ML22195A177

NSAL-14-2: Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties

ML22195A177
Person / Time
Site:	Beaver Valley, Millstone, Salem, Indian Point, Harris, Wolf Creek, Point Beach, Watts Bar, Sequoyah, Byron, Braidwood, Summer, Prairie Island, Seabrook, Surry, North Anna, Turkey Point, Diablo Canyon, Callaway, Farley, Robinson, South Texas, Cook, Comanche Peak
Issue date:	04/14/2014
From:	Kucharski R Westinghouse
To:	Bhatt S, Casner M, Pierce W NRC/NRR/DSS/SNSB, Tennessee Valley Authority
Santosh Bhatt NRR/DSS 301-415-6066
References
TVA-14-26, WAT-D-12062 NSAL-14-2
Download: ML22195A177 (10)

Text

Westinghouse Electric Company Nuclear Services 401 River Terminal Road Chattanooga, TN 37406 USA

©2014 Westinghouse Electric Company LLC All Rights Reserved Electronically Approved Records are Authenticated in the Electronic Document Management System WAT-D-12062 TVA-14-26 April 1, 2014 Mr. Mike Casner Mr. William Pierce Site Engineering Director Site Engineering Director Tennessee Valley Authority Tennessee Valley Authority Watts Bar Nuclear Plant Sequoyah Nuclear Plant P.O. Box 2000 EQB-1A P.O. Box 2000 - OPS 4A Spring City, TN 37381 Soddy Daisy, TN 37384 TENNESSEE VALLEY AUTHORITY WATTS BAR & SEQUOYAH NUCLEAR PLANTS Nuclear Safety Advisory Letter, NSAL-14-2, Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties Gentlemen:

Please find attached the following document for your use:

Nuclear Safety Advisory Letter NSAL-14-2, Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties dated March 03/31/2014.

The NSAL communicates the outcome of our safety evaluation in accordance with 10CFR 21.

This NSAL has been reviewed and approved by the PWROG Emergent Issue Group.

It applies to the current Westinghouse NSSS plants, Westinghouse NSSS plants with non-Westinghouse designed SGs, and the AP1000 plant.

This issue is not reportable under 10CFR Part 21 for the operating plants and AP1000 plants or 10 CFR 50.55(e) where Westinghouse has a contractual requirement to inform our AP1000 plant customers.

There is sufficient conservatism in the methods used in the analyses to compensate for this issue for both operating plants and the AP1000 plants.

There is possibly an operability concern for plants receiving this NSAL if there is insufficient margin in their technical specification value for containment peak

WAT-D-12062 TVA-14-26 April 1, 2014 Electronically Approved Records are Authenticated in the Electronic Document Management System pressure; therefore, information supporting an Operability Determination (OD) is provided as an Appendix A to the NSAL.

This issue may affect the Plant Technical Specifications. It is possible that the containment Integrated Leak Rate Test (IRLT) pressure may be exceeded for some operating plants and the AP1000 plant. A search for margin in analysis inputs or analysis methods may be necessary to try to reduce the peak pressure to the plant technical specification Pa value or the plant would need to make a technical specification change.

If you have any questions, please contact Ms. Linda Evans at 423-697-5078 or me at 423-697-5052.

Very truly yours, Ronald Kucharski Customer Projects Manager

/gt

Attachment:

NSAL-14-2 cc:

Ron Baumer TVA rnbaumer@tva.gov Eric Byrnes TVA enbyrnes@tva.gov Louis Belvin TVA lbbelvin@tva.gov Tom Carter TVA tjcarter@tva.gov Kasey Decker TVA kcdecker@tva.gov John Pope TVA jtpope@tva.gov Wes Daniel TVA dwdaniel@tva.gov Frank Koontz TVA fakoontz@tva.gov Todd Noe TVA ptnoe@tva.gov Chris Reneau TVA wcreneau@tva.gov Dave Lafever TVA dmlafever@tva.gov William Stackler TVA wtstackler@tva.gov Chris Carey TVA ccarey@tva.gov Jim Lemons TVA jflemons@tva.gov David Brown TVA dmbrown1@tva.gov William Geise TVA wbgeise@tva.gov OEProgram@tva.gov Linda Evans WEC Gerri Thurmond WEC Renee Giampole WEC

Westinghouse Non-Proprietary Class 3 Electronically approved records are authenticated in the electronic document management system Nuclear Safety Advisory Letter This is a notification of a recently identified potential safety issue pertaining to basic components supplied by Westinghouse.

