosed are stapled or single page documents in the following order:

1.  FOIA REQUEST
2.  Latest FOIA – (including acknowledgement letter)
3. FOIA Request  10/13/2019
4. Fee Estimate
5.  November 5, 2019, Administrative Closeout
6.  December 2, 2019, response from OGIS

Document 1 is the straightforward FOIA.

Document 2 is the associated acknowledgement letter

Document 3 clarifies (expands) Document 1

Document 4 states Leyse should not be classified as “Non-Excepted”

Document 5 is the administrative closing of Document 1

Document 6 is a response to Leyse from OGIS.  On page 2 OGIS has five bullets with analytical factors.  Leyse insists that the disclosure of the illegal and undisclosed to Leyse or the public of  relationships between NSAC and the NRC, will identify the existence of identifiable and malicious operations and activities of the NRC, will contribute to public awareness of NRC operations that are not otherwise disclosed to Leyse or the public, will not further any commercial interests of Leyse, and the public interest in these disclosures will exceed Leyse’s interest when the disclosures are publicized.

Robert H. Leyse

Simpson and upper head injection system (UHI)

Facts for EPRI and NRC

Simpson and upper head injection system (UHI)

From page 199 of Simpson’s book, NUCLEAR POWER FROM UNDERSEAS TO OUTER SPACE 

“Where this couldn’t be done or wasn’t enough, we proposed an upper head injection system.”  Simpson bragged about “... many competent engineers who were becoming available, including Retallick from our NERVA program.”  Actually, Retallick was not that competent, but upper head injection was subsequently well defended by John Taylor who had moved from Westinghouse to a senior position at the Electric Power Research Institute.



UHI and dissolved nitrogen - the Chexal Technical Review Panel and several related documents. 



This communication provides additional documents that are relevant to the Chexal Technical Review Panel that is the focal point of my prior letter dated August 9, 2016.  For convenience it includes three of the documents that are in the August 9, 2016 communication.

Chexal’s email that set up the Technical Review Panel is repeated below.  I was never allowed to participate in that review and I have never seen the panel’s report.  I have asked EPRI to send me that report and it appears unlikely the EPRI will ever provide that.

My communication of August 9, 2016, includes a reference that predates the Chexal directive by almost 2 years, NUCLEONICS WEEK, January 10, 1991.  The two documents below predate the Chexal email by 20 years. Leyse to Gallagher, October 1972, asks if, “… the effects  

of nitrogen saturation and outgassing have been considered in any aspect of design or operation …”

“

Gallagher to Leyse, November 1972, responds, “… the effects of nitrogen saturation have not been considered.”



Three months following Gallagher’s memo, quantitative data was provided by Poulson and Cleary as follows:







So, it is with the above background, that I remain infuriated by the NRC report. AEOD/E504, Failures in the Upper Head Injection System, February 28, 1984.  I then wrote the following on October 3, 1984, that is featured in my letter to you that is dated August 9, 2016:







I provided a copy of the above to the NRC (Keppler) under the condition that it would not expose me as the author. I blacked out portions of the document that identified its source.  Nevertheless, NRC contacted my employer.  That is documented as follows:





However, it was the responses to Leyse, October 3, 1984, that led to the McGuire discovery of no measurement of water level and likely nitrogen-filled accumulators.  Lang refers to an NSAC review that “… concluded that malfunction of the upper head injection system would not place the plant in jeopardy ….”

Lang then reported, “Lanning said that AEOD’s evaluation is similar but that detailed thermal hydraulic analysis will be performed for confirmation.” 

NSAC never admitted that the water-filled accumulator at McGuire was likely filled with nitrogen instead of water and it is likely that AEOD also did not uncover that situation.  Thus the two “separate and independent” organizations were in agreement.

GE ESBWR references and access by US public

Subject: FOIA Appeal NRC-2020-000053 Submitted
Date: 11/25/2019 12:33:12 PM Mountain Standard Time
From: admin@foiaonline.gov
To: bobleyse@aol.com
Sent from the Internet (Details)

This message is to notify you of a new appeal submission to the FOIAonline application. Appeal information is as follows:

• Appeal Tracking Number: NRC-2020-000053
• Request Tracking Number: NRC-2019-000390
• Requester Name: Robert Leyse
• Date Submitted: 11/25/2019
• Appeal Status: Submitted
• Description: Appeal the denial of information of each of the listed INPO, NSAC, or NSAC/INPO records referenced in ESBWR Licensing Topical Report NEDO-33262 (Rev. 2), submitted by GE Hitachi Nuclear Energy, which may be found in ADAMS as ML081560316

Cover letter for FOIA APPEAL, Mailed today, EXPRESS MAIL

November 19, 2019

Robert H. Leyse                                                                                                  P. O. Box 2850                                                                                                Sun Valley, ID 83353

FOIA Officer                                                                                                       U. S. Nuclear Regulatory Commission                                                            Mail Stop T-2 F43                                                                                      Washington, D. C. 20555-0001

FOIA Appeal

Enclosed are stapled documents in the following order:

1.  FOIA REQUEST

2.  NRC FORM 464  2019-000390  dated 10/02/2019

3.  RE: NRC-2019-OOO390  dated 11/1/2019

4. NRC FORM 464  2019-000390 Rev  dated 11/01/2019

Document 1 requests that the public be granted access to references that are cited in a vital GE licensing report; the public need this access so that assorted outside experts may ascertain their accuracy, etc.

Document 2 is the NRC’s erroneous denial.

Document 3 describes the errors in the denial of Document 2.

Document 4 is a correction of Document 2.

