ML031050221

Request for License Amendment Measurement Uncertainty Recapture Power Uprate, Enclosures 7

ML031050221
Person / Time
Site:	Hatch
Issue date:	12/19/2002
From:	Sumner H Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
HL-6328
Download: ML031050221 (133)

Text

Enclosure 7 Edwin I. Hatch Nuclear Plant Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications Introduction The Edwin I. Hatch Nuclear Plant is currently licensed to operate at a maximum rated thermal power (RTP) of 2763 MWt. This power level is supported by a number of analyses and evaluations performed with an RTP uncertainty of 2 2%, either through 10 CFR 50, Appendix K, "ECCS Evaluations Models" or Regulatory Guide 1.49. By applying a reduced thermal power uncertainty to these analyses, Plant Hatch can justify increasing the reactor thermal power level and still remain within the boundary of these specific analyses. Southern Nuclear Operating Company (SNC) is requesting approval to increase the licensed RTP by 1.5% to 2804 MWt.

This power increase will be accomplished by using a more accurate main feedwater flow measurement system to provide input into the calculated core thermal power (CTP) of each unit.

The 1.5% uprate is based upon reducing margin that is assumed in analyses to account for the measurement uncertainties associated with calculating the CTP of each unit. Plant Hatch's current accident and transient analyses include a minimum 2% margin on RTP to account for power measurement uncertainty. This power measurement uncertainty was originally required by 10 CFR 50, Appendix K, which required a 2% margin between the licensed power level and the power level assumed for the emergency core cooling system (ECCS) evaluations. In 2000, the NRC amended 10 CFR 50, Appendix K, to provide licensees the option of maintaining the 2% power margin or applying a reduced margin. If the licensee elects to apply a reduced margin, the new assumed power level has to account for measurement uncertainties in the instrumentation used in the core thermal power computation. The revised Appendix K rule has an effective date of July 31, 2000.

The feedwater flow measurement uncertainty is the most significant contributor to CTP measurement uncertainty. Based upon this fact and on the Appendix K rule change, SNC proposes a reduced power measurement uncertainty of 0.5% and an increase in RTP of 1.5%.

To accomplish the reduction in uncertainty and the increase in power, SNC will install a CrossflowTm system on both units.

The CrossflowTm system provides a more accurate measurement of feedwater flow than the currently installed instrumentation used for the CTP calculation at Plant Hatch. Combustion Engineering Topical Report CENPD-397-P-A, Rev. 01 'Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology,"'() documents the theory, design, and operating features of the CrossflowTm system and its ability to achieve increased flow measurement accuracy. In a safety evaluation dated March 20, 2000, the NRC approved CENPD-397-P-A, Rev. 01(1 for referencing in license applications for power uprates. The CrossflowTm system will provide a measured feedwater mass flow to within an assumed 0.42%

for both Hatch units. Tables E7-2 and E7-4 provide the total RTP uncertainty (95/95 power measurement uncertainty) for Units 1 and 2, respectively. The bounding resultant RTP HL-6328 E7-1 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications uncertainty is for Unit 2, which is - 0.461% RTP. Therefore, the total power measurement uncertainty required by Appendix K of 0.5% for each unit is justified. This value (0.5%) was used in the safety analyses supporting this license amendment request. The reduced power measurement uncertainty alleviates the need for the 2% power margin originally required by Appendix K, thereby allowing an increase in the RTP available for electrical generation.

Discussion This enclosure addresses the selected guidance items specified in Attachment 1 of NRC Regulatory Issue Summary (RIS) 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications,"(2) specifically to items 1,VII.2, VII.3, and V1I.4. The specific RIS 2002-03 guidance items are in bold and the applicable Plant Hatch information to follow:

L Feedwater flow measurement technique and power measurement uncertainty

1. A detailed descriptionof the plant-specificimplementation of the feedwater flow measurement technique and the power increase gained as a result of implementing this technique. This descriptionshouldinclude:

A. Identification(by document title, number, and date) of the approved topicalreport on the feedwater flow measurement technique The feedwater flow measurement system being installed at Plant Hatch Units 1 and 2 is the AMAGl~estinghouse CROSSFLOW ultrasonic flow measurement (UFM) system. The design of this advanced flow measurement system is addressed in detail by the manufacturer in topical report CENPD-397-P-A, Rev. 01.( 1)

The Unit 1 system consists of three flow-measurement devices installed on the "A' feedwater pipe, and a single device on the UB" feedwater pipe. The Unit 2 system (to be installed in the 2003 spring refueling outage) will consist of one device installed on each feedwater pipe. Each flow measurement device is composed of eight non-intrusive flow transducers (four primary/four redundant) strapped on the feedwater piping and a set of temperature sensing transducers connected via coax cable to a remote panel that acts as a data acquisition system and viewing station. An AMAG CORRTEMP ultrasonic temperature measurement system used to correct the current RTD-based feedwater temperature is included in each unit system.

HL-6328 E7-2 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications B. A reference to the NRCS approval of the proposed feedwater flow measurement technique NRC approval of the proposed feedwater flow measurement technique was granted via NRC letter dated, March 20, 2000, "Acceptance for Referencing of CENPD-397-P, Revision-01 -P, 'Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology" (TAC NO. MA5452).

C. A discussion of the plant-specific implementation of the guidelines in the topical report and the staff's letter/safety evaluation approving the topical report for the feedwater flow measurement technique The Crossflow UFM will be installed in accordance with the requirements of CENPD-397-P-A, Rev. 01l), Section 8.0, Crossflow Field Implementation.

This system will be used for fulltime online CTP determination. The system will be provided with a software layer that will integrate the systems to the plant process computer to correct the current venturi-based flow and current RTD-based temperature.

D. The dispositions of the criteria that the NRC staff stated should be addressed (i.e., the criteria included in the staff's approval of the technique) when implementing the feedwater flow measurement technique In approving Topical Report CENDP-397-P-A, Rev. OV), the NRC established four criteria to be addressed by each licensee in requesting a measurement uncertainty recapture power uprate license amendment. The four criteria and a discussion of how each will be satisfied for Plant Hatch follow:

Criterion 1 The licensee should discuss the development of maintenance and calibration procedures that will be implemented with the Crossflow UFM installation. These procedures should include process and contingencies for an inoperable Crossflow UFM and the effect on thermal power measurement and plant operation.

Implementation of the power uprate license amendment will include developing the necessary procedures and documents required for operation, maintenance, calibration, testing, and training at the uprated power level with the new Crossflow UFM system. Plant maintenance and calibration procedures will be revised to incorporate the Crossflow maintenance and calibration requirements prior to declaring the system operational and prior to increasing power above the current license thermal power level.

HL-6328 E7-3 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications Plant operation with the Crossf low UFM system out of service is discussed in Sections G and H below.

Criterion 2 For plants that currently have the Crossflow UFM installed, the licensee should provide an evaluation of the operational and maintenance history of the installed UFM and confirm that the instrumentation is representative of the Crossflow UFM and is bounded by the requirements set forth in Topical Report CENPD-397-P.

This criterion is not applicable to Plant Hatch Units 1 and 2. Plant Hatch currently uses flow venturies for the feedwater flow measurement contribution to the CTP computation. The installation and operation of the Crossf low system is in anticipation of approval of the proposed amendment. Installation of the systems will be completed prior to implementation of the requested license amendment.

Criterion 3 The licensee should confirm that the methodology used to calculate the uncertainty of the Crossflow UFM in comparison to the current feedwater flow instrumentation is based on accepted plant setpoint methodology (with regard to the development of instrument uncertainty). If an alternative methodology is used, the application should be justified and applied to both the venturi and the Crossflow UFM for comparison.

The Plant Hatch heat balance uncertainty study(3) was performed using ISA-RP67.04, Part 11-1994, "Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation"(4) as the instrument uncertainty determination methodology. This study provides the total CTP uncertainty by evaluating the various contributions of the variables from the measured instrumentation to the CTP. A baseline condition was established based upon heat balance process conditions. Each measurement variable was then varied independently by a nominal error to determine the sensitivity (weighting factor) of that error on the output CTP calculation.

The analysis provides a total measurement uncertainty for CTP assuming a Crossflow total uncertainty of 0.42% of actual flow. The calculation provides all weighting factors and other parameter uncertainties, so that computation of the final uncertainty, based upon the final vendor specification, could be easily determined.

For Plant Hatch, the primary feedwater flow measurement will be made with the flow venturies. The venturi measurement will be corrected on a very frequent basis automatically within the plant computer. Preliminary information from the UFM vendor, based upon current testing and HL-6328 E7-4 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications configuration, shows the maximum total feedwater flow error from the UFMs to be +/- 0.42% actual flow.

To provide a clear understanding of the uncertainties of the CTP computation, using the current configuration, as compared to the UFM corrected (after uprate) configuration, Tables E7-1 through E7-4 were created. Tables E7-1 and E7-2 address Unit 1, and Tables E7-3 and E7-4 address Unit 2. The tables provide a summary of the CTP input parameter uncertainties, the associated sensitivity, and the contributions of uncertainty to the CTP computation uncertainty.

Criterion 4 The licensee of a plant at which the installed Crossflow UFM was not calibrated to a site-specific piping configuration (Flow profile and meter factors not representative of the plant-specific installation) should submit additional justification. This justification should show that the meter installation is either independent of the plant-specific flow profile for the stated accuracy, or that the installation can be shown to be equivalent to known calibration and plant configurations for the specific installation, including the propagation of flow profile effects at higher Reynolds numbers. Additionally, for previously installed and calibrated Crossflow UFM, the licensee should confirm that the plant-specific installation follows the guidelines in the Crossflow UFM topical report.

Both Unit 1 and Unit 2 calibrations for the UFMs will be performed with a mockup laboratory installation, modeling the actual to-be-installed plant configuration. Therefore, the laboratory setup inherently models this configuration. Thus, the flow profile and meter factors are representative of the plant-specific installation, and no additional justification is required.

E. A calculation of the total power measurement uncertainty at the plant, explicitly identifying all parameters and their individual contribution to the power uncertainty Reference 3 (Attachment 1 to Enclosure 7) is an instrument uncertainty calculation and study that fully defines all parameters which contribute to the total power measurement uncertainty. It determines sensitivities of the uncertainty associated with the measurement of each parameter to the uncertainty of the total power computation, as well as the total power measurement uncertainty for various configurations, including the use of UFMs to correct the venturi measurement and current configuration. This calculation was performed in accordance with Reference 4.

F. Information to specifically address the following aspects of the calibration and maintenance procedures related to all instruments that affect the power calorimetric.

HL-6328 E7-5 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications

i. maintaining calibration Calibration and maintenance will be performed using site procedures developed from the Crossflow system technical and O&M manuals. All work will be performed in accordance with site work control procedures.

Verification of acceptable Crossf low system operation will be provided by local onboard system diagnostics.

Calibration of other instrumentation that contributes to the power calorimetric computation is performed periodically, with appropriate precision. M&TE, setting tolerances, calibration frequencies, and instrumentation accuracy were evaluated and accounted for within the uncertainty determination of Reference 1.

ii. controllingsoftware and hardware configuration Software and hardware configuration for the Crossflow system and all other instrumentation that affect the power calorimetric are controlled through the plant modification process, which will require evaluation of any changes necessary to Reference 3. Changes to software and/or hardware are evaluated through the 10 CFR 50.59 process.

iii. performing corrective actions Corrective actions involving maintenance will be performed by l&C maintenance personnel, qualified in accordance with Hatch I&C Training Program, and formally trained on the Crossflow system.

iv. reporting deficiencies to the manufacturer Reliability of the Crossf low system will be monitored by Plant Hatch Engineering personnel in accordance with the requirements of the Condition Reporting system. Although use of the Crossf low system is non-safety related for this application, the system is designed and manufactured under the vendor's quality control program.

v. receiving and addressing manufacturer deficiencyreports The Crossflow system purchase order includes the requirement that Westinghouse inform SNC of any deficiencies in accordance with maintenance agreement reporting requirements. Disposition of manufacturer deficiency reports are handled through the SNC Condition Reporting system.

G. A proposed allowed outage time for the instrument, along with the technicalbasis for the time selected The proposed allowed outage time for any ultrasonic flow meter is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, to provide sufficient time for troubleshooting, repair, calibration, and return of the system to operation. The instrument uncertainties computed within Reference 3 for the case of UFM corrected venturi flow measurement take credit for the fact that selected feedwater flow venturi uncertainties are compensated for by the UFM correction. These uncertainties include such HL-6328 E7-6 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications items as feedwater flow element errors, venturi fouling, and the following errors for the feedwater flow transmitters:

• time dependent drift,

• measurement and test equipment (M&TE) effect,

• setting tolerance effect, and

• static pressure effect.

Given that the outage time of the UFM begins during operation at or near 100% power, any fouling of the venturi will occur in the original feedwater flow reading and eliminated due to the UFM correction just prior to the instrument outage. M&TE and setting tolerance errors are introduced in the calibration process and do not change between calibrations at a given point on the instrument scale. Static pressure effect is constant for a single point on the transmitter scale at a certain static pressure value. Therefore, with the exception of feedwater flow transmitter drift, all other error terms remain approximately the same during the instrument outage, if process conditions remain approximately the same. Thus, these errors are compensated for by the UFM correction just prior to the outage. The feedwater flow transmitters are Rosemount transmitters, which have a very small instrument drift term (0.25% URL for 6 months). Drift is compensated for just prior to the instrument outage; thus, the only time-dependent drift term is the amount of drift occurring during the time of the instrument outage. The error during a 72-hour interval is considered to be negligible in comparison to other error terms, and is of no significance. Therefore, the 72-hour outage time is considered acceptable.

The outage time is considered applicable only if power is not significantly altered in this time interval. Therefore, if power is changed significantly during the outage time, the actions stated below will be implemented at the time of the change.

H. Proposedactions to reduce power level if the allowed outage time is exceeded, includinga discussion of the technicalbasis for the proposed reducedpower level If the outage time of the UFM is exceeded, the uncertainties of the secondary calorimetric computation will be considered as equal to the uncertainties as shown in Tables E7-1 and E7-3. The negative uncertainties are of concern, since these represent the worst case (reading a lower power level than actual). Therefore, the largest error is for Unit 1, or -1.228% RTP. Since the current licensed power level is based upon an error of 2% RTP, the power level will be reduced to an amount that is 0.5% above the current licensed level, or 2777 MWt for both units. Because the venturi instrumentation should be reading more accurately with the UFM correction just prior to the HL-6328 E7-7 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications instrument outage, and since the power reduction will be very small, leaving the process conditions very close to the pre-outage situation, the venturi readings will not be returned to a correction factor of 1.0, but left with the correction factor used just prior to the instrument outage.

VII. Other

2. A statement confirming that the licensee has identifiedall modifications associated with the proposed power uprate, with respect to the following aspects of plant operations that are necessary to ensure that changes in operator actions do not adversely affect defense in depth or safety margins:

A. emergency or abnormal operating procedures B. controlroom controls, displays(includingthe safety parameter display system) and alarms C. the controlroom plant reference simulator D. the operator trainingprogram SNC identified all significant modifications associated with this power uprate.

These include the following for each unit.

1) Install, test, calibrate, and start up the UFMs.

This modification will also install the necessary hardware and software to provide the UFM input to the plant computer and automatically correct the feedwater flow venturi readings to the UFM readings.

2) Rescale average power range monitors (APRMs) to provide 0 -100% power output for 0 - 2804 MWt input. This modification will also rescale the flow-referenced high power trip function to 0.57W + 56.8% RTP for two-loop operation and 0.57W + 56.8% - 0.57 AW RTP for single-loop operation. This modification will also rescale the Plant Computer to associate 100% RTP with 2804 MWt. (See Enclosure 1.)

3) Provide changes to the plant simulator, to provide the UFM readings on the plant computer and APRMs, to reflect the correct % RTP value to megawatt thermal correlations and show the corrections to the feedwater flow venturi readings.

4) Provide changes to operator training procedures and guides to reflect required actions during UFM outages, to reflect accuracy of the core thermal power and feedwater flow measurements under the two scenarios.

HL-6328 E7-8 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications

5) Change operating procedures to reflect the approved allowed outage time for the UFMs, detailing the steps that must be taken to reduce power level, in accordance with items I.1.G and I.1.H above.

6) Other modifications, not addressed by Technical Specifications, required to be performed in support of this project will be performed under the 10 CFR 50.59 evaluation process prior to implementation of the power uprate. (These include, but are not limited to, changes to the flow-referenced settings for rod block within the APRMs.)

3. A statement confirminglicensee intent to complete the modifications identifiedin Item 2 above (includingthe training of operators), prior to implementation of the power uprate.

SNC fully intends to complete the installation of all modifications identified in item 2 above, including the modifications to the simulator, prior to implementation of the proposed power uprate. In addition, SNC will provide operator training on the plant changes and operational aspects of the plant due to this project prior to implementation.

4. A statement confirminglicensee intent to revise existingplant operating procedures related to temporary operation above "fullsteady-state licensed power levels" to reduce the magnitude of the allowed deviation from the licensedpower level. The magnitude shouldbe reduced from the pre-power uprate value of2 percent to a lower value correspondingto the uncertaintyin power level creditedby the proposed power uprate application.

Existing plant operating procedures related to operation above "full steady-state licensed power levels" are based upon the guidance of NRC Memorandum SSINS No. 0200(5), which states: "itis permissible to briefly exceed the "full, steady-state licensedpower level"by as much as 2% for as long as 15 minutes. In no case should 102% power be exceeded, but lesserpower "excursions" for longerperiods should be allowed, with the above as guidance (i. e., 1% excess for 30 minutes, 1/2 for one hour, etc., should be allowed).'

The proposed power uprate of 1.5% is based upon justifications provided in Enclosure 8. NEDC-33056P( 6 ), "Thermal Power Optimization and Power Averaging Guidelines," Revision 1, February 2002 evaluates the effects of a 1.5% uprate on the various aspects of reactor power variation (bi-stable flow phenomena, accidents and transients, core flow measurements, and thermal limits) and concludes that the existing guidelines(5) are still acceptable for operation after implementation of the improved feedwater measurement and 1.5% uprate.

SNC proposes to follow the recommendations provided in Reference 6 for temporary operation above "lull steady-state licensed power levels."

HL-6328 E7-9 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications REFERENCES

1. CENPD-397-P-A, Rev. 01, "Improved Flow Measurement Accuracy Using CROSSFLOW Ultrasonic Flow Measurement Technology," dated May 2000.

2. NRC Regulatory Issue Summary 2002-03, "Guidance on the Content of Measurement Uncertainty Recapture Power Uprate Applications," January 31, 2002.

3. SINH-02-069, 'Uncertainty Study for the Heat Balance Computation Including an Evaluation of UFM Feedwater Flow Measurement Devices," Revision 0.

4. ISA-RP67.04, Part II - 1994, 'Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation," May 1995.

5. NRC Memorandum SSINS No. 0200, "Discussion of Licensed Power Level" (AITS F14580H2), August 22,1980, (NRC Inspection Manual: Inspection Procedure 61706: Core Thermal Power Evaluation).

6. NEDC-33056P 'Thermal Power Optimization and Power Averaging Guidelines,"

Revision 1, February 2002.

HL-6328 E7-1 0 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications TABLE E7-1 UNIT 1 PROCESS PARAMETER INPUTS TO CORE THERMAL POWER CURRENT CONFIGURATION USING VENTURIES WITH NO UFM CORRECTION Contribution to CTP Sensitivity Computation Parameter Uncertainty (%RTPlParameter Units) (eRTP)

Feedwater Flow (% DP Span) 1.5289 0.6988 1.068395 Feedwater Flow Element Errors 0.559 0.9983 0.55805

(% Actual Flow)

Dependent FW Flow Term 0.0784 0.6988 0.054786

(% DP Span)

Feedwater Temperature (OF) 1.2711 0.1309 0.166387 Reactor Pressure (% Span) 1.078 0.0568 0.06123 CRD System Flow (% DP Span) 2.3727 0.0049 0.011626 CRD System Flow Element Effects 1.4142 0.0035 0.00495

(% Actual Flow)

CRD Inlet Temperature (OF) 1.8144 0.0003 0.000544 RWCU Flow (% DP Span) 1.9002 0.0008 0.00152 RWCU Flow Element Effects 0.7071 0.0012 0.000849

(% Actual Flow)

RWCU Inlet Temperature (OF) 3.0429 0.0014 0.00426 RWCU Outlet Temperature (OF) 3.0429 0.0012 0.003651 Recirc Pump Power (% Span) 1.327 0.0030 0.003981 Correction Factor (MW) 0.550 0.0362 0.019910 Square Root Sum of the Squares (SRSS) Total Uncertainty 1.227% RTP RWCU Flow Bias (%DP Span) -1.7864 0.0008 -0.001429 Feedwater Flow Venturi Fouling +0.6000 0.9983 +0.598980

(% Actual Flow)

Total Bias Error -0.001, +0.599% RTP Total RTP Uncertainty -1.228, +1.826% RTP HL-6328 E7-11 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications TABLE E7-2 UNIT 1 PROCESS PARAMETER INPUTS TO CORE THERMAL POWER VENTURIES CORRECTED BY CROSSFLOW UFMS WITH 0.42% ACTUAL FLOW UNCERTAINTY Contribution to CTP Sensitivity Computation Parameter Uncertainty (%RTPlParameter Units) (+/-%RTP)

UFM Flow (% Actual Flow) 0.42 0.9983 0.419286 Feedwater Temperature (OF) 1.2711 0.1309 0.166387 Reactor Pressure (%Span) 1.078 0.0568 0.06123 CRD System Flow (%DP Span) 2.3727 0.0049 0.011626 CRD system Flow Element 1.4142 0.0035 0.00495 Effects (%Actual Flow)

CRD Inlet Temperature (0F) 1.8144 0.0003 0.000544 RWCU Flow (% DP Span) 1.9002 0.0008 0.001520 RWCU Flow Element Effects 0.7071 0.0012 0.000849

(% Actual Flow)

RWCU Inlet Temperature (OF) 3.0429 0.0014 0.00426 RWCU Outlet Temperature (OF 3.0429 0.0012 0.003651 Recirc Pump Power (%Span) 1.327 0.003 0.003981 Correction Factor (MW) 0.550 0.0362 0.019910 Square Root Sum of the Squares (SRSS) Total Uncertainty l 0.456 RWCU Flow Bias (% DP Span) -1.7864 0.0008 -0.001429 Total Bias Error -0.001% RTP Total RTP Uncertainty -0.457, +0.456% RTP HL-6328 E7-12

Enclosure 7 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications TABLE E7-3 UNIT 2 PROCESS PARAMETER INPUTS TO CORE THERMAL POWER CURRENT CONFIGURATION USING VENTURIES WITH NO UFM CORRECTION Contribution to CTP Sensitivity Computation Parameter Uncertainty (%RTP/Parameter Units) (+/-tRTP)

Feedwater Flow (% DP Span) 1.5135 0.6476 0.980143 Feedwater Flow Element Errors 0.559 0.9983 0.558050

(%Actual Flow)

Dependent FW Flow Term 0.0918 0.6476 0.059449

(% DP Span)

Feedwater Temperature (OF) 1.2711 0.1384 0.17592 Reactor Pressure (% Span) 1.078 0.0589 0.063494 CRD System Flow (%DP Span) 2.3727 0.0049 0.011626 CRD system Flow Element Effects 1.4142 0.0035 0.00495

(%Actual Flow)

CRD Inlet Temperature (OF) 1.8144 0.0003 0.000544 RWCU Flow (% DP Span) 1.9008 0.0008 0.001521 RWCU Flow Element Effects 0.7071 0.0012 0.000849

(%Actual Flow)

RWCU Inlet Temperature (OF) 3.0429 0.0014 0.00426 RWCU Outlet Temperature (OF) 3.0429 0.0012 0.003651 Recirc Pump Power (% Span) 1.327 0.003 0.003981 Correction Factor (MW) 0.550 0.0362 0.019910 Square Root Sum of the Squares (SRSS) Total Uncertainty 1.154% RTP RWCU Flow Bias (% DP Span) -1.5294 0.0008 -0.001224 Feedwater Flow Venturi Fouling +0.6000 0.9983 +0.598980

(%Actual Flow)

Total Bias Error J -0.001, +0.599% RTP Total RTP Uncertainty -1.155, +1.753% RTP HL-6328 E7-1 3

Enclosure 7 Request for License Amendment Measurement Uncertainty Recapture Power Uprate Plant Modifications TABLE E7-4 UNIT 2 PROCESS PARAMETER INPUTS TO CORE THERMAL POWER VENTURIES CORRECTED BY CROSSFLOW UFMS WITH 0.42% ACTUAL FLOW UNCERTAINTY Contribution to CTP Sensitivity Computation Parameter Uncertainty (%RTPlParameter Units) (eRTP)

UFM Flow (% Actual Flow) 0.42 0.9983 0.419286 Feedwater Temperature (OF) 1.2711 0.1384 0.17592 Reactor Pressure (% Span) 1.078 0.0589 0.063494 CRD System Flow (% DP Span) 2.3727 0.0049 0.011626 CRD system Flow Element 1.4142 0.0035 0.00495 Effects (% Actual Flow)

CRD Inlet Temperature (OF) 1.8144 0.0003 0.000544 RWCU Flow (% DP Span) 1.9008 0.0008 0.001521 RWCU Flow Element Effects 0.7071 0.0012 0.000849

(% Actual Flow)

RWCU Inlet Temperature (OF) 3.0429 0.0014 0.00426 RWCU Outlet Temperature (OF) 3.0429 0.0012 0.003651 Recirc Pump Power (% Span) 1.327 0.003 0.003981 Correction Factor (MW) 0.550 0.0362 0.019910 Square Root Sum of the Squares (SRSS) Total Uncertainty 0.460% RTP RWCU Flow Bias (% DP Span) -1.5294 l 0.0008 [ -0.001224 Total Bias Error [ -0.001% RTP Total RTP Uncertainty -0.461, +0.460% RTP HL-6328 E7-14

ATTACHMENT 1 TO ENCLOSURE 7 UNCERTAINTY STUDY FOR THE HEAT BALANCE COMPUTATION INCLUDING AN EVALUATION OF UFM FEEDWATER FLOW MEASUREMENT DEVICES Performed for: E. 1.Hatch Nuclear Plant SINH-02-069, Revision 0

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 1 of 89 TABLE OF CONTENTS 1.0 OBJECTIVE / SCOPE ........................................ 3 2.0 FUNCTIONAL DESCRIPTION ....................................... 3 3.0 METHODOLOGY ........................................ 4 4.0 DESIGN INPUTS ........................................ 6 5.0 ASSUMPTIONS AND ENGINEERING JUDGMENTS ....................................... 37 6.0 ANALYSIS ....................................... 39 7.0

SUMMARY

OF RESULTS ....................................... 83

8.0 CONCLUSION

S ....................................... 83

9.0 REFERENCES

....................................... 84 Attachment A Plant Computer Heat Balance Code Listing (22 pages)

Attachment B Rosemount 414L Temperature Transmitter Specifications (4 pages)

Attachment C Feedwater Flow Differential Pressure Transmitter Calibration Computations (1 page)

Attachment D Tabulation of On-Line Plant Computer Heat Balance Parameters (1 page)

Attachment E Hatch Feedwater Flow Correction and Digital Filtering Algorithms (1 page)

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 2 of 89 RECORD OF REVISIONS Rev. Date Paaes DescriDtion Oriainator 0 11/26/02 1-89 Initial Issue Kirk R. Melson

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 3 of 89 1.0 OBJECTIVE I SCOPE E. I. Hatch Nuclear Plant is performing a Minor Power Uprate Project, based on a reduction in the presently assumed 2% Heat Balance uncertainty. The purpose of this study is to evaluate the total measurement uncertainty of the heat balance computation under the following two conditions:

Case 1: current configuration, using feedwater flow venturis to measure feedwater flow, and Case 2: configuration after installation of Westinghouse Crossflow Ultrasonic Flowmeters (UFMs), which will be used to automatically correct the venturi readings on a continuous basis within the Plant Computer.

2.0 FUNCTIONAL DESCRIPTION The heat balance computation is performed within the Plant Computer per Reference 9.8.4. This computation derives the total core thermal power by computing the heat added to the different water sources flowing in and out of the reactor vessel. Specifically, the computation is performed as derived below.

Applying the 1st Law of Thermodynamics (Conservation of Energy), the thermal power generated by the reactor core is expressed as follows, assuming steady state conditions:

QCORE = QFW + QCR + QCU+ CF-Qp where: QCORE = Thermal Power Generated by Reactor Core QFW = Heat Applied to Feedwater QCR = Heat Applied to Control Rod Drive System Qcu = Heat Applied to Reactor Water Clean Up System CF = Correction Factor, Which Includes the Effects of Radiative Heat Loss and CRD Flow Corrections Qp = Heat Added by Recirculation Pumps This equation is verified via References 9.8.4 and 9.2.1.

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 4 of 89 3.0 METHODOLOGY This study is performed, using Reference 9.1 as the instrument uncertainty determination methodology. Since a number of different variables are measured with instrumentation in the heat balance equation, and since these variables are not equally weighted in the effects of the error propagation, weighting factors are assigned to each variable. This is done first by establishing a baseline condition, and then varying each measurement independently by a nominal error value to determine its effect on the output core thermal power calculation. For the CRD Inlet Temperature and the RWCU Inlet and Outlet Temperature measurements, because of the fact that thermocouples are used, and the fact that the spans of these instruments are so large in comparison to the small errors, +/-50 F is chosen for this nominal error value. For the Ultrasonic Flow Meters, since errors are significantly less than +/-1% of Actual Flow, the nominal value of +/-1% Actual Flow is used to establish the weighting factors. For all other parameters, a nominal error value of +/-5% Span is used to determine the weighting factors.

The baseline process conditions are established in the sub-sections of Section 6.1 per Figures 1.2-2 and 1.2-3 of Reference 9.8.1. The weighting factors are determined in Section 6.2. Total loop uncertainties of the various parameters that feed into the heat balance computation are performed within Section 6.3.

