ML20206C833

Summary of 980929 Meeting with Texas Utilities Electric Co Re Licensing Submittals Requesting Exemption to App K & Topical Rept Submitted in Support of Exemption Request .List of Attendees Encl

ML20206C833
Person / Time
Site:	Comanche Peak
Issue date:	04/26/1999
From:	Jaffe D NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
NUDOCS 9905030211
Download: ML20206C833 (19)

Text

.b als g k UNITED STATES s" j t

NUCLEAR REGULATORY. COMMISSION WASHINGTON. D.C. 30666 0001

\*****/ -- April 26, 1999 LICENSEE: Texas Utilities Electric Company.

FACILITY: Comanche Peak Steam Electric Station, Units 1 and 2

SUMMARY

OF MEETING WITH TEXAS UTILITIES ELECTRIC COMPANY.

REGARDING APPENDIX K EXEMPTION REQUEST AND TOPICAL REPORT THAT SUPPORTS THE EXEMPTION REQUEST On September 29,1998, members of the Nuclear Regulatory Commission (NRC) staff met with representatives of Texas Utilities Electric Company (TUE/ licensee) to discuss the licensing

. submittals requesting an exemption to Appendix K dated August 13,1998, and the topical report submitted in support of the exemption request dated July 17,1998.- Attachment 1 is the list of meeting attendees.

The NRC staff provided an introduction to the meeting and stated that this was a continuation of the dialogue that was started at the July 14,1998, meeting between the NRC and the licensee.

Since that meeting, the licensee has submitted the topical report on the Comanche Peak

- Steam Electric Station docket by letter dated July 17,1998, and submitted a request for exemption to portions of 10 CFR Part 50, Appendix K dated August 13,1998.

Prior to this meeting a phone conversation was held on September 23,1998, to ask the licenese questions that have arisen so far in the NRC staff's review. The purpose of the call was to focus the meeting on those aspects of the review that were not discussed in the submittal or to clarify the wording in the submittal. This meeting was conducted in two parts.

The first part covered the overall discussions and the non-proprietary answers to the questions the staff asked. The second portion of the meeting was closed since proprietary information contained in the topical report was discussed. The non-proprietary portion of the handout material is included as Attachment 2. Subsequent to the production of the handout material the licensee considered portions of the answer to Question 1 to be proprietary and the information was removed from the non-proprietary handout except for the list of quastions. The answer was discussed during the proprietary portion of the meeting along with the remaining questions.

The NRC staff later determined that the answer to Question 1 was non-proprietary and it is,

- therefore, included in Attachment 2.

The total number of questions answered in the handout material and discussed at the meeting was 30. Some responses were found to be adequate at the meeting. The other responses will require further review of the handout material and further clarification. The NRC staff will review the material and prepare a formal request for additional information (RAl) based on the lj further review. The licensee will then respond to the RAI and either supplement the exemption /

. request or revise the topical report as appropriate. 3

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The next step in the review process was discussed. The NRC staff stated that the technical review would continue with TUE as the pilot plant with a target date of December 1998 to

[OI complete a review of the topical report and the exemption request. The NRC staff stated that

- phone calls and meetings would be used to facilitate communication and to keep the review 9905030211 990426  ?

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2-schedule on track. The licensee plans to submit a power uprate license amendment request (LAR) in December 1998. The licensee acknowledged that an LAR may need to be supplemented depending on the final outcome of the exemption request.

The staff and the licensee acknowledged the benefit of the, meeting and an understanding of the schedule for the licensee's submittals and staff's review effort to approve such submittals.

L David H. Ja , Senior ject Manager, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-445 and 50-446 Attachments: 1. Meeting Attendees

2. Licensee Presentation (Non-proprietary) cc w/atts: See next page m

. '. April 26, 1999 2-track. The licensee plans to submit a power uprate license amendment request (LAR) in December 1998. The licensee acknowledged that an LAR may need to be supplemented depending on the final outcome of the exemption request.

The staff and the licensee acknowledged the benefit of the meeting and an understanding of l the schedule for the licensee's submittals and staff's review effort to approve such submittals.