Documents 2 and 4 each disclose that since INPO considers the reports to be confidential commercial information the NRC is denying public access to those documents.  That is a very erroneous denial.  Assorted public experts may prevent another Fukushima.  INPO, the NRC and GE are not infallible!



Robert H. Leyse

Also in preparation, page 1 is missing

NRC FORM 464 Part I
(04-2018)
RESPONSE TO FREEDOM OF
INFORMATION ACT (FOIA) REQUEST
U.S. NUCLEAR REGULATORY COMMISSION NRC RESPONSE NUMBER
RESPONSE
TYPE INTERIM FINAL
PART I.D -- COMMENTS
Signature - Freedom of Information Act Officer or Designee
2019-000390 Rev 1
✔
As an initial matter, we note that you requested these same records many years ago; in response to that request, which
was designated FOIA/PA-2008-0322, our office withheld in their entirety the records that the NRC was able to locate under
FOIA exemption 4, and identified identified 13 records that the NRC was unable to locate.
Upon receipt of this request, we reached out to staff in the Office of Nuclear Reactor Regulation (NRR) to ascertain whether
the NRC now has copies of these 13 records. NRR staff informed our office that, under the Memorandum of Agreement with
the Institute of Nuclear Power Operations (INPO), the NRC is not provided access to INPO Nuclear Network OE reports;
nor does the NRC have access to the records originated by the Nuclear Safety Analysis Center, which is a part of the
Electric Power Research Institute (EPRI). (Although we originally misconstrued the reference to NSAC in your request as
pertaining to the Nuclear Science Advisory Committee, which also uses the same acronym, we now understand that you
meant the EPRI group. We note that we were unable to locate any of the specified reports, based upon only a report
number, on EPRI's website. You may wish to reach out to EPRI directly.) NRR staff informed us that they were unable to
locate an INPO Significant Event Report with the title "In Preparation - Inventory Drain Down." NRR staff also informed us
that, although they were not able to locate a joint NSAC/INPO report entitled Significant Event Report 56-81, "Loss of
Station and Reserve Auxiliary Power," they did locate an INPO Significant Event Report 56-81 with the same name. NRR
staff is unfamiliar with any publication called an INPO Nuclear Network WE entry or report and NRR staff reached out to
their point of contact at INPO who informed them that there is no such publication.
With respect to the INPO records that NRC does maintain, as required by our FOIA regulations, we reached out to INPO
representatives to ascertain their disclosure views. INPO confirmed that these reports are considered to be confidential
commercial information. Accordingly, they are being withheld in their entirety under FOIA exemption 4. See Part II.
Stephanie A. Blaney Digitally signed by Stephanie A. Blaney
Date: 2019.11.01 10:56:14 -04'00'
NRC FORM 464 Part II
(04-2018)
U.S. NUCLEAR REGULATORY COMMISSION
RESPONSE TO FREEDOM OF
INFORMATION ACT (FOIA) REQUEST
NRC Form 464 Part II (04-2018)
NRC
DATE:
PART II.A -- APPLICABLE EXEMPTIONS
Exemption 1: The withheld information is properly classified pursuant to an Executive Order protecting national security information.
Records subject to the request are being withheld in their entirety or in part under the FOIA exemption(s) as indicated below (5 U.S.C. 552(b)).
Exemption 2: The withheld information relates solely to the internal personnel rules and practices of NRC.
Exemption 3: The withheld information is specifically exempted from public disclosure by the statute indicated.
Sections 141-145 of the Atomic Energy Act, which prohibits the disclosure of Restricted Data or Formerly Restricted Data (42 U.S.C. 2161-2165).
Section 147 of the Atomic Energy Act, which prohibits the disclosure of Unclassified Safeguards Information (42 U.S.C. 2167).
41 U.S.C. 4702(b), which prohibits the disclosure of contractor proposals, except when incorporated into the contract between the agency and the
Exemption 4: The withheld information is a trade secret or confidential commercial or financial information that is being withheld for the reason(s)
indicated.
The information is considered to be proprietary because it concerns a licensee's or applicant's physical protection or material control and
accounting program for special nuclear material pursuant to 10 CFR 2.390(d)(1).
The information is considered to be another type of confidential business (proprietary) information.
The information was submitted by a foreign source and received in confidence pursuant to 10 CFR 2.390(d)(2).
Exemption 5: The withheld information consists of interagency or intraagency records that are normally privileged in civil litigation.
Deliberative process privilege.
Attorney work product privilege.
Attorney-client privilege.
Exemption 6: The withheld information from a personnel, medical, or similar file, is exempted from public disclosure because its disclosure would result
in a clearly unwarranted invasion of personal privacy.
Exemption 7: The withheld information consists of records compiled for law enforcement purposes and is being withheld for the reason(s) indicated.
(A) Disclosure could reasonably be expected to interfere with an open enforcement proceeding.
(C) Disclosure could reasonably be expected to constitute an unwarranted invasion of personal privacy.
(D) The information consists of names and other information the disclosure of which could reasonably be expected to reveal identities of confidential
sources.
(E) Disclosure would reveal techniques and procedures for law enforcement investigations or prosecutions, or guidelines that could reasonably be
expected to risk circumvention of the law.
(F) Disclosure could reasonably be expected to endanger the life or physical safety of any individual.
Other:
PART II.B -- DENYING OFFICIALS
In accordance with 10 CFR 9.25(g) and 9.25(h) of the U.S. Nuclear Regulatory Commission regulations, the official(s) listed
below have made the determination to withhold certain information responsive to your request.
DENYING OFFICIAL TITLE/OFFICE RECORDS DENIED APPELLATE OFFICIAL
EDO SECY
Other:
2019-000390 Rev
11/01/2019
✔
✔
Stephanie A. Blaney FOIA Officer INPO reports ✔
Select Title/Office from drop-down list
Select Title/Office from drop-down list
Select Title/Office from drop-down list