This information is being provided so that you can conduct a review of this issue to determine if any action is required.

1000 Westinghouse Drive, Cranberry Township, PA 16066

© 2014 Westinghouse Electric Company LLC. All Rights Reserved.

Subject:

Westinghouse Loss-of-Coolant Accident Mass and Energy Release Calculation Issue for Steam Generator Tube Material Properties Number: NSAL-14-2 Basic Component: Containment Mass and Energy Release Analysis Date:

03/31/2014 Substantial Safety Hazard or Failure to Comply Pursuant to 10 CFR 21.21(a)

Transfer of Information Pursuant to 10 CFR 21.21(b)

Advisory Information Pursuant to 10 CFR 21.21(d)(2)

Yes No N/A Yes Yes

SUMMARY

The loss-of-coolant accident (LOCA) mass and energy (M&E) release analyses are sensitive to energy stored in the reactor coolant system (RCS) metal, including the steam generator (SG) tubes. Recently, it was determined that the input modification program database and the input modification program preprocessor were using the density for stainless steel in determining the mass of the SG tubes and the specific heat (Cp) of stainless steel for the stored metal energy. Since all current Westinghouse-designed SGs use either Alloy 600 or Alloy 690 material for the SG tubes, there is a deviation from as-built plant parameters. Additionally, four plants for which Westinghouse has completed LOCA M&E calculations have non-Westinghouse-designed steam generators that have tubes manufactured from Alloy 800 material.

Westinghouse is providing this notification as a matter of best practice to inform affected customers of this issue and its potential impact on peak containment pressure. The increase in the LOCA M&E release associated with this issue affects the plant specific containment LOCA blowdown and post-blowdown transient conditions. For all affected plants, it has been determined that the maximum increase in containment pressure due to correcting the model for the Alloy 600, Alloy 690, or Alloy 800 SG tube material is 0.3 psi. This issue will not cause containment pressure to exceed the design limit if it was left uncorrected. Therefore, this issue could not create a substantial safety hazard (SSH) as defined by 10 CFR Part 21 if left uncorrected, and it is not reportable under 10 CFR Part 21.

Additional information, if required, may be obtained from Robert M. Jakub, (412) 374-4627 Author:

Reviewer Manager:

Gary W. Whiteman Regulatory Compliance Mark. J. Whitney Regulatory Compliance James A. Gresham Regulatory Compliance Verifier:

Verifier:

Robert M. Jakub Containment and Radiological Analysis Richard B. Lukas Containment and Radiological Analysis

NSAL-14-2 Page 2 of 8 This Nuclear Safety Advisory Letter (NSAL) is applicable to LOCA M&E release calculations (specifically for containment integrity peak pressure calculations) performed for Westinghouse-designed pressurized water reactors (PWRs) utilizing the methodology documented in WCAP-10325-P-A and WCAP-8264-P-A, Revision 1 (References 1 and 2, respectively). Combustion Engineering nuclear steam supply system (NSSS) designs and boiling water reactor (BWR) designs are not affected by this issue.

The identified issue is in addition to the other issues previously identified in NSAL-06-6 (Reference 3) and NSAL-11-5 (Reference 4).

ISSUE DESCRIPTION The LOCA M&E release analyses are sensitive to energy stored in the RCS metal, including the SG tubes. Recently, it was determined that the input modification program database and the input preprocessor were using the density for stainless steel in determining the mass of the SG tubes and the stored energy based on a stainless steel Cp. Since all current Westinghouse-designed SGs use either Alloy 600 or Alloy 690 material for the SG tubes, there is a deviation from the as-built plant parameters.

Additionally, four plants for which Westinghouse has completed LOCA M&E calculations have non-Westinghouse-designed SGs that have tubes manufactured from Alloy 800 material.