ORIG. SIENED BY David H. Jaffe, Senior Project Manager, Section 1 Prriset Directorate IV & Decommissioning D% ion of Licensing Project Management Othee of Nuclear Reactor Regulation Docket Nos. 50-445 and 50-446 Attachments: 1. Meeting Attendees

2. Licensee Presentation (Non-proprietary) cc w/atts: See next page '

DISTRIBUTION:

HARD COPY:

'Dochet Filefg PUBLIC PD 4-1 r/f T. Polich OGC ACRS E-MAIL: I S. Collins /R.Zimmerman (SJC1/RPZ)

J. Zwolinski(JAZ)

S. Black (SCB)

S. Richards (SAR)

L. Berry (LGB)  :

T. Martin (SLM3) .

C. Doutt (CKD)

R. Caruso (RXC)

E. Wang (EYW)

N. Lauben (GNL1)

D. Lange (DJL)

K. Brockman, RIV Document Name: G:\CPFINAL\ CPS 92998.WPD *See previous concurrence OFC F)EWfh?M PDIV-1/LA EELB' SRXB* PDIM-j/SC NAME bbJIBer MbCalvo JWermlel RGNm DATE kf)b //B/99 03/26/99 03/29/99 $AN99 COPY YES/NO b NO

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YES/NO YES/NO YES/NO OFFICIAL RECORD COPY OjG030

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. track. The licensee plans to submit a power uprate license amendment request (LAR) in December 1998. The licensee acknowledged that a LAR may nead to be supplemented depending on the final outcome of the exemption request.

The staff and the licensee acknowledged the benefit of the meeting and an understanding of the schedule for the licensee's submittals and staff's review effort to approve such submittals.

' Timothy J.. Polich, Project Manager Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation l

l Attachments: 1. Meeting Attendees

2. Licensee Presentation (Non-Proprietary)

Docket Nos. 50-445 and 50-446 l

cc w/atts: See next page j HARD COPY:

l Docket File PUBLIC 1 PD 4-1 r/f T. Polich

! OGC ACRS E-MAIL:

S. Collins /R.Zimmerman (SJC1/RPZ)

J. Zwolinski(JAZ)

I S. Black (SCE) l S. Richt rds (SAR) ,

L. Berry (LGB)

T. Martin (SLM3)

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C. Doutt (CKD)  !

R. Caruso (RXC)

E. Wang (EYW)

N. Lauben (GNL1)

D. Lange (DJL) >

l K. Brockman, RIV /

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Document Name: G:\CPFINAL\ CPS 92998.WPD .

OFC PM/PD4 LA/PD4 EELB SRXI6 // PD4 NAME TPolich M LBerry ' Jealv JWe'rm SRichards  ;

l DATE 3 //6/99 / /99 J #I/99 3/h99 / /99 I

COPY YES/NO YES/NO YES/NO YES/NO YES/NO OFFICIAL RECORD COPY gg

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! TU Electric Company Comanche Peak, Units 1 and 2 l

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Senior Resident inspector Honorable Dale McPherson U.S. Nuclear Regulatory Commission County Judge P. O. Box 2159 P. O Box 851 Glen Rose, TX 76403-2159 Glen Rose,TX 76043 i

Regional Administrator, Region IV Office of the Governor .

U.S. Nuclear Regulatory Commission ' ATTN: John Howard, Director 611 Ryan Plaza Drive, Suite 400 Environmental and Natural Arlington,TX 76011 Resources Policy P. O. Box 12428 Mrs. JJanita Ellis, President Austin, TX 78711 Citizers Association for Sound Energy 1426 South Polk Arthur C. Tate, Director Dallas,TX 75224 Division of Compliance & Inspection Bureau of Radiation Control Mr. Roger D. Walker Texas Department of Health TU Electric 1100 West 49th Street Regulatory Affairs Manager Austin, TX 78756-3189 P. O. Box 1002 Glen Rose,TX 76043 Mr. C. Lance Terry i TU Electric George L. Edgar, Esq. Group Vice President Nuclear Morgan, Lewis & Bockius Attn: Regulatory Affairs Department 1800 M Street, N.W. P. O. Box 1002 Washington, DC 20036-5869 Glen Rose, TX 76043 Jim Calloway Public Utility Commission of Texas Electric Industry Analysis P. O. Box 13326 Austin, TX 78711-3326 l

