alea-institute/kl3m-data-sample-004-shuffled huggingface dataset:
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# Assessment of Critical Subcooled Flow Through Cracks in Large and Small Pipes Using TRACE and RELAP5 (NUREG/IA-0457)

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        - NUREG/IA-0457 (PDF – 4.37 MB)

## Publication Information

**Manuscript Completed:** November 2015**Date Published:** August 2017

**Prepared by:** Andrew Oussoren; J. Riznic; S.T. Revankar*

Canadian Nuclear Safety Commission 280 Slater Street P.O. Box 1046 Station B Ottawa, ON, Canada

*School of Nuclear Engineering, Purdue University W. Lafayette, IN, USA

Kirk Tien, NRC Project Manager

Division of Systems Analysis Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

**Prepared as part of:** The Agreement on Research Participation and Technical Exchange Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

**Published by:** U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 

Availability Notice

## Abstract

The thermal hydraulics system code TRACE has been used to predict subcooled critical flow in crack geometries. Three different experimental data sets were modelled in this study. The first experiment, performed at Purdue University, measured critical flow in slits with small section thicknesses similar to steam generator tubing. These results are also compared to model predictions using the RELAP5 code. The second experiment, conducted by Ontario Hydro (OH), measured critical flow rates in simulated circumferential cracks in thick-walled piping. The third experiment, from Atomic Energy of Canada Ltd. (AECL), measured critical flow through pressure cycling-induced fatigue cracks in thick walled vessels.

For the Purdue tests, TRACE predictions were similar to those obtained with RELAP5. For the thin walled samples in these tests a junction nodalization was found to be more suitable than explicitly modelling the section thickness of the samples. TRACE predictions for both the Purdue and OH tests were in reasonable agreement with measured leak rates, with calculated discharge coefficients of between 0.5 and 1.0 for most cases. However, the model of the OH tests showed a trend towards under prediction as pressure dropped below 8 MPa. There was no clear trend demonstrated with respect to subcooling. For the AECL experiments the flow rate was significantly over predicted, with discharge coefficients as low as 0.1. This result is consistent with the modelling performed by the original experimenters and is likely due to the complexity of the flow pathway in the fatigue cracks used in this test, as compared to the machined geometries of the other samples.

In general TRACE appears to be a suitable tool for prediction of critical flow rates in crack geometries. Comparison against additional experimental data is recommended.

Page Last Reviewed/Updated Tuesday, March 09, 2021
No. 86-867.

#### Session et ux. *v.*I. T. O. Corporation of Ameriport.

C. A. 3d Cir. Certiorari denied.
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# The Development and Assessment of TRACE Model for Maanshan Nuclear Power Plant LOCA (NUREG/IA-0404)

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        - Publication Information
        - Abstract

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        - NUREG/IA-0404 (PDF - 2.60 MB)

## Publication Information

**Manuscript Completed:** September 2011**Date Published:** September 2012

**Prepared by:** Jong-Rong Wang, Hao-Tzu Lin, Jung-Hua Yang*, Chunkuan Shih* Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C. 1000, Wenhua Rd., Chiaan Village, Lungtan, Taoyuan, 325, TAIWAN *Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Section 2, Kuang Fu Rd., HsinChu, TAIWAN

A. Calvo, NRC Project Manager

**Prepared as part of:** The Agreement on Research Participation and Technical Exchange Under the Thermal-Hydraulic Code Applications and Maintenance Program (CAMP)

**Published by:** Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 

Availability Notice

## Abstract

The U.S. NRC is developing an advanced thermal hydraulic code named TRACE for safety analyses of nuclear power plants (NPPs). According to TRACE user’s manual, NRC has ensured that TRACE will be the main code used in thermal hydraulic safety analysis in the future, without further development of other thermal hydraulic codes such as RELAP5 and TRAC. SNAP is a program with graphic user interface, which processes the input and output of TRACE. As TRACE user’s manual describes, TRACE has a greater simulation capability than the other old codes, especially for events like loss-of-coolant accident (LOCA).