Finally, each of the total loop uncertainties are multiplied by their respective weighting factors, and combined to form the total measurement uncertainty for the heat balance thermal power computation in Section 6.4.

There are two bias terms in the measurement, one of which result from the fact that the operating density of the RWCU water is different than the density assumed when the flow / differential pressure relationship was determined for the flow orifice. The other bias term is from the fouling of the Feedwater Flow venturis. This bias may or may not be present, but may vary up to a limit specified herein. Errors for these bias terms are performed separately from the random terms, which are combined via the Square Root of the Sum of the Squares (SRSS) methodology in accordance with Reference 9.1.1.

Where multiple signals are used to determine averages for use in heat balance functions, the uncertainty of the average signal, c, is determined using the methodology of Reference 6.1.1, Section K2.

c=+ (kl*a)2 +(k2*b)2 ...

k1 = k2 = 1/no. signals; a = b = single signal loop error

Attachment 1 to HL-6328, Enclosure 7 E. 1.HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 5 of 89 c= +/- NumberSignals* (a)2 a V (NumberSignals)2 JNumbe rSignals

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 6 of 89 4.0 DESIGN INPUTS 4.1 MISCELLANEOUS DESIGN INPUTS 4.1.1 Reference 9.8.10 shows that for BWR-4 designs, the moisture carryover content is effectively zero, and should be treated as such in the Core Thermal Power computations. This reference also states that the uncertainty of this value when used at a BWR-4 is negligible. Therefore, there are no uncertainties in this estimate, and none are included in this computation. The model for the heat balance calculation will be changed to reflect a moisture carryover content of zero, per Reference 9.8.10.

4.1.2 References 9.8.12 and 9.8.13, plant data was taken for a number of variables for the Extended Uprate Plant Performance tests. Data was recorded from plant instruments, as well as from Measurement & Test Equipment (M&TE) at various process points. This data is used to determine the upstream tap pressure of the feedwater flow venturis to be used for determining feedwater flow parameters.

Note that the M&TE readings are used, as opposed to the plant readings, as the M&TE readings are taken at the process points of interest and are more accurate than the plant instrumentation. (It should be noted here that these pressures are used to determine enthalpies and densities for the feedwater.

Neither enthalpy nor density is a strong function of pressure in this region of the steam tables. However, the operating pressures are extrapolated for correctness of the computations.)

In order to determine the upstream tap pressures, Reference 9.8.12 performed testing at the new operating power level of 2763 MWt. From Appendix A, sheet 6 of 10, parameters P074A_1 and P074B_1 are the two nozzle pressure readings, 1113.47034 and 1118.48291 psia respectively for Test 2-3; and 1114.03223 and 1119.10312 psia respectively forTest 4-1. In orderto determine a nominal Unit 1 nozzle pressure, these values are averaged.

The average value of these 4 readings is 1116.27215 psia. Therefore, the nominal Unit 1 nozzle pressure is determined to be as follows:

Pnozl= 1116 psia In a similar manner, Reference 9.8.13 is used to determine the nominal pressure. However, Reference 9.8.13 only performed testing up to a power level of 2708 MWt. Therefore, in order to determine nozzle pressure for Unit 2, the values are extrapolated. From Appendix C, sheet 8 of 13, parameters P074A_1 and P074B_1 are the two nozzle pressure readings, 1119.643 and 1120.057 psia respectively for Test 2-5. In order to determine a nominal Unit 1 nozzle pressure, these values are averaged to obtain 1119.85 psia. Document reviews and SNC interviews indicate that the test control for power level during test performance was not adequate, so the value for QT on Sheet 10 of 13 of

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 7 of 89 Appendix C, 2723.445 MWt is used for the actual power level of the test.

From Sections 6.2.9, it can be seen that errors in actual flow rate (in terms of %

Actual FW flow) are on a near 1:1 relationship to errors in percent of rated power. When normalized, this means that in terms of percent, % Actual FW Flow approximates % power. Also, flow is related to the differential pressure across the FW venturis by a square root relationship. Since the output pressure of the venturis is approximately 1050 psia per Figure 1.2-2 of Reference 9.8.1, the extrapolation of nozzle pressure to a power level of 2763 MWt is performed as follows. (Approximate values)

Q = K1 x Flow = K2x(DP)0 5 K2 = Q1/(DP1) 05

= 2723.445 / (1119.85-1050)°5

= 325.8633 Pnoz2 (Q2/K2) 2 + 1050psia

= (2763/325.8633) + 1050 psia

= 1121.894 psia Therefore, the nozzle pressure for Unit 2 at 100% power operating conditions is extrapolated from test results to be as follows.

Pnoz2 = 1122 psia 4.2 FEEDWATER FLOW UNCERTAINTY CONSIDERATIONS Per References 9.7.1 through 9.7.12, the Feedwater Flow instrument loop consists of flow elements, power supplies and transmitters, which provide direct input to the Plant Computer. The following is a detailed listing of the considerations necessary for each instrument in this instrument loop configuration.

Note that for Case 2, the uncertainties of the Feedwater Flow instrument loop are continuously compensated out, by continuous automatic correction of the Plant Computer Feedwater Flow reading to match those of the Ultrasonic Flow Meters. Therefore, all errors derived within this section apply to Case 1 only.

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 8 of 89 4.2.1 FEEDWATER FLOW ELEMENT AND FLUID DENSITY TAG NUMBERS: 1(2)C32-NO01A, B [9.7.1,7]

MANUFACTURER: Permutit [9.6.1,2]

Flow Span: 0 - 8 MIbrhr [9.3.1,2]

4.2.1.1 Per Reference 9.5.24, the required Reference Accuracy of the flow element is +/- 0.5% at a 95% confidence. However, References 9.6.2 and 9.6.13 show that the laboratories calibrated the flow elements to an accuracy of +/- 0.25%.

This uncertainty is a result of the potential inaccuracy of the test flow used in the establishment of the discharge coefficient of the flow element. This accuracy is therefore a direct result of the test flow inaccuracy, and is in terms of %Actual Flow. Subsequent inspections have revealed no reason to increase this uncertainty value. Since this error is input to the flow measurement via an error in the discharge coefficient, this error is compensated for when the venturi flow value is corrected to the UFM value.

RAFWFE = +/- 0.2500% Actual Flow 4.2.1.2 Per Section 11-11-3 of Reference 9.1.2, the installation effects for the flow element are dependent on the compliance with standards in the installation of the venturi. Ifthe guidelines of Figure 11-11-1 are met, the errors due to piping installation effects are limited to +/-0.5%, but if they are not, an additional

+0.5% is required to be added. Per References 9.6.1 and 9.6.2, the feedwater flow venturis are installed as a part of a GE assembly, which includes a long section of 18" piping, using flow straighteners in the upstream and downstream piping sections. Per the same references, the Beta ratios are nominally 0.5 for Unit 1 and 0.5161 and 0.5151 for the two flow elements in Unit 2. Reference 9.1.2 Figure 11-11-1, item (H) is the most correct figure for assessing the acceptability of the installation. A Beta ratio of approximately 0.5 would require a minimum of 5 diameters upstream and 2 diameters downstream of straight piping. This would equate to approximately 7.5' upstream and 3' downstream. Per the descriptions and construction details in References 9.6.1 and 9.6.2, these requirements are fully met by these installations. Per Section C.3 of Reference 9.1.1, this error is considered as an error to the discharge coefficient, which is proportional to the Actual Flow value. Therefore, IEFWFE + 0.5000% Actual Flow Because of the detailed design using flow straighteners and long sections of piping, any errors, due to installation effects, in the flow to differential pressure characteristic at 100% power operating conditions, are considered to be minor and capable of being compensated for by the UFM correction process.

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 9 of 89 4.2.1.3 Per References 9.3.1 and 9.3.2 and Assumption 5.3, the calibration tables for the associated transmitters are computed, taking into account the feedwater temperature at 100% power conditions. The calculations compute the correct densities and account for thermal expansion of the venturis. Additionally, per Attachment 1 to Reference 9.3.2, Attachment E to this calculation, and References 9.8.16 and 9.8.17, the plant computer compensates the feedwater flow signal for feedwater temperature changes, and this compensation accounts for changes in thermal expansion factor. The only possible expansion factor error would be due to errors in the temperature signal to the plant computer, which would affect the compensation algorithm.

Errors in the temperature signal are so small that the resulting error in compensation for thermal expansion factor (EF) is negligible.

EFFWFE = Negligible 4.2.1.4 Per Attachment 1 to Reference 9.3.2, Attachment E to this calculation, and References 9.8.16 and 9.8.17, the plant computer compensates the feedwater flow signal for feedwater temperature changes, and this compensation accounts for changes in feedwater density. The only remaining density error would be due to errors in the temperature signal to the plant computer, which would affect the compensation algorithm. Errors in the indicated feedwater temperature affect the accuracy of the flow reading because of the associated density errors.

Per Section 6.3.2 of this calculation, an error for each temperature instrument loop in the plant computer has been determined as follows.

TUFwT = +2.5421 OF In this case, per Attachment E, the plant computer takes the average of the two feedwater temperature measurements to use for compensation of the associated loop flow. Therefore, the uncertainty is expressed as follows for this average.

TLU (Avg FW Temp) = SRSS (TUs of 2 FW Temp Signals) / No. Signals

= (2x(TU of Single FW Temp Signal) 2 2/2

= TU of Single FW Temp Signal / (2) 2 TLUFwT.comp L= TUFWT/ (2) "2 +/- 1.7975°F

Attachment 1 to HL-6328, Enclosure 7 E. 1.HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 10 of 89 For conservatism, and to account for small errors in the temperature compensation algorithm within the plant computer, a +20F temperature variation around the nominal value would reasonably bound the normal operating scenario. Therefore, the density error is computed based on this variation.

Using normalized values from Section 6.1.1 for the nominal flow rates at 100% Power Operation, Unit 1 C(100% Power) = 71.9625% Flow Span A(100% Power) = 51.7860% DP Span Unit 2 C(100% Power) = 74.6875% Flow Span A(100% Power) = 55.7822% DP Span Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the nominal feedwater temperatures for the two Units at 100% power operating conditions is 397.50 F for Unit 1 and 425.1IF for Unit 2. Per Design Input 4.1.2, the nozzle pressures for Unit 1 and 2 during 100% power operation are 1116 psia and 1122 psia, respectively. Per Assumption 5.3, at the completion of this project, the calibration tables of the feedwater flow transmitters will be adjusted to the heat balance operating conditions. Per Assumption 5.4, the present heat balance conditions are treated as the baseline. Therefore, the following band of densities is considered for the two units, as linearly interpolated from Reference 9.8.2.

Unit 1 p1 -(1116 psia and 395.51F) = 54.1307 Ibm/ft3 p1 (1116 psia and 397.5 0F) = 54.0488 Ibm/ft 3 p1 +(l116 psia and 399.50 F) = 53.9671 Ibm/ft 3 Unit 2 p2-(1122 psia and 423.1IF) - 52.9711 Ibm/ft 3 p2(1122 psia and 425.1OF) = 52.8839 Ibm/ft 3 p2+(1122 psia and 427.1 °F) = 52.7970 Ibm/ft 3 Per Reference 9.8.4, the input signal is multiplied by a constant to determine mass flow rate. Therefore, holding mass flow rate at a constant value equal to the heat balance figures and holding all other terms in the flow equation constant except for DP and density, the following are the effects on the differential pressure reading.

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 11 of 89 The equation is derived from References 9.3.1 and 9.3.2.

DP = K/p Ki = DPiI x pi

= 51.7860% DP Span x 54.04881bm/ft 3

= 2798.9712% DP Span x Ibm/ft3 K2 = DP 2 1x p 2

= 55.7822% DP Span x 52.88391bm/ft 3

= 2949.9803% DP Span x Ibm/ft 3 Therefore, the change in DP readings due to the density changes from those assumed at calibration are as follows:

DP1 . = K, / pi-

= 2798.9712/ 54.1307

= 51.7076% DP Span PEr. = +DP 1 --DP,

= +51.7076- 51.7860

= -0.0784% DP Span DP1 + = K1 / pi+

= 2798.9712 / 53.9671

= 51.8644% DP Span PErj+ = +DP1 +-DP ,,

= +51.8644- 51.7860

= +0.0784% DP Span This is treated as a random uncertainty, as the temperature effect that causes this error is random.

PEFWFE1 = +/-0.0784% DP Span DP2 . = K2 / P2-

= 2949.9803 / 52.9711

= 55.6904% DP Span PEr2 = +DP 2 -DP2 ,

= +55.6904- 55.7822

= -0.0918% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 12 of 89 DP2+ = K2 / P2

= 2949.9803 / 52.7970

= 55.8740% DP Span PE2+ = +DP 2 +-DP2 1

= +55.8740- 55.7822

- +0.0918% DP Span PEFWFE2 = +/-0.0918% DP Span Per Reference 9.8.8, when the UFMs are used to correct the venturi flow readings, the correction factor is updated continuously (generally not longer than 1 minute between updates). Any temperature compensation error for the venturi reading is compensated out. Therefore, this error term is only applicable without UFM correction.

NOTE: Because this error is directly due to the error of the Feedwater Temperature measurement, this term is dependent with the Feedwater Temperature Total Loop Uncertainty. The Feedwater temperature measurement is the reference for the calculation of Feedwater density and Feedwater enthalpy. Since the change in either of these terms is not random with respect to a directional change in Feedwater temperature, a dependent relationship exists between the density and enthalpy functions. Therefore, this term must be treated separately from the other terms in this loop, until combined with the Feedwater Temperature Error.

4.2.1.5 Feedwater Flow venturi fouling is an accepted phenomenon throughout the nuclear industry. For Hatch, tracer tests performed in the early 1990's and several inspections showed excellent agreement between measured and indicated feedwater flows. Feedwater flow has also been compared to steam and condensate flows in order to trend possible feedwater fouling effects. At Hatch, no fouling trends were evident.

Research was performed to determine if any industry data exists which would help to determine the feedwater venturi fouling for this specific plant. Per Reference 9.8.11, 'The most common cause of changes in nozzle bias is the phenomenon of fouling. Fouling induced biases have proven to be difficult to predict, both in magnitude and in variation over a fuel cycle. Long term comparisons of data from LEFM, nozzles, and other plant instruments have confirmed the presence of fouling in at least 21 plants. The average value is 1%... The average change in bias observed in 4 BWR's with the LEFM is 0.6%." Since venturi fouling can occur during a cycle, and is dependent on water chemistry, and process temperatures and pressures, the value from Reference 9.8.11 is conservatively chosen to envelope fouling at Hatch.

PEbFWFE =- + 0.6% Actual Flow

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 13 of 89 Since UFMs actually correct the Feedwater Flow reading during power operation, this error is eliminated for Case 2.

4.2.2 FEEDWATER FLOW TRANSMITTER CONSIDERATIONS TAG NUMBER: 1(2)C32-NQ02A, B [9.7.1-12]

MANUFACTURER: ROSEMOUNT [9.5.23, 24]

MODEL NUMBER: 1151 DP6B22MB (Smart) [9.5.23, 24]

1C32-N002A SPAN: 0 - 2367.9 UH2O [Att. C]

1C32-N002B SPAN: 0 - 2357.6 UH2O [Aft. C]

2C32-NO02A SPAN: 0 - 2325.2 UH2O [Aft. C]

2C32-NO02B SPAN: 0-2310.1 "H20 [Aft. C]

4.2.2.1 Per Reference 9.4.1, the Analyzed Drift (DA) for the flow transmitter is +/-

1.447% DP Span for a period of 30 months between calibrations. This includes the effects of Drift (DR), Reference Accuracy (RA) and Measurement &Test Equipment (M&TE).

DAFwFr = +/-1.4770% DP Span 4.2.2.2 Setting Tolerance effects are due to the flexibility of the technician in the calibration process. The effect is bounded by the As-Left setting tolerance of the device. Per References 9.2.2 and 9.2.3, the As-Left setting tolerance is +/-

0.2500% Span for Unit 1 and +/- 0.1250% Span for Unit 2.

STFwF1 = +/- 0.2500% DP Span STFwFT2 = +/- 0.1250% DP Span 4.2.2.3 Per Reference 9.6.4, the flow transmitter has specifications for static pressure zero and span effect. Per Reference 9.3.1 & 9.3.2, the zero effect is fully calibrated out, and the span effect is compensated for in the establishment of the calibration parameters for the transmitters. The residual static pressure span effect is due to the fact that each transmitter responds slightly differently with respect to the span effect, and the correction procedure given merely corrects for the average transmitter response of the all transmitters produced. This effect is specified in Reference 9.6.4 as +/-

0.25% of input reading per 1000 psi. Per Design Input 4.1.2, the highest nominal operating nozzle pressure is for Unit 2. Since the pressures are close to each other from Unit 1 to Unit 2, the error for Unit 2 is conservatively applied to both units, using 1122 psia as the pressure value. Calibration is performed at approximately 14.7 psia. Conservatively using 100% span to compute this figure, we obtain the following Static Pressure Effect (SPE).

SPEFwFT = +/- 0.25% DP Span x (1122-14.7) / 1000 psi

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 14 of 89 SPEFwFT = +/- 0.2768% DP Span 4.2.2.4 Per Reference 9.6.4, the flow transmitter Power Supply Effect (PSEFWFr) is given as less than +/-0.005% URL per Volt, and the load effect is negligible.

Per References 9.5.1 and 9.5.2, the power supplies for the flow transmitters regulate the voltage supplied to the transmitter to 26.5 +/- 1.5 VDC. Per Reference 9.6.4, the URL is 100 psid (2773 "H2O). Use of the transmitter with the least calibrated span maximizes this uncertainty term; therefore, the span for 2C32-NO02B is used to compute this term. Therefore, the flow transmitter Power Supply Effect (PSEFwFT) is given as:

PSEFwF = +/- (0.005% URL / VDC)(1.5 VDC) x [2773 "H20 /

2310.1 "H20]

PSEFwF, = +/- 0.0090% DP Span (Only Use with No UFM Correction) 4.2.2.5 Per Reference 9.6.4, the flow transmitter Temperature Effect (TEFWFT) is given as +/- (0.2% URL + 0.18% Span) / 1000F, and the URL is 2773 "H20.

Use of the transmitter with the least calibrated span maximizes this uncertainty term; therefore, the span for 2C32-NO02B is used to compute this term. The transmitters are not accessible during operation, and are located in the Turbine Building per Reference 9.8.5. Per Reference 9.8.6, the maximum temperature during normal operation in this room is 100F, and the normal temperature is 700F. Therefore, the maximum difference in temperature between operating conditions and calibration conditions is 300F, which covers calibrations in the field or in the laboratory. Therefore, the temperature effect is computed as follows:

TEFwFr =+ (((0.2%)(2773 "H20) / 2310.1 "H20] + 0.18% Span] x (300F/1 000F)

TEFwFp +0. 1260% DP Span 4.2.3 ULTRASONIC FLOW METER (UFM) UNCERTAINTY CONSIDERATIONS 4.2.3.1 Per Reference 9.8.3, "On the basis of the staff's review of the Topical Report CENPD-397-P, Revision 01 (Proprietary and Non-Proprietary), the staff concludes that the CROSSFLOW UFM is designed and tested to achieve the flow measurement uncertainty of 0.5 percent or better, with a 95 percent confidence interval." However, Westinghouse is making changes to the design of the Crossflow UFM system to provide significantly better accuracy.

Based on Reference 9.8.3, the Crossflow UFM is at least accurate to 0.5 percent of actual flow. At the point of issuance of this evaluation, based on preliminary laboratory results, the accuracy for the Hatch specific Crossflow system is anticipated to be +/-0.42% actual flow. Therefore, the Total

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 15 of 89 parameter Uncertainty (TU) is derived as follows:

TUUFMXFLOB = +/- 0.4200% Actual Flow (95% Confidence Factor) 4.3 FEEDWATER TEMPERATURE UNCERTAINTY CONSIDERATIONS Per References 9.7.1 and 9.7.13-9.7.23, the Feedwater Temperature instrument loop to the Plant Computer consists only of the Temperature Element and the Temperature Transmitter, which directly feeds the Plant Computer. The temperature transmitters are calibrated individually. Per discussions with plant personnel, once every 5 to 6 years, a loop calibration check is performed, using an oil bath for the temperature element. The plant computer reading is checked to be within +/-0.50F of a precision measurement in the oil bath for a 3-point calibration. If the measurement differs by more than the criteria, then the RTD is replaced. For conservatism, no credit is taken for reductions in uncertainty due to this loop calibration.

4.3.1 FEEDWATER TEMPERATURE ELEMENT TAG NUMBER: 1(2)B21 -NO41 A-D [9.7.1, 13]

MANUFACTURER: Rosemount [9.5.11, 13]

TYPE: 177 L - 200 Q Platinum RTD [9.5.11, 13]

4.3.1.1 Per References 9.5.11-14, the system accuracy for platinum temperature elements, the associated leads and temperature transmitter is +/- 0.30 0F. This term is addressed in Section 4.3.2.1.

4.3.1.2 The RTD has no adjustment and therefore cannot be calibrated. Therefore, the errors that can be introduced during calibration (Setting Tolerance and M&TE) do not apply to this device.

STFwTE = N/A M&TEFWTE = N/A 4.3.1.3 RTD lead wire effects are negligible with 4-wire RTDs. Per References 9.7.14 and 9.7.15, these are 4-wire RTDs. Therefore, RTD Lead Wire Effects are negligible for this application.

LWFWTE = N/A 4.3.1.4 RTD Self-Heating Effects are negligible if used with flowing fluids. Therefore, Self-Heating Effects are negligible for this application.

SHFWTE = N/A

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 16 of 89 4.3.2 FEEDWATER TEMPERATURE TRANSMITTER TAG NUMBER: 1(2)B21-N602A-D [9.7.1, 9.7.13-23]

MANUFACTURER: Rosemount [9.5.12,14]

MODEL NUMBER: 414H [9.5.12,14]

SPAN: 150OF [9.2.6, 7]

SPAN: -150mVDC [9.2.6, 7]

The Vendor Technical Manuals do not contain performance specifications for these devices, and Rosemount no longer has supporting specification sheets. However, Reference 9.6.11 contains technical information on a similar product from Rosemount.

In the absence of specification type information for the 414H transmitter, specifications for the 414L are used.

4.3.2.1 Per References 9.5.12 and 9.5.14, the reference accuracy of the temperature sensor and transmitter as a matched pair is given as +/-0.30F for 3a. For this evaluation, we must consider 2a uncertainties, or +/-0.20F. Per Reference 9.6.11, the Model 414L temperature transmitters are accurate to +/-0.1% with platinum RTDs. The range of these transmitters is 1500F, which gives a transmitter error of +/-0.1 50F. Therefore in accordance with the data sheets, in order to provide the error of the total system, the Reference Accuracy is conservatively assigned as follows:

RAFw7 = +/- 0.20000F 4.3.2.2 Setting Tolerance effects are due to the flexibility of the technician in the calibration process. The effect is bounded by the As-Left setting tolerance of the device. Per References 9.2.6 and 9.2.7, the As-Left setting tolerance is +

0.3000 mVdc. Therefore, STwlT = +/- (0.3000/149.96) x 150OF STFwrU = +/- 0.3001 OF 4.3.2.3 Measurement and Test Equipment effects (M&TE) are errors introduced during the calibration process and are constant at a given point on the calibration curve throughout an operating cycle. M&TE is chosen to be at least as accurate as the equipment being calibrated, and generally considerably more accurate. Therefore, in order to provide the technician with flexibility in the choice of M&TE, the M&TE uncertainty is conservatively set equal to the device Reference Accuracy.

M&TErwrr =T + 0.20000 F

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 17 of 89 4.3.2.4 Reference 9.6.11 does not specify any drift uncertainty values. Because of the redundancy in signals, and the use of these signals in different applications, significant drift would be detected and corrected. For conservatism, the magnitude of the drift uncertainty is set equal to the Reference Accuracy term. Therefore, DRFwT-r = +/- 0.20000F 4.3.2.5 Per Reference 9.6.11, the temperature transmitter Power Supply Effect (PSE) is given as less than or equal to +/-0.01 % for a +/-10% line change. Per References 9.7.14 and 9.7.15, the 120 VAC instrument bus powers the temperature transmitters. Per Section 8.7.3 (Unit 1) and Section 8.3.1.1.4 (Unit 2) of Reference 9.8.1, the instrument AC power system is regulated to within +/- 10%. Therefore, the temperature transmitter Power Supply Effect is given as:

PSEFwTT = +/- (0.0001) x (150 0F)

PSEFwTT = +/- 0.01 500F 4.3.2.6 Per Reference 9.6.11, the temperature transmitter Temperature Effect (TE) is given as +/- 0.05 0F / OF. Per Reference 9.8.5, the temperature transmitters are located in the Turbine Building. Per Reference 9.8.6, the maximum temperature during normal operation in this room is approximately 100 0F, and the normal temperature is 70 0F. However, because of the location of the devices, a maximum temperature of 1200F is conservatively used for computing temperature effect for these devices. The maximum difference in temperature between operating conditions and calibration conditions is 500F, which covers calibrations in the field or in the laboratory. Therefore, the temperature effect is computed as follows:

TER= +/- (0.05OF/ 0F) x (500F)

TEFw1- = +/- 2.50000F 4.3.3 UFM FEEDWATER TEMPERATURE 4.3.3.1 The Westinghouse Crossflow UFMs are temperature compensated within the Crossflow system, using separate temperature measurement equipment from the existing installed feedwater temperature sensors. The error of that temperature is included within the overall uncertainty for the UFM as given in Section 4.2.3.1 above. There are no additional UFM temperature errors that need to be considered, and for Case 2, the UFM temperature error is NOT a dependent term with the Feedwater temperature error (which is used in the enthalpy determination), since different instruments are used to supply the temperature input to Feedwater density and the enthalpy calculation.

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 18 of 89 4.4 REACTOR PRESSURE UNCERTAINTY CONSIDERATIONS Per References 9.7.16, 20, 24 and 25, the Reactor Pressure transmitter is a direct input to the Plant Computer, with no intermediate devices. Therefore, the pressure transmitter and Plant Computer are the only devices in the loop for which instrument uncertainty must be considered.

4.4.1 REACTOR PRESSURE TRANSMITTER TAG NUMBER: 1(2)C32-NOO5A, B [9.7.16,20, 24, 25]

MANUFACTURER: Rosemount [9.5.5, 6]

MODEL NUMBER: 1151GP9 [9.5.5,6]

URL: 3000 PSIG [9.6.4]

SPAN: 0-1200 PSIG* [9.2.2, 9.2.3]

• Includes 14 psi (Ul) and 14.5 psi (U2) sensing line pressure offset. [9.2.2, 9.2.3]

4.4.1.1 Drift analysis has shown gauge pressure transmitters to perform more accurately than differential pressure transmitters, in terms of % Span. Also, Rosemount 1151 transmitters perform at least equally with, if not more accurately, than Rosemount 1153 transmitters in terms of drift. There has been no specific drift analysis for Rosemount 1151 gauge pressure transmitters, but specific drift analyses were prepared for Rosemount 1153 Range Code 9 transmitters (Reference 9.4.2), and for Rosemount 1151 differential pressure transmitters (Reference 9.4.1). (Note however that Reference 9.4.2 limits application of the conclusions to transmitters with Tumdown Factors less than or equal to 2, which is not the case for these transmitters.) Reference 9.4.2 computes the Analyzed Drift (DA) for Rosemount 1153 Range Code 9 gauge pressure transmitters to be lower than that shown in Reference 9.4.1 for the Rosemount 1151 differential pressure transmitters. Therefore, for conservatism, the drift value for the Rosemount 1151 differential pressure transmitters is used for this application.

Per Reference 9.4.1, the Analyzed Drift (DA) is applied to these transmitters as +/- 1.447% DP Span for a period of 30 months between calibrations. This includes the effects of Drift (DR), Reference Accuracy (RA) and Measurement & Test Equipment (M&TE).

DARPT = +/- 1.4470% Span 4.4.1.2 Setting Tolerance effects are due to the flexibility of the technician in the calibration process. The effect is bounded by the As-Left setting tolerance of the device. Per References 9.2.2 and 9.2.3, the As-Left setting tolerance is +

0.2500% Span.

STRPT = +/- 0.2500% Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 19 of 89 4.4.1.3 Per Reference 9.6.4, the pressure transmitter Power Supply Effect (PSERpT) is given as less than +/-0.005% URL per Volt, and the load effect is negligible.

Per References 9.5.3 and 9.5.4, the power supplies for the pressure transmitters are rated to 26.5 +/- 1.5VDC. Per Reference 9.6.4, the URL is 3000 psig. Therefore, the pressure transmitter Power Supply Effect (PSERPT) is given as:

PSERPT = +/- (0.005% URL / VDC) x (1.5 VDC) x (3000psig I 1200 psig)

PSERPT = +/- 0.01875% Span 4.4.1.4 Per Reference 9.6.4, the pressure transmitter Temperature Effect (TEFwFT) is

+/- (0.4% URL + 0.36% Span) / 100F, and the URL is 3000 psig. The transmitters are located in the Reactor Building at elevation 158' per Reference 9.8.5. Per Reference 9.8.7, the maximum temperature during normal operation at this location is 100 0F. Per Reference 9.8.6, the normal temperature in the area (assumed for calibration conditions) is 70 0F.

Therefore, the maximum difference in temperature between operating conditions and calibration conditions is 300F, which covers calibrations in the field or in the laboratory. Therefore, the temperature effect is computed as follows:

TERPT = + (((0.4%)(3000 psig) / 1200 psig] + 0.36% Span](301F/1001F)

TERPT = + 0.4080% Span 4.5 CONTROL ROD DRIVE FLOW UNCERTAINTY CONSIDERATIONS Per References 9.7.16, 20, and 26-29, the Control Rod Drive Flow instrument loop consists only of the flow element and the flow transmitter, which is directly input to the Plant Computer.