The LOCA containment response analysis is performed to demonstrate that the energy released into containment from a LOCA will not result in the calculated pressure exceeding acceptance criteria established for a given plant. Historically, the containment design pressure has been used by utilities as the basis for setting the Containment Integrated Leak Rate Test (ILRT) Technical Specification (TS) pressure (Pa) value. A revision to 10 CFR 50, Appendix J has allowed licensees to reduce their ILRT Pa to the LOCA calculated peak containment pressure, with little or no margin. As a result, a small increase in the calculated peak containment pressure could result in exceeding a plants TS limit for the ILRT Pa.

TECHNICAL EVALUATION The magnitude of the LOCA calculated peak containment pressure for Westinghouse-designed PWRs is a function of the containment heat sinks and engineered safety features equipment, which includes the fan coolers, containment spray system and passive systems like the ice condenser containments and the AP10001 plant. Depending upon the effectiveness of each of these systems, the containment peak pressure time can vary from the early blowdown period to the longer term post-reflood period. Given this range of performance differences, the penalty chosen for this safety significance determination must be bounding, and therefore can be overly conservative in some applications, depending upon the limiting break location. Westinghouse has performed sensitivity studies assessing the impact of the subject change using the current LOCA M&E methodology. The sensitivity studies showed an approximate increase in the LOCA calculated containment peak pressure of less than 0.2 psi for this issue for SG tubes manufactured with Alloy 600 and Alloy 690 material and 0.3 psi for SGs manufactured with Alloy 800 material. This increase will not result in exceeding the plants containment design pressure for any Westinghouse-designed PWR with a Westinghouse-supplied LOCA M&E and containment analysis.

However, for plants where the utility has set the ILRT TS pressure (Pa) close to the LOCA calculated peak containment pressure, with little or no margin, the calculated peak containment pressure analysis of record (AOR) value with the 0.2 psi or 0.3 psi penalty applied may result in an adjusted peak pressure higher than the TS controlled ILRT Pa.

1 AP1000 is a trademark or registered trademark in the United States, of Westinghouse Electric Company LLC, its subsidiaries, and/or affiliates. This mark may also be used and/or registered in other countries throughout the world.

All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks or registered trademarks of their respective owners

NSAL-14-2 Page 3 of 8 SAFETY SIGNIFICANCE Conservatism exists in the Westinghouse LOCA M&E methodology (Reference 1) to justify that a SSH could not exist. These conservatisms are presented in Reference 2, Section 5.1, Model Conservatisms, which has been approved by the United States Nuclear Regulatory Commission (U.S. NRC). The NRCs Safety Evaluation Report (SER) is attached at the beginning of Reference 1. Reference 1 has shown that the conservatisms in the LOCA M&E methodology result in an increase in the calculated containment pressure that is on the order of 6 or more psi greater than what would be calculated using less conservative LOCA M&E releases. In addition, if the less conservative techniques of Reference 1 were applied to the Reference 2 methodology, a similar conservatism would be seen. It is noted that approximately 4 psi of the 6 psi margin was used in the Reference 4 notification that some plants are still operating under as a justification for continued operation (JCO). Thus, there is still 2 psi of conservatism that can be used to offset the use of the incorrect SG tube material properties and, if necessary, the impact of Reference 3. Therefore, the containment structural integrity will not be challenged and this issue could not create a SSH if it were left uncorrected.

For plants where Westinghouse holds the containment AOR, it has been determined that if the LOCA M&E analysis issues were left uncorrected, no plant would exceed its containment design pressure acceptance criteria. Furthermore, known conservatisms in the methodology would more than compensate for the SG tube material issue such that the containment peak pressure would not exceed the containment design pressure and in some cases, the ILRT Pa value would not be exceeded. Therefore, the affected plants would remain safe in the event of a postulated large LOCA.

Appendix A contains information that can be used as input to an Operability Determination.

AFFECTED PLANTS This issue applies to all Westinghouse-designed PWRs using a Westinghouse-supplied LOCA M&E calculation. Additionally, this issue applies to four plants where Westinghouse has completed LOCA M&E calculations that have non-Westinghouse-designed SGs with tubes manufactured with Alloy 800 material. Combustion Engineering-designed PWR and BWR designs are not affected by this issue. Plant analyses that have been revised, have not been revised, and are in progress of being revised are listed below.