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TEXAS UTILITIES ELECTRIC- NRR MEETING REGARDING l I

COMANCHE PEAK STEAM ELECTRIC STATION UNIT 2 EXEMPTION FROM APPENDIX K AND TOPICAL REPORT TO SUPPORT THE EXEMPTION HELD ON SEPTEMBER 29.1998 ,

I MEETING ATTENDEES l Name Oraanization Phone

~ C. Doutt NRC/NRR/HICB 301-415-2847 l T. Polich NRC/NRR/PD4-1 301-415-1038 J. Donoghue NCR/NRR/SRXB 301-415-1131 R. Caruso NCR/NRR/SRXB 301-415-1813 E. Wang NCR/NRR/PGEB 301-415-1076 N. Lauben* NCR/RES/RPSB 301-415-6762 l R. Walker TUE 254-897-8233 F. Maddy TUE 254-897-8551 i W. Choe TUE 214-812-4371 C. Hastings Caldon 412-341-9920 l E. Hauser Caldon 412-341-9920 H. Estrada Caldon 410-849-3324 .

J, Regan Key Tech /Caldon 410-385-0200 H. Fontecilla APS/ Virginia Power 703-838-2314 M. Philips Winston and Strawn 202-371-5729 W. Horn Winston and Strawn 202-371-5729 C. Brinkman" ABB-CE 301-881-7040 l

• Only present for a portion of the proprietary part of the meeting

" Only present for the non-proprietary part of the meeting  !

ATTACHMENT 1

Responses to NRC Questions: September 29,1998 Non Proprietary NRC Staff Questions

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1. Describe Caldon's understanding of the background for 1.02 being ascribed just for.

instrument uncertainty in power determination

2. On page 5-2 of the Topical Report, explain the justification for the use of PTC-6.

3. Describe how the LEFM/ is used in calorimetric power determinations.

5. Who is responsible and how are Calibration, Maintenance, and Training performed and achieved? ,

6. How will monitoring, verification, and error reponing be handled? l

9. Clarify that the 0.5%'used in the Topical Report is 95% confidence level (20).

12. Does cross flow = transverse velocity?

29. How is the LEFM used currently to provide correction factors to the venturis? Is the l correction determined on the basis of the absolute accuracy or the repeatability of the l LEFM?

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Responses to NRC Questions: September 29.1998

- Question 1:

Describe Caldon's understanding of the background for 1.02 being ascribedjust for instrument uncertainty in power determination.

Answer:

Caldon proposes to increase current licensed power by 1% for plants using Appendix K evaluation models without any requirement to reanalyze ECCS performance if the plants utilize a new technology for determining thermal power. The new technology provides on-line verificction ofinstrument accuracy and is capable of accuracies sufficient to ensure that there is a higher level of certainty that 1.02% of the current licensed power will not be exceeded than is currently being provided, and, hence, a higher level of certainty that the criteria of 10 CFR 50.46 (b) will not be exceeded.

As set forth in the Opinion of the Commission in the ECCS rulemaking proceeding, RM-50-1. December 28,1973,Section I.A. of Part 50, Appendix K, specifies the following for the initial conditions to be used in Appendix K evaluation models:

For the heat sources...it shall be assumed that the reactor has been operating continuously at a power level at least 1.02 times the licensed power level (to allow for such uncertainties as instrumentation error), with the maximum peaking f factor allowed by the technical specifications. A range of power distribution

\

shapes and peaking factors representing power distributions that may occur over the core lifetime shall be studied and the one selected should be that which results I in the most severe calculated consequences for the spectrum of postulated breaks and single failures analyzed.

, The question has been raised as to what the phrase "such uncertainties as" implies and whether operating at a power level 1% higher than the current licensed power could have any effect on the continued validity of the current Appendix K analyses of ECCS  !

performance. A review of the regulatory history, including the interim acceptance criteria, 4 the record of the ECCS rulemaking proceeding, implementation of Appendix K, and operating experience, has turned up nothing that would suggest that anything other than the need to account for the uncertainty of determining the thermal power at which the reactor is operating led to the adoption of 1.02 times the maximum licensed power for an initialcondition for Appendix K ECCS evaluations.

The NRC staff, in its concluding statement, had recommended essentially the same initial conditions as were adopted by the Commission. However, the staff had recommended "that the reactor shall be assumed to have been operating continuously at a power level no lower than 1.02 times maximum licensed power level (to allow for instrument error) ..."