4.5.1 CONTROL ROD DRIVE FLOW ELEMENT AND FLUID DENSITY TAG NUMBERS: 1(2)C11-N003 [9.7.26, 28]

MANUFACTURER: Badger [9.5.20, 22]

Span: 0-200"H 20 [9.2.4, 5]

Flow Span: 0-100 GPM [9.5.20, 22]

(0-50,000 Ibm/hr on Plant Computer) 4.5.1.1 Per Reference 9.7.47, the rated accuracy of the flow element is +/-1%. This uncertainty is a result of the potential inaccuracy of the test flow used in the establishment of the discharge coefficient for the flow element. This accuracy is therefore a direct result of the test flow inaccuracy, and is in terms of %Actual Flow.

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 20 of 89 RACRDFE = +/- 1.0000% Actual Flow 4.5.1.2 Per Section 11-11-3 of Reference 9.1.2, the installation effects for the flow element are dependent on the compliance with standards in the installation of the flow nozzle. If the guidelines of Figure 11-11-1 are met, the errors due to piping installation effects are limited to +/-0.5%, but if they are not, an additional +/-t0.5% is required to be added. Per References 9.5.20 and 9.5.22, the line size is 2 inches. Per References 9.7.41 and 9.7.42, there are 10 diameters of straight pipe from the nearest bend to the entrance of the nozzle, and approximately 5 diameters of straight pipe after the nozzle until the next bend. Reference 9.1.2 Figure 11-11-1, item(C) is the most correct figure for assessing the acceptability of the installation. Figure l1-I1-1, item (C), requires more than 12 diameters of straight piping before the nozzle, no matter what the Beta ratio is. Therefore, the existing installation does not meet the requirements, and the additional +/-0.5% is added. Per Section C.3 of Reference 9.1.1, this error is considered as an error to the discharge coefficient, which is proportional to the Actual Flow value. Therefore, IECRDFE + 1.0000% Actual Flow 4.5.1.3 Per References 9.2.4, 9.2.5, 9.5.20, 9.5.22 and Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the flow nozzles are sized for operating conditions at approximately 100% power. The process condition used to size the flow nozzle is Specific Gravity - 1, and process temperature = 150 0F. Per Reference 9.8.14, the design temperature range for this system is 400F to 1500F. This is a small temperature band to consider, and the nozzle is closely matched to it. The small change in temperature will not significantly affect the characteristics of the flow orifice itself. Therefore, the effect on the measurement from the thermal expansion factor of the nozzle is negligible.

EFCRDFE Negligible 4.5.1.4 The CRD System Flow reading is not temperature (density) compensated within the Plant Computer. Changes in operating temperature of the CRD system water affect the accuracy of the reading because of the associated changes in density from that assumed for the calibrations of the transmitters.

Per References 9.5.20 and 9.5.22, the base condition for the design and calibration of the orifice plates is approximately a Specific Gravity of 1.

However, during operation, the source of the CRD water can either be the Condensate Storage Tank or the Condensate System. The water from these sources has a wide range of possible temperatures; therefore, the design temperature range is considered inthis analysis. Per Reference 9.8.14, the design temperature range for this system is 400F to 150 0F. Therefore, the errors from calibration conditions are considered, and the largest error conservatively applied in both directions as a random term. The base

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 21 of 89 condition assumed for calibration is Specific Gravity = 1, which implies the Saturated Water condition at 680F (Reference 9.8.2).

Pbase = 62.3208 Ibm/ft 3 Using normalized values from Section 6.1.4 for the nominal flow rates at 100% Power Operation, for Unit 2 are:

CRD Flow (U2) = 60.4706% Flow Span CRD Flow (U2) = 36.5669% DP Span The temperature band to be considered is 400 F to 150 0F, at a pressure of approximately 1050 psia, corresponding to the reactor pressure.

p-(1 050 psia and 400F) = 62.6566 Ibm/ft 3 Pbase = 62.3208 Ibm/ft3 p+(1 050 psia and 1500F) = 61.3874 Ibm/ft 3 Per Reference 9.8.4, the input signal is multiplied by a constant to determine mass flow rate. Therefore, holding mass flow rate at a constant value equal to the heat balance figures and holding all other terms in the flow equation constant except for DP and density, the following are the effects on the differential pressure reading.

DP= K/p K2 = DP2 1x p2

= 36.5669% DP Span x 62.32081bm/ft3

= 2278.8785% DP Span x Ibm/ft 3 Therefore, the change in DP readings due to the density changes from calibration conditions are as follows:

DP 2 = K2 /P2-

= 2278.8785 / 62.6566

= 36.3709% DP Span PEr2 = +DP2 . - DP2 i

= +36.3709- 36.5669

= -0.1960% DP Span DP2+ = K2 / p2+

= 2278.8785/61.3874

= 37.1229% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 22 of 89 PEr2+ = +DP2 +-DP2 !

= +37.1229- 36.5669

= +0.5560% DP Span The largest value is conservatively applied in both directions.

PECRDFE = +/-0.5560% DP Span 4.5.2 CONTROL ROD DRIVE FLOW TRANSMITTER TAG NUMBER: 1(2)C11-N004 [9.7.16,20,26-29]

MANUFACTURER: General Electric [9.5.21, 23]

MODEL NUMBER: 555111 BCAA3ABA [9.5.21, 23]

Span: 0-200"H 2 0 [9.2.4, 5]

Span: 10-5OmADC [9.2.4, 5]

Flow Span: 0-100 GPM [9.5.20, 22]

(0-50,000 Ibm/hr on Plant Computer) 4.5.2.1 Per Reference 9.6.7, the Reference Accuracy, including the effects of linearity, hysteresis and repeatability, for the G.E. 555 transmitter is +/- 0.4% of Span. Therefore, RACRDFT = +/- 0.4000% DP Span 4.5.2.2 Setting Tolerance effects are due to the flexibility of the technician in the calibration process. The effect is bounded by the As-Left setting tolerance of the device. Per References 9.2.4 and 9.2.5, the As-Left setting tolerance is +/-

0.5000% Span.

STCRDFr = +/- 0.5000% DP Span 4.5.2.3 Measurement and Test Equipment effects (M&TE) are errors introduced during the calibration process and are constant at a given point on the calibration curve throughout an operating cycle. M&TE is chosen to be at least as accurate as the equipment being calibrated, and generally more accurate. Therefore, in order to provide the technician with flexibility in the choice of M&TE, the M&TE uncertainty is conservatively set equal to the device Reference Accuracy.

M&TECRDFT= +/- 0.4000% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 23 of 89 4.5.2.4 Per Reference 9.6.7, the flow transmitter has specifications for static pressure effect as +/- 0.4% Span / 500 psi from 100% to 50% Span, or +/- 0.4%

Span to +/- 1.0% Span / 500 psi from 49% to 20% Span. Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the operating pressure for both units is approximately 1050 psia. Also per these figures, the normal flow rate of the CRD system is 3x104 lb/hr, which per Section 6.1.4, corresponds to 60 GPM, which is 60% Flow Span and 36% DP Span. The static pressure effect is conservatively assigned as +/- 1.0% Span / 500 psi. Calibration is performed at 14.7 psia. Therefore, SPECRDFT = +/-1.0% DP Span x (1 050-14.7) / 500 psi SPECRDFT = +/- 2.0706% DP Span 4.5.2.5 Per Reference 9.6.7, no power supply effect is specified for these transmitters. Given the large magnitude of the other accuracy specifications for this device, power supply effects are negligible in comparison. Therefore, PSEcRDFT = Negligible 4.5.2.6 Per Reference 9.6.7, the flow transmitter has specifications for temperature effect as +/- 1% Span / 100F from 100% to 50% Span; or +/- 1% Span to +/- 2%

Span / 100OF Span from 49% to 20% Span. Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the normal flow rate of the CRD system is 3x104 lb/hr, which per Section 6.1.4, corresponds to 60 GPM, which is 60% Flow Span and 36%

DP Span. Therefore, the normal temperature effect is conservatively assigned as +/- 2.0% Span / 100F. Per Reference 9.8.5, the transmitters are located at elevations 87 and 111 within the Reactor Building. Per Reference 9.8.7, the maximum normal operating temperature at these locations is 104 0F, and the normal temperature (assumed for calibration conditions) is 70 0F. Therefore, the maximum difference in temperature between operating conditions and calibration conditions is 34 0F, which covers calibrations in the field or in the laboratory. Therefore, the temperature effect is computed as follows:

TECRDFT = +/- 2.0% DP Span x (340F) / 100OF TEcRDFr = +/- 0.6800% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 24 of 89 4.6 CONTROL ROD DRIVE INLET TEMPERATURE UNCERTAINTY CONSIDERATIONS Per References 9.7.16, 22, and 26-29, the Control Rod Drive Inlet Temperature instrument loop consists only of the temperature element (thermocouple), which directly feeds its signal to the Plant Computer.

4.6.1 CONTROL ROD DRIVE INLET TEMPERATURE ELEMENT TAG NUMBER: 1(2)C11-N061 [9.7.16, 22, 26-29]

MANUFACTURER: Omega [9.5.7, 8]

TYPE: Type T Thermocouple [9.5.7, 8]

RANGE: 0-7520F* [9.8.9]

• References 9.5.7 and 9.5.8 show a range of 0-9000 F for these thermocouples.

However, it also identifies these as Omega Type T Thermocouples. Per Reference 9.8.9, a type T thermocouple maximum output is at 752 0F. Since these thermocouples are reading CRD System temperature, which is normally 123 0F, the range of these thermocouples is established as 0 to the maximum temperature, 752 0F.

4.6.1.1 Per Reference 9.8.9, the limits of error for an Omega Type T thermocouple are +/- 1.00C or 0.75%, whichever is greater, when above 0C. The normal temperature for the CRD water is 123.20F or 123.5 0F per Figures 1.2-2 and 1.2-3 of Reference 9.8.3. For comparison, 0.75% of 520C (125.6 0F) is 0.39 0C. Therefore, +/- 1.0CC (+/- 1.8000 0F) is used. The Reference Accuracy (RA) of the temperature element is:

RAcRDTE +/- 1.80000 F 4.6.1.2 The thermocouple has no adjustment and therefore cannot be calibrated.

Therefore, the errors that can be introduced during calibration (Setting Tolerance and M&TE) do not apply to this device.

STCRDTE = N/A M&TECRDTE = N/A

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 25 of 89 4.7 REACTOR WATER CLEANUP (RWCU) FLOW UNCERTAINTY CONSIDERATIONS Per References 9.7.17, 20, and 30-33, the RWCU Flow instrument loop consists only of the flow element and the flow transmitter, which is directly input to the Plant Computer.

4.7.1 RWCU FLOW ELEMENT AND FLUID DENSITY TAG NUMBERS: 1(2)G31-N035 [9.7.30, 32]

MANUFACTURER: GE / Vickery Simms [9.5.27, 9.6.3]

SPAN: 0-300 GPM [9.5.27, 9.6.3]

FLOW SPAN: 0 - 200" H2 0 [9.5.27, 9.6.3]

4.7.1.1 The Reference Accuracy of the flow element is given by the Unit 2 data sheet (Reference 9.5.27) to be +/-0.5%. Per References 9.6.3 and 9.6.12, the orifice bore calculations are identical for the Unit 1 and Unit 2 flow elements. This accuracy is a result of the potential inaccuracy of the test flow used in the establishment of the discharge coefficient for the flow element. This accuracy is therefore a direct result of the test flow inaccuracy, and is in terms of %Actual Flow.

RACUFE = + 0.5000% Actual Flow 4.7.1.2 Per Section 11-11-3 of Reference 9.1.2, the installation effects for the flow element are dependent on the compliance with standards in the installation of the flow nozzle. If the guidelines of Figure 11-11-1 are met, the errors due to piping installation effects are limited to +/-0.5%, but if they are not, an additional +/-0.5% is required to be added. Per Reference 9.5.27, the line size is 4 inches. Per References 9.7.43 and 9.7.44, there are at least 10.5 feet of straight piping upstream of the orifice and at least 2 %/2feet of straight piping downstream. This equates to at least 31 diameters upstream and at least 7 diameters downstream. Reference 9.1.2 Figure 11-I1-1, item(C) is the most correct figure for assessing the acceptability of the installation. Per References 9.6.3 and 9.6.12, the Beta ratio is 0.6813 or 0.6613. Figure (C) requires approximately 21 straight upstream diameters, and approximately 3 straight pipe diameters downstream. The installation meets the requirements of Figure (C), and the additional uncertainty does not require consideration.

Per Section C.3 of Reference 9.1.1, this error is considered as an error to the discharge coefficient, which is proportional to the Actual Flow value.

Therefore, IECUFE E

- +/- 0.5000% Actual Flow

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 26 of 89 4.7.1.3 Per References 9.6.3 and 9.6.12, the flow nozzles are sized for operating conditions at approximately 100% power. The temperature used to size the flow nozzle is 545 0F, which is close to the approximate 533.7 0F or 531.2 0F for the 100% power condition. The small change in temperature will not significantly affect the characteristics of the flow orifice itself. Therefore, the effect on the measurement from the thermal expansion factor of the nozzle is negligible.

EFcUFE = Negligible 4.7.1.4 The RWCU System Flow reading is not temperature (density) compensated within the Plant Computer. Changes in operating temperature of the RWCU system inlet water affect the accuracy of the reading because the associated changes in density affect the relationship of the flow to the differential pressure for the flow orifice.

There are two specific sub-categories of density uncertainties related to these flow elements. One difference is due to the fact that the nominal heat balance density conditions at 100% power differ from the assumed conditions for the calibrations of the transmitters, which produces a bias in the measurement. The second part of this uncertainty is due to the normal random variations in density during 100% power operation. This is the random portion of this error, and it is computed separately.

Bias References 9.6.3 and 9.6.12 are the bore diameter computations for the two subject flow elements. These computations use a specific gravity of 0.74 (pressure = 1178 psig and temperature = 5450F). The calibration procedures for the transmitters (References 9.2.8 and 9.2.9) use the differential pressure span determined by the bore diameter calculation as the endpoints. Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, these conditions differ from the nominal heat balance conditions, thereby producing a bias in the flow reading. In order to convert the specific gravity to density, the following equation is used. The base condition for this computation is the density used in the calibration, not the heat balance condition.

Po = SG / vfo where: SG = the specific gravity vfo = the specific volume at the Ref. Temp. of 601F (References 9.6.3 and 9.6.12 show Ref. Temp. of 600F.)

P0 = 0.74/0.016033

= 46.1548 Ibm/ft 3

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 27 of 89 The conditions as shown by the heat balance are (1050 psia and 531.2 0 F) for Unit 1 and (1050 psia and 533.70F) for Unit 2. These conditions yield the following densities.

p1 (1050 psia and 531.21F) = 47.2456 Ibm/ft 3 p2(1050 psia and 533.7 0F) = 47.0788 Ibm/ft 3 Using normalized values from Section 6.1.6 for the nominal flow rates at 100% Power Operation, RWCU Flow (U1) = 87.9626% Flow Span RWCU Flow (U1) = 77.3741% DP Span RWCU Flow (U2) = 88.2742% Flow Span RWCU Flow (U2) = 77.9234% DP Span Per Reference 9.8.4, the input signal is multiplied by a constant to determine mass flow rate. Therefore, holding mass flow rate at a constant value equal to the heat balance figures and holding all other terms in the flow equation constant except for DP and density, the following are the effects on the differential pressure reading.

DP = K/p Ki = DPij x pO

= 77.3741% DP Span x46.1548 Ibm/ft 3

= 3571.1861% DP Span x Ibm/ft3 K2 = DP 2 ixpo

= 77.9234% DP Span x 46.15481bm/ft 3

= 3596.5389% DP Span x Ibm/ft 3 Therefore, the changes in DP readings due to the density changes, from those used for calibration, are as follows:

DPp = K / pi

= 3571.1861 / 47.2456

= 75.5877% DP Span PEb1cu = +DP,-DP11

= +75.5877- 77.3741

= -1.7864% DP Span DP 2 K2 / P2

= 3596.5389 /47.0788

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 28 of 89

= 76.3940% DP Span PEb2cu = +DP2 -DP2 1

= +76.3940- 77.9234

= -1.5294% DP Span Random In order to quantify the normal variations in density of the RWCU inlet water, a spreadsheet containing Plant Computer data for 19 runs of the heat balance on Hatch Unit 2 is included as Attachment D. These runs were recorded once per minute over a 19-minute period on November 1, 2001.

There is only 1 computer point indication for the RWCU System inlet temperature. This means that 19 RWCU System inlet temperature readings were taken during that period of time. A statistical account of the data from that spreadsheet is listed below.

Parameter FW Temp

____ ____ (Deg F)

Mean 536.3998 St Dev 0.032517 2 Std Devs 0.065034 Min 536.324 Max 536.485 Range 0.161 No. l Readings 1 Per Figure 1.2-2 of Reference 9.8.1, the readings above differ from the nominal values at 100% power. The mean is 2.6998 0F higher than the nominal reading. However, note that the maximum value minus the minimum value is 0.1 61 OF and that a 2 standard deviation figure is only 0.065034 0F.

Given all of the above information, a +50 F temperature variation around the nominal value bounds the normal operating scenario. Therefore, the density error is computed based on this variation. The base condition for this computation is the heat balance condition.

Using a similar approach to that above taken for the bias determination, the random errors are determined as follows:

p = SG/vf The temperature band to be considered for Unit 1 is 531.2 +/- 50F.

p1 -(1 050 psia and 526.2 0F) =47.5656 Ibm/ft3

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 29 of 89 pl (1050 psia and 531.2 0F) = 47.2456 Ibm/ft 3 pl+(1050 psia and 536.2 0F) = 46.9131 Ibm/ft 3 The temperature band to be considered for Unit 2 is 533.7 +/- 50 F p2-(1050 psia and 528.7 0F) = 47.4077 Ibm/ft 3 p2(1050 psia and 533.7 0F) = 47.0788 Ibm/ft 3 p2+(1050 psia and 538.7 0F) = 46.7486 Ibm/ft 3 Using normalized values from Section 6.1.6 for the nominal flow rates at 100% Power Operation, RWCU Flow (U1) = 87.9626% Flow Span RWCU Flow (Ul) = 77.3741% DP Span RWCU Flow (U2) = 88.2742% Flow Span RWCU Flow (U2) = 77.9234% DP Span Per Reference 9.8.4, the input signal is multiplied by a constant to determine mass flow rate. Therefore, holding mass flow rate at a constant value equal to the heat balance figures and holding all other terms in the flow equation constant except for DP and density, the following are the effects on the differential pressure reading.

DP = K/p K3 = DP 1, x pi

= 77.3741 % DP Span x 47.24561bm/ft 3

= 3655.5858% DP Span x Ibm/ft3 K4 = DP 21x p2

= 77.9234% DP Span x 47.07881bm/ft 3

= 3668.5402% DP Span x Ibm/ft3 Therefore, the changes in DP readings, due to the density changes from those used for calibration, are as follows:

DP. = K3 / pi-

= 3655.5858 / 47.5656

= 76.8536% DP Span PEr 1 = +DPt-DPli

= +76.8536- 77.3741

= -0.5205% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 30 of 89 DP,+ = K 3 / pI+

= 3655.5858 /46.9131

= 77.9225% DP Span PEri+ = +DP1+-DP11

= +77.9225- 77.3741

= +0.5484% DP Span The larger of the errors is used and treated as a random uncertainty.

PECUFEl I= +/-0.5484% DP Span DP2 = K 4 /p 22-

= 3668.5402 /47.4077

= 77.3828% DP Span PEr2 = +DP2 . - DP 2 1

= +77.3828- 77.9234

= -0.5406% DP Span DP2+ = K 4 /p2+

= 3668.5402 / 46.7486

= 78.4738% DP Span PEr2+ = +DP2+-DP21

= +78.4738- 77.9234

= 0.5504% DP Span The larger of the errors is used and treated as a random uncertainty.

PECUFE2 = +/-0.5504% DP Span 4.7.2 RWCU FLOW TRANSMITTER TAG NUMBER: 1(2)G31 -N036 [9.7.17, 20, 30-33]

MANUFACTURER: BARTON [9.5.18,19]

MODEL NUMBER: 764 [9.5.18,19]

SPAN: 0-300 GPM (0-200"H 2 0) [9.5.18,19]

4.7.2.1 Per Reference 9.4.3, the Analyzed Drift (DA) for the flow transmitter is +/-

1.577% DP Span for a period of 30 months between calibrations. This includes the effects of Drift (DR), Reference Accuracy (RA) and Measurement & Test Equipment (M&TE).

DAcuF7 = +1.577% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 31 of 89 4.7.2.2 Setting Tolerance effects are due to the flexibility of the technician in the calibration process. The effect is bounded by the As-Left setting tolerance of the device. Per References 9.2.8, 9.2.9, 9.2.11 and 9.2.12, the As-Left setting tolerance is +/- 0.5000% Span.

STcuFr = +/- 0.5000% DP Span 4.7.2.3 Per Reference 9.6.5, the flow transmitter has a static pressure effect of 0.5% of maximum span per 1000 psig. Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the enthalpy of the water at the inlet to the RWCU is computed using a pressure of 1050 psia. Therefore, the effect is computed conservatively using 1050 psia during operating conditions. Calibration is performed at 14.7 psia. Conservatively using 100% span to compute this figure, we obtain the following Static Pressure Effect (SPE).

SPEcuFr = +/- 0.5% DP Span x (1050 -14.7) / 1000 psi SPEcuF7 = +/- 0.5177% DP Span 4.7.2.4 Per Reference 9.6.5, the flow transmitter Power Supply Effect (PSE) is given as less than +/-0.05% Span per Volt, and the load effect is shown as 0.1%

Span for a 100-ohm change. The load effect is only significant when changing between calibration and operating conditions, since the loads on the transmitter are different. The 100-ohm change is judged to be an adequate value for this computation. Per Reference 9.5.28 and Assumption 5.4.1, the power supplies are identified as GE/B&W/Bailey 570-06. Per Reference 9.6.6, these power supplies have an output of 52.5 VDC +/- 8%.

Therefore, the flow transmitter Power Supply Effect (PSEcuFr) is given as:

PSEcuFT = +/- {[(0.05% Span / VDC)(52.5 VDC)(.08)] 2 + [0.1%

2 0 Span] } .5 PSEcuFT = +/- 0.2326% DP Span 4.7.2.5 Per Reference 9.6.5, the flow transmitter Temperature Effect (TE) is given as

+/- 1.0% maximum span / 100F, for the temperature range of 400F to 150 0F.

Per Reference 9.8.5, the transmitters are located at elevation 158' within the Reactor Building. Per Reference 9.8.7, the maximum normal operating temperature at these locations is 100 0F. Therefore, the maximum difference in temperature between operating conditions and calibration conditions is 500F. Therefore, the temperature effect is computed as follows:

TEcuFr = 1.0% Span x (500 F/1 00F)

TEcuFT = 0.5000% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 32 of 89 4.8 RWCU INLET I OUTLET TEMPERATURE UNCERTAINTY CONSIDERATIONS Per References 9.7.18, 19, 22, 23, and 30-33, the RWCU Inlet and Outlet Temperature instrument loops consist only of the temperature elements, which are directly input to the Plant Computer.

4.8.1 RWCU INLET / OUTLET TEMPERATURE ELEMENTS TAG NUMBER: 1(2)G31-N004, N015 [9.7.18,19, 22, 23, 30-33]

MANUFACTURER: NECI [9.5.16-17, 5.2]

TYPE: Cu / Con T/C [9.5.16-17, 5.2]

RANGE: 0-6000 F [9.5.16-17, 5.2]

4.8.1.1 Per References 9.5.15, 9.5.16 and 9.5.17, the rated accuracy of these thermocouples is 0.75%. This agrees with References 9.7.45 and 9.7.46.

The normal temperature for the RWCU water is 4340 F or 436.8 0F per Figures 1.2-2 and 1.2-3 of Reference 9.8.3. For conservatism, 2250C (437 0F) is used for the computation. Therefore, the Reference Accuracy (RA) of the temperature element is:

RAcUTE = +/- (0.75/100) x 225 0C x (1.81F/ 0C)

RACUTE = +/- 3.03750 F 4.8.1.2 The thermocouple has no adjustment and therefore cannot be calibrated.

Therefore, the errors that can be introduced during calibration (Setting Tolerance and M&TE) do not apply to this device.

STRWCUTE = N/A M&TERWCUTE = N/A

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 33 of 89 4.9 RECIRCULATION PUMP POWER UNCERTAINTY CONSIDERATIONS Per References 9.7.16,17, 20, 21, and 35-40, the Recirculation Pump power signal is fed directly to the Plant Computer from the watt transducers.

4.9.1 PUMP WATT TRANSDUCER TAG NUMBER: 1(2)B31-R771, R772 [9.7.16,17,20, 21, 35-40]

MANUFACTURER: Ohio Semitronics [9.5.9,10]

MODEL NUMBER: PC5-004B [9.5.9,10]

RANGE: 0-8.4 MW 4.9.1.1 Per Reference 9.6.8, the accuracy of the PC5 Watt Transducer is +/-0.5% Full Scale, including the effects of power factor, linearity, repeatability, and current sensor. Therefore, the Reference Accuracy, based on a 100% reading, is shown as follows.

RARPWT = +/- 0.5000% Span 4.9.1.2 Setting Tolerance effects are due to the flexibility of the technician in the calibration process. The effect is bounded by the As-Left setting tolerance of the device. Per Reference 9.2.10, the As-Left setting tolerance is +/- 0.5%

Span.

STRPWT = +/- 0.5000% Span 4.9.1.3 Measurement & Test Equipment (M&TE) are errors introduced during the calibration process. The M&TE effect is a function of the accuracy of the test equipment used during calibration. M&TE is chosen to be at least as accurate as the equipment being calibrated, and generally more accurate.

This value is conservatively assigned to be equal to the As-Left setting tolerance of the transducers, in order to maximize the flexibility in choice of test equipment by the technician.

M&TERPWT = +/- 0.5000% Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 34 of 89 4.9.1.4 Per Reference 9.6.8, the Temperature Effect of the PC5 Watt Transducer is +/-1.0% Reading, +/-0.10% Full Scale for a temperature range of -100C to 600C. Per Reference 9.8.5, the transmitters are located at elevation 158' within the Reactor Building. Per Reference 9.8.7, the maximum normal operating temperature at these locations is 100F. Therefore, the ambient temperature of these devices is well within these limits. For conservatism, the two terms are combined via SRSS assuming a 100% scale reading.

TERPWT +/-+/- [(1 .)2 + (0.1)2]1/2 % Span TERPWT = +/- 1.0050% Span 4.10 MISCELLANEOUS HEAT BALANCE ANALYSIS INPUTS 4.10.1 PLANT COMPUTER (PC) INPUT UNCERTAINTIES 4.10.1.1 Per Section 2.3.2.1 of Reference 9.6.9, the Analog Input card with the highest values of gain accuracy and linearity is the AC 4050 to AC 4060. The Gain Accuracy is shown to be +/- 0.01% Full Scale. The linearity is given as 0.01 5%

Full Scale. The scale ranges for each signal are programmed into the processor, such that the input spans are equal to the full-scale values. The errors are combined in SRSS fashion.

RApc = +/- [(0.01)2 + (0.015)2]112 % Span RApc = +/- 0.0180% Span 4.10.1.2 Setting Tolerance effects are errors introduced during the calibration process.

Per Section 2.3.1.5 of Reference 9.6.9, the analog input cards incorporate a separate monitoring channel, which allows the user to conduct periodic checks and calibration routines on the system. This allows self-monitoring and calibration by the Plant Computer. Because of this, and the fact that these routines are performed on all Plant Computer input hardware, the calibration process is highly accurate. Therefore, the values of Setting Tolerance and M&TE are negligible with respect to the other error terms.

STpc = M&TEPc = Negligible 4.10.1.3 Because of the internal system performance checks during operation, significant instrument drift would be detected very quickly within the Plant Computer. Therefore, instrument drift is not applicable for the Plant Computer.

DRPc = N/A

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 35 of 89 4.10.1.4 Per Reference 9.6.9, no temperature effect specification is listed for the Plant Computer. Because the Plant Computer equipment is kept in a very controlled temperature environment, the variation in temperature is very small. For these reasons, the temperature effect for the Plant Computer is negligible when compared to the other uncertainty terms associated with the instrument loops.

TEpC = Negligible 4.10.1.6 The Resolution effect for the Plant Computer input card is conservatively treated as +/- 1 LSB. Per Reference 9.6.9, the A/D converters for the Plant Computer are 14-bit converters. However, per discussions with plant personnel, only 12 bits are used in the conversion process. The value is computed as follows:

RESPc= +/- 1 x [1/212] X 100%

RESpc= +/- 0.0244% Span 4.10.1.7 The computational error of the Plant Computer involves the errors in the algorithms and conversions which are produced in the Plant Computer software. Per References 9.6.10 and 9.8.4, the steam table utility functions are performed in accordance with the ASME steam tables or Keenan &

Keyes. These are very accurate sources, and correct for the applications, such that any potential minor error in the tables themselves are negligible with respect to other Heat Balance errorlterms. Other functions are performed digitally to many significant digits. Therefore, the computational errors associated with the Plant Computer are negligible with respect to the other instrument uncertainties present in the instrument loops.

COMPC =p Negligible

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 36 of 89 4.10.2 HEAT BALANCE CORRECTION FACTOR (CF) INPUT UNCERTAINTIES 4.10.2.1 Per Reference 9.8.10, the normal QRAD (radiative heat loss) used for Hatch is 1.1 MW. Per Reference 9.2.1, a Correction Factor (CF) of 2.0 MW is added to the manual heat balance computation to account for radiative heat losses and CRD Flow correction. Therefore, 0.9 MW is conservatively added to the manual total power computation to account for CRD Flow correction and other non-instrumented losses. Per Reference 9.8.4, the term used with the heat balance computation on the Plant Computer is labeled QRADX. At the present time, the value of this constant is 1.7 MW, and is intended to account for QRAD (radiative heat losses) and other non-instrumented losses, as documented in Reference 9.8.18. The QRAD term included in QRADX is 1.1 MW, and an additional approximate 0.6 MW is included to account for other non-instrumented losses, such as:

a) Recirculation Pump Seal Inflow, b) Reference Leg Keepfill System Flow, c) Inflow Through RWCU Seals, d) RWCU Leakage, and e) Significant Variations in CRD Temperature. (This term was included prior to installation of CRD Temperature measurement on the plant computer, and no longer exists.)