Plant analyses that have been revised:

Angra 1 North Anna 1 & 2 Beaver Valley 2 Prairie Island 1 & 2 Byron/Braidwood 1 & 2 South Texas 1 & 2 Diablo Canyon 1 & 2 Surry 1 & 2 H.B. Robinson 2 Indian Point 2 & 3 Wolf Creek

NSAL-14-2 Page 4 of 8 Plant analyses that have not been revised:

Plant analysis that are in progress and will be revised:

D.C. Cook 1 & 2 AP1000 plants:

The calculated peak containment pressure resulting from using the corrected long-term LOCA M&E releases would be higher than reported in the Design Control Document (DCD), Revision 19 (Reference 5), but would still remain less than the containment design pressure. The containment evaluation model is currently being updated to address various design changes. A substantial amount of additional metal heat sink mass has been identified, some of which will be included in the revised model. The condensation of steam on these additional heat sinks would result in a lower calculated peak containment pressure. This would provide analysis margin to offset the corrected LOCA M&E releases. Therefore, the containment structural integrity will not be challenged and a SSH does not and could not exist for these plants. This issue does not represent a significant breakdown in the Westinghouse 10 CFR 50 Appendix B quality assurance program and it is also not reportable under 10 CFR 50.55(e).

Plant analyses that are not impacted:

Maanshan 1 & 23 Vandellos II3 NRC AWARENESS The NRC is not aware of the current issue of the use of an incorrect material for the SG tubes in the LOCA M&E release calculations. Westinghouse has determined that this issue is not reportable under 10 CFR Part 21 because the identified deviation could not result in a SSH if it were left uncorrected. Also, for the AP1000 plants, it has been determined that this issue is not reportable under 10 CFR 50.55(e).

RECOMMENDED ACTIONS

1.

Determine the estimated impact by adding the estimated 0.2 psi or 0.3 psi penalty to the current licensing basis AOR calculated containment LOCA blowdown and post-blowdown peak pressure result values. Add any other known containment pressure penalties that may have been determined from plant specific evaluations.

2 Seabrook has an analysis that was performed in 2012 to address the issues identified in References 3 and 4. If this analysis has not become the current licensing basis AOR, then the current AOR is not impacted by this SG tube material issue because the historical AOR was performed prior to the use of the input processor.

3 Maanshan 1 and 2 and Vandellos II have an analysis that was performed prior to the use of the input processor.

Almaraz 1 & 2 Point Beach 1 & 2 A.W. Vogtle 1 & 2 R.E. Ginna Asco 1 & 2 Ringhals 3 Beaver Valley 1 Salem 1 & 2 Callaway 1 Seabrook 12 Comanche Peak 1 & 2 Sequoyah 1 & 2 Hanbit 1 & 2 (Yonggwang 1 and 2)

Shearon Harris 1 J.M. Farley 1 & 2 Turkey Point 3 & 4 Kori 1, 2, 3, & 4 V.C. Summer Krsko Watts Bar 1 & 2 Millstone 3

NSAL-14-2 Page 5 of 8

2.

Determine the impact of the error on the containment peak pressure and containment peak temperature/equipment qualification (EQ). A plant specific analysis can be performed to determine these impacts, in addition to the potential sump temperature and ultimate heat sink input impacts.

3.

Determine the available margin that may exist between the current analysis input assumptions versus actual plant conditions that can be used to offset the estimated penalty.

4.

Revise the affected analyses, as required, based on revised analysis input assumptions to address these issues to offset the penalty.

5.

If it is ultimately determined that sufficient offsetting margin cannot be identified, an alternative approach may be considered such as a more realistic analysis methodology.

6.

Assess the impact on the TS Pa.