In preparation, related to following GE ESBWR entry

NRC FORM 464 Part I
(04-2018)
RESPONSE TO FREEDOM OF
INFORMATION ACT (FOIA) REQUEST
U.S. NUCLEAR REGULATORY COMMISSION NRC RESPONSE NUMBER
RESPONSE
TYPE INTERIM FINAL
REQUESTER: DATE:
DESCRIPTION OF REQUESTED RECORDS:
PART I. -- INFORMATION RELEASED
The NRC has made some, or all, of the requested records publicly available through one or more of the following means:
(1) https://www.nrc.gov; (2) public ADAMS, https://www.nrc.gov/reading-rm/adams.html; (3) microfiche available in the NRC Public
Document Room; or FOIA Online, https://foiaonline.regulations.gov/foia/action/public/home.
Agency records subject to the request are enclosed.
Records subject to the request that contain information originated by or of interest to another Federal agency have been referred to
that agency (See Part I.D -- Comments) for a disclosure determination and direct response to you.
We are continuing to process your request.
See Part I.D -- Comments.
PART I.A -- FEES
AMOUNT
You will be billed by NRC for the amount indicated.
You will receive a refund for the amount indicated.
Fees waived.
Since the minimum fee threshold was not met,
you will not be charged fees.
Due to our delayed response, you will not be
charged search and/or duplication fees that
would otherwise be applicable to your request.
PART I.B -- INFORMATION NOT LOCATED OR WITHHELD FROM DISCLOSURE
We did not locate any agency records responsive to your request. Note: Agencies may treat three discrete categories of law
enforcement and national security records as not subject to the FOIA ("exclusions"). See 5 U.S.C. 552(c). This is a standard
notification given to all requesters; it should not be taken to mean that any excluded records do, or do not, exist.
We have withheld certain information pursuant to the FOIA exemptions described, and for the reasons stated, in Part II.
Because this is an interim response to your request, you may not appeal at this time. We will notify you of your right to appeal any of
the responses we have issued in response to your request when we issue our final determination.
You may appeal this final determination within 90 calendar days of the date of this response. If you submit an appeal by mail,
address it to the FOIA Officer, at U.S. Nuclear Regulatory Commission, Mail Stop T-2 F43, Washington, D.C. 20555-0001. You may
submit an appeal by e-mail to FOIA.resource@nrc.gov. You may fax an appeal to (301) 415-5130. Or you may submit an appeal
through FOIA Online, https://foiaonline.regulations.gov/foia/action/public/home. Please be sure to include on your submission that it
is a “FOIA Appeal.”
PART I.C -- REFERENCES AND POINTS OF CONTACT
You have the right to seek assistance from the NRC's FOIA Public Liaison by submitting your inquiry at https://www.nrc.gov/reading-rm/
foia/contact-foia.html, or by calling the FOIA Public Liaison at (301) 415-1276.
If we have denied your request, you have the right to seek dispute resolution services from the NRC's Public Liaison or the Office of
Government Information Services (OGIS). To seek dispute resolution services from OGIS, you may e-mail OGIS at ogis@nara.gov, send
a fax to (202) 741-5789, or send a letter to: Office of Government Information Services, National Archives and Records Administration,
8601 Adelphi Road, College Park, MD 20740-6001. For additional information about OGIS, please visit the OGIS website at
https://www.archives.gov/ogis.
2019-000390 1
✔
Robert Leyse 10/02/2019
Each of the listed INPO, NSAC, or NSAC/INPO records referenced in ESBWR Licensing Topical Report NEDO-33262
(Rev. 2), submitted by GE Hitachi Nuclear Energy, which which may be found in ADAMS as ML081560316.

FOIA for docs referenced in GE ESBWR Licensing Topical Report NEDO-33262

Subject: FOA request
Date: 8/7/2019 8:21:04 AM Mountain Standard Time
From: bobleyse@aol.com
To: FOIA.Resource@nrc.gov

FOIA REQUEST
The following public document, NEDO 33262, written by General Electric Company and its foreign affiliate, references dozens of documents from INPO and/or NSAC. NRC likely has these documents in its files. Under this FOIA I am requesting that NRC place these documents into its PDR for access by the public. The public needs these documents in order to assess NEDO-33262.
Submittal of ESBWR Licensing Topical Report NEDO-33262, ESBWR Human Factors Engineering Operating Experience Review Implementation Plan (OER), Revision 2.
ML081560316
2008-05-31

The following list of INPO, NSAC and NSAC/INPO documents is compiled from NEDO-33262. These documents must be placed in NRC’s PDR under this FOIA.

INPO SOER 88-3

INPO SOER 85-4

INPO SOER 87-2

INPO SOER 84-7

INPO Significant Operating Experience Reports

INPO Significant Operating Experience Report 84-7, Pressure Locking and Thermal Binding ofGate Valves, December 14, 1984.

INPO Significant Event Reports

INPO Significant Event Report xx-9 1, - In Preparation - Inventory Drain Down, 1991.

INPO Significant Event Report 26-89, Loss of Residual Heat Removal Capability Due ToCommon Mode Failure of Flow Control Valves, October 4, 1989.

INPO Significant Event Report 11-89, Inadvertent Introduction of Hydrogen Into The Instrumentand Station Air Systems, April 11, 1989.

INPO Significant Event Report 5-89, Lack of Control of Testing Disables or Challenges SafetySystems, March 3, 1989.

INPO Significant Event Report 36-88, Loss of Residual Heat Removal Due to Misleading VisualIndication of Water Level, November 30, 1988.