The adjustment to QRADX for the non-instrumented losses is assessed periodically to ensure that a conservative value is being used. Therefore, any errors in these measurements are in the conservative direction, such that they cause the Core Thermal Power measurement to be higher than actual.

Therefore, these uncertainties do not need to be considered in this calculation.

Section 4.5.1.4 of this calculation computes error terms that fully account for errors in CRD flow indication, due to density differences between the calibration assumed values and the actual process values. Therefore, CRD flow measurement errors are fully accounted for. Since the manual heat balance Correction Factor (CF) is larger than the Plant Computer adjustment, QRADX, even more conservatism exists with this measurement, and no uncertainties from the additional non-instrumented losses need to be considered in this calculation.

4.10.2.2 Per Section 4.10.2.1, QRAD (radiative heat losses) used for Hatch is 1.1 MW, both in the manual and Plant Computer generated heat balance computations. This value is derived from Reference 9.8.10, which also states that the uncertainty in this value is large, up to 50% of the value, or +/-0.55 MW. Therefore, this uncertainty is considered in the computation of measurement uncertainty for the core thermal power computation.

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 37 of 89 TUORAD = +0.5500 MW 5.0 ASSUMPTIONS AND ENGINEERING JUDGMENTS 5.1 Per Reference 9.5.28, the power supply for the Unit 2 RWCU flow transmitter instrument loop is a GE/Bailey/B&W Model 570-06. The data sheet series S-18454 for Unit 1 was not available for reference, but this same model number is shown in NUCLEIS for Unit 1. Since the designs for this instrumentation appears to be identical between units, and since the model is the same in NUCLEIS, it is assumed that the Unit 1 power supply is the same model number.

5.2 Per References 9.5.16-17, the Unit 2 RWCU Temperature Element is an NECI Cu / Con thermocouple with a range of 0-6000F. The data sheet series S-1 8454 for Unit 1 was not available for reference. Reference 9.5.15 shows the purchase specification requirements to be the same for Unit 1 as shown on References 9.5.16-17. Since the designs for this instrumentation appear to be identical between units, it is assumed that the Unit 1 thermocouples have similar performance characteristics to the Unit 2 thermocouples.

5.3 Per References 9.3.1 and 9.3.2 and per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the calibration tables for the transmitters are not performed for the exact heat balance conditions. It is assumed that at the completion of this project, the transmitter calibration calculations, References 9.3.1 and 9.3.2, will be revised to reflect the projected heat balance conditions after the power uprate.

5.4 Because the present heat balance conditions will not change significantly, this analysis is performed using the present heat balance conditions. Because the power uprate is of such a small magnitude (only 1.5%), the uncertainty values derived herein are valid after power uprate.

5.5 References 9.5.7 and 9.5.8 show a range of 0-9000 F for the CRD System Inlet thermocouples. However, these references also identify these as Omega Type T thermocouples. Per Reference 9.8.9, a type T thermocouple maximum output is at 7520 F. Since these thermocouples are reading CRD System temperature, which is normally 1230 F, the range of these thermocouples is established as 0 to the maximum temperature, 752 0F.

5.6 Although specific gravities and enthalpies are a function of both pressure and temperature, they are significantly more affected by pressure for steam and by temperature for subcooled water. When determining the weighting factors for the steam pressure uncertainties, the effects on the steam enthalpy were considered. However, the Plant Computer also uses steam pressure in the determination of enthalpy for the feedwater temperature, CRD water temperature, and RWCU water temperature. Because pressure has such an insignificant affect on enthalpy and since the relative contributions to the thermal

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 38 of 89 power computation are very small, these effects are considered negligible.

5.7 Various of the vendor documents for flow elements researched by this calculation express accuracy for the flow elements in terms of %. Although flow element output is actually a differential pressure reading, the error in the flow element is generally considered to be the error of the calibration process, which establishes a given flow rate and determines an appropriate discharge coefficient, which would allow conversion of the differential pressure from the flow element to the correct flow rate, within the stated accuracy. Since the error is in the discharge coefficient, and since the error of the flow element is zero at zero flow conditions, the error is proportional to actual flow rate, not a span value. Therefore, the % accuracy values are established as % Actual Flow for the purposes of this calculation. Interpretation of this error as proportional to flow versus differential pressure is conservative at the subject flow rates.

5.8 Per recent industry direction with regard to Reference 9.1.1, random errors that share a common cause (such as common environment or M&TE) are not considered dependent. Therefore, the dependent errors are limited to those errors that directly affect more than one factor of the heat balance computation.

For this calculation, the density errors for the Feedwater Flow Element, which are produced via the temperature compensation algorithm within the plant computer, are considered dependent with the Feedwater Temperature error used for the development of Feedwater enthalpy. Therefore, these two errors are added prior to the SRSS process, per Reference 9.1.1. This only applies to Case 1, since UFM correction eliminates the Feedwater Flow density error, which is based on Feedwater Temperature.

5.8 Per Reference 9.8.8, the Feedwater Flow venturi instrument loop will be continuously corrected, based on the readings of the Westinghouse Crossflow UFMs. New correction factors are implemented no less often than once per minute. This operation reduces the measurement error for Feedwater Flow to that of the UFMs alone. All errors of the Feedwater Flow venturi instrument loop are calibrated out, to within the error of the UFMs. Therefore, no Feedwater Flow venturi instrument loop errors are considered in Case 2.

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 39 of 89 6.0 ANALYSIS Computations are performed to an accuracy of several significant digits, but presented in this study rounded to four decimal places in most cases. Hand verification of this study utilizing the rounded values could result in slightly different results due to round off errors. Final answers are rounded to three decimal places.

6.1 ESTABLISHMENT OF BASELINE CONDITIONS Per Section 2.0, QCORE = QFW + QCR + Qcu + CF - Qp where: QCORE = Thermal Power Generated by Reactor Core QFW = Heat Applied to Feedwater to Steam Process QCR = Heat Applied to Control Rod Drive System QCU = Heat Applied to Reactor Water Clean Up System CF = Correction Factor, Which Includes the Effects of Radiative Heat Loss and CRD Flow Corrections QP = Heat Added by Recirculation Pumps Figures 1.2-2 and 1.2-3 of Reference 9.8.1 establish the baseline Reactor System Heat Balance conditions for 100% rated conditions (2763 MWt) for each Unit. The equations for figuring the heat balance, in process units, are identical, since the physical parameters are the same. There are small differences, however, in the baseline heat balance 100% power conditions.

Where the exact parameters are given directly on the figures, no computations are required. However, where the exact parameters are not given, computations are performed to determine the baselines.

6.1.1 BASELINE CONDITIONS FOR FEEDWATER FLOW Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the following values are obtained for 100% Power operating conditions.

Feedwater Flow 11.514 Mlbm I hr (Unit 1) 5.757 Mlbm / hr (Unit 1) per loop 11.950 Mlbm / hr (Unit 2) 5.975 Mlbm / hr (Unit 2) per loop These values need to be converted to units of differential pressure, in order to apply the uncertainty values and observe the affect on the heat balance

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 40 of 89 calculation. In order for this to be done, density must be determined.

Per Design Input 4.1.2, the nominal feedwater flow upstream venturi tap pressures are as follows.

Pnoz1= 1116 psia Pnoz2= 1122 psia Attachment C is a tabular listing of the values with References 9.3.1 and 9.3.2, as computed using the upstream tap pressures shown above, and the values of other parameters defined by Figures 1.2-2 and 1.2-3 of Reference 9.8.1.

The value, hs, is the full-scale differential pressure, as computed at operating conditions, which produces an indication of 8 Mlbm/hr in each feedwater flow loop. As computed in Attachment C, the four hs values are as follows:

1C32N002A 2405.70 inWC 1C32N002B 2395.17 inWC 2C32N002A 2362.53 inWC 2C32N002B 2347.23 inWC Since each of the differential pressure values correspond to the same flow rates, and since most of the errors for the flow loops are expressed in % DP Span, the analysis can be performed on any of the instrument loops. In a normalized fashion, given that the density of the feedwater does not change, the flow is related to the differential pressure as a square root function, such that:

C = (A)1/2 Where C is a normalized fraction of the full-scale flow rate of 8 Mlbm/hr. A is the normalized fraction of full-scale differential pressure.

Unit 1 The heat balance flow rate through each feedwater flow loop is 5.7570 Mlbm/hr, as shown in Figure 1.2-3 of Reference 9.8.1. This corresponds to a C value of C(100% Power) = 5.7570/ 8 = 0.719625 Expressed as a percentage, C(100% Power) = 71.9625% Flow Span Therefore, the A value is computed as follows:

A(100% Power) = c2 = (0.719625)2 = 0.517860

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 41 of 89 Expressed as a percentage, A(1 00% Power) = 51.7860% DP Span Unit 2 The heat balance flow rate through each feedwater flow loop is 5.9750 Mlbm/hr, as shown in Figure 1.2-2 of Reference 9.8.1. This corresponds to a C value of C(100% Power) = 5.9750/8 = 0.746875 Expressed as a percentage, C(100% Power) = 74.6875% Flow Span Therefore, the A value is computed as follows:

A(100% Power) = c2 = (0.746875)2 = 0.557822 Expressed as a percentage, A(1 00% Power) = 55.7822% DP Span 6.1.2 BASELINE CONDITIONS FOR FEEDWATER TEMPERATURE Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the following values are obtained for 100% Power operating conditions.

Feedwater Temperature 397.5 0F (Unit 1) 425.1'F (Unit 2)

Also, per the same references, the baseline enthalpies of the feedwater are 373.4 BTU/lbm (Unit 1) and 403.2 BTU/lbm (Unit 2).

6.1.3 BASELINE CONDITIONS FOR REACTOR PRESSURE Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the following values are obtained for 100% Power operating conditions.

Reactor Pressure 1050 psia (Both Units)

For saturated conditions, per Reference 9.8.2, the baseline condition of steam enthalpy is 1191 BTU/lbm.

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 42 of 89 6.1.4 BASELINE CONDITIONS FOR CONTROL ROD DRIVE (CRD) SYSTEM FLOW Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the following values are obtained for 100% Power operating conditions.

Control Rod Drive System Flow 30,000 Ibm/hr (Both Units)

The temperatures are as shown in Section 6.1.5, which yield densities of the following, given a pressure of approximately 1050 psia.

p1 (1050 psia and 123.2 0F) = 61.8582 Ibm/ft3 p2(1050 psia and 123.5 0F). = 61.8525 Ibm/ft3 Therefore, flow rate is equal to the following:

Flow (vol) = Flow (mass) x (7.48052 gal / ft3 )/ (p x (60 min/hr))

Flowl (vol) = Flowl (mass) x (7.48052 gal / ft3)/ (PlCRD x (60 min/hr))

Flowl (vol) = (30000 Ibm/hr) x (7.48052 gal/ft3 )/ (61.8582 Ibm/ft 3 x (60 min/hr))

Flowl (vol) = 60.4651 GPM Flow2 (vol) = Flow2 (mass) x (7.48052 gal / ft3 )/ (p2 CRD x (60 min/hr))

Flow2 (vol) = (30000 lbmlhr) x (7.48052 gal / ft3)/ (61.8525 Ibm / ft3 x (60 min/hr))

Flow2 (vol) = 60.4706 GPM The differential pressure transmitters are calibrated to measure 0-100 GPM,.

Therefore, expressing each flow rate as in units of % Flow Span, and then in %

DP Span, CRD Flow (Ul) = 60.4651% Flow Span CRD Flow (Ul) = (0.604651)2 x 100% DP Span CRD Flow (Ul) = 36.5603% DP Span CRD Flow (U2) = 60.4706% Flow Span CRD Flow (U2) = (0.604706)2 x 100% DP Span CRD Flow (U2) = 36.5669% DP Span 6.1.5 BASELINE CONDITIONS FOR CONTROL ROD DRIVE (CRD) SYSTEM TEMPERATURE Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the following values are obtained for 100% Power operating conditions.

Control Rod Drive Sys Temp 123.2 0F (Unit 1) 123.5 0F (Unit 2)

Attachment 1 to HL-6328, Enclosure 7 E. 1.HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 43 of 89 6.1.6 BASELINE CONDITIONS FOR REACTOR WATER CLEANUP (RWCU)

SYSTEM FLOW Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the following values are obtained for 100% Power operating conditions.

RWCU Flow 100,000 Ibm/hr (Both Units)

The flow rate is obtained in the inlet to RWCU from the Reactor. Therefore, the inlet temperatures are used for temperature determination. As seen in Section 6.1.7, the inlet temperatures are 531.2 0 F (Unit 1) and 533.71F (Unit 2).

Using a pressure of 1050 psia for these measurements, per Reference 9.8.2, the densities are determined as follows:

p1 (1050 psia and 531.2 0 F) = 47.2456 Ibm/ft3 p2(1050 psia and 533.7 0F) = 47.0788 Ibm/ft3 Therefore, flow rate is equal to the following:

Flow RWCU (vol) = Flow (mass) x (7.48052 gal / ft3)/ (p x (60 min/hr))

Flow1 RWCU (vol) = Flowl (mass) x (7.48052 gal / ft3 )/ (p1 x (60 min/hr))

Flow1 RWCU (vol) = (100000 Ibm / hr) x (7.48052 gal I ft3)/ (47.2456 Ibm / ft3 x (60 min/hr))

Flow1 Rwcu (vol) = 263.8877 GPM FIOW2RWCU (vol) = Flow2 (mass) x (7.48052 gal / ft3 )/ (q2 x (60 min/hr))

Flow2RWCU (vol) = (100000 Ibm / hr) x (7.48052 gal / ft )/ (47.0788 Ibm / ft3 x (60 min/hr))

Flow2RWCU (vol) = 264.8227GPM The differential pressure transmitters are calibrated to measure 0-300 GPM.

Therefore, expressing each flow rate as in units of % Flow Span, and then in %

DP Span, RWCU Flow (Ul) = (263.8877 GPM / 300 GPM) x 100% Flow Span RWCU Flow (Ul) = 87.9626% Flow Span RWCU Flow (Ul) = (0.879626)2 x 100% DP Span RWCU Flow (Ul) = 77.3741% DP Span RWCU Flow (U2) = (264.8227GPM / 300 GPM) x 100% Flow Span RWCU Flow (U2) = 88.2742% Flow Span RWCU Flow (U2) = (0.882742)2 x 100% DP Span I RWCU Flow (U2) = 77.9234% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 44 of 89 6.1.7 BASELINE CONDITIONS FOR REACTOR WATER CLEANUP (RWCU)

SYSTEM INLET AND OUTLET TEMPERATURES Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the following values are obtained for 100% Power operating conditions.

RWCU Inlet Temperature 531.2 0F (Unit 1) 533.70 F (Unit 2)

RWCU Outlet Temperature 434.0F (Unit 1) 436.8 0F (Unit 2) 6.1.8 BASELINE CONDITIONS FOR RECIRCULATION PUMP POWER Unit 1 From Figure 1.2-3 of Reference 9.8.1, the Reactor Recirculation System Flow rate is 34.2E6 Ibm/hr, with a delta H of 0.8 Btu/Ibm. Computing the recirculation pump power released to the water, Power Delivered = 34.2E6 Ibm / hr x 0.8 BTU / Ibm = 27.36E6 BTU / hr Converting this term to units of Watts, we obtain the following:

Power Delivered = 27.36E6 BTU / hr x (1hr/60 min) x (17.5796 Watts

/(BTU/min))

= 8.0163 MW Using an efficiency of 93% for the pumps (per Reference 9.2.1), this would equate to continuous power outputs from each of the two pumps of:

Watt Transducer Output (Each Pump) = (8.0163MW / 0.93) / 2 = 4.3098 MW Unit 2 From Figure 1.2-2 of Reference 9.8.1, the Reactor Recirculation System Flow rate is 34.3E6 Ibm/hr, with a delta H of 0.8 Btuflbm. Computing the recirculation pump power released to the water, Power Delivered = 34.3E6 Ibm / hr x 0.8 BTU / Ibm = 27.44E6 BTU / hr Converting this term to units of Walts, we obtain the following:

Power Delivered = 27.44E6 BTU / hr x (1hr/60 min) x (17.5796 Watts

/(BTU/min))

= 8.0397 MW

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 45 of 89 Using an efficiency of 93% for the pumps (per Reference 9.2.1), this would equate to continuous power outputs from each of the two pumps of:

Watt Transducer Output (Each Pump) = (8.0397MW / 0.93) /2 = 4.3224 MW.

6.2 DETERMINATION OF WEIGHTING FACTORS Inthe determination of the weighting factors, 100% thermal power is computed based on a value of 2763 MW thermal, per Figures 1.2-2 and 1.2-3 of Reference 9.8.1. Each parameter in the equation is varied by a nominal value to determine these factors, per the methodology of Section 3.0.

6.2.1 FEEDWATER FLOW WEIGHTING FACTOR For the Feedwater Flow measurement, all errors are expressed in terms of %DP Span, except for the Reference Accuracy and Installation Effect for the Flow Element itself, which are expressed in terms of % Actual Flow. All of the errors are random. The weighting factors for the errors expressed in % Actual Flow are determined in Section 6.2.9. The weighting factors for the errors expressed in % DP span are derived below.

Unit 1 Per Section 6.1.1, the nominal Feedwater Flow and differential pressure for Unit 1 are:

A(100% Power) = 51.7860% DP Span C(100% Power) = 71.9625% Flow Span A nominal +/-5% uncertainty value results in a band as follows:

46.7860% DP Span < A < 56.7860% DP Span The square root function produces the following in terms of % Flow Span.

68.4003% Flow Span < C < 75.3565% Flow Span.

The error in this value is as follows:

+ 3.3940/ -3.5622% Flow Span Because of the very nearly equal values, for symmetry, the largest uncertainty value is used in both directions. Therefore, a 5% DP Span error for the feedwater flow signals results in the following:

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 46 of 89

+ 3.5622% Flow Span Converting this value to process units,

+ 284976.0000 Ibm/hr The two loop flow rates are summed in the Plant Computer, with each of the loops having uncertainties as expressed above. This error is SRSS'ed to obtain the values shown.

Total Flow Error = ((284976.0000)2 + (284976.0000)2)1/

= +403016.9242 Ibm/hr Per Section 6.1.3 and Figure 1.2-3 of Reference 9.8.1, the difference in enthalpy from feedwater to steam in the reactor at 100% power is as follows:

DH = 1191-373.4 = 817.6 BTU/lbm Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW) = Error (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= +/-403016.9242 Ibmlhr x 817.6 BTU/lbm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +/-96.5432 MW Converting to a percentage, Error (%) (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) +/-3.4941% RTP The weighting factor is therefore determined as follows:

WFFWF1 = Error (% RTP) / Error (% DP Span)

= 3.4941 % RTP / 5% DP Span

= 0.6988% RTP / % DP Span Unit 2 Per Section 6.1.1, the nominal feedwater flow and differential pressure for Unit 2 are:

A(100% Power) = 55.7822% DP Span C(100% Power) = 74.6875% Flow Span

Attachment 1 to HL-6328, Enclosure 7 E. 1.HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 47 of 89 A nominal +/-5% uncertainty value results in a band as follows:

50.7822% DP Span < A < 60.7822% DP Span The square root function produces the following in terms of % Flow Span.

71.2616% Flow Span < C < 77.9629% Flow Span.

The error in this value is as follows:

+ 3.2754 / -3.4259% Flow Span Because of the very nearly equal values, for symmetry, the largest uncertainty value is used in both directions. Therefore, a 5% DP Span error for the feedwater flow signals results in the following:

+ 3.4259% Flow Span Converting this value to process units,

+ 274072.0000 Ibm/hr The two loop flow rates are summed in the Plant Computer, with each of the loops having uncertainties as expressed above. This error is SRSS'ed to obtain the values shown.

Total Flow Error = + ((274072.0000)2 + (274072.0000)2)12

= + 387596.3395 Ibm/hr Per Section 6.1.3 and Figure 1.2-2 of Reference 9.8.1, the difference in enthalpy from feedwater to steam in the reactor at 100% power is as follows.

DH = 1191-403.2 = 787.8 BTU/lbm Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation.

Error (MW) = Error (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= +/-387596.3395 Ibm/hr x 787.8 BTU/Ibm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

- +/-89.4650 MW

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 48 of 89 Converting to a percentage, Error (%) = +/- (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = + 3.2380% RTP The weighting factor is therefore determined as follows.

WFFwF2 = Error (% RTP) / Error (% DP Span)

= 3.2380% RTP /5% DP Span

= 0.6476% RTP I % DP Span 6.2.2 FEEDWATER TEMPERATURE WEIGHTING FACTOR The Feedwater Temperature measurement is only used in the Heat Balance equation for determining Feedwater enthalpy. A 5% error for the Feedwater Temperature measurement equates to an error of +/-7.50F, since the span of the instrument is 150 0F.

Unit 1 Per Figure 1.2-3 of Reference 9.8.1, the Feedwater Temperature at 100% power conditions is 397.50F, and the nominal Feedwater Flow is 11514000 Ibm/hr. Per Design Input 4.1.2, the nominal upstream tap pressure for Unit 1 is as follows.

Pnozi = 1116 psia The temperature range of concern is 397.5 +/- 7.50F, or 3900 F S TFW < 4050F Assuming no change in pressure, the enthalpy computed for these temperature values are as shown, per Reference 9.8.2.

H(390 0F, 1116psia) = 365.4224 BTU/lbm H(397.5 0F, 111 6psia) = 373.4300 BTU/Ibm H(405 0F, 111 6psia) = 381.4726 BTU/lbm The thermal power computation for the feedwater is performed by the following equation:

QFW= DH (BTU/lbm) x Flow (Ibm/hr) x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 49 of 89 The DH value is the difference between the steam enthalpy in the reactor and the feedwater enthalpy. For this exercise, steam enthalpy is constant, so the only change is in feedwater enthalpy. Therefore, the error in the power signal is determined by setting the DH term equal to the error in the enthalpy of the feedwater and solving for the resulting power error.

DQFwr(-) = (365.4224-373.43)BTU/lbm x 11514000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= -27.0138 MW DQFWT(+) = (381.4726-373.43)BTU/Ibm x 11514000 Ibm/hr (1hr/60 min) x (17.5796 Walts /(BTU/min)) x (1 MW / 1 E6 W)

= +27.1319 MW Due to the symmetry of the error, the error is conservatively established as the largest value, expressed in both directions.

Error (MW) = +/- 27.1319 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +/-0.9820% RTP The weighting factor is therefore determined as follows:

WFFwT1 = Error (% RTP) / Error (IF)

= 0.9820% RTP / 7.5 0F

= 0.1309% RTP /OF Unit 2 Per Figure 1.2-2 of Reference 9.8.1, the Feedwater Temperature at 100% power conditions is 425.1 0F, and the nominal Feedwater Flow is 11950000 Ibm/hr. Per Design Input 4.1.2, the nominal upstream tap pressure for Unit 2 is as follows.

Pnoz2 = 1122 psia The temperature range of concem is 425.1 +/- 7.5 0F, or 417.6°F < TFW < 432.6 0F Assuming no change in pressure, the enthalpies computed for these temperature values are as shown, per Reference 9.8.2.

H(417.6 0 F, 1122psia) = 395.0752 BTU/lbm H(425.1 0 F, 1122psia) = 403.2295 BTU/Ibm

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 50 of 89 H(432.61F, 1122psia) = 411.4183 BTU/Ibm The thermal power computation for the feedwater is performed by the following equation:

QW = DH (BTU/lbm) x Flow (Ibm/hr) x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

The DH value is the difference between the steam enthalpy in the reactor and the feedwater enthalpy. For this evaluation, steam enthalpy is constant, so the only change is in feedwater enthalpy. Therefore, the error in the power signal is determined by setting the DH term equal to the error in the enthalpy of the feedwater and solving for the resulting power error.

DQFwT(-) = (395.0752-403.2295)BTU/lbm x 11950000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= -28.5504 MW DQFWT(+) = (411.4183-403.2295)BTU/lbm x 11950000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +28.6712 MW Due to the symmetry of the error, the error is conservatively established as the largest value, expressed in both directions.

Error (MW) = +/- 28.6712 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +/-1.0377% RTP The weighting factor is therefore determined as follows:

WFFw,2 = Error (% RTP) / Error (OF)

= 1.0377% RTP / 7.5 0F

= 0.1384% RTP /OF 6.2.3 REACTOR PRESSURE WEIGHTING FACTOR The Reactor Pressure measurement error is only considered in the Heat Balance computation for determining Steam enthalpy. (See Assumption 5.6.) A 5% error for the Reactor Pressure measurement equates to an error of +/-60 psi, since the span of the instrument is 1200 psi. The Steam enthalpy is used both in the determination of power from feedwater and from the CRD system flow.

Therefore, both these items are included in this assessment.

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 51 of 89 Unit 1 The steam enthalpy is figured from the saturation tables. Per Reference 9.8.2, the steam enthalpies for the different applicable values of steam pressure are as follows:

H( 11 0 psia) = 1188.68 BTU/lbm H(1 050 psia) = 1191.00 BTU/lbm H(990 psia) = 1193.26 BTU/lbm Therefore, a positive Reactor Pressure error causes a negative steam enthalpy error, because saturation conditions are assumed.

The thermal power computations for the Feedwater and CRD System are performed by the following equation:

Qs = DH (BTU/lbm) x Flow (Ibm/hr) x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

The DH value is the difference between the steam enthalpy in the reactor and the feedwater enthalpy (or the CRD System water enthalpy). For this evaluation, feedwater and CRD enthalpies are constant, so the only change is in steam enthalpy. Therefore, the error in the thermal power measurements is determined by setting the DH term equal to the error in the enthalpy of the steam and solving for the resulting power error. Since the same enthalpy error exists for both systems, the total error due to the steam enthalpy error is computed by combining flow rates for the Feedwater and CRD systems. Per Figure 1.2-3 of Reference 9.8.1, Flow = 11514000 Ibm/hr + 30000 Ibm/hr

= 11,544,000 Ibm/hr DQs(-) = (1193.26-1191)BTU/lbm x 11,544,0001bm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +7.6440 MW DQs(+) = (1188.68-1191)BTU/lbm x 11,544,00O0bm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= -7.8470 MW Due to the symmetry of the error, the error is conservatively established as the largest value, expressed in both directions.

Errors (MW) = +/- 7.8470 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error(%) = +0.2840% RTP

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 52 of 89 The weighting factor is therefore determined as follows:

WFsp = Error (% RTP) / Error (% Span)

= 0.2840% RTP /5 % Span

= 0.0568% RTP / % Span Unit 2 The steam enthalpy is figured from the saturation tables. Per Reference 9.8.2, the steam enthalpies for the different applicable values of steam pressure are as follows:

H(1 110 psia) = 1188.68 BTU/lbm H(1050 psia) = 1191.00 BTU/lbm H(990 psia) = 1193.26 BTU/lbm Therefore, a positive Reactor Pressure error causes a negative steam enthalpy error at saturated conditions.

The thermal power computations for the Feedwater and CRD System are performed by the following equation:

QS = DH (BTU/lbm) x Flow (Ibm/hr) x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

The DH value is the difference between the steam enthalpy in the reactor and the feedwater enthalpy (or the CRD System water enthalpy). For this exercise, feedwater and CRD enthalpies are constant, so the only change is in steam enthalpy. Therefore, the error in the thermal power measurements is determined by setting the DH term equal to the error in the enthalpy of the steam and solving for the resulting power error. Since the same enthalpy error exists for both systems, the total error due to the steam enthalpy error is computed by combining flow rates for the Feedwater and CRD systems. Per Figure 1.2-2 of Reference 9.8.1, Flow = 11950000 Ibm/hr + 30000 Ibm/hr

= 11,980,000 Ibm/hr DQs(-) = (1193.26-1191)BTU/lbm x 11,980,00O0bm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +7.9327 MW DQs(+) = (1188.68-1191)BTU/lbm x 11,980,000 Ibm/hr (1hr/60 min) x (17.5796 Wafts /(BTU/min)) x (1 MW / 1E6 W)

= -8.1433 MW Due to the symmetry of the error, the error is conservatively established as the largest value, expressed in both directions.

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 53 of 89 Errors (MW) = +/-8.1433 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +/-0.2947% RTP The weighting factor is therefore determined as follows:

WFsp2 = Error (% RTP) I Error (% Span)

= 0.2947% RTP /5 % Span

= 0.0589% RTP / % Span 6.2.4 CRD SYSTEM FLOW WEIGHTING FACTOR For the CRD System Flow measurement, all errors are expressed in terms of

%DP Span, except for the Reference Accuracy and Installation Effect for the Flow Element itself, which are expressed in terms of % Actual Flow. All of the errors are random. The weighting factors for each of these types of errors are derived below.

Weighting Factors for % DP Span Errors Unit 1 Per Section 6.1.4, the nominal CRD System Flow at 100% Power operation is:

C (U1)= 60.4651% Flow Span A (U1)= 36.5603% DP Span A nominal +/-5% uncertainty value results in a band as follows:

31.5603% DP Span< A < 41.5603% DP Span The square root function produces the following in terms of % Flow Span.

56.1786% Flow Span < C < 64.4673% Flow Span.

The error in this value is as follows:

+ 4.0022 / -4.2865% Flow Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 54 of 89 Because of the very nearly equal values, for symmetry, the worst-case uncertainty value is used in both directions. Therefore, a 5% DP Span error for the feedwater flow signals results in the following:

ErrorcRD = +/- 4.2865% Flow Span x (100 GPM)

ErrorCRD = +/- 4.2865 GPM Converting this value to process units, since the density is held constant during this exercise, the value is multiplied by the ratio of mass flow to volumetric flow at nominal 100% power conditions.