The Maanshan NPP operated by Taiwan Power Company is the only Westinghouse-PWR in Taiwan. The rated core thermal power is 2775 MWt. The reactor coolant system has three loops, each of which includes a reactor coolant pump and a steam generator. The pressurizer is connected to the hot-leg piping in loop 2. The Maanshan NPP TRACE model has been successfully established [1]-[3]. This study focuses on the establishment of the TRACE large break LOCA (LBLOCA) model for Maanshan NPP. Then, the TRACE LBLOCA analysis results compare with the FSAR data [4]. In this study, the LBLOCA is defined as a double-ended guillotine in cold-leg. The break is located in loop 1, which is one of the two loops that don’t have a pressurizer. For a LBLOCA analysis, the most important parameter is the peak cladding temperature (PCT). As defined by the 10 CFR 50.46 regulation [5], the PCT does not exceed 1477.6 K (2200°F). In this LBLOCA, the peak cladding temperature of TRACE calculated was 1358.8 K (1986°F).

Page Last Reviewed/Updated Tuesday, March 09, 2021
## In the Court of Appeals Second Ap pellate District of Texa s at Fort Worth

___________________________ No. 02

- 22

- 00141

- CV

___________________________

On Appeal from the 322nd District Court Tarrant County, Texas Trial Court No. 322

- 644937

- 18
Before Kerr, Birdwell, and Bassel, JJ. Memorandum Opinion by Jus tice Kerr 

### I N THE 

### I NTEREST OF 

### A.J., A.T. , J.T. , AND 

### J.J., C HILDREN2 MEMORANDUM OPINION Appellan t is J. J. ’ s fath e r . 1 T he Department of Family and Protective Services (DFPS) removed J.J. and his three half

- s iblings from Mo ther and placed the m with their maternal grandmother while seeking family reunification and , alternatively,

te rmination of par ental rights

. A t the trial ’ s conclusion, t he trial court made the grandmother the children ’ s permanent managing conservator and made Father a possessory conservator
. F ather ’ s brief is not a mod el of clari ty
. But t o the e x tent we can ide ntify his
issues , w e do not reach his argument that “ the ground s for termination were not
proven ” because the trial court did not actually terminate Father ’ s parental rights

. 2
And w e over rule Father ’ s due

- process com plaint s beca use he d id no t preserve them

. 3
See Tex. R. App. P. 33.1 ( setting out preservation

- of

- error requirements) ; Tex. R. Civ.

1 We use aliases to refer to the child and his family members. See Tex. Fam. Code Ann. § 109.002(d); Tex. R. App. P. 9.8(b)(2).

2 Fa ther argues tha t the ev idenc e was legally and factually insufficient to terminate his parental rights , b ut his rights were not terminated. We encourage greater attention to the record in the future. F urt her, a s DFPS note s , Fat her ’ s brief doe s not address the trial cou rt ’ s conservator ship order , and

— in any event

— the record b efore
us does not refle ct an abuse of discretion

. See I n r e J

. M

. , N o . 0 2

- 1 7

- 0 0 1 5 6

- C V , 2 0 1 7 W L 4 5 4 2 6 7 4 , a t * 1 , * 5 ( T e x . A p p
. — F o r t W o r t h O c t
. 1 2 , 20 17, no pet.) (mem. op .)

.
3 Father com plains that his d ue

- proces s right s were violated because he did not receive any services from DFPS , but he did not raise or receive a ruling on this issue at trial. He also co mplains th at he w as not contacted by his counsel be fore trial , but h e does not explain how he was harmed by any la ck of contact in light o f the record and the trial ’ s outcome
. Cf. Tex. R. App. P. 38.1, 44.1

.3 P. 251 (setting out continuance requirements) ; In re L.M.I. , 11 9 S.W.3d 7 07 , 711 (Tex. 2003) ; In re J.P.

- L. , 592 S.W.3d 559 , 575 (Tex. App

. — Fort W orth 2019, pet. denied)

. We affirm the trial court ’ s judgment.

/s/ Elizabeth Kerr Elizabeth Kerr Justice
Delivered: August 26, 2022
Proposed Rule: Mandatory Reporting of Greenhouse Gas Emissions 

Meeting Documentation 

Date Staff Contact 

Information (Name, email, phone #) 

Name of Organization (Attach list of all attendees names and emails if possible) 

Summary of Discussion (e.g., applicability, reporting requirements, etc.) 