INPO Significant Event Report 35-87, Non-lsolable Reactor Coolant System Leak, November12,1987.

INPO Significant Event Report 3 5-86, Extended Loss of Shutdown Cooling due to Steam Bindingof Shutdown Cooling Pumps, October 24, 1986.

INPO Significant Event Report 31-86, Loss of Residual Heat Removal Flow Due To InadvertentDraining Of The Reactor Coolant System, September 3, 1986.

INPO Significant Event Report 23-86, Loss of Decay Heat Removal Due To Inadequate ReactorCoolant System Level Control, July 3, 1986.

INPO Significant Event Report 17-86, Loss Of Shutdown Co. , Flow, May 27, 1986.

INPO Significant Event Report 79-84, Loss Of Shutdown Cooling Due to Inaccurate Level Indication, November 14, 1984.

INPO Significant Event Report 71-84, Residual Heat Removal Pump Damage Caused ByOperation With Suction Valve Closed, October 2, 1984.

INPO Significant Event Report 60-83, Loss of Residual Heat Removal (RHR) Cooling DuringReactor Vessel Drain Down, August 30, 1983.

INPO Significant Event Report 59-83, Residual Heat Removal (RHR) Pump Suction ValveClosure Due To Control Circuitry Design, August 18, 1983.

INPO Significant Event Report 13-83, Unplanned Radioactive Release and Loss of ShutdownCooling, February 25, 1983.

NSAC/INPO Significant Event Report 95-81, Automatic Valve Closure Causing Loss ofShutdown Decay Heat Removal, November 25, 1981.

NSAC/1NPO Significant Event Report 91-81, Steam Voiding in the Reactor Coolant SystemDuring Decay Heat Removal Cooldown, October 6, 1981.

NSAC/INPO Significant Event Report 89-81, Level Instrumentation Oscillations Due ToReference Leg Flashing, October 23, 1981.

NSAC/INPO Significant Event Report 87-81, Inadequate Reactor Coolant System (RCS) WaterLevel Indication, October 19, 1981.

NSAC/INPO Significant Event Report 78-81, Erroneous Indication. Reactor Vessel LevelCauses Loss of RHR, October 1, 1981.

INPO Nuclear Network, WE 655 ENR PAR 90-061, Residual Removal Flow FluctuationsDuring Drawing of Vacuum in the Reactor Coolant System, September 19, 1990.

INPO SERS 42-81 and 5-89.

INPO SERS 17-88 and 36-87

NSAC Report. 146.

INPO/NSAC Significant Operating Experience Report 80-5, Potential Loss of Coolant Accident(LOCA) From A Single Electrical Failure, September 23, 1980.

INPO Significant Experience Report 11-88, Inadvertent Disablement of The Automatic StartCapability For All Site Diesel Generators, May 6, 1988.

INPO Significant Experience Report 25-85, Emergency Diesel Generator Failed To SupplyEmergency Bus Due To Non-emergency Trip, June 3, 1985.

INPO Significant Experience Report 73-83, Loss of All AC Power (Blackout), October 27, 1983.

NSAC/INPO Significant Event Report 56-81, Loss of Station and Reserve Auxiliary Power,August 56, 1981.

INPO SOERs 87-2 and 85-1

NSAC Reports 52 and 43

INPO SOER 82-4, and SER 31-81 and SER 5-90.

INPO SER 38-85

INPO SER 72-84

INPO SOER 85-1.

INPO SOER 82-2

NSAC REPORT 88

INPO SER 63-84 and 2-82

INPO Significant Operating Experience Report 87-2, Inadvertent Draining of Reactor Vessel toSuppression Pool at B WRs, March 19, 1987.

INPO Significant Operating, Experience Report 82-4, Improper Alignment of Spray System ToResidual Heat Removal System, May 19, 1982.

INPO Significant Operating Experience Report 82-2, Inadvertent Reactor Pressure VesselPressurization, Apr. 28, 1982.

INPO Significant Event Report 7-91, Failure to Control Valve Lineup Status Resulting in aReactor Vessel Coolant Drain Down, April 2, 1991.

INPO Significant Event Report 19-90, Monitoring Plant Evolutions Using Inoperable ControlBoard Indications, November 21, 1990.

INPO Significant Event Report 5-90, Premature Lifting and Excessive Blowdown of ResidualHeat Removal Relief Valves, February 3, 1990.

INPO Significant Event Report 39-87, Undetected Loss of Reactor Coolant Due To Release ofDissolved Gases, December 29, 1

INPO Significant Event Report 4-86, Internal Flooding of An Emergency Core Cooling System(ECCS) Pump Room, January 6, 1986.

INPO Significant Event Report 37-83, Supplement 2, Inadvertent Draining of Reactor PressureVessel To Suppression Pool, October 9, 1985.

INPO Significant Event Report 37-83, Inadvertent Draining of Reactor Vessel to SuppressionPool, June 9, 1983.

NSAC/INPO Significant Event Report 85-81, Inadvertent Discharge From Reactor CoolantSystem to Containment Sump, September 25, 1981.

NSAC/INPO Significant Event Report 64-81, Reactor Coolant Leak Due To Technician's Error,August 14, 1981.

NSAC/INPO Significant Event Report 31-81, Inadvertent Containment Spray, April 29, 1981.

NSAC/INPO Significant Event Report 1-81, January 16,1981.

INPO Nuclear Network Entry WE 496, EAR TYO 90-005, RPV Was Pressurized at Low VesselMetal Temperature Condition During Refueling Outage, March 1, 1990.