ErrorcRD = +/- 4.2865 GPM x (30000 Ibm/hr)/ 60.4651 GPM ErrorcRD = +/- 2126.7640 Ibm/hr Per Section 6.1.3 and Figure 1.2-3 of Reference 9.8.1, the difference in enthalpy from the CRD System to steam in the reactor at 100% power is as follows:

DH = 1191-94 = 1097 BTU/lbm Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW) = ErrorCRD (Ibm/hr) x DH -x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= _2126.7640 Ibm/hr x 1097 BTU/lbm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= _0.6836 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +/-0.0247% RTP The weighting factor is therefore determined as follows:

WFCRDF1 = Error (% RTP) I Error (% DP Span)

= 0.0247% RTP /5% DP Span

= 0.0049% RTP / % DP Span Unit 2 Per Section 6.1.4, the nominal CRD System Flow is:

C (U2)= 60.4706% Flow Span A (U2)= 36.5669% DP Span

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 55 of 89 A nominal +/-5% uncertainty value results in a band as follows:

31.5669% DP Span < A < 41.5669% DP Span The square root function produces the following in terms of % Flow Span.

56.1844% Flow Span < C < 64.4724% Flow Span.

The error in this value is as follows:

+4.0018 / -4.2862% Flow Span Because of the very nearly equal values, for symmetry, the worst case uncertainty value is used in both directions. Therefore, a 5% DP Span error for the feedwater flow signals results in the following:

ErrorcRD = +/- 4.2862% Flow Span x (100 GPM)

ErrorcRD +/- 4.2862 GPM Converting this value to process units, since the density is held constant during this exercise, the value is multiplied by the ratio of mass flow to volumetric flow at nominal 100% power conditions.

ErrorcRD = +/- 4.2862 GPM x (30000 Ibm/hr)/ 60.4706 GPM ErrorcRD = +/- 2126.4218 Ibm/hr Per Section 6.1.3 and Figure 1.2-2 of Reference 9.8.1, the difference in enthalpy from the CRD System to steam in the reactor at 100% power is as follows:

DH = 1191-94.3 = 1096.7 BTU/lbm Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW) = ErrorcRD (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= +2126.4218 lbm/hrx 1097 BTU/lbm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +/-0.6833 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +/- 0.0247% RTP

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 56 of 89 The weighting factor is therefore determined as follows:

WFCRDF2 = Error (% RTP) / Error (% DP Span)

= 0.0247% RTP / 5% DP Span

= 0.0049% RTP / % DP Span Since the errors are equivalent to the significant digits considered by this study, the Weighting Factors for the Unit 1 CRD System Flow is considered equal to that of the Unit 2 CRD System.

WFcRDF = 0.0049% RTP / % DP Span Weighting Factors for % Actual Flow Errors Unit 1 Per Section 6.1.4, the nominal CRD System Flow at 100% Power operation is:

C (U1)= 60.4651% Flow Span A nominal +/-5% Actual Flow uncertainty value results in a band as follows:

0.95*60.4651 % Flow Span < C < 1.05*60.4651 % Flow Span 57.4418% Flow Span < C < 63.4884% Flow Span.

The error in this value is as follows:

+ 3.0233 / -3.0233% Flow Span ErrorcRD = 3.0233% Flow Span x (100 GPM)

ErrorCRD = + 3.0233 GPM Converting this value to process units, since the density is held constant during this exercise, the value is multiplied by the ratio of mass flow to volumetric flow at nominal 100% power conditions.

ErrorcRD = +/- 3.0233 GPM x 30000 lbm/hr/ 60.4651 GPM ErrorcRD = +/- 1500.0223 Ibm/hr Per Section 6.1.3 and Figure 1.2-3 of Reference 9.8.1, the difference in enthalpy from the CRD System to steam in the reactor at 100% power is as follows:

DH = 1191-94 = 1097 BTU/lbm

Attachment 1 to HL-6328, Enclosure 7 E. 1.HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 57 of 89 Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW) = ErrorcRD (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= 1500.0223 Ibm/hr x 1097 BTU/lbm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +0.4821 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = _0.0174% RTP The weighting factor is therefore determined as follows:

WFCRDFlAF = Error (% RTP) / Error (% Actual Flow)

= 0.01 74% RTP /5% Actual Flow

= 0.0035% RTP / % Actual Flow Unit 2 Per Section 6.1.4, the nominal CRD System Flow is:

C (U2)= 60.4706% Flow Span A nominal +/-5% Actual Flow uncertainty value results in a band as follows:

0.95*60.4706% Flow Span < C < 1.05*60.4706% Flow Span 57.4471 % Flow Span < C < 63.4941 % Flow Span.

The error in this value is as follows:

+3.0235 / -3.0235% Flow Span ErrorcRD = +/- 3.0235% Flow Span x (100 GPM)

ErrorcRD = +/- 3.0235 GPM Converting this value to process units, since the density is held constant during this exercise, the value is multiplied by the ratio of mass flow to volumetric flow at nominal 100% power conditions.

ErrorcRD = +/- 3.0235 GPM x (30000 Ibm/hr)/ 60.4706 GPM ErrorcRD = +/- 1499.9851 Ibm/hr

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 58 of 89 Per Section 6.1.3 and Figure 1.2-2 of Reference 9.8.1, the difference in enthalpy from the CRD System to steam in the reactor at 100% power is as follows:

DH = 1191-94.3 = 1096.7 BTU/lbm Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW)= ErrorcRD (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= +/-1499.9851 Ibm/hrx 1096.7 BTU/lbm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

- +/-0.4820 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +/- 0.0174% RTP The weighting factor is therefore determined as follows:

WFcRDF2AF = Error (% RTP) / Error (%Actual Flow)

= 0.01 74% RTP / 5% Actual Flow

= 0.0035% RTP / % Actual Flow Since the errors are equivalent to the significant digits considered by this study, the Weighting Factors for the Unit 1 CRD System Flow is considered equal to that of the Unit 2 CRD System.

W FcRDFAF WA 0.0035% RTP / % DP Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 59 of 89 6.2.5 CRD SYSTEM INLET TEMPERATURE WEIGHTING FACTOR The CRD System Temperature measurement is only used to compute the enthalpy of the CRD System water, which is injected into the reactor during normal operation. The weighting factor is in terms of %RTP / OF, so a nominal variation of +/-50F is used to determine the weighting factor. Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the inlet temperatures of the CRD System to the reactor are 123.2 0 F (Unit 1) and 123.5 0 F (Unit 2). Because of the closeness in values, a nominal value of 123.5 0 F is used to determine the corresponding error in the thermal power measurement. Per Reference 9.8.2, H(1 050 psia, 128.5 0F) = 99.0990 BTU/Ibm H(1050 psia, 123.5 0F) = 94.3000 BTU/Ibm H(1 050 psia, 118.5 0F) = 89.1575 BTU/lbm The thermal power computations for the CRD System are performed by the following equation:

QCRDT = DH (BTU/Ibm) x Flow (Ibm/hr) x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

The DH value is the difference between the steam enthalpy in the reactor and the CRD System water enthalpy. For this evaluation, the steam enthalpy is constant, so the only change is in CRD System water enthalpy. Therefore, the error in the thermal power measurements is determined by setting the DH term equal to the error in the enthalpy of the CRD System water and solving for the resulting power error. Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, CRD System Flow = 30000 Ibm/hr DQCRDT(+) = (99.099-94.3)BTU/lbm x 30000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1 E6 W)

= +0.0422 MW DQCRDT(-) = (89.1575-94.3)BTU/lbm x 30000 Ibm/hr (I hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1 E6 W)

= -0.0452 MW Due to the symmetry of the error, the error is conservatively established as the largest value, expressed in both directions.

ErrorcRDT (MW) = +/- 0.0452 MW

Attachment 1 to HL-6328, Enclosure 7 E.1. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 60 of 89 Converting to a percentage,

'Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) +/- 0.0016% RTP The weighting factor is therefore determined as follows:

WFCRDT = Error (% RTP) / Error (OF)

= 0.0016% RTP / 5 'F

= 0.0003% RTP / OF 6.2.6 RWCU SYSTEM FLOW WEIGHTING FACTOR For the RWCU System Flow measurement, all errors are expressed in terms of

%DP Span, except for the Reference Accuracy and Installation Effect for the Flow Element itself, which are expressed in terms of % Actual Flow. All of the errors are considered random. The weighting factors for each of these types of errors are derived below.

Weighting Factors for % DP Span Errors Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the nominal RWCU System Flow at 100% Power operating conditions is 100000 Ibm/hr and the differential enthalpy is 112.6 BTU/lbm. The flow loops are of the same configuration and range. Therefore, both units are analyzed identically.

Per Section 6.1.6, the nominal flow rate at 100% power corresponds to the following values in terms of % Flow Span and % DP Span.

RWCU Flow (Ul) 87.9626% Flow Span RWCU Flow (Ul) 77.3741% DP Span RWCU Flow (U2) 88.2742% Flow Span RWCU Flow (U2) 77.9234% DP Span Because of the square root function, more effects are seen from DP errors at lower nominal flow rates. Therefore, the lowest (Unit 1) nominal flow rate is used, and the weighting factor is applied to both units.

C (U1)= 87.9626% Flow Span (263.8878 GPM)

A (U1)= 77.3741% DP Span A nominal +/-5% uncertainty value results in a band as follows:

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 61 of 89 72.3741 % DP Span< A+a 5 82.3741% DP Span The square root function produces the following in terms of % Flow Span.

85.0730% Flow Span < C+c < 90.7602% Flow Span.

The error in this value is as follows:

+ 2.7976/-2.8896% Flow Span Because of the very nearly equal values, for symmetry, the worst case uncertainty value is used in both directions. Therefore, a 5% DP Span error for the feedwater flow signals results in the following:

ErrorRwcu = +/- 2.8896% Flow Span x (300 GPM)

ErrorRwcu = +/- 8.6688 GPM Converting this value to process units, since the density is held constant during this exercise, the value is multiplied by the ratio of mass flow to volumetric flow at nominal 100% power conditions.

ErrorRwcu = +/- 8.6688 GPM x (100000 Ibm/hr)/ 263.8878 GPM ErrorRwcu = +/- 3285.0325 Ibm/hr Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the difference in enthalpy from the RWCU System (inlet to outlet) at 100% power is as follows:

DH = 525.5-412.9 = 112.6 BTU/lbm Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW) = ErrorD (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= +/-3285.0325 Ibm/hr x 112.6 BTU/lbm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= _0.1084 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +/-0.0039% RTP

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 62 of 89 The weighting factor is therefore determined as follows:

WFCUF = Error (% RTP) / Error (% DP Span)

= 0.0039% RTP / 5% DP Span

= 0.0008% RTP / % DP Span Weighting Factorsfor % Actual Flow Errors Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the nominal RWCU System Flow at 100% Power operating conditions is 100000 Ibm/hr and the differential enthalpy is 112.6 BTU/lbm. The flow loops are of the same configuration and range. Therefore, both units are analyzed identically.

Per Section 6.1.6, the nominal flow rate at 100% power corresponds to the following values in terms of % Flow Span.

RWCU Flow (Ul) 87.9626% Flow Span RWCU Flow (U2) 88.2742% Flow Span Because a given % of Actual Flow is greater for larger flow rates, Unit 2 is used to establish the weighting factor.

C (U2)= 88.2742% Flow Span (264.8226 GPM)

A nominal +/-5% Actual Flow uncertainty value results in a band as follows.

0.95*88.2742% Flow Span < C+c < 1.05*88.2742% Flow Span 83.8605% Flow Span < C+c < 92.6879% Flow Span The error in this value is as follows.

+ 4.4137/ -4.4137% Flow Span ErrorRwcu = +/- 4.4137% Flow Span x (300 GPM)

ErrorRwcu = +/-13.2411 GPM Converting this value to process units, since the density is held constant during this exercise, the value is multiplied by the ratio of mass flow to volumetric flow at nominal 100% power conditions.

ErrorRwcu = +/- 13.2411 GPM x (100000 Ibm/hr)/ 264.8226 GPM ErrorRwcu = +/- 4999.9887 Ibm/hr

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 63 of 89 Per Figures 1.2-2 and 1.2-3 of Reference 9.8.1, the difference in enthalpy from the RWCU System (inlet to outlet) at 100% power is as follows:

DH = 528.6-416.0 = 112.6 BTU/lbm Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW) = ErrorD (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1 E6 W)

= +/-4999.9887 Ibm/hr x 112.6 BTU/Ibm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1 E6 W)

= +/-0.1650 MW Converting to a percentage, Error (%) (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) +/-0.0060% RTP The weighting factor is therefore determined as follows:

WFCUFAF = Error (% RTP) / Error (% DP Span)

= 0.0060% RTP / 5% DP Span

= 0.0012% RTP / % DP Span 6.2.7 RWCU SYSTEM INLET / OUTLET TEMPERATURE WEIGHTING FACTORS With respect to the heat balance computation, the Inlet and Outlet Temperature measurements for the RWCU System are only used to compute the enthalpies of the RWCU System water at the inlet and outlet. The weighting factor is in terms of %RTP / OF, so a nominal variation of +/-51F is used to determine the weighting factor.

Unit 1 Per Figure 1.2-3 of Reference 9.8.1, the nominal temperatures and enthalpies of interest are as follows:

RWCU Inlet Temperature, Enthalpy 531.2 0F, 525.5 BTU/Ibm RWCU Outlet Temperature, Enthalpy 434.0 0F, 412.9 BTU/lbm

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 64 of 89 The enthalpies of the water in an error band around these nominal temperatures, assuming a pressure of 1050 psia are as follows, per Reference 9.8.2:

Inlet H(1050 psia, 536.2 0F) = 531.8310 BTU/lbm Inlet H(1050 psia, 526.2 0F) = 519.3722 BTU/lbm Outlet H(1050 psia, 439.0OF) = 418.3910 BTU/lbm Outlet H(1050 psia, 429.0OF) = 407.4100 BTU/lbm The thermal power computations for the RWCU System are performed by the following equation:

QCRDT = DH (BTU/lbm) x Flow (Ibm/hr) x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

The DH value is the difference between the inlet and outlet enthalpy values. For this exercise, the inlet and outlet temperature errors are assessed separately, with weighting factors determined for each. Therefore, only one of the enthalpies is varied at a time. Therefore, the error in the thermal power measurements is determined by setting the DH term equal to the error in the enthalpy and solving for the resulting power error. Per Figure 1.2-3 of Reference 9.8.1, RWCU System Flow = 100,000 Ibm/hr DQRWCTI(-) = (519.3722-525.5)BTU/Ibm x 100,000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= -0.1795 MW DQRWcT;(+) = (531.831 -525.5)BTU/lbm x 100,000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +0.1855 MW Due to the symmetry of the error, the error is conservatively established as the largest value, expressed in both directions.

ErrorRwcT, (MW) = +/- 0.1855 MW Converting to a percentage, Error (%) (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) _0.0067% RTP

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 65 of 89 The weighting factor is therefore determined as follows:

WFcuTlI = Error (% RTP) / Error (OF)

= 0.0067% RTP /5 OF

= 0.0013% RTP / OF DQRWCTO(-) = (407.41-412.9)BTU/lbm x 100,000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1 E6 W)

= -0.1609 MW DQRWcTo(+)= (418.391-412.9)BTU/lbm x 100,000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +0.1609 MW ErrorRwcTo (MW) = +/- 0.1609 MW Converting to a percentage, Error (%) (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) +/-0.0058% RTP The weighting factor is therefore determined as follows:

WFcuTlo = Error (% RTP) / Error (OF)

= 0.0058% RTP /5 OF

= 0.0012% RTP / OF Unit 2 Per Figure 1.2-2 of Reference 9.8.1, the nominal temperatures and enthalpies of interest are as follows:

RWCU Inlet Temperature, Enthalpy 533.7°F, 528.6 BTU/lbm RWCU Outlet Temperature, Enthalpy 436.8°F, 416.0 BTU/ibm The enthalpies of the water in an error band around these nominal temperatures, assuming a pressure of 1050 psia are as follows, per Reference 9.8.2:

Inlet H(1050 psia, 538.7°F) = 534.9685 BTU/Ibm Inlet H(1050 psia, 528.7 0F) = 522.4497 BTU/lbm Outlet H(1 050 psia, 441.8 0 F) = 421.4662 BTU/lbm Outlet H(1050 psia, 431.8°F) = 410.4782 BTU/lbm

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 66 of 89 The thermal power computations for the RWCU System are performed by the following equation:

QCRDT = DH (BTU/Ibm) x Flow (Ibm/hr) x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

The DH value is the difference between the inlet and outlet enthalpy values. For this exercise, the inlet and outlet temperature errors are assessed separately, with weighting factors determined for each. Therefore, only one of the enthalpies is varied at a time. Therefore, the error in the thermal power measurements is determined by setting the DH term equal to the error in the enthalpy and solving for the resulting power error. Per Figure 1.2-2 of Reference 9.8.1, RWCU System Flow = 100,000 Ibm/hr DQRWC-f(-) = (522.4497-528.6)BTU/lbm x 100,000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= -0.1802 MW DQRWCTi(+) = (534.9685-528.6)BTU/lbm x 100,000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +0.1866 MW Due to the symmetry of the error, the error is conservatively established as the largest value, expressed in both directions.

ErrorRWCT-1 (MW) = +/- 0.1866 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP

= +/-0.0068% RTP The weighting factor is therefore determined as follows:

WFcUT2I = Error (% RTP) / Error (OF)

= 0.0068% RTP /5 OF

= 0.0014% RTP / OF DQRWCTo(-) = (410.4782-416)BTU/lbm x 100,000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= -0.1618 MW DQRwcTo(+) = (421.4662-416)BTU/lbm x 100,000 Ibm/hr (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +0.1602 MW

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 67 of 89 Due to the symmetry of the error, the error is conservatively established as the largest value, expressed in both directions.

ErrorRwcTO (MW) = +/- 0.1 618 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +0.0059% RTP The weighting factor is therefore determined as follows:

WFcuT2o = Error (% RTP) / Error (OF)

= 0.0059% RTP /5 OF

= 0.0012% RTP / OF Because of the similarity between Units 1 and 2 and the relatively small contribution to the heat balance computation, the weighting factors are assigned as the worst case for use in both units.

WFcuT = 0.0014% RTP / OF WFcuro = 0.0012% RTP / OF 6.2.8 RECIRCULATION PUMP POWER WEIGHTING FACTOR The range of the watt transducers for the recirculation pumps is 0-8.4 MW. The weighting factor for this parameter is merely a matter of expressing this term as a percentage of Rated Thermal Power.

100% Span = 8.4 MW The 100% Rated Thermal Power for the Reactor is 2763 MW. Therefore, the weighting factor is determined as follows:

WFRPP= 8.4 MW / 2763 MW = 0.0030% RTP / % Span 6.2.9 ULTRASONIC FEEDWATER FLOW AND FEEDWATER FLOW ELEMENT WEIGHTING FACTOR The Ultrasonic Feedwater flow uncertainties and the Feedwater Flow Element uncertainties are listed in terms of % Actual Flow. Since each of these uncertainties are significantly less than 1% Actual Flow, a nominal value of 1%

Actual Flow is used to determine the weighting factor, rather than the nominal

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 68 of 89 5% Span used for the majority of the parameters.

Unit 1 Per Figure 1.2-3 of Reference 9.8.1, the nominal Feedwater Flow rate is 11514000 Ibm / hr. Therefore, a 1% Actual Flow uncertainty is computed.

Error (Ibm/hr) = 0.01 x 11514000 Ibm /hr

= 115140 Ibm / hr Per Section 6.1.3 and Figure 1.2-3 of Reference 9.8.1, the difference in enthalpy from feedwater to steam in the reactor at 100% power is as follows:

DH = 1191-373.4 = 817.6 BTU/Ibm Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW) = Error (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= +/-115140 Ibm / hr x 817.6 BTU/lbm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +27.5819 MW Converting to a percentage, Error (%) (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) +/-0.9983% RTP / % Actual Flow The weighting factor is therefore determined as follows:

WFUFM1 = Error (% RTP) / Error (% Flow Span)

= 0.9983% RTP / 1% Actual Flow

= 0.9983% RTP / % Actual Flow Unit 2 Per Figure 1.2-2 of Reference 9.8.1, the nominal Feedwater Flow rate is 11950000 Ibm / hr. Therefore, a 1% Actual Flow uncertainty is computed.

Error (Ibm/hr) = 0.01 x 11950000 Ibm / hr

= 119500 Ibm / hr Per Section 6.1.3 and Figure 1.2-2 of Reference 9.8.1, the difference in enthalpy from feedwater to steam in the reactor at 100% power is as follows:

DH = 1191-403.2 = 787.8 BTU/lbm

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 69 of 89 Therefore, the error in flow rate indication would produce the following error in terms of the thermal power computation:

Error (MW) = Error (Ibm/hr) x DH x (1hr/60 min) x (17.5796 Watts

/(BTU/min)) x (1 MW / 1E6 W)

= +/-119500 Ibm / hr x 787.8 BTU/lbm x (1hr/60 min) x (17.5796 Watts /(BTU/min)) x (1 MW / 1E6 W)

= +/-27.5830 MW Converting to a percentage, Error (%) = (Error (MW) / 2763 MW thermal) x 100% RTP Error (%) = +/-0.9983% RTP / % Flow Span The weighting factor is therefore determined as follows:

WFUFM2 = Error (% RTP) / Error (% Flow Span)

= 0.9983% RTP / 1% Flow Span

= 0.9983% RTP / % Flow Span Since the Unit 1 and Unit 2 weighting factors are equivalent, they are combined as one term.

WFUFM = 0.9983% RTP / % Flow Span Additionally, the Feedwater Flow Venturi Fouling bias terms are expressed in the same units. Therefore, the conversion factor is the same.

WFFWFB = 0.9983% RTP / % Flow Span Finally, the Reference Accuracy and Installation Effects for the Feedwater Flow Venturi are expressed in the same units. Therefore, the conversion factor is the same.

WFFWFE = 0.9983% RTP / % Flow Span 6.2.10 CORRECTION FACTOR ERROR WEIGHTING FACTOR The error in correction factor is in units of MW. Therefore, the weighting factor is merely determined using the conversion from MW to % RTP.

WFqrad= 100% RTP/ 2763 MW

= 0.0362% RTP/ MW

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 70 of 89 6.2.11 WEIGHTING FACTOR

SUMMARY

The following table is created from the values derived in Sections 6.2.1 through 6.2.10 above.

Parameter Term Value Units Feedwater Flow - Unit 1 WFFWF1 0.6988  % RTP / % DP Span Feedwater Flow - Unit 2 WFFwF2 0.6476  % RTP / % DP Span Feedwater Flow Venturi Fouling WFFWFB 0.9983  % RTP /  % Actual Flow Feedwater Flow Element Effects WFFWFE 0.9983  % RTP /  % Actual Flow Feedwater Temperature - Unit 1 WFFwT1 0.1309  % RTP / OF Feedwater Temperature - Unit 2 WFFwT2 0.1384  % RTP / OF Reactor Pressure - Unit 1 WFsp, 0.0568  % RTP /  % Span Reactor Pressure - Unit 2 WFsP2 0.0589  % RTP /  % Span CRD System Flow WFCRDF 0.0049  % RTP / % DP Span CRD System Flow Element Effects WFCRDFAF 0.0035  % RTP / % Actual Flow CRD Inlet Temperature WFCRDT 0.0003  % RTP / oF RWCU Flow WFcuF 0.0008  % RTP /  % DP Span RWCU Flow Element Effects WFcUFAF 0.0012  % RTP /  % Actual Flow RWCU Inlet Temperature WFcu-i 0.0014  % RTP / OF RWCU Outlet Temperature WFcuTo 0.0012  % RTP / OF Recirc Pump Power WFRPP 0.0030  % RTP /  % Span UFM WFUFM 0.9983  % RTP /  % Actual Flow Correction Factor (QRAD) WFORAD 0.0362  % RTP / MW Table 1 - Weighting Factor Summary Table

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 71 of 89 6.3 INDIVIDUAL INSTRUMENT LOOP UNCERTAINTY COMPUTATIONS 6.3.1 FEEDWATER FLOW UNCERTAINTY COMPUTATION Case 1 - No UFM Correction The following non-zero terms are derived in Section 4.2 and 4.10 for the Feedwater Flow instrument loops. These errors apply only to Case 1, since the continuous UFM Correction effectively eliminates all these error terms for Case 2.

RAFWFE = +/- 0.2500% Actual Flow IEFWFE +/- 0.5000% Actual Flow PEFWFE1 = +/- 0.0784% DP Span (Dependent with FW Temp)

PEFWFE2 = +/- 0.0918% DP Span (Dependent with FW Temp)

PEbFWFE = + 0.6000% Actual Flow (Bias, Treat Separately)

DAFwFr = +/-1.4770% DP Span STFwFTr = +/- 0.2500% DP Span STFwFT2 = +/- 0.1250% DP Span SPEFwFr = +/- 0.2768% DP Span PSEFwF7 = +/- 0.0090% DP Span TEFwF - +/- 0.1260% DP Span RApC = +/- 0.0180% Span RESpC + 0.0244% Span

+/-

The Plant Computer uncertainties are expressed in generic percent span terms, which apply directly to this loop, since the DP signal feeds the Plant Computer directly.

Three random terms are computed herein, which are separately considered for the total heat balance uncertainty. The first term is the total of the random dependent loop errors that are expressed in terms of % DP Span. The second is the total of the random loop errors that are expressed in terms of % Actual Flow. The third is the Feedwater Flow element density error (PEFWFE), which is expressed in terms of % DP Span, but which is a dependent term with the Feedwater Temperature uncertainty.

Therefore the terms combine in an SRSS fashion to produce Feedwater Flow uncertainties (TLU) for each single flow loop (A and B).

Unit 1 TLUFWF1 = SRSS(DAFWFT, STFWFT1, SPEFwFr, PSEFwFT, TEFwFr, RApc, RESPC)

= + 1.5289% DP Span TLUFWF1AF = SRSS(RAFWFE, IEFWFE)

= +/- 0.5590% Actual Flow PEFWFE1 = +/- 0.0784% DP Span (Dependent with FW Temp)

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 72 of 89 In addition, there is a bias term, due to fouling of the feedwater flow elements, which must be considered.

PEbFwFE = + 0.6000% Actual Flow (Bias, Treat Separately)

Unit 2 TLUFWF2 = SRSS(DAFwFr, STFwFr2, SPEFwFT, PSErwFT, TEFwFT, RApc, RESpc)

= +/- 1.5135% DP Span TLUFWF2AF = SRSS(RAFWFE, IEFWFE)

= +/- 0.5590% Actual Flow PEFWFp = +/- 0.0918% DP Span (Dependent with FW Temp)

PEbFWFE = + 0.6000% Actual Flow (Bias, Treat Separately)

Case 2 - Continuous UFM Correction Under this scenario, the UFM is used to automatically correct the FW flow indication in the Plant Computer on a continuous basis. Therefore, all uncertainty terms are eliminated from the flow venturi instrument loops. The only remaining uncertainty value is for the Westinghouse Crossflow UFMs.

Therefore, TUUFMXFLO - +/- 0.4200% Actual Flow (95% Confidence Factor) 6.3.2 FEEDWATER TEMPERATURE UNCERTAINTY COMPUTATION The following non-zero terms are derived in Section 4.3 and 4.10 for the Feedwater Temperature instrument loops.

RAFw-rr ++/-0.20000 F STFw-rr +/- 0.3001 OF M&TEFw-r = +/- 0.20000 F DRFwrr = +/- 0.20000 F PSEFwTT = +/- 0.01500 F TEFwrr = +/- 2.50000F

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 73 of 89 The span of the temperature indication is 1500F. The Plant Computer uncertainties are expressed in generic percent span terms, which apply directly to this loop, since the temperature signal feeds the Plant Computer directly. Therefore, the PC Conversions are made as shown below.

RApcFWT = +/- RApc x 150OF / 100% Span

= +/- 0.0180% Span x 1501F / 100% Span

= +/- 0.02700F RESpcPwT = + RESpc x 150'F / 100% Span

= +/- 0.0244% Span x 150OF / 100% Span

= +/- 0.03660 F The terms combine in an SRSS fashion to produce a Feedwater Temperature uncertainty for each indication. Units 1 and 2 are identically configured.

TUFWT = SRSS(RAFwTT, STFwTT, M&TEFwrr, DRFwTT, PSEFwTT, TEFw-r, RAPCFWT, RESpcFwT)

= +2.5421 OF The real parameter of concern is the average feedwater temperature between the loops. There are 4 feedwater temperature measurements, two on each feedwater loop, which are used in the determination of the average, and each sensor has the same configuration and range. Therefore, when the Plant Computer actually determines the value for average feedwater temperature, the Total Loop Uncertainty in the determination is shown as follows:

TLU (Average FW Temp) = SRSS (TUs of 4 FW Temp Signals) / No. Signals

= (4x(TU of Single FW Temp Signal) )112/4

= TU of Single FW Temp Signal /2 TLUFwT = TUFWT/ 2 = +1.2711 OF 6.3.3 REACTOR PRESSURE UNCERTAINTY COMPUTATION The following non-zero terms are derived in Section 4.4 and 4.10 for the Reactor Steam Pressure instrument loops.

DARPT +/- 1.4470% Span STRPT +/- 0.2500% Span PSERPT = +/- 0.0188% Span TERPT +/- 0.4080% Span RApc = +/- 0.0180% Span RESpc = +/- 0.0244% Span

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 74 of 89 The uncertainties for this measurement are not changed for any of the analyzed scenarios, and the configuration is identical from Unit 1 to Unit 2. Therefore, only one set of Total Uncertainty is determined.

TUsp = SRSS(DARPT, STRpT, PSERPT, TERPT, RApc, RESpc)

= +/- 1.5245% Span There are 2 Reactor Pressure measurements that are averaged to determine Reactor Pressure, and each measurement loop has the same configuration and range. Therefore, when the Plant Computer actually determines the value, the Total Loop Uncertainty in the determination is shown as follows:

TLUsp = SRSS (TUs of 2 Rx Pressure) / No. Signals

= (2x(TU of Single RX Pressure Signal)2)"2/2

= TU of Single Rx Pressure / (2)1~2 TLUSP = TUsp / (2)1/2 = +/- 1.0780% Span 6.3.4 CONTROL ROD DRIVE (CRD) FLOW UNCERTAINTY COMPUTATION The following non-zero terms are derived in Section 4.5 and 4.10 for the CRD System Flow instrument loops.