Written CommenProvidedDid they submit to 

2/22-3/1/10 email and telephone correspondence 

 Mollie Averyt, ICF International, 

maveryt@icfi.com

202-862-1569 on behalf of EPA under Contract EPW-07-068, Task Order 68 

Phil Bolin, Mitsubishi Electric Power Products, Inc. (MEPPI) 724-778-5235 and David Giegel, Mitsubishi Electric Power Products, Inc. (MEPPI) 724-778-5255 

Topics included: - how disbursements are measured.--MEPPI is currently 

employing the use of an accurate digital scale. A bottle is weighed before filling and after filling and the weight after filling is considered the disbursement.

- options for accurately determining nameplate capacity of 

equipment before it is shipped to the customer. - it would be very difficult and impractical to weigh the 

shipping unit 

- determining variability of nameplate capacity among a 

certain make/model is a plausible option; thought that internal volume is very, very close to the same from one unit to the next unit. However, bushings may vary more than other parts of the breaker. 

- OEMs have no experience with flow meters in using 

SF6. Focus is instead on the pressure rise that monitors 

the transfer of the gas from a bottle to a breaker. Pressure gauges, however, vary in accuracy, usually rated at 1.5% to 2%; however, it has not been a concern at the facility if the breaker is not precisely filled to a certain partial charge—it is more of a concern at installation that the breaker is precisely filled to its full charge. 

- The process by which self-sealing valves connect the bottle 

to the breaker for filling. The pressure gauge may be on the regulator on the bottle, or may be temporarily connected to the breaker or may be permanently mounted on the breaker. 

- Filling of breakers--breakers are filled with SF6 gas to a full 

charge as part of the testing procedures; the equipment is then evacuated and filled to a nominal charge for shipment at a later stage in the process. 

No 
Date Staff Contact 

Information (Name, email, phone #) 

Name of Organization (Attach list of all attendees names and emails if possible) 

Summary of Discussion (e.g., applicability, reporting requirements, etc.) 

Written CommenProvidedDid they submit to 

- Overcharge in the field--when a breaker is filled to its 

functioning charge out in the field sometime is may be overfilled by up to 3 psig. On an 85 psig rated pressure breaker, this overfill will place an extra 3% of gas in the breaker above nameplate. If a leak occurs, it must leak down to the first alarm level (10% below nameplate capacity) in order to cause an alarm. By initially overfilling by 3%, the breaker will need to leak 13% (3% plus 10%) of its gas in order to leak down to the first alarm level which is 10% below nameplate capacity).
EXHIBIT 32.2

Certification of Principal Financial Officer

Pursuant to 18 U.S.C. 1350

(Section 906 of the Sarbanes-Oxley Act of 2002)

I, George Z. Moratis, Executive Vice President and Chief Accounting Officer (principal financial and accounting officer) of InPhonic, Inc. (the Registrant), certify that to the best of my knowledge, based
upon a review of the Annual Report on Form 10-K, as amended for the period ended December 31, 2006 of the Registrant (the Report):

(1) The Report fully complies with the requirements of Section 13(a) of the Securities Exchange Act of 1934, as amended; and

(2) The information contained in the Report fairly presents, in all material respects, the financial condition and results of operations
of the Registrant.



/s/ GEORGE Z. MORATISName:

George Z. MoratisDate:

May 31, 2007
Exhibit 13.2

ChipMOS TECHNOLOGIES (Bermuda) LTD.

CERTIFICATION

Pursuant to 18 U.S.C. §1350, the undersigned, Shou-Kang Chen, Chief Financial Officer of ChipMOS TECHNOLOGIES (Bermuda) LTD. (the
Company), hereby certifies, to his knowledge, that the Companys Annual Report on Form 20-F for the year ended December 31, 2010 (the Report) fully complies with the requirements of Section13(a) or 15(d), as
applicable, of the Securities Exchange Act of 1934, and that the information contained in the Report fairly presents, in all material respects, the financial condition and results of operations of the Company.