NSAC Report 129

INPO SOER 85-1

INPO SER 9-86, 51-81, 72-84, 92-84, 9-86, 31-88

INPO Significant Operating Experience Reports

INPO Significant Operating Experience Report 87-2, Inadvertent Draining of Reactor Vessel ToSuppression Pool at BWRs, March 19, 1987.

INPO Significant Operating 'Experience Report 85-1, Reactor Cavity Seal Failure, January 10,1985.

INPO Significant Event Report 1-91, Spent Fuel Pool Overflow Events. January 4, 1991.

INPO Significant Event Report 17-90, Reactor Coolant System Temperature Below AnalyzedLimit for an Extended Time Period, October 24, 1990.

INPO Significant Event Report 15-89, Internal Flooding Resulting From Freeze Plug Failures,June 9, 1989.

INPO Significant Event Report 31-88, Reactor Cavity Seal Failure From Deflation andInadequate Design, October 27, 1988.

INPO Significant Event Report 3-88, Inadvertent Draining of Reactor Vessels Due ToProcedural Content and Usage Deficiencies, February 12, 1988.

INPO Significant Event Report 7-87, Pressurization of Vessel During Cold Shutdown, March 19,1987.

INPO Significant Event Report 4-87, Pipe Break and Condensate Storage Tank Draining, March9, 1987.

INPO Significant Event Report 40-86, Spent Fuel Pool Leakage, December 24, 1986.

INPO Significant Event Report 8-86, Inadvertent Drainage of Refueling Shield Tank, February24, 1986.

INPO Significant Event Report 41-85, Containment Spraying Events, September 19, 1985.

INPO Significant Event Report 38-85, Reactor Vessel Partially Drained Due To InadvertentActuation of the Automatic Depressurization System (ADS) While in Shutdown, August 12, 1985.

INPO Significant Event Report 92-84, Partial Drain of Spent Fuel Storage Pool To Spent FuelShipping Cask Pit Due To Deflated Seal, December 27, 1984.

INPO Significant Event Report 72-84, Reactor Cavity Seal Ring Failure, October 3, 1984.

INPO Significant Event Report 72-84, Supplement 1, Reactor Cavity Seal Ring Failure, April18, 1985.

INPO Significant Event Report 72-84, Supplement 2, Reactor Cavity Seal Failure, February 13,1986.

INPO Significant Event Report 63-84, Over pressurization of Reactor Vessel During ColdShutdown, Aug. 30, 1984.

INPO Significant Event Report 46-83, Inadvertent Initiation of Low Pressure Coolant Injection(LPCI), July 1, 1983.

INPO Significant Event Report 2-82, Cold Pressurization of Reactor Coolant System, January 7,1982.

NSAC/INPO Significant Event Report 76-81, Loss of Primary Coolant To reactor BuildingSump, September 25, 1981.

NSAC/INPO Significant Event Report 61-81, Inadvertent Spent Fuel Pool Overflow, August 12,1981.

NSAC/INPO Significant Event Report 51-81, Spent Fuel Pool Watertight Gate Seals, July 28,1981

INPO Nuclear Network Entry OE 4629, Low Level in Spent Fuel Pool due to Loss of Air toTransfer Canal Weir Gate Bladder, June 4, 1991.

INPO SER 17-90.

INPO SER 15-83

NSAC Report 129

INPO Significant Event Report 1-88

INPO Significant Event Report 21-86

INPO Significant Event Report 59-81

INPO Significant Event Report 31-83

INPO Significant Event Report 31-83

INPO Significant Event Report 5-86

INPO Significant Event Report 15-91, Fuel Mispositioning Events DL3 to Fuel Bundle SelectionErrors, June 11, 1991.

INPO Significant Event Report 10-88, Fuel Assembly Lifted With Upper InteMal, April 21,1988.

INPO Significant Event Report 5-86, Dropped New Fuel Assembly, January 15, 1986.

INPO Significant Event Rep~rt 21-86, Dropped Fuel Assembly, June 16, 1986.

INPO Significant Event Report 31-85, Inadvertent Fuel Bundle Movement, June 27, 1985.

INPO Significant Event Report 31-83, Irradiated Fuel Assembly Dropped From Fuel HandlingCrane, June 6, 1983.

INPO Significant Event Report 15-83, Fuel Handling Error, March 11, 1983.

INPO Significant Event Report 43-82, Fractured Fuel Assembly Guide Tubes, July 19, 1982.

INPO Significant Event Report 59-81, Dropped Fuel Assembly, August 11, 1981.

INPO Nuclear Network Entry OE 4167, Fuel Assemblies Withdrawn With Upper Intenals,October 5, 1990.

INPO Nuclear Network Entry OE 4112, Fuel Assemblies Withdrawn With Upper Internals -Update to OE's 4167, 4177, and 4187, October 26, 1990.

INPO Nuclear Network Entry OE 4113, Fuel Assemblies Withdrawn With Upper lnteMals -Update to OE4167(message replaced OE 4177), October 27, 1990.

INPO Nuclear Network Entry OE 4114, Fuel Assemblies Withdrawn With Upper Internals -Update to OE4167 and 4177(message replaced OE 4187), October 27, 1990.

9th International Conference on Boiling and Condensation Heat Transfer

9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado

MICROSCALE PHASE CHANGE HEAT TRANSFER TO WATER
Robert H. Leyse*
INZ Inc., P. O. Box 2850, Sun Valley, ID 83353
bobleyse@aol.com

ABSTRACT
Leyse has pioneered the field of microscale phase change heat transfer to water at ultra-high power density with fine platinum wires, 7.5 μm diameter, that are joule heated in pressurized degassed deionized water. Each wire functions simultaneously as a heat transfer element and as a resistance thermometer as originated by Nukiyama (1934). These experiments cover the pressure range from 200 to 4000 PSIA and the heat flux range from very low to 4000 W/cm2 while bulk water temperature is maintained in the range of 20 oC. These investigations cover two separate situations: Case (i) constant pressure and varying power, and Case (ii) somewhat constant power and varying pressure. Limited explorations reveal a significant impact of dissolved nitrogen (saturated) at 1000 PSIA.