NSAL-14-2 Page 6 of 8 Appendix A: Information to be Used as Input to an Operability Determination Long-term Containment Peak Pressure Design Basis Analyses The post-blowdown LOCA M&E releases, when calculated with a more mechanistic LOCA analysis code like WCOBRA-TRAC, are significantly lower than those calculated using the WCAP-10325-P-A methodology. As a result, the calculated post-blowdown containment pressure for the large dry and sub-atmospheric containment designs would be significantly reduced if the more accurate LOCA M&E releases were used.

Additionally, because of the lower energy release rate, the time of ice bed melt-out in the ice condenser containment design would be extended. This would result in a potentially lower sump temperature and peak containment pressure, since the residual heat removal system would be able to remove energy from the containment sump over a longer period of time.

Separate from the non-mechanistic assumptions are inherent input, initial condition, and model conservatisms, which are sufficient to offset the impact of the identified error for all plants. Section 5.1 of WCAP-10325-P-A documents the conservatisms that are inherent in the methodology. These conservatisms include modeling aspects and initial conditions assumptions that result in a peak calculated containment pressure that is a minimum of 6 psi higher than what would be calculated with more realistic input values. These inherent conservatisms also apply to the methodology described in WCAP-8264-P-A, Revision 1. If these more realistic input values were applied to the WCAP-10325-P-A or WCAP-8264-P-A, Revision 1 LOCA M&E release calculation methodology, a similar reduction in the calculated peak containment pressure would be observed. The input, initial condition, and model assumption conservatisms include:

Core power and primary side fluid temperatures apply an uncertainty.

The RCS system volume used in M&E release calculation has an additional 1.4% for uncertainty and 1.6% of the system volume is added for thermal expansion.

SG parameters are skewed to maximize available energy. These assumptions are based on full power, maximum Tavg (with applied uncertainty), 0% SG tube plugging (assuming no fouling), maximum SG level plus uncertainty with an additional 10% increase in secondary mass.

Decay heat is maximized and a two sigma uncertainty has been applied to ensure conservatism.

Core stored energy is maximized by assuming the conditions at the most limiting time in life and maximum core fluid temperatures.

The moisture carryover fraction correlation used in the reflood transient for a double-ended pump suction (DEPS) break was developed for ECCS type applications for a double-ended cold leg (DECL) break. At the beginning of the reflood phase in an M&E energy release transient for a DEPS break, the initial core temperature is greatly reduced compared to a DECL for a peak clad temperature analysis. This core temperature reduction would greatly reduce the carryover fraction for a DEPS case by as much as 50%. In turn, the releases to the containment would be reduced.

Since the conservatisms in the LOCA M&E release calculation methodology offset the estimated penalty due to the effect of the error, it was determined that a SSH does not exist. As discussed above, the WCAP-10325-P-A and WCAP-8264-P-A, Revision 1 methodology contains modeling and initial condition assumption conservatisms that result in a calculated peak pressure that is 6 psi higher than the peak pressure that would be determined from a more realistic analysis. This 6 psi would offset the resulting increase in the containment peak Pa associated with this issue. Therefore, if a more realistic

NSAL-14-2 Page 7 of 8 analysis were performed, Pa would not increase above the current value, and there would be no impact on the 10 CFR 50, Appendix J, Type A, B, and C tests.

Discussion of Potential Impacts on other Design Basis Analyses and Evaluations that use the LOCA M&E Releases as an Input Containment Peak Temperature Analyses The containment peak temperature analyses are performed to confirm that the upper limit of the containment liner design temperature is not exceeded. The containment vapor temperature can be higher than the upper limit of the containment liner design temperature for a short period of time, since the thermal response time of the steel is much slower due to the surface heat transfer rate and conduction.

The containment peak temperature for a large break LOCA is calculated using the containment peak pressure LOCA M&E releases as an input; however, the containment model initial conditions are adjusted to calculate an upper bounding containment temperature response. The containment temperature will increase between 1.0°F and 1.5°F per psi increase in the steam partial pressure. Therefore, the total estimated increase in the containment peak temperature, considering this issue, is estimated to be less than 1°F. Equipment qualification (EQ) programs typically include approximately a 15°F peak temperature margin as discussed in IEEE Std 323-2003, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations. Typically, the overall limiting containment temperature is associated with a main steam line break (MSLB) during the early portion of the profile (i.e.,

approximately the first 10 to 20 minutes) and the LOCA transient is limiting for the remainder of the profile (i.e., approximately 30 to 120 days).