Date: June 3, 2011



/s/ Shou-Kang ChenName:

Shou-Kang ChenTitle:

Chief Financial Officer


The
foregoing certification is being furnished solely pursuant to 18 U.S.C. §1350 and is not being filed as part of the Report or as a separate disclosure document.
HomeNRC LibraryDocument CollectionsNUREG-Series PublicationsPublications Resulting from International Agreements

# Assessment of RELAP5/MOD3.2 Using LOFT Large Break LOCA Test, LP–02–6 (NUREG/IA-0139)

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        - NUREG/IA-0139 (PDF 3.58 MB)

## Publication Information

**Date Published:** August 1998

**Prepared by:** T. S. Choi, J. H. Lee, B. S. Park, C. S. Cho, J.Y. Park/KNFC Y. S. Bang, S. W Seul, H. J. Kim/KINS

Korea Nuclear Fuel Company DogJin-Dong 150 Yusong Gu, Daejeon City Korea

Korea Institute of Nuclear Safety Advanced Reactor Dept. P.O. Box 114 Yusung, Taejon 305–600 Korea

**Prepared as part of:** The Agreement on Research Participation and Technical Exchange under the International Thermal-Hydraulic Code Assessment and Maintenance Program (CAMP)

Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Availability Notice

## Abstract

The LOFT experiment LP-02-6 was simulated using the RELAP5/MOD3.2 code to assess its capability to predict the thermal-hydraulic phenomena in LBLOCA of the PWR. The reactor vessel was modeled with two core channels and split downcomer for a base calculation. The results of the base calculation show that the code can not predict the early bottom-up quenching which is a distinguished phenomenon of the experiment LP-02-6, mainly due to the deficiency of break flow model.

The discharge coefficient sensitivity study was performed to show that the calculated subcooled break flow which might significantly affect the early bottom-up quenching is dependent on the coefficient. More detailed modeling of the cross flow in the split downcomer was performed, but, resulted in little improvement on the predictability of bottom-up quenching. Additional calculation using the RELAP5/MOD3.1 instead of RELAP5/MOD3.2 showed that there is no large difference between the versions in the simulation of LBLOCA.

Page Last Reviewed/Updated Tuesday, March 09, 2021
Exhibit 32 CERTIFICATIONS OF PRINCIPAL EXECUTIVE OFFICER AND PRINCIPAL FINANCIAL OFFICER PURSUANT TO 18 U.S.C. SECTION 1350, AS ADOPTED PURSUANT TO SECTION 906 OF THE SARBANES-OXLEY ACT OF 2002 In connection with the Annual Report of Micromet, Inc. (the Company) on Form 10-K for the fiscal year ended December 31, 2010 as filed with the Securities and Exchange Commission on the date hereof (the Report), I, Christian Itin, President and Chief Executive Officer of the Company, certify, pursuant to 18 U.S.C. Section 1350, as adopted pursuant to Section 906 of the Sarbanes-Oxley Act of 2002, that to my knowledge: 1.  The Report fully complies with the requirements of Section 13(a) or 15(d) of the Securities Exchange Act of 1934; and 2.  The information contained in the Report fairly presents, in all material respects, the financial condition of the Company at the end of the period covered by the Report and results of operations of the Company for the period covered by the Report.  Date: March 3, 2011

          

          /s/ Christian ItinChristian Itin President and Chief Executive Officer (principal executive officer)
 This certification accompanies the Report and shall not be deemed filed by the Company with the Securities and Exchange Commission and is not to be incorporated by reference into any filing of the Company under the Securities Act of 1933, as amended, or the Securities Exchange Act of 1934, as amended (whether made before or after the date of the Report), irrespective of any general incorporation language contained in such filing. 
 In connection with the Annual Report of Micromet, Inc. (the Company) on Form 10-K for the fiscal year ended December 31, 2010 as filed with the Securities and Exchange Commission on the date hereof (the Report), I, Barclay A. Phillips, Senior Vice President and Chief Financial Officer of the Company, certify, pursuant to 18 U.S.C. Section 1350, as adopted pursuant to Section 906 of the Sarbanes-Oxley Act of 2002, that to my knowledge: 1.  The Report fully complies with the requirements of Section 13(a) or 15(d) of the Securities Exchange Act of 1934; and 2.  The information contained in the Report fairly presents, in all material respects, the financial condition of the Company at the end of the period covered by the Report and results of operations of the Company for the period covered by the Report.  Date: March 3, 2011