Concluding Statement of Position of the Regulatory Staff, RM-50-1, at 40,109.

Regulatory Guide 1.49, Rev.1, issued in December 1973, concurrent with the issue of the f( Commission's opinion in RM-50-1, recommended the use of an assumed power level 1.02 1

B Responses to NRC Questions: September 29,1998 times the proposed licensed power for analyses and evaluations of all normal, transient and

( accident conditions necessary to evaluate the adequacy of the facility. The staff explained the purpose of using 1.02 times the proposed licensed power as follows:

... analyses in support of the proposed licensed power are made for a slightly higher power to allow for possible instrument errors in determining the power level. The regulatory staff has concluded that a margin of 2% is adequate for this purpose.

Nowhere in the Commission's opinion is any reason given for the addition of the words "such uncertainties as."

The record shows that the initial conditions of Appendix K had their inception in the ECCS evaluation models approved with the issuance of the interim acceptance criteria. 36 Fed. Reg.12247, June 1971. See for example WCAP-7422L, Westinghouse PWR Core Behavior Following a Loss-of-Coolant Accident. Id at Appendix A, Part 3. Here,1.02 times the licensed power was used to account for uncertainties in determining the operating power level and the worst possible power distributions and maximum peaking factors were determined. The maximum peaking factors for which ECCS performance were found to be acceptable were placed in technical specifications for the plants.

The manner in which the initial conditions have been applied since the promulgation of Appendix K is also instructive. Initially, when Appendix K was issued, reactors then operating under the interim acceptance criteria were required to perform ECCS evaluations using approved Appendix K evaluation models. Since the worst possible power distributions and maximum peaking factors allowed by the then current technical specifications were used, unless, of course, the Appendix K evaluations supported different values. (As specified in 10 CFR 50.36, the technical specifications must be e derived from the analyses and evaluations in the safety analyses reports and amendments y

thereto. Hence, the maximum peaking factors allowed by plant technical specifications must be consistent with the nuximum peaking factors used in the ECCS evaluations which demonstrate that the ECCS performance satisfies the criteria of 10 CFR 50.46(b).

In general, in applying initial conditions for the interim acceptance criteria and those required by Appendix K,Section I.A., the following were incorporated in approved ECCS evaluation models:

1. Reactor power assumed to have been continuously at 1.02 times the licensed power (or proposed licensed power in case of amendments) to account for uncertainties in determining thermal power.

2. A cosine curve representing the power distribution shape resulting in the worst consequences (normally the highest calculated peak clad temperatures).

The cosine curve is worse than any other power shape occurring at any time in core life. This shape can only occur during a return to power and for a l

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Responses to NRC Questions: September 29.1998 short time thereafter and is not possible in continuous power operation.

Hence it is an extremely conservative assumption; and

3. The maximum peaking factor in the technical specifications (or proposed to be placed in the technical specifications);

These three assumptions have then been used in the Appendix K evaluation models as inputs to calculate the initial stored energy in the fuel, fission heat, and decay heat from actinides and fission products. None of these input assumptions are affected by increasing the operating power provided that the operating power does not exceed the 1.02 times maximum licensed power assumed in the evaluation of the ECCS performance showing that the criteria of 10 CFR 50.46(b) will not be exceeded. This is consistent with the objective stated in the opinion of the Commission in RM-50-46 dealing with conservatism:

1 (1) Stored Heat. The assumption of 102% of maximum power, highest allowed peaking factor, and the highest estimated thermal resistance between the UO 2and the cladding [ calculated using the above input assumptions]

provides a calculated stored heat that is possible but unlikely to occur at the time of the hypothetical accident... Opinion of the Commission, RM-50-1, December 28,1973 at 27, A-4.

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Thus, it appears that the approval of the Caldon proposal resides solely in demonstrating that there is a sufficiently high probability that the power level assumed for evaluation of

( ECCS performance in existing Appendix K evaluations will not be exceeded in operation at a 1% increase in licensed power level with the proposed improved technology for determining plant operating power.

A review of the Standard Review Plan (SRP) was also conducted to identify references to e application of the 2% margin to initial conditions for accidents. The results, summarized j in Attachment 1, indicate that the 2% margin for initial conditions is required for analysis of 12 accidents in SRP Chapter 15. Of these 12 accidents, the license is permitted to use less than 2% margin in 9 cases provided the lower margin can be justified by the applicant.