RACRDFE +/- 1.0000% Actual Flow

+

IECRDFE +/- 1.0000% Actual Flow PECRDFE = +/- 0.5560% DP Span RACRDFT = +/- 0.4000% DP Span STcRDFT +/- 0.5000% DP Span M&TECRDFr = +/- 0.4000% DP Span SPECRDFT = +/- 2.0706% DP Span TECRDFT = +/- 0.6800% DP Span RApc = +/- 0.01 80% Span RESpc = +/- 0.0244% Span The uncertainties for this measurement are not changed for any of the analyzed scenarios, and the configuration is identical from Unit 1 to Unit 2. Therefore, only one set of Total Loop Uncertainty is determined.

Two random terms are computed herein, which are separately considered for the total heat balance uncertainty. The first term is the total of the loop errors that are expressed in terms of % DP Span. The second is the total of the random loop errors that are expressed in terms of % Actual Flow.

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 75 of 89 The Plant Computer uncertainties are expressed in generic percent span terms, which apply directly to this loop, since the DP signal feeds the Plant Computer directly. Therefore the terms combine in an SRSS fashion to produce a CRD System Flow uncertainty (TLU).

TLUCRDF = SRSS(PECRDFE, RACRDFT, STCRDFT, M&TECRDFT, SPECRDFT, TECRDFT, RApc, RESpc)

= +/- 2.3727% DP Span TLUCRDFAF = SRSS(RACRDFE, IECRDFE)

= +/- 1.4142% Actual Flow 6.3.5 CRD SYSTEM TEMPERATURE UNCERTAINTY COMPUTATION The following non-zero terms are derived in Section 4.6 and 4.10 for the CRD System Temperature instrument loops.

RACRDTE = +/- 1.80000 F Per Section 4.6.1, the span of these thermocouples is 0-752 0F. Therefore, the PC uncertainty conversions are made as shown below.

RAPCCRDT = +/- RApc x 7520F / 100% Span

= +/- 0.0180% Span x 7520 F / 100% Span

= +/- 0.13540 F RESPCCRDT = +/- RESpc x 7520 F / 100% Span

= +/- 0.0244% Span x 7520 F / 100% Span

+/- 0.18350 F The configuration of the CRD System Temperature measurement does not change for any of the scenarios analyzed, and Unit 1 is identically configured with Unit 2. Therefore, only one Total Loop Uncertainty (TLU) value is necessary.

TLUCRDT = SRSS(RAcRDTE, RApccRDT, RESPCCRDT)

- +/- 1.81440 F

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 76 of 89 6.3.6 REACTOR WATER CLEANUP (RWCU) SYSTEM FLOW UNCERTAINTY COMPUTATION The following non-zero terms are derived in Section 4.7 and 4.10 for the RWCU System Flow instrument loops.

RACUFE = +/- 0.5000% Actual Flow IECUFE = +/- 0.5000% Actual Flow PECUFE1 = +/- 0.5484% DP Span PECUFE2 = + 0.5504% DP Span PEb1CU = -1.7864% DP Span (Bias, Treat Separately)

PEb2cu = -1.5294% DP Span (Bias, Treat Separately)

DAcuFT = +/- 1.5770% DP Span STcuFr = +/- 0.5000% DP Span SPEcuFT = +/- 0.5177% DP Span PSEcuFr = +/- 0.2326% DP Span TEcuFT = +/- 0.5000% DP Span

-RApc = +/- 0.01 80% Span RESpc = +/- 0.0244% Span Two random terms are computed herein, which are separately considered for the total heat balance uncertainty. The first term is the total of the loop errors that are expressed in terms of % DP Span. The second is the total of the random loop errors that are expressed in terms of % Actual Flow.

The Plant Computer uncertainties are expressed in generic percent span terms, which apply directly to this loop, since the DP signal feeds the Plant Computer directly. Therefore the terms combine in an SRSS fashion to produce a RWCU System Flow uncertainty (TLU).

TLUCUF SRSS(PECUFE, DAcuFr, STcuFT, SPEcuFT, PSEcuFT, TEcuFr, RApc, STpc, M&TEpc, RESpc)

TLUCUFAF SRSS(RACUFE, IECUFE)

Unit 1 TLUCUF1 = SRSS(PECUFEI, DAcuF7, STcuF7, SPEcuFr, PSEcdFr, TEcuFr, RApc, RESpc)

= +/- 1.9002% DP Span TLUCUFl AF SRSS(RACUFE, IECUFE)

= +/- 0.7071 % Actual Flow

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 77 of 89 Unit 2 TLUcUF 2 = SRSS(PECUFE9, DAcuFT, STcuF7, SPEcuF7, PSEcuFr, TEcuFr, RApc, RESPc)

= +/- 1.9008% DP Span TLUCUF2AF= SRSS(RACUFE, IECUFE)

= +/- 0.7071% Actual Flow For the RWCU flow uncertainty, an additional bias exists, for which the value is defined by the Process Error.

PEb1cu = -1.7864% DP Span (Bias, Treat Separately)

PEb2cu = -1.5294% DP Span (Bias, Treat Separately) 6.3.7 RWCU INLET / OUTLET TEMPERATURE UNCERTAINTY COMPUTATION The following non-zero terms are derived in Section 4.8 and 4.10 for the RWCU Inlet and Outlet System Temperature instrument loops.

RACUTE - +/- 3.03750 F Per Section 4.8.1, the span of these thermocouples is 0-600 0F. Therefore, the PC uncertainty conversions are made as shown below.

RAPCCUT = +/- RApc x 6000 F / 100% Span

+/- 0.0180% Span x 6000°F / 100% Span

+/- 0.10800 F RESPccuT = +/- RESpc x 6000 F / 100% Span

= +/- 0.0244% Span x 6000 F / 100% Span

= +/- 0.14640 F The configuration of the RWCU Inlet or Outlet Temperature measurement does not change for any of the scenarios analyzed, and Unit 1 is identically configured with Unit 2. Therefore, only one Total Loop Uncertainty (TLU) value is necessary.

TLUcUT = SRSS(RAcUTE, RAPCCUT, RESPCCUT)

= +/- 3.0429°F

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 78 of 89 6.3.8 RECIRCULATION PUMP POWER UNCERTAINTY COMPUTATION The following non-zero terms are derived in Section 4.9 and 4.10 for the Recirculation Pump Power instrument loops.

RARPWT = +/- 0.5000% Span STRPWT = +/- 0.5000% Span M&TERPWT = +/- 0.5000% Span TERPWT = +/- 1.0050% Span RApc = +/- 0.0180% Span RESPC = +/- 0.0244% Span The uncertainties for this measurement are not changed for any of the analyzed scenarios, and the configuration is identical from Unit 1 to Unit 2. Therefore, only one set of Total Loop Uncertainty is determined.

TLURpp = SRSS(RARPwT, STRPWT, M&TERpwT, TERPWT, RApc, RESpc)

= +/- 1.3270% Span 6.3.9 CORRECTION FACTOR UNCERTAINTY COMPUTATION There is only one uncertainty term to be entered for the correction factor, so no computations are necessary.

TUQRAD = +/-0.5500 MW

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 79 of 89 6.3.10 INDIVIDUAL PARAMETER TOTAL LOOP UNCERTAINTY

SUMMARY

The following table is created from the values derived in Sections 6.3.1 through 6.3.9 above.

Parameter Term Value Units Feedwater Flow - Unit 1 TLUFWF1 + 1.5289  % DP Span Feedwater Flow Elem Eff. - Unit 1 TLUFWF1AF +/- 0.5590  % Actual Flow Dependent FW Flow - Unit 1 PEFWFE1 + 0.0784  % DP Span Feedwater Flow - Unit 2 TLUFWF2 + 1.5135  % DP Span Feedwater Flow Elem Eff. - Unit 2 TLUFWF2AF +/- 0.5590  % Actual Flow Dependent FW Flow - Unit 2 PEFWFE2 + 0.0918  % DP Span Feedwater Venturi Fouling PEbFWFE + 0.6000  % Actual Flow Feedwater Temperature TLUFWT + 1.2711 OF Reactor Pressure TLUsP +/- 1.0780  % Span CRD System Flow TLUCRDF +/- 2.3727  % DP Span CRD System Flow Element Effects TLUCRDFAF +/- 1.4142  % Actual Flow CRD Inlet Temperature TLUCRDT +/- 1.8144 OF RWCU Flow (Unit 1) TLUCUF1 + 1.9002  % DP Span RWCU Flow Element Eff. (Unit 1) TLUCUF1AF +/- 0.7071  % Actual Flow RWCU Flow (Unit 2) TLUcUF2 + 1.9008  % DP Span RWCU Flow Element Eff. (Unit 2) TLUCUF2AF +/- 0.7071  % Actual Flow RWCU Flow Bias (Unit 1) PEb1cu -1.7864  % DP Span RWCU Flow Bias (Unit 2) PEb2cu -1.5294  % DP Span RWCU Inlet / Outlet Temperature TLUCUT + 3.0429 OF Recirc Pump Power TLURPP + 1.3270  % Span UFM - Westinghouse uCrossflow" TUUFMXFLOB +/- 0.4200  % Actual Flow Correction Factor TUORAD + 0.5500 MW Table 2 - Individual Parameter Total Loop Uncertainty Summary Table

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 80 of 89 6.4 TOTAL HEAT BALANCE MEASUREMENT UNCERTAINTIES The total heat balance measurement uncertainties are computed for 2 scenarios at 100% Power:

1. No Ultrasonic Flow Measurement (UFM) devices are installed. In this case, there is no correction to the venturi measurements, but operating in the current configuration, with accurate transmitter calibrations.

2. Westinghouse "Crossflow" Ultrasonic Flow Meters are used to correct the Feedwater Flow signals in the Plant Computer on a continuous basis.

The measurement uncertainty for each of these configurations is computed in the sub-sections to follow. NOTE: In addition to the random uncertainties computed in the sections below, there are two sets of bias terms needing consideration. The first is for the Feedwater Flow Venturi fouling, which is only a factor in Case 1, where UFMs are not used. This term may and may not be present, and could vary from a value of zero up to the value computed. This term is computed in Section 6.4.1.

The second bias is for the RWCU system flow measurement. Given the nature of this bias, which is negative, it is always present in the measurement, as performed, at 100% power heat balance conditions. This bias term is computed here and then repeated in each of the sections below.

TLUb = (WFCUF XPEbcu)

Unit 1 TLUb1 = (WFcUF X PEb1cu)

= -0.0014% RTP Unit 2 TLUb2 = (WFcuF X PEb2cu)

= -0.0012% RTP

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 81 of 89 11- 6.4.1 CASE 1 - NO ULTRASONIC FLOW METERS USED FOR CORRECTION OF FEEDWATER FLOW All random errors are combined in SRSS fashion, and bias terms are computed separately. The total measurement uncertainties of each parameter, which are derived in Section 6.3, are combined with weighting factors, which are derived in Section 6.2, in proportion to their total contribution to the calculation of uncertainties for the thermal power computation. Feedwater Temperature uncertainty and Feedwater Flow Element PE (due to density correction) are treated as dependent variables and are added prior to the SRSS computation.

TLURTP(Case 1) +/-((WFFWF X TLUFWF) + (WFFwFE X TLUFWFAF) + 2WFFWF X PEFWFE + WFFwT X TLUFWT)2 + (WFsp x TLUsp) + (WFcRDF X TLUCRDF) 2 + (WFCRDFAF X TLUCRDFAF) 2 + (WFCRDT X TLUCRDT) 2 + (WFcUF x TLUCUF) 2 + (WFCUFAF X TLUCUFAF) 2 +

(WFcuri x TLUCUT) 2 + (WFcUTO x TLUCUT) 2 + (WFRPP X TLURPP) 2 + (WFQRAD x TUQRAD) 2 )1' 2 Unit 1 TLURTP-U Ci = +/-((WFFWF1 X TLUFWF1) 2 + (WFFWFE X TLUFWFlAF) 2 + (WFFWF1 X PEFWFE1 + WFFWTI X TLUFWT) 2 + (WFspi x TLUSp) 2 +

(WFCRDF x TLUCRDF) 2 + (WFCRDFAF X TLUCRDFAF) 2 + (WFCRDT X TLUCRDT) 2 + (WFCUF X TLUcUF1) + (WFCUFAF X TLUCUF1AF) 2 + (WFcuTi x TLUCUT) + (WFcuTo X TLUCUT) 2 +

(WFRPP x TLURPP) 2 + (WFQRAD x TUQRAD) 2 ) 1' 2

= + 1.2273% RTP TLUbl = (WFcUF x PEb1 cu)

= -0.0014% RTP TLUFWFOUL = (WFFwFB X PEbFWFE)

= +0.5990% RTP Unit 2 TLURTP-U2C1 = +/-((WFFWF 2 X TLUFWF 2 ) 2 + (WFFWFE X TLUFWF2AF) 2 + (WFFWF2 X PEFWFE2 + WFFWT2 X TLUFwT) 2 + (WFsp2 X TLUsp) 2 +

(WFCRDF X TLUCRDF) 2 + (WFCRDFAF X TLUCRDFAF) 2 + (WFCRDT X TLUCRDT) 2 + (WFcUF x TLUcuF2) 2 + (WFCUFAF X TLUCuF2 AF)2 + (WFcun x TLUcuT) 2 + (WFcuTo X TLUCUT) 2 +

(WFRpp X TLURPP)f + (WFQRAD x TUQRAD) 2 )11 2

+/- 1.1542% RTP TLUb 2 = (WFcuF X PEb2cu)

= -0.0012% RTP TLUFWFOUL (WFFWFB X PEbFWFE)

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 82 of 89

= +0.5990% RTP 6.4.2 CASE 2- WESTINGHOUSE "CROSSFLOW" ULTRASONIC FLOW METERS USED FOR CORRECTION OF FEEDWATER FLOW ON A CONTINUOUS BASIS All random errors are combined in SRSS fashion, and bias terms are computed separately. The total measurement uncertainties of each parameter, which are derived in Section 6.3, are combined with weighting factors, which are derived in Section 6.2, in proportion to their total contribution to the calculation of uncertainties for the thermal power computation.

TLURTp(case 2) +/-((WFUFM X TUUFMXFLO) 2 + (WFFWT X TLUFWT) 2 + (WFsp X TLUsp) 2 + (WFCRDF X TLUCRDF) 2 + (WFCRDFAF X TLUCRDFAF) 2

+ (WFCRDT X TLUCRDT) 2 + (WFCUF X TLUCUF) 2 + (WFCUFAF X TLUCUFAF) 2 + (WFcuni x TLUCUT) 2 + (WFcuTo x TLUCUT)2 +

(WFRPP x TLURPP) 2 + (WFORAD x TUORAD) 2 )112 Unit 1 TLURTP.U1C2 +/-((WFUFM X TUUFMXFLO) 2 + (WFFwTt X TLUFWT) 2 + (WFspi X TLUsp) 2 + (WFCRDF X TLUCRDF) 2 + (WFCRDFAF X TLUCRDFAF) 2

+ (WFCRDT X TLUCRDT) 2 + (WFCUF x TLUCUFI) 2 + (WFCUFAF X TLUCUF1AF) 2 +(WFcuTi x TLUCUT) 2 +(WFcUT. x TLUCUT) 2 +

(WFRpp X TLURPP) 2 + (WFQRAD X TUQRAD)2)

TLUFRTP-U10C2 +/- 0.4559% RTP (For Cross Flow Uncertainty of +/-0.42%)

TLUb1 = (WFcuF X PEb1cu)

= -0.0014% RTP Unit 2 TLURTP-U2C2 = +/-((WFUFM X TUUFMXFLO) 2 + (WFFWT2 X TLUFW,) 2

+ (WFsp2 X TLUsp) 2 2

+ (WFcRDF X TLUCRDF) + (WFCRDFAF X TLUCRDFAF) 2 2

+ (WFcRDT X TLUCRDT) + (WFCUF X TLUcuF 2 ) + (WFCUFAF X TLUCUF2AF) 2 +(WFCUTi x TLUCUT) 2 + (WFcuTo X TLUcuT) 2 +

(WFRPP X TLURPP) 2 + (WFORAD x TUORAD) 2) 1'2 TLURTP-U2C2 = +/- 0.4598% RTP (For Cross Flow Uncertainty of +/-0.42%)

TLUb2 = (WFcUF x PEb2cu)

= -0.0012% RTP

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 83 of 89 7.0

SUMMARY

OF RESULTS This evaluation analyzes the measurement uncertainty of the heat balance at E.

I. Hatch Nuclear Plant, given 2 possible cases. Table 3 provides a summary of the total heat balance measurement uncertainties for each of the 2 analyzed cases.

Bias Random Bias Random CaeUnit 1 Uncert Unit 2 Uncert Case R Unit 1r l Unit 2

%RT) (RP (%RP) (%RTP)

1. No UFM Correction -0.001 +/-1.227 +0.001 /_+_1.154

2. Westinghouse "Crossflow" UFM to Continuously -0.001 +/- 0.456 -0.001 +/- 0.460 Correct the Venturi Reading Table 3 - Total Heat Balance Measurement Uncertainty Summary

8.0 CONCLUSION

S This study evaluates the measurement uncertainty of the heat balance computation at E. I. Hatch Nuclear Plant under the following two conditions:

Case 1: current configuration, using feedwater flow venturis to measure feedwater flow, and Case 2: configuration after installation of Westinghouse Crossflow Ultrasonic Flowmeters (UFMs), which will be used to automatically correct the venturi readings on a continuous basis within the Plant Computer.

The total heat balance measurement uncertainties of these two configurations are calculated and displayed in Table 3. These uncertainties only apply to approximately 100% power operating conditions. Although these uncertainties were computed for pre-power uprate conditions, due to the small amount of the power uprate, the uncertainties are still valid after the power uprate.

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 84 of 89

9.0 REFERENCES

9.1 Methodology and Industry Standards 9.1.1 ISA-RP67.04, Part II - 1994, "Methodologies for the Determination of Setpoints for Nuclear Safety Related Instrumentation", May 1995 9.1.2 ASME Interim Supplement 19.5 on Instruments and Apparatus, Application Part II of Fluid Meters, Sixth Edition, 1971 9.2 Procedures 9.2.1 34SV-SUV-025-OS, 'Core Heat Balance," Rev. 10.3 9.2.2 57CP-CAL-069-1S, "Rosemount Model 1151 AP, DP, AND GP Transmitters," Rev. 26.4 9.2.3 57CP-CAL-069-2S, "Rosemount Model 1151 Transmitters," Rev. 31.4 9.2.4 57CP-CAL-019-1S, "GE Type 555/556 Pressure Transmitter," Rev. 15.0 9.2.5 57CP-CAL-01 9-2S, "Bailey Type BQ and GE Type 555 & 556 Pressure Transmitters," Rev. 17 ED 3 9.2.6 57CP-CAL-052-1 S, "Rosemount Temperature / Flow Transmitter," Rev.

6.3 9.2.7 57CP-CAL-052-2S, "Rosemount Temperature Transmitter," Rev. 7.1 9.2.8 57CP-CAL-1 03-1 S. "ITT Barton Model 764 Differential Pressure

c. Transmitter," Rev. 19 ED 3 9.2.9 57CP-CAL-103-2S, "ITT Barton Model 764 Differential Pressure Transmitter," Rev. 15 ED 5 9.2.10 57CP-CAL-292-ON, "Ohio Semitronics PC5 Watt Transducer," Rev. 0 ED 1

9.2.11 571T-G31-002-1 S, "RWCU System Differential Flow Instrument FT&C,"

Rev. 0 ED 1 9.2.12 571T-G31 -002-2S, "RWCU System Differential Flow Instrument FT&C,"

Rev. 0 ED 2 9.3 Calculations 9.3.1 SINH 90-019, "Feedwater Flow Transmitter Calibration," Rev. 2 9.3.2 SINH 95-007, "Feedwater Flow Transmitter Calibration," Rev. 0 9.4 Drift Studies 9.4.1 SNC-007, "30-Month Drift Analysis for Rosemount 1151 Series Differential Pressure Transmitters, with Range Codes 4-8," Performed by EXCEL Services Corporation, Rev. 0 9.4.2 SNC-009, u30-Month Drift Analysis for Rosemount 1153 Series B or D Pressure Transmitters, with Range Code 9," Performed by EXCEL Services Corporation, Rev. 0 9.4.3 SNC-001, "30-Month Drift Analysis for Barton 764 Differential Pressure Transmitters," Performed by EXCEL Services Corporation, Rev. 0

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 85 of 89 9.5 Instrument Data Sheets 9.5.1 A-16466, Sheet C32B, Rev. 0 9.5.2 A-26466, Sheet C32A, Rev. 0 9.5.3 A-1 6466, Sheet C32D, Rev. 0 9.5.4 A-26466, Sheet C32C, Rev. 0 9.5.5 A-16440, Sheet C32B, Rev. 0 9.5.6 A-26440, Sheet C32A, Rev. 0 9.5.7 A-16454, Sheet C1 A, Rev. 0 9.5.8 A-26454, Sheet C 1A, Rev. 0 9.5.9 A-16472, Sheet B31A, Rev. 0 9.5.10 A-26472, Sheet B31 C, Rev. 0 9.5.11 S-1 8450, Sheet 18, Rev. 13 9.5.12 S-1 8450, Sheet 22, Rev. 8 9.5.13 S-29056, Sheet 20, Rev. 3 9.5.14 S-29056, Sheet 25, Rev. 18 9.5.15S-15210, Rev. 0 9.5.16 S-28165, Sheet 2, Rev. 23 9.5.17S-28165, Sheet 4, Rev. 5 9.5.18 S-1 9593, Sheet 53, Rev. 2 9.5.19 S-42348, Sheet 60, Rev. 2 9.5.20 S-1 8452, Sheet 2, Rev. 1 9.5.21 S-1 8452, Sheet 4, Rev. 1 9.5.22 S-23679, Sheet 2, Rev. 6 9.5.23 S-23679, Sheet 4, Rev. 0 9.5.24 S-1 8453, Sheet 26, Rev. 16 9.5.25 S-1 8453, Sheet 3, Rev. 22 9.5.26 S-27943, Sheet 3, Rev. 9 9.5.27 S-28165, Sheet 10, Rev. 21 9.5.28 S-28165, Sheet 22, Rev. 21 9.6 Vendor Manuals 9.6.1 SNC Vendor Manual SX-1 6821, "Final Instru. Manual Feedwtr. Flow Meter," Rev. 0 9.6.2 SNC Vendor Manual S-32271, "Instruction Manual," Dated Nov. 1, 1977 9.6.3 SNC Vendor Manual S-1 9238, Elements, Tab 12, "Orifice Bore Calculations for 1G31 -N035," Dated June 9,1972 9.6.4 SNC Vendor Manual S-81290, Rev. 1, Rosemount Product Manual 00809-0100-4593, "Model 1151 Smart Pressure Transmitters," Rev. Al, Dated 1996 9.6.5 SNC Vendor Manual SX-1 9672, Tab 6, "Installation and Operation Manual, Model 764 Differential Pressure Electronic Transmitter," Manual No. 88C4, Dated 1988 9.6.6 SNC Vendor Manual S-30696, Power Supplies, Tab 6, B&W/Bailey/GE Manual No 4532K30-01 OA, "Instructions, Type 570-06, 07 Isolated Power

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 86 of 89 Supply," Not Dated 9.6.7 SNC Vendor Manual S-30696, Transmitters, Tab 2, "G.E./Babcock &

Wilcox/Bailey Type 555 Differential Pressure Transmitter Factory Styles 1, 2 and 3," Manual No. 198 4532K16-300D, Rev. 2/70 9.6.8 SNC Vendor Manual S-53168, Ohio Semitronics Manual PC5TPH-RO,

'Watts Transducers - Installation Instructions," Rev. 1 9.6.9 SNC Vendor Manual S-44130, "Analogic Data Acquisition System Instruction Manual," General Electric Publication No. GEY-5657A, Volume 11, Part 1, Dated September 1989, and Supplement 1, Dated February 1993 9.6.10 SNC Vendor Manual S-71259, "PCRS Plant Computer System Operation and Maintenance Instruction Manual, Volume 2 Binder 2," General Electric Manual GEK-97227-2A, April 1999 9.6.11 Rosemount Data Sheet, "Model 414L Linear Bridges and Accessories,"

Fax from Rosemount Dated 11/28/2001 (Attachment B) 9.6.12 SNC Vendor Manual S-30698, Elements, Tab 7, "Orifice Bore Calculations for 2G31 -N035," Dated 6/9/72 9.6.13 SNC Vendor Manual S-1 6270, "Feedwater Flow Meter Section Purchase Specification," dated 8/25/70 9.7 Drawings 9.7.1 Drawing H-1 1604, "Piping & Instrumentation Diagram, Condensate &

Feedwater System, Sheet 3," Rev. 39 9.7.2 Drawing H-1 7844, uFeedwater Control System C32 Elementary Diagram, Sheet 3 of 8," Rev. 18 9.7.3 Drawing H-43844, "Process Computer Replacement System 1C95 Elementary Diagram, Sheet 5 of 8," Rev. 2 9.7.4 Drawing H-17845, "Feedwater Control System C32 Elementary Diagram, Sheet 4 of 6," Rev. 24 9.7.5 Drawing H-43850, "Process Computer Replacement System 1C95 Elementary Diagram, Sheet 11 of 16," Rev. 2 9.7.6 Drawing H-43851, "Process Computer Replacement System 1C95 Elementary Diagram, Sheet 12 of 16," Rev. 2 9.7.7 Drawing H-21038, 'Turbine Building Condensate & Feedwater System P&ID, Sheet 3 of 3," Rev. 3 9.7.8 Drawing H-27521, "Feedwater Control System 2C32 Elementary Diagram, Sheet 3 of 6," Rev. 13 9.7.9 Drawing H-51557, "Process Computer Replacement System 2C95 Elementary Diagram, Sheet 5 of 16," Rev. 2 9.7.10 Drawing H-27522, "Feedwater Control System 2C32 Elementary Diagram, Sheet 4 of 6," Rev. 16 9.7.11 Drawing H-51563, "Process Computer Replacement System 2C95 Elementary Diagram, Sheet 11 of 16," Rev. 3 9.7.12 Drawing H-51564, "Process Computer Replacement System 2C95 Elementary Diagram, Sheet 12 of 16," Rev. 3

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 87 of 89 9.7.13 Drawing H-21038, 'Turbine Building Condensate & Feedwater System P&ID, Sheet 3 of 3," Rev. 3 9.7.14 Drawing H-1 7752, "Nuclear Boiler Process Instrumentation System B21 Elementary Diagram," Rev. 24 9.7.15 Drawing H-27465, "Nuclear Boiler Process Inst. Sys. 2B21A Elementary Diagram, Sheet 1 of 1," Rev. 22 9.7.16 Drawing H-43844, "Process Computer Replacement System 1C95 Elementary Diagram Sheet 5 of 16," Rev. 2 9.7.17 Drawing H-43845, "Process Computer Replacement System 1C95 Elementary Diagram Sheet 6 of 16," Rev. 2 9.7.18 Drawing H-43850, "Process Computer Replacement System 1C95 Elementary Diagram Sheet 11 of 16," Rev. 2 9.7.19 Drawing H-43851, "Process Computer Replacement System 1C95 Elementary Diagram Sheet 12 of 16," Rev. 2 9.7.20 Drawing H-51557, uProcess Computer Replacement System 2C95 Elementary Diagram Sheet 5 of 16," Rev. 2 9.7.21 Drawing H-51558, "Process Computer Replacement System 2C95 Elementary Diagram Sheet 6 of 16," Rev. 2 9.7.22 Drawing H-51563, "Process Computer Replacement System 2C95 Elementary Diagram Sheet 11 of 16," Rev. 3 9.7.23 Drawing H-51564, "Process Computer Replacement System 2C95 Elementary Diagram Sheet 12 of 16," Rev. 3 9.7.24 Drawing H-1 7844, "Feedwater Control System C32 Elementary Diagram, Sheet 3 of 6," Rev. 18 9.7.25 Drawing H-27521, "Feedwater Control System 2C32 Elementary Diagram, Sheet 3 of 6," Rev. 13 9.7.26 Drawing H-1 6065, "Control Rod Drive System P&ID, Sheet 2," Rev. 40 9.7.27 Drawing H-17115, "Control Rod Drive Hyd. Instr. Sys. - C11 Elementary Diagram, Sheet 3 of 3," Rev. 7 9.7.28 Drawing H-26007, "Control Rod Drive System P&ID, Sheet 2," Rev. 33 9.7.29 Drawing H-27518, "Control Rod Drive Hyd. Instr. Sys. - 2C11 B Elementary Diagram, Sheet 3 of 3," Rev. 9 9.7.30 Drawing H-1 6188, "Reactor Water Clean-Up System P&ID, Sheet 1," Rev.

57 9.7.31 Drawing H-1 7177, "Reactor Water Clean-Up System G31 Elementary Diagram, Sheet 2 of 4," Rev. 37 9.7.32 Drawing H-26036, "Reactor Water Clean-Up System P&ID, Sheet 1," Rev.