INTRODUCTION
One reviewer of an early Leyse paper1 remarked, “The study has used interesting ultra high heat fluxes over a wide pressure range. New transition paths are demonstrated in the results. The author does not claim fundamental explanations of the phenomena, but he challenges the theoreticians as well as the numerical modelers to use their skills instead for further study of the phenomena. However, because the report contains new results that are interesting enough to excite new theories in the field of boiling phenomena, I recommend its publication.” To date neither Leyse nor anyone else has come forth with related new theories in the field of boiling phenomena.

APPARATUS AND PROCEDURES
The heat transfer element is a fine platinum wire, 7.5 μm diameter by 1.14 mm long. It is installed within the lower end of a vertical stainless steeel tube, 0.3 inches internal diameter. by 8 inches long. The arrangement of the platinum wire and ancillary support apparatus is detailed in Figure 1 in which the pyrex tube is replaced by the stainless steel tube for the high pressure runs of this presentation. The microscale heat transfer element is welded to platinum terminals, 0.020 inch diameter. The local water temperature is measured with a chromel-alumel thermocouple that is sheathed withn a 0.020 inch diameter stainless steel assembly. This thermocouple junction is about 0.020 inches below the horizontal fine platinum wire. The assembly is filled with degassed, demineralized water and is pressurized with a pneumatic hand pump. Pressure is monitored with a Rosemount recording pressure transducer .
The heat transfer element is powered with a programmed direct current power source. The power measurement is atraightforward. A 10 ohm precision resistance is in series with platinum element. Voltage is measured across the platinum element and across the 10 ohm resistance. The product of these two voltages divided by 10 yields the power in watts to the platinum element. All data, the two voltages, the pressure, and the water temperature is recorded every 0.1 second in an excel spreadsheet which facilitates data handling, plotting and analysis.
The programmed power supply controls the total voltage drop across the two resistances. This is a satisfactory arrangement for the runs in which power is programmed during the series of runs at constant pressure, Case (i). As will be discussed later, it is an acceptable, although possibly a less satisfactory arrangement for runs at somewhat constant heat flux, Case (ii). (The runs at somewhat constant heat flux were an afterthought; a worthwhile afterthought). A run at constant pressure is completed in about 10 seconds, the power increases for 5 seconds and then decreases for 5 seconds (Figure 2).
Preparations for a run consist of equipment setup including the fillng and venting processes to insure a “solid” system with degassed, demineralized water. The programmed power supply is set (checked) for a linear increase and decrease to a with the 10 ohm resistor and a dummy resistor, usually three ohms. Following these preparations, the run proceeds; data are recorded every 0.1 seconds as power increases for about 5 seconds and then decreases for about 5 seconds and the run is complete.