Attachments: l

1. Summary Table of Review of Chapter 15 of Standard Review Plan. I

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Responses to NRC Qusstions: September 29.1998 Attachment 1 to Question 1 Section of Chapter 15 Reference to 102% initial Permission to use power level condition below 102% if justified 15.1 1 through 4: Decrease in Yes Yes Feed temp etc.

15.1.5 Steam sys piping failures No in and out of containment 15.1.5 A Rad Consequences No 15.2.1-5 Loss of Load etc Yes No 15.2.6 Loss of Emergency AC to Yes Yes Station Auxiliaries 15.2.7 Loss of Normal Feed Flow Yes Yes l

15.2.8 Feed System Pipe Breaks No '

-PWR 15.3.12 Loss of forced reactor Yes Yes ,

coolant Flow 15.3.3 4 Reactor coolant pump Yes Yes l

, j motor seizure 15.4.1 Uncontrolled Control rod No withdrawal- subcritical/ low ,

power j 15.4.2 Uncontrolled Withdrawal No at Power

( 15.4.3 Control Rod Malfunction Yes Yes 15.4.4 5 Startup ofinactive loop Yes No and flow controller malfunction-BWR 15.4.6 CVCS reduces boron Yes No concentration PVQt y 15.4.7 Inadvertert Loading of No fuel assembly 15.4.8 Spectrum of rod ejection No accidents - PWR 15.4.8 A Rad Consequences of No Ejection 15.4.9 Spectrum of rod drop No accidents - BWR 15.4.9 A Rad consequences No 15.5.1-2 Inadvertent Operation of Yes Yes ECCS that increases inventory 15.6.1 Inadvertent opening of Yes Yes PWR or BWR pressure relief valve i 15.6.2 Rad consequences of the No j failure of smalllines carrying i l

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Responses to NRC Questions: September 29.1998 Attachment I to Question 1

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Section of Chapter 15 Reference to 102% initial Permission to use powerlevel condition below 102% if justified primary coolant outside containment 15.6.3 Rad consequences of No steam generator tube rupture -

PWR 15.6.4 Rad Consequences of No Main Steam Line Failure Outside containment- BWR 15.6.5 LOCA resulting from Yes Yes spectrum of pipe breaks within RC pressure bounda.y 15.6.5 Apps A, B, C, D: Various No rad consequences of LOCA 15.7.3 Rad releases due to liquid No containing tank failures 15.7.4 Rad consequences of fuel No handling accidents 15.7.5 Spent fuel cask drop No accidents l

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Responses to NRC Questions: September 29,1998 Question 2:

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On page 5-2 of the Topical Report, explain thejustification for the use of PTC-6.

Answer:

The context of the reference to PTC-6 is repeated here from page 5-2:

"Immediately after it is calibrated, a flow nozzle is capable of providing measurement accuracies in the i0,5% range, providing the differential pressure and fluid temperature measurements are made with laboratory grade, calibrated instruments (see for example the discussion of turbine heat rate testing in ASME-PTC-6, Reference 9)."

PTC-6 is referred to for purposes ofillustration only and does not apply to the use of the LEFM/ for thermal power measurement.

Attachments:

None.

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Responses t2 NRC Questions: September 29,1998 Question 3:

' Describe how the LEFM/ is used in calorimetric power determinations.

Answer:

The proposed use of the LEFM/ is for direct measurement of feedwater mass flow and temperature, and indirect measurement of feedwater enthalpy, for the thermal power determination. This determination would be used directly to calibrate the nuclear instruments in lieu of the existing instrumentation. At the discretion of the licensee, the LEFM/ may also be used for calorimetric calculation of reactor coolant flow, and for setting non-safety-related setpoints for which thermal power is an input. The increased accuracy as compared to the existing instrumentation would be beneficial in these applications.

At Comanche Peak, the LEFM is currently used for the secondary calorimetric calculation only. The secondary calorimetric is used as input for the daily calibration of NIS and the cross-correlation N16 system.

In some plants, feedwater flow and/or temperature instruments are used as direct inputs to the reactor protection system or another automatic safety function. In these cases, those instruments are classified as safety-related, and would continue to be used for these functions. The LEFM/ is not being proposed for these functions. Its use would be

-( limited to power determination and the non-safety-related uses of calorimetric power discussed above.