38 9.7.33 Drawing H-27731, "Reactor Water Clean-Up System 2G31 A Elementary Diagram, Sheet 2 of 3," Rev 20 9.7.34 Drawing S-1 5072, "P&ID, Reactor Recirculation System," Rev. 9 9.7.35 Drawing H-17863, "Reactor Recirc. Pump & M.G. Set 'A' Sys. B31 Elementary Diagram, Sht. 4 of 10," Rev. 14 9.7.36 Drawing H-17905, "Reactor Recirc. Pump & M.G. Set 'B' Sys. B31 Elementary Diagram, Sht. 4 of 10," Rev. 17 9.7.37 Drawing S-26003, "Reactor Recirculation System P&ID Sheet 1," Rev. 27

Attachment 1 to HL-6328, Enclosure 7 E. l. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 88 of 89 9.7.38 Drawing S-26004, "Reactor Recirculation System P&ID Sheet 2 Electrical Diagram," Rev. 6 9.7.39 Drawing H-27484, "Reactor Recirc. Pump and M.G. Set 'A' Sys. 2B31 Elementary Diagram, Sheet 4 of 10," Rev. 17 9.7.40 Drawing H-27493, "Reactor Recirc. Pump and M.G. Set 'B' Sys. 2B31 Elementary Diagram, Sheet 4 of 10," Rev. 16 9.7.41 Drawing S-15093, "Control Rod Drive Hydraulic System - Arrangement Master Controls Area," Rev. C 9.7.42 Drawing S-25307, "Control Rod Drive Hydraulic System - Arrangement Master Controls Area," Rev. 3 9.7.43 Drawing H-1 6888, "Reactor Water Clean-Up System Discharge Piping from Pumps COO1A&B," Rev. 2 9.7.44 Drawing H-26854, "RWCU System From Reactor Recir. Loop to Non-Regen. Heat Exchanger," Rev. 6 9.7.45 Drawing S-1 5382, "Temperature Element Assembly Drawing - General Use, Sheet 1," Rev. 7 9.7.46 Drawing S-27302, "PPD - Temp Element, Sheet 1," Rev. 15 9.7.47 Drawing S-17364, "Flow Nozzle Assy CRD Hydraulic Cont. Sys.," Rev. 2 9.8 Miscellaneous 9.8.1 E. I. Hatch Nuclear Power Plant, Unit 2 Final Safety Analysis Report (FSAR), Rev. 19 9.8.2 ASME Steam Tables, 5th Edition, 1967 9.8.3 NRC Safety Evaluation Report, Dated March 20, 2000, Response to Crossflow Topical Report CENPD-397-NP-A Rev. 01 9.8.4 E. I. Hatch Process Computer Program Listing and Description for Heat Balance Computation (Attachment A) 9.8.5 NUCLEIS Equipment Listings, Dated 11/12/01 to 11/13/01 9.8.6 SNC Manual S-25193, "BWR Equipment Environmental Interface Data,"

GE Document No. 22A2928, Rev. 2 9.8.7 SNC EQ Specification No. SS-2102-238, "Environmental Qualification Requirements for Safety-Related Class 1E Equipment, Components, and Instrumentation," Rev. 9 9.8.8 Advanced Measurement & Analysis Group, Inc. (AMAG) Document SRS-7132-06, "Software Requirements Specifications For Hatch Algorithm and Correction Layer Software Development Project," September 2002, Rev.

00 Draft C 9.8.9 The Temperature Handbook, Omega Complete Temperature Measurement Handbook and Encyclopedia, Volume 27,1989 9.8.10 General Electric Nuclear Energy Report, "Impact of Steam Carryover Fraction on Process Computer Heat Balance Calculations," Dated September 2001 9.8.11 Caldon Report PR-244, "Caldon Experience in Nuclear Feedwater Flow Measurement," Rev. 15, Dated June, 2001 9.8.12 Generating Plant Performance, Southem Company Services, Plant Field

Attachment 1 to HL-6328, Enclosure 7 E. I. HATCH NUCLEAR PLANT SINH-02-069, Rev. 0 HEAT BALANCE UNCERTAINTY EVALUATION Sheet 89 of 89 Services Test Report No. EWO: H184-BW of Test Series HA01 06, uEdwin 1.Hatch Nuclear Plant, Unit 1 Extended Uprate Plant Performance Test, May, 1999," Issued September 24,1999 9.8.13 Generating Plant Performance, Southern Company Services, Plant Field Services Test Report No. EWO: H184-BW of Test Series HA0204, "Edwin I. Hatch Nuclear Plant, Unit 2 Extended Uprate Plant Performance Test, November, 1998," Issued September 23,1999 9.8.14 SNC Document No. S-1 5131, G.E. Design Specification, uControl Rod Drive System," Rev. 8 9.8.15 Hatch Feedwater Flow Correction and Digital Filtering Algorithms (Attachment E), R. L. Miller, Dated October 17, 2002 9.8.16 G.E. Nuclear Energy Letter from R. E. Kingston, to K. S. Folk,

Subject:

Hatch-2 Cycle 13 3D MONICORE Databank Revision, Dated November 8, 1995 9.8.17 G.E. Nuclear Energy Letter from C. J. Paone, to K. S. Folk,

Subject:

Hatch 1 and 2 Feedwater Flow Coefficients, Dated May 2,1990 9.8.18 Southern Nuclear Corporation Response to Significant Occurrence Report C09505439

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing C GENERAL ELECTRIC COMPANY PROPRIETARY INFORMATION C

C Copyright General Electric Company,1984,1987,1988,1989,1991,1992,19 93

,1 9 9 4 C

C These materials are copyrighted and proprietary trade secret subject C matter, all rights reserved. Publication has been made in a limited, C copyright sense, and does not affect or limit any rights, obligations, C or remedies with respect to activities violating legal or equitable C rights, including trade secret rights, that are not equivalent to any C of the exclusive rights within the general scope of the copyright laws.

C Use or copying of all or any portion of these programs including the C preparation of derivative works is prohibited except with the expressed C written authorization from General Electric Company.

C CG3SRO1CTPSB C

SUBROUTINE G3SRO1CTPSB (LU ,ERRORCODE)

C C PURPOSE:

C C G3P6 (alias G3MCIL) performs heat balance calculation. It also C

C INPUT DESCRIPTION:

C C ARGUMENTS:

C C Lu - File code of error logging file C

C GLOBAL VARIABLES:

C C G3P6CONSTe - Plant constants Common Block C COMDAS - Common Block - PSC values and status C

C OUTPUT DESCRIPTION:

C C ARGUMENTS:

C C ERRORCODE - Return error code; O=Good, l=Bad C

C GLOBAL VARIABLES:

C C SRlCOM - Common Block with Heatbalance data C

C SUBROUTINES CALLED:

C C G3CONVERT C LWRWERR C LCGETERTM C LCERTOD C HGP C HFP C HPTL C HGSF C HSCF Page 1 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing C

C SPECIAL NOTES:

C C (1) THE DATABASE OF PLANT DATA IS ASSUMED TO CONTAIN STATUS WORDS C FOR EACH THERMODYNAMIC SENSOR WHICH HAS THE VALUE 'GOOD',

C 'REMO', 'FAIL', OR 'SUBS' TO INDICATE STATUS OF THE SENSOR C SIGNAL TO BE RESPECTIVELY GOOD, REMOVED FROM SCANNING BY THE DAS, C FAILED, AND REMOVED FROM SCANNING WITH A SUBSTITUTE VALUE SUPPLIED C BY THE OPERATOR.

C C (2) PLANT THERMODYNAMIC SENSOR DATA C

C THE FOLLOWING SCALARS AND ARRAYS ARE DEFINED TO CONTAIN THE C PLANT THERMODYNAMIC SENSOR READINGS RETURNED TO THE PROGRAM C FROM THE DAS IN THE "COMDAS" LABEL COMMON.

C C PSCDATA(I), I=1,NSC - PLANT THERMODYNAMIC SENSOR READINGS C PSCSTAT(I), I=1,NSC - PLANT THERMODYNAMIC SENSOR SCAN STATUS C

C PSCDATA / PSCSTAT ARRAY assignment for a plant with 2 clean-up C flow branches (NCUB=2), 2 feedwater branches (NFWB=2),

c 2 recirculation loops (NPUMP=2) is as follows:

C C 1. WCR Control rod system flow C 2. TCR Control rod system flow temperature C 3. WCUA Cleanup system flow, branch A C 4. WCUB Cleanup system flow, branch B C 5. TCU1 Cleanup system inlet temperature

_1 C 6. TCU2 Cleanup system exit temperature C 7. MWPA Recirc pump A motor power C 8. MWPB Recirc pump B motor power C 9. PRG Reactor pressure C 10. DPM Core pressure drop C 11. RWL Reactor water level C 12. TFWA1 Feedwater temperature 1, branch A C 13. TFWA2 Feedwater temperature 2, branch A C 14. TFWB1 Feedwater temperature 1, branch B C 15. TFWB2 Feedwater temperature 2, branch B C 16. GMWE Generator gross power C 17. WT Total core flow C 18. WDA1 Driving flow 1, loop A C 19. WDA2 Driving flow 2, loop A C 20. WDB1 Driving flow 1, loop B C 21. WDB2 Driving flow 2, loop B C 22. TDA1 Inlet temperature 1, loop A C 23. TDA2 Inlet temperature 2, loop A C 24. TDB1 Inlet temperature 1, loop B C 25. TDB2 Inlet temperature 2, loop B C 26. WFWA Feedwater flow, branch A C 27. WFWB Feedwater flow, branch B C

C WAVX(I), I=1,NFWB - Digitally Averaged C Feedwater Branch Flows In Nfwb Branches C RAPDATA(I), I=1,NAP - APRM Channel Readings C For NAP APRM Channels, Scan Units Page 2 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing C RAPSTAT(I), I=1,NAP - Status Words For APRM C Channel Sensors C

C (3) IT IS ASSUMED THAT IF ANY OF THE PLANT SENSORS ARE REMOVED FROM C SCANNING IT SHALL BE POSSIBLE FOR THE OPERATOR TO MANUALLY INSERT C A SUBSTITUTE VALUE FOR THE SENSOR READING INTO THE DBMS, THE C SUBSTITUTE VALUE OCCUPYING THE POSITION IN THE DATABASE WHERE C THE NORMAL VALUE WOULD BE STORED IF THE SENSOR WERE GOOD AND C BEING SCANNED.

C C (4) IT IS FURTHER ASSUMED THAT THE DAS SHALL MAINTAIN THE VALUES C IN THE DBMS IN THE UNITS SPECIFIED BY THE I/O LIST.

C CONVERSION TO STANDARD INTERNAL UNITS WILL BE PERFORMED IN THE C THIS PROGRAM.

C C (5) IN ADDITION TO THE SENSORS ABOVE THE FOLLOWING STATUS WORDS ARE C ASSUMED TO EXIST IN THE DATABASE:

C C FWBA(I), I=1,NFWB - FEEDWATER BRANCH ACTIVE STATUS HAVING THE C VALUE 'ACTI' IF ACTIVE AND THE VALUE 'PASS' C IF NOT ACTIVE. IF SIGNALS DO NOT EXIST, SET C THE VALUE TO 'ACTI' AS AN ASSUMED VALUE.

C C RLA(I), I=1,NPUMP - RECIRC LOOP ACTIVE STATUS HAVING THE VALUE C 'ACTI' IF ACTIVE AND THE VALUE 'PASS' IF C NOT ACTIVE. IF THE STATUS SIGNALS DO NOT C EXIST, SET THE VALUE TO 'ACTI' AS AN ASSUMED C VALUE.

C C ELOC - EQUALIZER LINE OPEN/CLOSED STATUS HAVING THE VALUE 'OPEN' C IF THE EQUALIZER LINE IS OPEN AND HAVING THE VALUE 'SHUT' C IF THE EQUALIZER LINE IS NOT OPEN. IF THE SIGNAL IS NOT C PRESENT, AND AN EQUALIZER LINE IS PRESENT, SET THE VALUE C TO 'OPEN' AS AN ASSUMED VALUE. IF THE EQUALIZER LINE IS C NOT PRESENT, SET THE VALUE TO 'SHUT'.

C C PROCEDURE:

C C The procedure is outlined in the special note above. Refer C to 22A6701AA specification for more detail.

C C REQUIREMENTS REF:

C C This module was developed from and conforms to the functional C requirements in the General Electric specification 22A6701AA, rev. 0.

C IMPLICIT NONE INCLUDE 'PDL:G3SROlCTPSB.INC/LIST' INCLUDE '(DF NSSIDB NSSICLAS)/LIST' INCLUDE '(DF NSSIDBLIVEPD)/LIST' INCLUDE 'GPRM:G3NSSIPRM.GPM/LIST' Page 3 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing INCLUDE 'GCMN:G3P6CONST.GMN/LIST' INCLUDE 'Lcmn:G3SRlCOM.cmn/list' INCLUDE 'LCMN:G3COMDAS.CMN/LIST' Character TEXT*50  ! Error text Character*23 asciitime  ! System Time in ASCII format Integer*4 eristime(2)  ! System Clock Time in Eris Time Format Integer*4 LU  ! File code of error logging file Integer*4 ERROR-CODE  ! Return Error Code; 0=good, l=bad Real*4 G3CONVERT  ! Unit conversion Real*4 TFWSUB (NFWBZ)  ! Feedwater temperature Real*4 HFWSUB(NFWBZ)  ! Feedwater entholphy Real*4 WFWSUB (NFWBZ)  ! Feedwater flow Real*4 TDT(NPUMPZ) ! Recirc inlet temperature Real*4 TEM Local for limit comparision Real*4 PUMPPOWER  ! Local variable for Pump Power Real*4 FEEDFLOW Local variable for Feedwater Flow Real*4 delta Local veriable Parameter (delta=l.Oe-20) ! Small enough to test (.eq. 0)

Integer*4 ILFLAG(NSCZ)  !PSC sensor status array Integer*4 I, J, K !Loop counter Integer*4 IP  ! Local variable Integer*4 N3  ! Local variable Integer*4 INDX  ! Local variable Integer*4 LOC  ! Local variable Integer*4 NUMP  ! No of recirc pump local counter Integer*4 JLOOP  ! Loop counter Integer*4 P_LOC  ! Local variable Integer*4 WDLOC  ! Local variable C PSC sensor status defination Byte GOOD /0/,  ! PSC sensor status evluation 1 FAIL /1/,

1 SUBS /2/,

1 REMO /3/

Integer*4 1 ACTI /'ACTI'/,

1 PASS /'PASS'/,

1 SHUT / 'SHUT'!,

1 OPEN /'OPEN'/

C PSC sensor criticality code Integer*4 Critical  ! Critical Sensor, not use last value Integer*4 Uselastgv  ! Non critical sensor, use last value Integer*4 Usezero  ! Non critical sensor, use 0.0 value Parameter (Critical = 0)

Parameter (Use_lastgv = 1)

Parameter (Use-zero = 2)

Page 4 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing C PSC array index for each type of PSC data Integer*4 iWCR  ! Control rod system flow Integer*4 iTCR Control rod system flow temperature Integer*4 iWCU  ! Cleanup system flow, branch A Integer*4 iTCU  ! Cleanup system inlet temperature Integer*4 iMWP  ! Recirc pump A motor power Integer*4 iPR  ! Reactor pressure Integer*4 iDPM  ! Core pressure drop Integer*4 iRWL  ! Reactor water level Integer*4 iTFW  ! Feedwater temperature 1, branch A Integer*4 iGMWE  ! Generator gross power Integer*4 iWT  ! Total core flow Integer*4 iWD Driving flow 1, loop A Integer*4 iTD Inlet temperature 1, loop A Integer*4 iWFW  ! Feedwater flow, branch A Logical*4 FirstTime/ .true./

C IREC (RECIRC LOOP ACTIVE FLAG) BITS SETTING Integer*4 BITVALUE(5)/l, 2, 4, 8, 16/  ! BIT 0, 1, 2, 3, 4 ON C Controlling Flags To send or Not to send the Warning Messages:

Logical*4 ALLLOOPFLAG Logical*4 ALL_FWB_FLAG Logical*4 WDLOOPFLAG(NPUMPZ)

Logical*4 TDLOOPFLAG (NPUMPZ)

Logical*4 EPSWD3DFLAG(NPUMPZ)

Logical*4 EPSTD3DFLAG(NPUMPZ)

Logical*4 EPSWFWFLAG(NFWBZ)

C External (Heatbalance calculation)

Real*4 HGP Real*4 HFP Real*4 HPTL Real*4 HGSF Real*4 HSCF External HGP External HFP External HPTL External HGSF External HSCF C +---------------------+

C I First Get System Time to stamp error log I C +-----------------------+

call lcgetertm (eristime)  ! Get system time in ERIS form Page 5 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing call lcertod( eristime, asciitime)  ! Convert ERIS time in ASCII C +----------------------+

C I If it is first time, perform some sanity check: I C +----------------------+

if( FirstTime ) then FirstTime = .false.

! Set default values for these error messages falgs all_loop flag = .tr=Le.  ! recirc loops availability flag allfwb flag = .trtLe.  ! feedwater branch availability flag do i = 1, npumpz wdloop flag(i) = .true.  ! driving flow not available flag tdloop flag(i) = .true.  ! driving temp not available flag epswd3dflag(i) = .true.  ! two driv. fl. differ more flag epstd3dflag(i) = .true.  ! two driv temp differ more flag end do do i = 1, nfwbz epswfwflag(i) = .true. ! digital av and inst.fwf comp. flag end do

! Check Against Assumed Capacities For NFWB, NPUMP and NCUB IF(NCUB .LT.1 .OR. NCUB .GT. 3) THEN TEXT =' NUMBER OF CLEANUP BRANCHES MUST BE 1 TO 3' CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT GO TO 9000 <<<<<<<<<<<<< FATAL ERROR EXIT >>>>>>>>>>>>>>>>>>

END IF IF(NFWB .LT. 1 .OR. NFWB .GT. 3) THEN TEXT = ' NUMBER OF FEEDWATER BRANCHES MUST BE 1 TO 3' CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT GO TO 9000 ! <<<<<<<" FATAL ERROR EXIT >>>>>>>>>>>>>>>>>>

END IF IF(NPUMP .LT. 1 .OR. NPUMP .GT. 5) THEN TEXT = ' NUMBER OF RECIRC LOOPS MUST IBE 1 TO 5' CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT GO TO 9000 ! <<<<<<<<<<<<< FATAL ERROR EXIT >>>>>>>>>>>>>>>>>>

END IF

! calculate a PSC array index for each type of PSC data iWCR = 1  ! Control rod system flow iTCR = 2  ! Control rod system flow temperature iWCU = 3  ! Cleanup system flow, branch A Page 6 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing iTCU = iWCu + ncub  ! Cleanup system inlet temperature iMWP = iTCU + 2  ! Recirc pump A motor power iPR = iMWP + npump  ! Reactor pressure iDPM = iPR + 1  ! Core pressure drop iRWL = iDPM + 1  ! Reactor water level iTFW = iRWL + 1  ! Feedwater temperature 1, branch A iGMWE = iTFW + 2*nfwb  ! Generator gross power iWT = iGMWE + ngen  ! Total core flow iWD = iWT + 1  ! Driving flow 1, loop A iTD = iWD + 2*npump  ! Inlet temperature 1, loop A iWFW = iTD + 2*npump  ! Feedwater flow, branch A end if C +-----------------------+

C I Start with Good CTP calculation status I C +--------------------+

KCTP = 0  ! Init CTP calc status to GOOD CTP = 0.0  ! Initialize CTP to 0.0 C +--------------------+

C I First Check CONSTANTS.CED datasets values I C +--------------------+

DO I = 1, NFWB IF( FWNM(I) .EQ. 0.0 ) THEN  ! Bad Nominal FW Flow TEXT = 'FWNM('//char(48+i)//') 0 in CONSTANTS.CED file' CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT GO TO 9000 ! <<<<<<<<<<<<< FATAL ERROR EXIT >>>>>>>>>>>>>>>>>>>

END IF END DO DO I = 1, NPUMP IF( WDNM(I) .EQ. 0.0 ) THEN  ! Bad nominal maximum driving flow TEXT = 'WDNM('//char(48+i)//') = 0 in CONSTANTS.CED file' CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT GO TO 9000 ! <<<<<<<<<<<<< FATAL ERROR EXIT >>>>>>>>>>>>>>

END IF END DO C +---------------------------+

C I Convert the PSC data to the right units of measurement I C +----------------------+

! G3CONVERT converts the new thermodynamic data from DAS if

! necessary from metric to english units of measure. All NSS i calulations and data bank constants are in english units.

PSCDATA( iWCR) = G3CONVERT(0,1,PSCDATA( iWCR) )

! P60PTCRD NSSIDB item is set to TRUE if CRD temperature Page 7 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing sensor exist in which case the plant DAS input is used.

Otherwise, P60PTCRD is FALSE and the CRD temperature input is obtained from the CT2 dataset value of CONSTANTS.CED file.

if( P60PTCRD ) then  ! plant sensor exist PSCDATA( iTCR) = G3CONVERT(0,5,PSCDATA( iTCR) else PSCDATA( iTCR) = CT2 plant sensor does not exist PSCSTAT( iTCR) = GOOD end if DO i = 0, NCUB-1 PSCDATA( iWCU +i) = G3CONVERT(O,1,PSCDATA( iWCU +i))

END DO PSCDATA( iTCU) = G3CONVERT(0,5,PSCDATA( iTCU))

PSCDATA( iTCU+l) = G3CONVERT(0,5,PSCDATA( iTCU+l))

! PR is gauge pressure; do NOT convert PR to absolute pressure

! here.

PSCDATA( iPR) = G3CONVERT(0,7,PSCDATA( iPR))

DPM is a delta pressure.

D----------------------D-P PSCDATA( iDPM) = G3CONVERT(0,7,PSCDATA( iDPM))

PSCDATA( iRWL) = G3CONVERT(0,6,PSCDATA( iRWL))

PSCDATA( iWT) = G3CONVERT(0,1,PSCDATA( iWT))

DO I=0,2*NPUMP -1 PSCDATA( iWD +i) = G3CONVERT(0,1,PSCDATA( iWD +i))

PSCDATA( iTD +i) = G3CONVERT(0,5,PSCDATA( iTD +i))

END DO DO I=O, 2*NFWB -1 PSCDATA( iTFW +i) = G3CONVERT(0,5,PSCDATA( iTFW +i))

END DO DO I=O, NFWB -1 PSCDATA( iWFW +i) = G3CONVERT(O,1,PSCDATA( iWFW +i))

WAVX( i+1) = G3CONVERT(O,1,WAVX( i+1))

END DO C +---_---------+

C I Calculate ILFLAG and PSC arrays I C +---------------- +

! Load Sensors Reading and Sensor Status to PSC and ILFLAG arrays.

! Use PSC sensor criticality code in assigning last good value.

Page 8 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing do i = 1, nsc ilflag( i ) = pscstat( i )

if( (pscstat( i ) .eq. good) .and.

1 (pscdata(i) .ge. rll~i)) .and.

1 (pscdata(i) .le. rul(i)) ) then psc( i ) = pscdata( i )  ! good value, use it else if( (pscstat( i ) .eq. subs) .and.

1 (pscdata(i) .ge. rll(i)) .and.

1 (pscdata(i) .le. rul(i)) ) then psc( i ) = pscdata( i )  ! good sub value, use it else

! Use criticality code to assign last good value.

if ( critical_psci(i) .eq. critical ) then psc( i ) = -1.0  ! do not use last good value else if ( critical_psci(i) .eq. uselastgv ) then psc( i ) = ct(i)  ! use last good value else if ( criticalpsci(i) .eq. use_zero ) then psc( i ) = 0.0  ! set value to 0.0 else psc( i ) = -1.0  ! assign bad value.

end if if ( ilflag (i) .eq. good ) then ilflag Mi) = fail !set status to fail end if end if end do C +-----------------------+

C I Calculate the failed sensor list, IFSL, and update CT array. I C +---------------------------+

nfsl = 0 do i=l,nfsedz ifsl(i) = 0  ! initialize to zero end do do i=l,nsc if( ilflag(i) .gt. good ) then  ! not good status nfsl = nfsl + 1 if( nfsl .le. nfsedz ) ifsl(nfsl) = i else ct(i) = psc(i)  ! update latest good scan values end if end do Page 9 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing C +-+

C l Check aprm readings and add any failed APRM Identification to the l C failed sensor list. I C +----------------------------+

do i=l,nap if(rapdata(i) .lt. rll(nsc+l) .or. rapdata(i) .gt. rul(nsc+l) 1 .or. rapstat(i) .eq. fail .or. rapstat(i) .eq. remo) then rap(i) = 0.

nfsl = nfsl + 1 if( nfsl.le.nfsedz ) ifsl(nfsl) = nsc+i else rap(i) = rapdata(i) end if end do C +-----___________________+

C l Calculate Recirc Loops active flags and l C l Feedwater Branches open/close Flags C +-------------------+

if(epspump .eq. 0.0) epspump = 0.1 !set default to 0.1 mw if(epswdflw .eq. 0.0) epswdflw = 0.05 !set default to 5% of wdnm Use Recirc Pump Power and drive flow in calculation of Recirc loop active flag do i= 1, npump p-loc = imwp-l+i wdloc iwd + (i-l)*2 if (eusclppw .ne. 0) then !dvd by nssidb eu scaling constant pumppower = psc(p_loc)/eusclppw else pumpypower = psc(p_loc) end if ppw(i) = pumpypower if ( (ilflag(p-loc) .eq. good) .or.

1 (ilflag(p-loc) .eq. subs) ) then

!pump power signal is good, use pump power for testing if ( pump_power .gt. epspump ) then rla(i) = acti  ! recirc loop is active else rlaWi) = shut  ! recirc loop is inactive ppw(i) = 0.0  ! set pump power to zero end if else if ( (psc(wdloc) .gt. epswdflw*wdnm(i) ) .or.

1 (psc(wdloc+l) .gt. epswdflw*wdnm(i) ) ) then

! drive flow greater than minimum, loop is active.

rla(i) = acti Page 10 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing else rla(i) = shut ppw(i) = 0.0 end if end do Use instantaneous Feedwater Flow for Feedwater branch active flag if(epsfwflw .eq. 0.0) epsfwflw = 0.01 !set default to 1% of fwnm loc = iwfw - 1 do i= 1, nfwb feedflow = psc(loc+i) if (feedflow.gt.-1.1 .and.feedflow .lt. -0.9)KCTP=2 !critical and

!FAIL or REMO fwba(i) = acti if (feedflow .lt. epsfwflw

• fwnm(i) ) then fwba(i) = shut feedwater branch is closed end if end do

! Calculate Number of Active Recirc Loops

! and Number of Active Feedwater Branches if all loops are active set irec with all bits off if loop 1 is inactive set irec with bit # 0 on if loop 2 is inactive set irec with bit # 1 on if loop 3 is inactive set irec with bit # 2 on if loop 4 is inactive set irec with bit # 3 on if loop 5 is inactive set irec with bit # 4 on irec = 0  ! start with all pumps active no_a_rcloop = npump  ! start with all pumps active do i=l,npump if(rla(i) .ne. acti) then irec = irec + bit value(i) no_arcloop = no a-rc loop - 1 end if end do no_a fw br = nfwb  ! start with all fw branches active do i=1, nfwb if(fwba(i) .ne. acti) then no_a_fw_br = no_a_fwbr - 1 end if end do Send Warning Message if Number of Active FW branches less than NFWB OR Number of Active Recirc Loops less than NPUMP Page 11 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing if(no a fw-br .lt. nfwb) then if ( allfwbflag ) then write(text, '(a,i3)')

1 ' NO OF ACTIVE FEEDWATER BRANCHES LESS THAN ', NFWB CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT allfwb flag = .false.

end if else all_fwbflag = .true.

end if if(no_a_rc_loop .lt. npump) then if ( allloop_flag ) then write(text, '(a,i3)')

1 NO OF ACTIVE RECIRC LOOPS LESS THAN ', NPUMP CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT allloop-flag = .false.

end if else all_loopflag = .true.

end if C +-----------------------+

C I Calculate Measured or Operator Supplied Core Flow I C +-------------------------- +

WTOPS = -1.0  ! Operator supplied Core Flow WT = -1.0  ! Measured Core Flow LOC = iWT  ! WT flow PSC base index IF( ILFLAG(LOC) .EQ. good ) WT = PSC(LOC)

IF( ILFLAG(LOC) .EQ. subs ) WTOPS = PSC(LOC)

C +------------------------+

C I Apply Engineering Unit Scaling Constants and Decode PSC array I C +--------------------------+

! control rod system flow IF (EUSCLWCR .NE. 0) THEN WCR = PSC(iWCR) / EUSCLWCR ! Dvd by NSSIDB EU Scaling Constant ELSE WCR = PSC(iWCR)

ENDIF

! control rod system flow temperature TCR = PSC(iTCR)

! cleanup system flow Page 12 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing WCU = 0 DO I=O,NCUB-1 IF (EUSCLWCU .NE. 0) THEN WCU = WCU+PSC(iWCU+I)/EUSCLWCU IDvd By NSSIDB EU Scaling Constant ELSE WCU = WCU+PSC(iWCU+I)

END IF END DO

! cleanup system temperature TCU1 = PSC(iTCU)

TCU2 = PSC(iTCU+1)

! reactor dome pressure, pressure drop and water level PR = PSCUiPR) + 14.7  ! added 14.7 for abs. pressure (PSIA).

DPM = PSC(iDPM)

RWL = PSC(iRWL)

! generator power LOC = iGMWE IF( NGEN .EQ. 1)THEN GMWE(1) = PSC(LOC)

GMWE(2) = 0.

ELSE GMWE(1) = PSC(LOC)

GMWE(2) = PSC(LOC+1)

END IF C +---------------+

C I Set Equalizer OPEN/SHUT Flags I C +-----____________+

IF (IEQLZR .EQ. 1 )THEN IEQL=1 !Equalizer Line Is Shut ELSE IF (IEQLZR .EQ. 2 )THEN IEQL=O lEqualizer Line Is Open ELSE IF (IEQLZR .EQ. 0 )THEN IF (ELOC .EQ. OPEN )THEN IEQL=0 ELSE IEQL=1 END IF END IF C +------------------+

C I CALCULATE JET PUMP DRIVING FLOW I C +---------------- +

Page 13 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing WD = 0.

DO I = 1, NPUMP  ! Number of Recirc Pump J = iWD +(I-1)*2  ! Drive Flow Index IF( RLA(I) .eq. ACTI ) TH

!Set the bad reading to the other good reading.

IF( (ILFLAG(J) .GT. 0) .AND. (ILFLAG(J+1) .EQ. 0) ) THEN PSC(J) = PSC(J+l)  ! next one is "good" status END IF IF( (ILFLAG(J+l).GT. 0) .AND. (ILFLAG(J) .EQ. 0) ) THEN PSC(J+l) = PSC(J)  ! previous one is "good" status END IF

!If both readings are bad, set WD = -1.0 IF( (PSC(J) .LE. 0.0) .AND. (PSC(J+l) .LE. 0.0) ) THEN IF ( WDLOOPFLAG(I) ) THEN WRITE(TEXT, '(A,I3,A)')

1 BOTH DRIVING FLOWS OF ACTIVE LOOP',I,' ARE BAD.'

CALL LWRWERR (TEXT,-'DUMY')

CALL SEND LOG( '3DFMC028', TEXT WD LOOP FLAG(I) = .false.