          

          /s/ Barclay PhillipsBarclay Phillips Senior Vice President and Chief Financial Officer (principal financial officer)
 This certification accompanies the Report and shall not be deemed filed by the Company with the Securities and Exchange Commission and is not to be incorporated by reference into any filing of the Company under the Securities Act of 1933, as amended, or the Securities Exchange Act of 1934, as amended (whether made before or after the date of the Report), irrespective of any general incorporation language contained in such filing.
UNPUBLISHED

UNITED STATES COURT OF APPEALS

FOR THE FOURTH CIRCUIT

No. 98-2790

LAURIE B. FOOTE; IRA T. FOOTE, JR.,

Plaintiffs - Appellants,

versus

DONALD H. BARTON; DAWSON HOESED; LANDON DEAL;

PEGGY SINGLETON; EDDIE LANCE; MIKE THORPE;

DAVID BABB; KELVIN PHILLIPS; RICHARD WORLEY;

PAUL AVERETTE; TRANSYLVANIA COUNTY BOARD OF

EDUCATION; TRANSYLVANIA COUNTY,

Defendants - Appellees.

Appeal from the United States District Court for the Western District of North Carolina, at Asheville. Lacy H. Thornburg, District

Judge. (CA-98-117-1)

Submitted: March 11, 1999 Decided: March 16, 1999

Before WIDENER and LUTTIG, Circuit Judges, and PHILLIPS, Senior

Circuit Judge.

Affirmed by unpublished per curiam opinion.

Laurie B. Foote, Ira T. Foote, Jr., Appellants Pro Se. Ronald G.

Blanchard, Hendersonville, North Carolina; Cynthia Stubbs Lopez,

ROBERTS & STEVENS, P.A., Asheville, North Carolina; Grady Michael

Barnhill, Patrick T. Gillen, WOMBLE, CARLYLE, SANDRIDGE & RICE,

Charlotte, North Carolina, for Appellees.
2

Unpublished opinions are not binding precedent in this circuit.

See Local Rule 36(c).

PER CURIAM:

Laurie and Ira Foote appeal the district court’s order dismissing with prejudice their action under 42 U.S.C.A. § 1983; 42

U.S.C.A. § 1985; and 42 U.S.C.A. § 1986 (West Supp. 1998), and

dismissing without prejudice any state supplemental claims. We

have reviewed the record and the district court’s opinion accepting

the recommendation of the magistrate judge and find no reversible

error. Accordingly, we affirm on the reasoning of the district

court. See Foote v. Barton, No. CA-98-117-1 (W.D.N.C. Nov. 2,

1998). In addition, we have reviewed the record and find the

pleadings do not set forth a claim for denial of equal protection.

See Sylvia Dev. Corp. v. Calvert County, Md., 48 F.3d 810, 819 (4th

Cir. 1995). We dispense with oral argument because the facts and

legal contentions are adequately presented in the materials before

the court and argument would not aid the decisional process.

AFFIRMED
HomeNRC LibraryDocument CollectionsNews Releases1998

# Note to Editors: Beaver Valley Meeting on November 16, 1998

----

----

I-98-123

November 10, 1998

**NOTE TO EDITORS:**

The Nuclear Regulatory Commission staff will meet with officials of Duquesne Light Company on Monday morning, November 16, to discuss the preliminary results of the NRC's Operational Safety Team Inspection at the Beaver Valley nuclear power station in Shippingport, Pa. The inspection was conducted October 26-29.

An Operational Safety Team Inspection is conducted to verify that the plant is being operated safely and in conformance with regulatory requirements, and that the organizations that control and support plant operations are functioning effectively to ensure operational safety.

Following the NRC's presentation, Duquesne Light will discuss Unit Restart Plan Performance.

The meeting, which begins at 10:30, will be held at Beaver Valley's Emergency Response Facility (conference room A) at the site. It is open to the public for observation.

Page Last Reviewed/Updated Thursday, March 25, 2021