Attachments:

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Responses to NRC Questions: September 29,1998 Question 5:

Who is responsible and how are Calibration, Maintenance, and Training performed and I achieved?

i Answer:

Calibration and Maintenance Calibration and maintenance is performed by I&C using site procedures. The site procedures are developed using the CALDON technical manuals. All work is performed in accordance with site work control procedures.

Routine preventive maintenance procedures include physical inspections, power supply checks, back-up battery replacements, and internal oscillator frequency verification.

Ultrasonic signal verification and alignment procedures which involve digital oscilliscopes with the MFM will be replaced by automatic set-up in the EFM/. Signal verification will still be possible by review of signal quality measurements performed and displayed by the GFM/.

Training

( I&C personnel must be qualified per the I&C training program on the EFM system  ;

before work or calibration may be performed. Formal training from Caldon was provided l to site personnel. Formal training on the LEFM/ system will be provided by Caldon. j l

Attachments:

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Responses to NRC Questions: September 29.1998 Question 6: 1

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How will monitoring, verification, and error reponing be handled?

Answer:

Though this application is not safety-related, the GFM/ system is designed and manufactured under Caldon's Quality Control Program , which provides for configuration control, deficiency reponing and correction, and maintenance. Specific examples of quality measures undenaken in the design, fabrication and testing of the GFM/ systeni are provided in the Topical Repon, Section 6.4 and Table 6.1. Table 6.1 lists the error bounding, validation and verification procedures planned for the LEFM/

system.

At Comanche Peak, the GFM system is included in the System Health Plan and the preventative maintenance program. The system is monitored by the System Engineer for reliability. As a plant system, all equipment problems fall under the site work control process. All conditions that are adverse to quality are documented under the ONE/ SMART form program. The software falls under TU Electric's Appendix D QA program with a software QA plan in place. The current software was verified and validated and is under Caldon's Verification and Validation Program. Caldon's Verification and Validation Program provides procedures for deficiency reponing for engineering action and notification of holders of V&V software.

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The Comanche Peak EFM/ System will likewise be under Caldon's V&V Program, and procedures will be maintained for user notification ofimportant deficiencies.

Attachments:

None.

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Responses to NRC Questions: September 29,1998 Question 9:

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Clarify that the 0.5% used in the Topical Report is 95% confidence level (2a).

Answer:

The 0.5% mass flow uncertainty stated for the chordal LEFM and the LEFM/ is a 2 standard deviation (20) uncertainty; that is, it represents a 95% confidence interval. This is intended to be a bounding approximation. This subject is discussed further in response to Question 13.

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Responses to NRC Questions: September 29,1998 l Question 12:

Does cross flow = transverse velocity?

Answer:

Yes. For the purposes of this report, the terms are used interchangeably.

Attachments:

None.

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29,1998 Responses to NRC Questions: E-T Question 29:

How is the GFM used curmntly to provide correction factors to the venturis? Is the correction determined on the basis of the absolute accuracy or the repeatability of the EFM?

Answer:

The GFM is used at Comanche Peak to directl'y calibrate the nuclear instrumentation.

The correction factor is used only to keep the venturi c&libration contemporary for use in the event that the LEFM is unavailable.

A correction factor is calculated in accordance with a plant procedure which has its methodology based on approved calculation. A minimum of 50 separate two hour data sets of FW mass flow rate from the LEFM and venturis are recorded in a spreadsheet.

The percent difference of each data set, the average percent difference of all data sets, and the standard deviation of all data sets is calculated. The correction factor is calculated from the average percent difference plus the two standard deviation mvgin.

The spreadsheet calculation is independently reviewed, documented by a TE, and given to the System Engineering Computer Gmup to implement in the appropriate plant computer software under an approved change process.

The Plant Computer multiplies the feedwater flow rate as determined by the venturis by g the correction factor. This correction is displayed on the plant computer as " NET LEFM

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CORRECTED POWER" and is available for use when the LEFM is out of service. The LEFM is used directly when it is in service. The 2 standard deviation margin used in the correction factor calculation prevents this corrected MWth from being equal to the MWth calorimetric power determined directly from the LEFM.

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The correction is based on the absolute accuracy of the LEFM but a high degree of repeatability is also required.

Attachments:

None.

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