END IF WD = -1.0  ! No Jetpump Driving Flow GO TO 1180  ! Bad WD value ELSE WDLOOPFLAG(I) = .true.

END IF

!Now compare both readings

!----------------C--------

TEM = ABS( PSC(J) - PSC(J+1) ) / WDNM(I)

IF (EpsWD3D .eq. 0.0) EpsWD3D = 0.05  ! Set default to 0.05 IF( TEM .GT. EpsWD3D ) THEN

! two readings are different to the level which

! exceeds the 5t range, so set to a bad reading IF ( EPSWD3DFLAG(I) ) THEN WRITE(TEXT, I(A,I3,A)')

1 ' DRIVE FLOWS OF LOOP',I,' DIFFER MORE THAN LIMIT' CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT )

WRITE(TEXT, '(A,F6.4,A,F6.4)')

1 ' DIFFERNCE = ', TEM, ' LIMIT = ', EPSWD3D Page 14 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT EPSWD3DFLAG(I) = .false.

END IF WD = -1.0 GO TO 1180 Bad WD value ELSE EPSWD3DFLAG(I) = .true.

END IF If we are here, Both readings are good. Take the average WD = WD + 0.5*( PSC(J) + PSC(J+1)

END IF END DO 1180 CONTINUE C +----------------------+

C ICALCULATE RECIRC FLOW INLET TEMPERATURE AND ENTHALPY I

+

C +--------------------------

TD = 0.0 HD = 0.0

! If the driving flow is bad, set Temperature TD

! and Enthalpy HD to bad IF(WD .LE. 0.0) THEN TD = -1.0  ! bad temp.

HD = -1.0  ! bad enthalpy GO TO 1220 END IF DO I = 1, NPUMP  ! Number of Recirc Pump J = iTD +(I-1)*2  ! Inlet Tempt. Index IF( RLA(I) .eq. ACTI ) THEN

!Set the bad reading to the other good reading.

IF( (ILFLAG(J) .GT. 0) .AND. (ILFLAG(J+I) .EQ. 0) ) THEN PSC(J) = PSC(J+l)  ! next one is "good" status END IF IF( (ILFLAG(J+l).GT. 0) .AND. (ILFLAG(J) .EQ. 0) ) THEN PSC(J+l) = PSC(J)  ! previous one is "good" status END IF

!If both readings are bad, set TD = -1.0 Page 15 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing IF( (PSC(J) .LE. 0.0) .AND. (PSC(J+l) .LE. 0.0) ) THEN IF ( TDLOOPFLAG(I) ) THEN WRITE(TEXT, '(A,I3,A)')

' BOTH INLET TEMP OF ACTIVE LOOP ',I,' ARE BAD.'

CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( 13DFMC028, TEXT TDLOOPFLAG(I) = .false.

END IF TDT(I) = 0.0  ! Bad Inlet Temperature GO TO 1215 ELSE TDLOOPFLAG(I) = .true.

END IF

!Now compare both readings TEM = ABS( (PSC(J)/PSC(J+1)) - 1.0)

IF (EpsTD3D .eq. 0.0) EpsTD3D = 0.03  ! Set default to 0.03 IF( TEM .GT. EpsTD3D ) THEN two readings are different to the level which

! exceeds the 3% range, so set to a bad reading IF ( EPSTD3DFLAG(I) ) THEN WRITE(TEXT, '(A,I3,A)')

1' INLET TEMP OF LOOP',I,' DIFFER MORE THAN LIMIT' CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT WRITE(TEXT, '(A,F6.4,A,F6.4)')

1' DIFFERNCE = ', TEM, ' LIMIT = ', EPSTD3D CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT EPSTD3DFLAG(I) = .false.

END IF TDT(I) = 0.0  ! Bad Inlet Temperature GO TO 1215 ELSE EPSTD3DFLAG(I) = .true.

END IF

! If we are here, Both readings are good. Take the average TDT(I) = 0.5*(PSC(J)+PSC(J+1))

ELSE TDT(I) = 0.0 END IF 1215 CONTINUE Page 16 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing END DO  ! End of Inlet Temp. Computation Loop

! Calculate Inlet Temp and Enthalpy as long as

! one loop has good reading JLOOP = 0 TD = 0.0 DO I= 1, NPUMP IF (TDT(I) .GT. 0.0) THEN TD = TD + TDT(I)

JLOOP = JLOOP +1 END IF END DO

! Test if all loops signal are bad IF (JLOOP .EQ. 0) THEN

! All Inlet Termprature Readings aare bad.

! Set Tepmerature and Enthalpy to bad.

TD = -1.0 HD = -1.0 ELSE TD = TD / JLOOP IF (IETCAL .EQ. 0) THEN HD = HSCF(PR,TD,HFT,TDIV,BTC)

ELSE HD = HPTL(PR, TD)

END IF END IF 1220 CONTINUE C +------------------+

C I Calculate Feedwater Temperature. I C +-------------------+

! If only one of the redundant feedwater temperature sensors is

! failed, set the failed sensor reading equal to the non-failed

! sensor reading.

DO I=1, NFWB IF( FWBA(I) .EQ. ACTI ) THEN  ! Branch is Active J = iTFW +(I-1)*2  ! FW temperature PSC base index IF( (ILFLAG(J) .GT. 0) AND. (ILFLAG(J+l) .EQ. 0) ) THEN PSC(J) = PSC(J+l)  ! next one is "good" status Page 17 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing END IF IF( (ILFLAG(J+l).GT. 0) .AND. (ILFLAG(J) .EQ. 0) ) THEN PSC(J+1) = PSC(J)  ! previous one is "good" status END IF

!If both readings are bad for any one of the active

!branch, then set the CTP flag to unkown IF( (PSC(J) .LE. 0.0) .AND. (PSC(J+l) .LE. 0.0) ) THEN KCTP = 2  ! 2 = CTP unknown TFWSUB(I) = 0.0  ! No Good Data ELSE  ! Branch is Inactive TFWSUB(I) = (PSC(J)+PSC(J+1))/2.  ! Take average value END IF ELSE  ! Branch is Inactive TFWSUB(I) = 0.0  ! set 0.0 for inactive branch END IF END DO C +-+---____________

C I Calculate feedwater flow rate. I C +---------------- +

LOC iWFW-1  ! = NCUB+5*NPUMP+2*NFWB+NGEN+B  ! FW flow PSC base index -1 DO I = 1, NFWB IF( FWBA(I) .EQ. ACTI ) THEN

!If the reading is bad for the active

!branch, then set the CTP flag to unkown IF( PSC(LOC+I) .LT. 0.0 ) THEN  ! failed or removed KCTP =2 2 =CTP unknown ELSE TEM = ABS( WAVX(I) - PSC(LOC+I) ) / FWNM(I)

IF(TEM .GT. EPSWFW) THEN ILFLAG(LOC + I) = 4  ! bad FW digital reading KCTP = 2  ! 2 = CTP unknown NFSL = NFSL + 1 IF (NFSL .LE. NFSEDZ) IFSL(NFSL) = NSC - NFWB + I IF ( EPSWFWFLAG(I) ) THEN  ! Send Warning Message WRITE(TEXT, '(A,I3,A)')

' FEEDWATER FLOW BRANCH ',I, 2 ' DIFFER MORE THAN LIMIT' CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT WRITE(TEXT, '(A,F6.4,A,F6.4)')

' DIFFERNCE = ', TEM, ' LIMIT = ', EPSWFW CALL LWRWERR (TEXT,'DUMY')

CALL SEND LOG( '3DFMC028', TEXT Page 18 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing EPSWFWFLAG(I) = .FALSE.

END IF ELSE EPSWFWFLAG(I) = .TRUE.

END IF END IF ELSE WAVX(I) = 0.0 END IF WFWSUB(I) = WAVX(I)

END DO C ----------------------- +

C Evaluate the status of other sensors required for C Heat balance calculation. If bad, set KCTP = 2 C +--------------------+

! Check that the following PSC data required for CTP

! calculation is available: WCR, TCR, WCUW( ncub ), TCU1, TCU2,

! PPW( npump ), PRG. Note, PSC data for TFW1/2( nfwb ) and

! WFW( nfwb ) have all ready been checked as reflected by

! KCTP = 2 if they are unavailable.

DO I= 1, iPR IF (PSC(i) .EQ. -1.0 ) KCTP = 2  ! 2 = CTP unknown END DO C +------------------------------------+

C I Set CTP calculations to default values for unknown CTP condition C +---------------------------------+

IF (KCTP .EQ. 2) THEN CTP = 0.0  ! RP6( 2)

HG = 0.0  ! RP6( 10)

HS = 0.0  ! RP6( 11)

HF = 0.0  ! RP6( 12)

HFG = 0.0  ! RP6( 13 WFW = 0.0  ! RP6( 16 TFW = 0.0  ! RP6( 17)

HFW = 0.0  ! RP6( 18 QFW = 0.0  ! RP6( 19 QP = 0.0  ! RP6( 23)

HCU1 = 0.0  ! RP6( 32)

HCU2 = 0.0  ! RP6( 33 QCU = 0.0  ! RP6( 34)

HCR = 0.0  ! RP6( 37 )

QCR = 0.0  ! RP6( 38 )

Page 19 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing GO TO 8000 1 Rturn with good status END IF C +-----------------------------+

C I If we are here, we got good status and reading of scan data.l C +--------------------------+

! Calculate Feedwater flow by adding all active branches.

! Use Degital Average Feedwater flow.

WFW = 0.0 DO I=1,NFWB IF( FWBA(I) .EQ. ACTI ) WFW = WFW + WAVX(I)

END DO C +------------------+

C I Calculate energy balance terms. I C +---------------+

! If IETCAL .EQ. 0 then use old STEAM TABLE which is based on FITS to

! either KEENAN and KEYES or ASME data; otherwise use new STEAM TABLE

! which is based on the ASME (or PANACEA model) data.

! saturated steam enthalpy IF (IETCAL .EQ. 0) THEN HG = HGSF(PR,HGC,PLHG)  ! saturated steam ELSE HG = HGP(PR)

END IF saturated liquid steam enthalpy IF (IETCAL .EQ. 0) THEN HF = HFC (1) + (HFC(2) + HFC(3)*PR)*PR  ! saturated liquid ELSE HF = HFP(PR)

END IF HFG = HG - HF vaporization

! Subcooled water (Feedwater) enthalpy DO I = 1, NFWB IF (FWBA(I) .EQ. ACTI) THEN IF (IETCAL .EQ. 0) THEN HFWSUB(I) = HSCF (PR, TFWSUB(I), HFT, TDIV, BTC)

ELSE HFWSUB(I) = HPTL (PR, TFWSUB(I))  ! subcool liquid Page 20 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing END IF ELSE HFWSUB(I) = 0.0  ! inctive branch END IF END DO

! Control Rod Drive Flow enthalpy IF (IETCAL .EQ. 0) THEN HCR = HSCF (PR,TCR,HFT,TDIV,BTC)  ! Cont. Rod Driv ELSE HCR = HPTL(PR, TCR)

END IF

! Cleanup loop inlet enthalpy IF (IETCAL .EQ. 0) THEN HCU1 = HSCF (PR,TCU1,HFT,TDIV,BTC)  ! Cleanup loop inlet ELSE HCU1 = HPTL(PR, TCU1)

END IF

! Cleanup loop exit enthalpy IF (IETCAL .EQ. 0) THEN HCU2 = HSCF (PR,TCU2,HFT,TDIV,BTC)  ! Cleanup loop exit ELSE HCU2 = HPTL(PR, TCU2)

END IF HS = HG - FM*HFG IF (Cl .EQ. 0.0) C1 = 3.413  ! MBTU/MWH CONVERSION FACTOR QCR = WCR*(HS-HCR)/C1 QCU = WCU*(HCUl-HCU2)/Cl Energy added to Recirc Pumps QP = 0.

DO I=1,NPUMP QP = QP + PPW(I)*ETAl END DO Feedwater water JLOOP = 0 TFW = 0.

HFW = 0.

IF (SRlOPT1) THEN  ! simple averaging Page 21 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment A Plant Computer Heat Balance Code Listing DO I = 1, NFWB IF (FWBA(I) .EQ. ACTI) THEN TFW = TFW + TFWSUB(I)

HFW = HFW + HFWSUB(I)

JLOOP = JLOOP + 1 END IF END DO IF ( JLOOP .GT. 0 ) THEN  ! Avoid devid by zero TFW = TFW/JLOOP HFW = HFW/JLOOP ELSE TFW = 0.0 HFW = 0.0 END IF ELSE DO I = 1, NFWB  ! wfwsub(i) = 0 for inactive branches TFW = TFW + TFWSUB(I)*WFWSUB(I)

HFW = HFW + HFWSUB (I)*WFWSUB (I)

END DO IF (WFW .GT. 0.001) THEN  ! Avoid devid by zero TFW = TFW / WFW HFW = HFW / WFW ELSE TFW = 0.0 HFW = 0.0 END IF END IF QFW = WFW*(HS-HFW)/Cl C +-___________________________________

C I Now calculate Core Thermal Power I C +-__________________________+

CTP = QFW+QCR+QCU+QRADX-QP C +__-_____________+

C! I Return Status I C +__-_____________+

8000 continue ERRORCODE = 0 RETURN  ! NORMAL RETURN 9000 CONTINUE ERRORCODE = 1  ! G3P6 WILL TERMINATE RETURN  ! FATAL ERROR RETURN END Page 22 of 22

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment B Rosemount 414L Temperature Transmitter Specifications HIV-2-2Ol 12:24 RSMia Uff NXLE;R 61288828e2 P.01 MODEL 414l 1.

I LINEAR BRIDGES t~ IM AND ACCESSORIES

_ ¢S@tiT <'>>. :_ ' I i

t II i

i MIL1JVCLrTS I I

Ii posi FexaNOxe 7671 4' I To RIL az - m 6 co. 2f.

-54 i

TMPERATLRE TrIMPERA'¢LRE t

Linear mv/degree output for computer input or digital indication

_._ to 0.1% with platinum RTD's Accurate

..i with a miallvolt readout device, tfat de'ice DESCRIPIION becomes adirectreading temperature indicator.

The Model 414L Linear Bridge* converts The millivolt/degree output is ideal for'use the resistance of a platinum resistlae tn-in computer systemsbecause Me signal does 30t peraiwre sensor to a millivolt per degree output require storage space for correction factors.

signal. The output is zero millivolts it zero degrees (F or C). The Slope Of the ou-put Is Differentil temperature measurements can one millivolt per degree withineha non4-nesrity be radeby using twolinearbridges as Sb'OWi an tolerance shown in the ordering information.

When the output of a liner bridge is interfaced page 2.

"Rosemount J% tNTe .0 &*3*

Page 1 of 4

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment B Rosemount 414L Temperature Transmitter Specifications HOJ-28-201 12:24 ROSErMT Nu-EAR b1FM '

ORDERING INFORMATION Available from stock except as noted.

41M4L UNIEAR RIDGE (

CODE INPUT VOLTAGE 2 l15 VAC CODE I SENSOpt Ro_

nBridg^eson are properly trimrned to mate with sensors A 100 n purchased the same order. Provide model and serial C*B 200 n numbers for Rosemount-sensorv-purchas edwd-tereut-C 400 n order. Provide complete RvsT Informaton for platinum

• D goo a sensors not manufactured by Rosemount.

• E 2000 n CalibraUon data sheets accompanying each bridge 6JnI 50 completely tdentify the mating sensors.

CODE I LOAD RESISTANCE A 10 KS?2 F Load resistance means the Input resistance B 25 xn of your DVM, DPM. A/D convertor or recorder.

C 50 (Note: Slide wire recorders are usually infinite E 500 Kr? ohms at null. Specify 10 megohms as the load.)

F l megn G 10 meg CODE I RANGE NON-LINEARITY (MAX) ,

A -lO0 to +500- F *0. 03%

B 0 to +100 r tO.06% (

D -350 to +1000'F 0. 17%

• E 0 to -1350- F *0. 10%

F -200 to +500 C *0. 17%

G -100 to +200 C '0. 03%

1 O to +200 C A0. 03%

J O to +500' C *0. 06%

• X 0 to +750 C *0 10%

414SL- 3- A- G- A - TYPICAL MODEL NUMBER ACCESSORIES t These options are not stacked-._

SPECIFY ITEM MODEL NO 200n R. stocked for temperature range "A" only.

Socket ............................. 4.....

20-8 Single Channel Chassis ..... 4205 Six Channel Chassis ..... 420L Selector Switch Kit ..... 420-14 420L with Selector Switch Kit Installed (5 Channel Capacity),... 420L-14 R osem ount Inc. PHONE!

s POSTOOF69C4 oIMNEPOUS BOX13519 MINNESOTA955L3i5 (612)941 5560 TWX 910-57&3103 TELEX-2940tn CAsBLROSEMPOUNT Page 2 of 4

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment B Rosemount 414L Temperature Transmitter Specifications IJc¢W-kd1 1d-*4 XtjtLAjU I NULL&HK b1.dlC.ucW r.,

TYPICAt IIJCTAI I ATI^1U

XMODEL 414L UNCAR BRIDGE.

OAC CHA 5 MCR

• _4-WIRE PLATINUM RtESSTANCC TEMPERATURte SENSOR SPECIFICATIONS )IFFERENTIAL OPERATION Dlfferential Temperatures can be measured ACCURACY 40.1%of span y ConnectiDg the negattve output terminals of OUTPUT I mv/F or*C wo 414L31s and connecttng the readout between INPUT POWER leir positive terminals.

118 VAC, 6l0%, 50/60 Hz or 28 VDC. SC ,VAC EFFECT OF INPUT POWER VARIATIONS Output will not shift by more than 0. 01% for

'o To*K 2 a *10% llne change.

AMBIENT TEMPERATURE 40 to 140'F.

EFFECT OF AMBIENT TEMPERATURE

• 0 005/F change in ambient temperature NOTE: The readout device must have onc LEADWIRE EFFECT iegobm or greater Input Impedance for aT Output will not change by more than 0 05' peration. The absolute temperature readings for a 62nehange in 3n of balanced sensor lead- Ion% each bridge are not affected by the AT wires. conection.

2 Page 3 of 4

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment B Rosemount 414L Temperature Transmitter Specifications t~-29-2W0l 12:25 ROSEMONT NUCLEA 6128288260 P.04

- DIMENSIONAL DR R ~ - w m y w ACCESSURILE FAOK -IO -- _.

tosr 3-II 5oUMS us COA 91 MODEL 420-8 SOCKET rfr INo OR15IOU The Model 420-8 Socket offers a simple way to mount the Model 414L in cabinets or on the baekof panels. A protective strip, marked 4th a connection diagram, covers the power termi-nals.

MODEL 4205 CHASSIS The Model 420S Chassis includes a line cord and fully-nclosed 115 volt power connectionfor bench or panel mounting one Model 414L3 bridge.

A terminal strip provides connection points for the sensor and signal leadwir*s.

MODEL 420L CHASSIS orztYnesAnt hag &uaee-*ie line cord, switch, fuseandpilotllght. e chas ltsin a stand-ard 19-inrh relay rack and has rear-mounted terminal stripe.

I. MODEL 420-14 SWITCH KIT A modification Idt for the 420L Chassis, the Model 420-14 replaces the 6th socket and pro-vides switch selected output of the other five bridges to a common pair of output terminals.

MODEL 420S CHASSIS Available as a kit or factory insfalled_

, { _._. __. - -- ,.. -- ,__

II _,_

-_j

• I MOOEL 420. CHASS1S MODCL 420.-1 BWITCH 3

Page 4 of 4

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment C

_- Feedwater Flow Differential Pressure Transmitter Calibration Computations Unit 1 Transmitter A Tap Set 1 Unit 1 Transmitter B Tat Set 1 Feedwater Temp 397.5 deg F Feedwater Temp 397.5 deg F Total FW Flow 1.15140E+07 Ibm/hr Total FW Flow 1.15140E+07 Ibm/hr FW Flow per Loop 5.75700E+06 Ibm/hr FW Flow per Loop 5.75700E+06 Ibm/hr Upstream Tap Pressure 1116 sia Upstream Tap Pressure 1116 psia vif 0.0185018 Ibm/ft3 vfl 0.0185018 Ibm/t 3 D1 54.04879525 ft3/lbm D1 54.04879525 R3/lbm Vt 0.0185018 Ibmlf 3 Vt 0.0185018 Ibm/ft3 gr 5.75700E+06 Ibm/hr qr 5.75700E+06 Ibm/hr a 32.137 W/sec 2 n 32.137 ft/sece Alpha 9.16900E-06 inrin/deg F Alpha 9.16900E-06 in/in/deg F Fa 1.006051499 Fa 1.006051499 d 7.737 inches d 7.7405 inches Beta 0.4999 Beta 0.4999 _

2 mu 2.77800E-06 Ibf-sec/f 2

mu 2.77800E-06 Ibf-sec/f Rr 3.53321 E+07 Rr 3.53161 E+07 Cavg 0.9918 Cavg 0.9933 . -

Ravg 4.2191 OE+06 Ravg 4.90260E+06 Cr 0.994339104 Cr 0.99562174 hr 1245.816943 inWC hr 1240.363154 inWC hs 2405.70 inWC hs __2395.17 invC SPE 1.45 1//I1000psi SPE 1.45 0/%/10psi CF 0.015968851 CF 0.01596885 hsc 2367.889084 inWC hsc 2357.523221 inWC Unit 2 Transmitter A Tar Set 2 Unit 2 Transmitter B Tar Set 2 Feedwater Temp 425.1 de F Feedwater Temp 425.1 deg F Total FW Flow 1.19500E+07 Ibm/hr Total FW Flow 1.19500E+07 Ibm/hr FW Flow per Loop 5.97500E+06 Ibm/hr FW Flow per Loop 5.97500E+06 Ibm/hr Upstream Tap Pressure 1122 psia Upstream Tap Pressure 1122 psia v2 0.018909356 Ibm/ft3 vf2 0.018909356 Ibm/ft3 D2 52.88387399 ft3 bm D2 52.88387399 ft3 bm Vt 0.018909356 Ibm/f 3 vt 0.018909356 Ibm/ft3 gr 5.97500E+06 Ibm/hr gr 5.97500E+06 Ibm/hr

__32.137 ft/sec2 _ 32.137 ft/sec2 Alpha 9.22600E-06 in/in/deg F Alpha 9.22600E-06 in/in/deg F Fa 1.006600064 Fa 1.006600064 d 7.775 inches d 7.775 inches Beta 0.5161 Beta 0.515 2

mu 2.57500E-06 Ibf-sec/t 2 mu 2.57500E-06 Ibf-sec/f Rr 3.93675E+07 Rr 3.93675E+07 Cavq 0.9968 Cavg 1.0003 Ravg 4.41140E+06 Ravg 4.18720E+06 Cr 0.999376607 . - Cr 1.002952824 hr 1317.869158 inWC hr 1309.336864 inWC hs 2362.53 inWC hs 2347.23 inWC SPE 1.45 //1 000psi SPE 1.45 %/1000psi CF 0.01605585 CF 0.016055851 hsc 2325.19226 inWC hsc 2310.138243 inWC Page 1 of 1

Hatch Heat Balance Uncertaic., Evaluation SINH-02-069, Rev. 0 C Attachment D Tabulation of On-Line Plant Computer Heat Balance Parameters B020 B023 B024 B025 B030 B031 8032 8033 ACRDTEMP B014 Date/Time B015 B016 8018 B019 3 387 536 404 438.673 1034.619 426.889 426.413 426.786 427.089 125.247 0.026 11/1 000 5.957 6 062 0.09 3.326 3.391 536.404 438 589 1034.802 426.986 426 511 426.942 427.206 125.355 0.026 11/1 0.01 5.91 6026 0.091 3313 3.413 536.404 438 673 1034.802 426.869 426.433 426.766 427.011 125.247 0.026 11/1 0 02 5 876 5.986 009 3.313 3.409 536.404 438.673 1034.435 426.928 426.433 426 825 427.069 125.355 0.026 11/1 0.03 5.904 6.1 009 3.33 3394 536.324 438.673 1034.435 426.791 426.374 426.688 426.972 125.355 0.026 11/1 0 04 5.928 604 009 3.329 34 536.404 438.504 1034 619 426.732 426.374 426.649 426.933 125.355 0026 11(1 0:05 5.938 6032 0.09 3314 3377 536.324 438.589 1034.435 426.908 426 433 426.805 427.089 125.247 0.026 11/1 0.06 5.925 6 077 0.091 3326 3.391 536 404 438 504 1034.802 426.928 426.511 426.864 427.089 125.355 0026 11/1 0.07 5.975 6.053 009 3.314 3.4 536.404 438.673 1034.802 427.025 426 55 426 962 427.225 125.355 0026 11/1 0 08 5.938 6.055 0.09 3.335 3.388 536.404 438.589 1034.619 42683 426.452 426.766 427.05 125.247 0.026 11/1 009 5.958 6.056 009 3314 3397 536.404 438.589 1034.802 426.791 426.413 426.727 427.011 125 247 0.026 11/1 0:10 5.884 6051 0.091 3.314 536 404 438.673 1034.802 427.006 426.491 426.922 427.167 125.247 0.026 11/1 0-11 5.91 6.027 009 3.308 3392 3.388 536.404 438.589 1034 985 427.006 426.491 426.922 427.206 125.247 0.026 11/1 0:12 5.949 6075 009 3343 34 536.404 438.589 1034.802 427.045 42655 426.942 427.186 125.247 0.026 11/1 0:13 5.971 6004 0.09 3.36 3.377 536.404 438.673 1034.802 427.006 426 53 426.922 427.225 125.247 0.026 11/1 0:14 5.949 6.054 0.09 3.359 3.393 536.485 438 673 1034.985 426.928 426.472 426.805 427.069 125.355 0 026 11/1 0:15 5.906 6.032 0.09 3 343 3402 536.404 438.758 1034.985 426 869 426.452 426.786 427.03 125 247 0.026 11/1 0:16 5.929 6.057 009 3.361 3.394 536.404 438.589 1034.802 427.025 426.55 426.903 427.186 125.355 0.026 11/1 0:17 5.952 6.054 009 3.353 3.408 536.404 438.504 1034.985 426.791 426 413 426.747 427.05 125.355 0026 11/1 0:18 5.914 6024 0.09 3.346 Recirc Recirc RWCU RWCU Rx FW TemF FW TemF FW Temp FW Tem Ail 4 FW CRD Syst CRD FW Flow FW Flow RWCU D s Flow Parameter Loop A, Loop B, Flow PP Mtr A PP Mtr B InIt Tmp Outl Tmp Pressure FW Tem FW Tem FW Tem FW Temp TempsT Deg F PSIG DegF Temp, Deg F MIbm/hr Mlbm/hr Mlbm/hr Mlbmuhr Pow MW Pow MW Deg F 438 6198 1034.754 426.9133 426.4656 426.8278 427.0981 426.8262 125.2981579 0.026 Mean 5.930158 6 045526 0.090158 3.331632 3.394789 536.3998 0 St Dev 0 027811 0.026364 0.000375 0 018133 0.009641 0.032517 0 070163 0.18168 0.09523 0.057547 0 093708 0.089824 0.246183 0 055402831 0.140325 0.36336 0.190459 0.115095 0.187417 0.179648 0.492366 0.110805662 0 2 Std Devs 0.055623 0.052727 0.000749 0 036266 0 019282 0.065034 125.247 0 026 Min 5.876 5 986 0.09 3.308 3.377 536 324 438.504 1034.435 426.732 426.374 426.649 426.933 426.374 3 413 536.485 438.758 1034.985 427.045 426 55 426.962 427.225 427.225 125.355 0.026 Max 5.975 6.1 0 091 3.361 0 036 0.161 0.254 0.55 0.313 0.176 0 313 0 292 0.851 0.108 0 Range 0 099 0 114 0.001 0 053 19 19 19 19 19 19 19 19 76 19 19 No. Readings 19 19 19 19 4.3224 533.7 436.8 1035 425.1 425.1 425.1 425.1 425.1 123 5 0.03 Heat Bal Value 5.975 5.975 0.1 4.3224 Page 1 of 1

Hatch Heat Balance Uncertainty Evaluation SINH-02-069, Rev. 0 Attachment E Hatch Feedwater Flow Correction and Digital Filtering Algorithms

', rhe equation for calculating Feedwater Flow for a single loop is FWLoopNFlow_ corrected = [1 + C1 * (AT) + C2 * (AT) * *2]

• FWLoopNFlow Output point ID's are C51 C7001 for Loop A and C51 C7002 for Loop B (used later) where N = Loop Identifier (Aor B)

AT = (FWLoopTempN - RatedFWTemp)

FWLoopTempN = (FWLoopTemplN + FWLoopTemp2N)12 Average FW Temperature C1 and C2 are constants based on the expansion properties of the venturi throat.

INPUT Point ID's for the algorithm are FWLoopTemplA = B030 FOR LOOP A FWLoopTemp2A = B031 FWLoopTemplB = B032 FOR LOOP A FWLoopTemp2B = B033 K...FWLoopAFlow = B015 B015 and B016 are RAW Feedwater Flows in mlb/hr FWLoopBFlow= B016 UNIT CONSTANTS FROM THE POINT DEFINITION DATABASE UNIT C1 C2 Rated FW Temp 1 -0.3485E-03 -0.4156E-06 393.0 2 -0.3858E-03 -0.4481E-06 424.0 Each Corrected FW Loop Flow is'composed" once every 4 seconds. On the same frequency, the each corrected FW Loop Flow is "smoothed"/digitally filtered according to the following algorithm*:

C51C7001 _sm = [prevC5lC7001 sm * (Sm _factor -1)] + C5lC7001 _sm]I Sm _ factor Smoothing factor is 15 for both units.

• (database algorithm ID"N20" - GE Doc # 23A5250, sheet 184 section 30.4.1.12)

Page 1 of 1 RL Miller 12/16/2002