Figure 2. Timing of Run at 1000 PSIA
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9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado
RUNS AT CONSTANT PRESSURE AND VARYING POWER
The top plot in Figure 3 covers all of the data for eight runs at constant pressure ranging from 200 to 6000 PSIA. At subcritical pressures 200, 1000 and 2500 PSIA, the inception of nucleate boiling begins at progressively increasing temperatures, 240, 307 and 349 oC. The plots are nearly identical for the increasing and decreasing data sets. (At 3000 PSIA there is a unique situation that is detailed in a later discussion, Figures 4 and 5.)
At supercritical pressures there is completely different characteristic. The temperature of the heat transfer element increases vastly with only a modest inclrease in heat flux, in contrast to the nucleate boiling charactersitc at subcritcial presssures where only a modest increase in temperature yields a substantial increase in heat flux.. At 6000 PSIA the increasing and decreasing plots are almost identical beyond 550 oC, and are somewhat together over their entire span. The divergence between increasing and decreasing data sets progressively increases for runs at the pressure data sets from 6000, 5000, 4300 to 3600 PSIA.
The plots for the increasing power only are the mid set in Figure 3. Note that the departures from natural circulation are somewhat similar for the runs at 3600 and 4300 PSIA in the temperature range from the critical temperature to about 550 oC. This is in contrast to the runs at 5000 and 6000 PSIA. The 3000 PSIA case is clearly distinct.
The consistently parallel plots in the lower set in Figure 3also include the 3000 PSIA case. It is interesting that at all pressures, the return to natural circulation without phase change heat transfer is at progressively lesser subcritical temperatures, although the 6000 PSIA case is very close to the critical temperature. It is also interesitng that the 3000 PSIA case fits very well into the trends.
See Figure 4 for futher discussion of the subcritical data. Note that the transition to phase change heat transfer is 65 oC greater than the saturation temperature at 200 PSIA and that it becomes very little at 2500 PSIA. Moving to Figure 5, at 3000 PSIA there is a complex transiation to phase change heat transfer. At a heat flux of about 2700 W/cm2 there is a transition to nucleate boiling at very nearly the saturation temperature. Within about 0.3 seconds there is a departure from the nucleate boiling. And, within another 0.1 seconds the peak temperature jumps by about 480 oC to 876oC. The 876 oC point is at the peak of the programmed power. The return to non-phase change heat transfer proceeds more slowly, about 0.9 seconds. The return is at 300 oC and that is about 60 oC less than the departure.
RUNS AT CONSTANT PRESSURE AND VARYING POWER WITH DISSOLVED NITROGEN
The apparatus also has been operated as described with the water saturated with disolved nitrogen. A supply of nitrogen under high pressure was valved to the apparatus for several days at 1000 PSIA such that the water became saturated with disolved nitrogen. Leyse discovered that the presence of dissolved nitrogen changed the heat transfer during non phase change heat transfer (natural circulation). It also alterted the point of departure to phase change heat transfer. He received a U. S. Patent2 for his process. The plots in Figure 6 are copied from the patent. The following text is copied from the patent; the refernces to FIG. 4 and FIG. 5 are for the plots in Figure 6.
The plot of power applied versus resistance of the sensor element when immersed in degassed water at approximately 1000 psi is shown in FIG. 4. A corresponding plot of water saturated with nitrogen at approximately 1000 psi shown in FIG. 5 reveals several aspects which quantify the presence of dissolved nitrogen in the nitrogen saturated water relative to the degassed water. Each curve has a region of linear increasing slope S and a knee K at which the slope abruptly increases. With degasssed water the linear slope S is 0.26 watts/ohm while for water saturated with nitrogen the slope S is 0.19 watts/ohm. With degassed water, the coordiantes of the knee K are 8.7 ohms and 0.97 watts while with water saturated with nitrogen, the coorddinates of the knee are 8.0 ohms and 0.66 watts. Similar calibrations may be produced for intermediate concentrations of dissovled nitrogen.
It is not necessary to present this discovery as plots of heat flux vs. temperature in order to have an operational and patentable device. However, it is clear that the heat transfer coefficient during natural circulation is substantially less with dissolved nitrogen. It is also clear that the transition from natural circulation to phase change heat transfer occurs at a substantially lower heat flux for the case with dissolved nitrogen. It appears that the transition from natural circulation to phase change heat transfer has a somewhat rounded shape with dissolved nitrogen in contrast to the relatively sharp transition with degassed water. Further investigations are proposed.
9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado
Figure 3 Microscale Phase Change Heat Transfer to Subcooled Water – Three Plots
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9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado
Figure 4 Four runs at constant pressures ranging from 200 to 3000
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Figure 5 Unique near-critical run at 3000 PSIA Tsat = 368.53 oC
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9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado
9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado
RUNS AT SUBSTANTIALLY CONSTANT POWER AND VARYING PRESSURE
These runs were completed as an afterthought. Each of four runs has the total voltage set at a fixed value. (Recall that the total voltage is the sum of that across the 10 ohm resistance plus that across the platinum element.) A run proceeds as follows: The apparatus is pressurized to about 6000PSIA, power is turned on, and pressure is steadily reduced to about 200 PSIA over a period of about 20 seconds. See Figures 7 and 8. Figure 7 is a plot of heat flux during each run; the runs are coded according to the peak heat flux in each; 2630, 2930, 3010 and 3360 W/cm2. Figure 8 is a plot of the corresponding temperature of the heat transfer element during each run.
For run 2630, the plot is smooth over the entire span and the heat flux is within 30 W/cm2 of 2600 W/cm2 for the span from about 6000 PSIA to 1000 PSIA. The corresponding temperature plot, Figure 8, is also smooth and relatively flat.
Runs 2930 and 3010 each have a distinct upward step of about 200 oC at about 3800 and 4200 PSIA respectively. Next, each has a steep temperature increase up to the critical temperature at which point the temperature “turns around” and a steep decrease follows until there is a step decrease at about 2400 PSIA for each. The step decreases each terminate very near to the saturation temperature.
Run 3360 has the same character as runs 2930 and 3010, although the upward step is at a higher pressure than is covered in these investigations.
The heat flux plots, Figure 7, reflect the varying resistance of the platinum element. An increase in resistance of the element (temperature) leads to an increased voltage drop and an increased power. This is a consequence of the control by fixed voltage across the series circuit of the 10 ohm resistance and the platinum element. As the resistance of the platinum element increases, its fraction of the total voltage increases. Although this leads to less amperage in the circuit, the net impact is an increase in power to the platinum element. The circuit design turns out to be fortuitously tailored for this investigation because if the heat flux was indeed held constant over the pressure range it would take a multitude of runs to discover the step changes as well as the turnaround at or near the critical pressure.