As discussed above, if a more realistic analysis were performed, Pa would not increase above the current value due to this issue, and the containment peak temperature would not be impacted. Temperature is a function of pressure; therefore, if pressure does not increase, temperature will not increase.

Containment Equipment Qualification Analyses Containment EQ analyses are performed to confirm that the equipment pressure and temperature test envelopes are not exceeded during the time period for which the equipment is required to perform its safety function. These analyses use the long-term LOCA containment integrity M&E releases as an input.

The issue identified above affects the calculated energy release rates during the first hour following a LOCA event. Typically, the margin between the calculated pressure and temperature conditions from those contained in the EQ pressure and temperature profiles is the greatest during the first few hours of the LOCA event. Typically, the limiting short-term containment conditions from an EQ standpoint are associated with a MSLB. Exceeding the EQ pressure and temperature envelopes is typically a concern well beyond the first hour of a LOCA M&E release. Therefore, this issue should have very little impact on the long-term EQ.

As discussed above, if a more realistic analysis were performed, Pa would not increase above the current value due to these issues, and the containment peak temperature would not be impacted. Temperature is a function of pressure. Therefore, if pressure does not increase, temperature will not increase. If the peak pressure and temperature are not impacted, then there would be no impact on long-term pressure and temperature.

NSAL-14-2 Page 8 of 8 Containment Sump Temperature Analyses Containment sump temperature analyses are performed to confirm the minimum net positive suction head available (NPSHa) for the emergency core cooling system (ECCS) recirculation and containment spray system (CSS) pumps is always greater than the required value (NPSHr). The NPSHa is the sum of the atmospheric pressure and static fluid heads, minus the sum of the head losses and vapor pressure heads.

In general, the NPSHa decreases as the pump flow rate and sump temperature increase due to the corresponding increases in head loss and vapor pressure, respectively. However, in accordance with U.S.

NRC Regulatory Guide 1.1, the NPSHa is calculated assuming the atmospheric pressure and vapor pressure are equal. This calculation technique results in a slight increase in NPSHa as sump temperature increases due to reduced friction losses.

The containment sump temperature response is calculated using the containment peak pressure LOCA M&E releases as an input; however, the energy released as steam is minimized in these analyses to maximize the calculated sump temperature. The sump temperature is not expected to increase due to the impact of this issue.

If a more realistic analysis was performed, the overall initial internal energy of the RCS would be lower and the active containment heat removal systems would be more efficient. This would result in a lower M&E release to the containment atmosphere and the sump and an increase in the energy removed from the containment. The result would be a lower sump temperature transient.

Ultimate Heat Sink Analyses Analyses are performed to determine the energy load on the ultimate heat sink. The issue identified above can affect the calculated energy release rates during the first hour following a LOCA event. The energy in the containment atmosphere and the containment sump is eventually transferred to the plants ultimate heat sink. Typically, the loads on the ultimate heat sink are the highest shortly after the transfer to sump recirculation. The increase in the peak calculated temperature for the ultimate heat sink is not expected to be significant; however, the duration of the temperature increase may be longer than currently analyzed.

If a more realistic analysis was performed, the energy transferred to the ultimate heat sink would be reduced as compared to the design basis analysis; therefore, the ultimate heat sink analyses would not be impacted.

REFERENCES

1. WCAP-10325-P-A, Westinghouse LOCA Mass and Energy Release Model for Containment Design - March 1979 Version, May 1983 (Proprietary, WCAP-10326-A, Non-Proprietary).

2. WCAP-8264-P-A, Revision 1, Topical Report - Westinghouse Mass and Energy Release Data for Containment Design, August 1975.

3. NSAL-06-6, LOCA Mass and Energy Release Analysis, June 6, 2006.

4. NSAL-11-5, Westinghouse LOCA Mass and Energy Release Calculation Issues, July 25, 2013.

5. APP-GW-GL-700, Revision 19, AP1000 Design Control Document, June 2011.

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