In Figure 8 the plots of all runs are very close together at pressures from about 1400 PSIG to termination of the runs at about 300 PSIG. This because the heat transfer is by nucleate boiling, and at any given pressure the temperature varies relatively little with heat flux. Clearly, the plots would be very similar to these even if constant heat flux had been achieved over that pressure range. This is consistent with the nucleate boiling regimes of the plots in Figure 4 in two respects; one, at any given pressure the temperature varies relatively little with heat flux, and two, the difference between the platinum temperature and the saturation temperature decreases as the saturation temperature increases.
APPLICATIONS WILL LIKELY PRECEDE UNDERSTANDING OF THESE PROCESSES
During 2007, J. F. Zhao3 reported “Boiling is a very complex and illusive process because of the interrelation of numerous factors and effects as the nucleate process, the growth of the bubbles, the interaction between the heater’s surface with liquid and vapor, the evaporation process at the liquid-vapor interface, and the transport process of the vapor and hot liquid away from the heater’s surface. For a variety of reasons, fewer studies have focused on the physics of the boiling process than have been tailored to fit the needs of engineering endeavors. As a result, the literature has been flooded with the correlations involving several adjustable, empirical parameters. These correlations can provide quick input to design, performance and safety issues and hence are attractive on a short term basis. However, the usefulness of the correlations diminishes very quickly as parameters of interest start to fall outside the range of physical parameters for which the correlations were developed. Thus the physics of the boiling process itself is not properly understood yet, and is poorly represented in the most correlations, despite of almost seven decades of boiling research.”
This paper reveals multiple discoveries in the field of phase change heat transfer that are substantially more complex and illusive than the pool boiling that has been reported to date. The author has not hypothesized controlling physical mechanisms governing the heat transfer behavior, and he has not articulated a path to gain a fundamental understanding of the physics governing his high heat flux processes. An understanding of the impact of dissolved gases may be the first penetration. Applications in microscale chemical engineering such as isotope separat ion may yield insights. Discoveries in phase change heat transfer have been revealed via physical experiments only, and that is likely to be the continuing source even though supercomputers are everywhere.
9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado
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Figure 8 Four Runs at Substantially Constant Heat Flux 20030040050060070080090001000200030004000500060007000PRESSURE PSIATemperature oC2630 W/cm22930 W/cm23010 W/cm23360 W/cm2TsaturationPcritical
9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado
APPARATUS FOR MONITORING THE TIMING AND PATTERNS OF CIRCULATION
Apparatus is visualized for determining the timing and patterns of circulation. Two phases are planned. The first phase will utilize the present assembly with two opposed elements. One element will be will powered while the opposite element monitors temperature. The test procedure will include pulsed application of high heat flux and precise timing of the response of the temperature element.
The second phase will include a complex assembly with perhaps eight platinum elements. One element will be the heat transfer element while others will be resistance thermometers. At least two of the elements will be vertical. A series of runs will have the heat transfer source rotated among the elements. Circulation patterns will be inferred via data analysis. Among several challenges is the requirement to support and deliver power to the fine platinum wires without unduly disturbing the flow patterns; fine copper wires under tension may work.
Details of the campaign are very incomplete. These phase two activities will be iterative; a series of assemblies will be modified from run to run, etc. There are no design tools to guide the arrangements of the platinum elements. The analysis of the measurements may lead to further schemes; practical applications are unlikely.
Measurements to date have been at 0.1 second intervals. Future capability will include 0.001 second intervals and perhaps faster. A faster recording capability will yield further detailing of at least three aspects of the runs to date; the jump from about 480 oC to 876oC at 3000 PSIA, the increasing and the decreasing step changes during the runs at substantially constant power, and the sharpness of the transitions from natural circulation to nucleate boiling or supercritical conditions.
The visualized run series will include at least five interelated features:
1. Ramped runs with durations including and exceeding the ten seconds of runs to date.
2. Runs with step power inputs up to 4000 W/cm2 and precise timing of responses of temperature detectors.
3. Runs at system pressures from one atmosphere to 400 atmospheres.
4. Runs at system bulk temperatures form 20oC to 370oC.
5. Runs with constant power and varying pressure. (Power inputs up to 5000 W/cm2 that are held constant during controlled uniform rate of system pressure decrease from 10,000 PSIA to 200 PSIA over 10 seconds and 20 seconds.
COMMENTS
At subcritical pressures this work did not explore heat transfer regimes beyond nucleate boiling. At the time it was believed that achieving 4000 W/cm2 at 1000 PSIG was remarkable. Leyse believed that burnout was less likely at one third of the critical pressure and therefore the programmed power was restricted to lesser peak heat fluxes during the initial runs at higher subcritical pressures. The results at 3000 PSIA with the brief time, about 0.3 seconds, in the nucleate boiling regime, followed by the fantastic jump across the supercritical temperature arena with very little increase in power justifies the tame approach in setting peak voltage. The exploration of the supercritical arena also proceeded with caution; however, 4000 W/cm2 was achieved at 6000 PSIA.
Bakhru and Lienhard4, 1972, asserted in their publication, BOILING FROM SMALL CYLINDERS, that, “Nucleate boiling does not occur on the small wires” and “Three modes of heat removal are identified
for the monotonic curve and described analytically: a natural convection mode, a mixed film boiling and
natural convection mode, and a pure film boiling mode.” However, although those wires are three to ten times the diameter of the 7.5 micron platinum wires of this work; this work clearly revealed nucleate boiling from the small wires. Balhru and Lienhard only performed experiments at low pressures; it would be a relatively easy experiment to deploy those wires at higher pressures in order to reveal a transition to nucleate boiling.
Finally, the Leyse procedures that have deployed a steady increase of heat flux and relatively fast recording with a very sensitive heat transfer element have led to highly significant discoveries that have been lost for decades in the popular procedures of holding fixed steady powers for long times with massive heat transfer elements.
9th International Conference on Boiling and Condensation Heat Transfer
April 26-30, 2015 – Boulder, Colorado
REFERENCES
1. Leyse, R. H., Microscale heat transfer to subcooled water, 200-6000 psia, 0-3,500 W/cm2 , Annals of the New York Academy of Sciences, Volume 974, MICROGRAVITY TRANSPORT PROCESSES IN FLUID, THERMAL, BIOLOGICAL, AND MATERIALS SCIENCES, pages 261–273, October 2002.
 2. Leyse, R. H., Method for monitoring for the presence of dissolved gas in a fluid under pressure, United States Patent 5,621,161 April 15, 1997.
3. Zhao, J. F., Wan, S. X., Liu, G., Li, D., Lu, Y. H., and Yan, N., Lateral Motion and Departure of Vapor Bubbles in Nucleate Pool Boiling on Thin Wires in Microgravity, Proceedings of the Fifth International Conference on Fluid Mechanics, Tsinghua University Press & Springer, Aug. 15-19, 2007.
4. Bakhru, N. and Lienhard, J. H., Boiling from small cylinders, Int. J. Heat and Mass Transfer, vol. 15, pp. 2011-2025, 1972.