ML082831447

Request for Additional Information Regarding Analysis in Support of Control Rod Insertion Following a Cold Leg Loss-of -Coolant Accident (Tac MD9396)

ML082831447
Person / Time
Site:	Watts Bar
Issue date:	10/14/2008
From:	John Lamb Watts Bar Special Projects Branch
To:	Campbell W Tennessee Valley Authority
Lamb John G./NRR/DORL, 415-3100
References
TAC MD9396
Download: ML082831447 (7)

Text

October 14, 2008 Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION REGARDING ANALYSIS IN SUPPORT OF CONTROL ROD INSERTION FOLLOWING A COLD LEG LOSS-OF-COOLANT ACCIDENT (TAC NO. MD9396)

Dear Mr. Campbell:

By letter dated August 1, 2008 (Agencywide Documents Access and Management System Accession No. ML082180091), Tennessee Valley Authority (TVA) proposed a license amendment to change the Watts Bar Nuclear Plant (WBN), Unit 1, Technical Specifications (TSs). One proposed change would revise TS 5.9.5 for the core operating limits report. Your application contained Westinghouse Topical Report WCAP-16932-P, Control Rod Insertion Following a Cold Leg LOCA [Loss of Coolant Accident] for Watts Bar Unit 1. The proposed TS change would, in part, allow credit for the negative reactivity provided by the insertion of the rod cluster control assemblies following a postulated cold leg LOCA.

The staff has reviewed the information provided by TVA and has determined that additional information is required to complete its evaluation of the proposed license amendment. The specific questions are detailed in the enclosed request for additional information (RAI). Based on discussions with your staff, we understand that you plan to respond to the enclosed RAI within 45 days of the date of this letter.

If you have any questions regarding this issue, please feel free to contact me at (301) 415-3100.

Sincerely,

/RA/

John G. Lamb, Senior Project Manager Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosure:

RAI cc w/enclosure: See next page

Mr. William R. Campbell, Jr.

Chief Nuclear Officer and Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801

SUBJECT:

WATTS BAR NUCLEAR PLANT, UNIT 1 - REQUEST FOR ADDITIONAL INFORMATION REGARDING ANALYSIS IN SUPPORT OF CONTROL ROD INSERTION FOLLOWING A COLD LEG LOSS-OF-COOLANT ACCIDENT (TAC NO. MD9396)

Dear Mr. Campbell:

By letter dated August 1, 2008 (Agencywide Documents Access and Management System Accession No. ML082180091), Tennessee Valley Authority (TVA) proposed a license amendment to change the Watts Bar Nuclear Plant (WBN), Unit 1, Technical Specifications (TSs). One proposed change would revise TS 5.9.5 for the core operating limits report. Your application contained Westinghouse Topical Report WCAP-16932-P, Control Rod Insertion Following a Cold Leg LOCA [Loss of Coolant Accident] for Watts Bar Unit 1. The proposed TS change would in part allow credit for the negative reactivity provided by the insertion of the rod cluster control assemblies following a postulated cold leg LOCA.

The staff has reviewed the information provided by TVA and has determined that additional information is required to complete its evaluation of the proposed license amendment. The specific questions are detailed in the enclosed request for additional information (RAI). Based on discussions with your staff, we understand that you plan to respond to the enclosed RAI within 45 days of the date of this letter.

If you have any questions regarding this issue, please feel free to contact me at (301) 415-3100.

Sincerely,

/RA/

John G. Lamb, Senior Project Manager Watts Bar Special Projects Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-390

Enclosure:

RAI cc w/enclosures: See next page DISTRIBUTION:

PUBLIC RidsNrrPMJLamb M. Hartzman, NRR RidsNrrDorlDpr LP-WB R/F RidsNrrLABClayton RidsOgcRp RidRgn2MailCenter RidsNrrDorlLpwb RidsNrrDeEmcb RidsAcrsAcnwMailCenter RidsNrrPMPMilano ADAMS Accession Number: ML082831447 *via memorandum OFFICE LP-WBB/PM LP-WB/LA EMCB/BC LP-WB/BC NAME JLamb BClayton KManoly* LRaghavan DATE 10/ /08 10/ /08 10/ 06 /08 10/ 14 /08 OFFICIAL RECORD COPY

William R. Campbell, Jr.

Tennessee Valley Authority WATTS BAR NUCLEAR PLANT cc:

Mr. Gordon P. Arent Mr. Michael A. Purcell New Generation Licensing Manager Senior Licensing Manager Tennessee Valley Authority Nuclear Power Group 5A Lookout Place Tennessee Valley Authority 1101 Market Street 4K Lookout Place Chattanooga, TN 37402-2801 1101 Market Street Chattanooga, TN 37402-2801 Mr. Ashok S. Bhatnagar Senior Vice President Ms. Beth A. Wetzel, Manager Nuclear Generation Development Corporate Nuclear Licensing and and Construction Industry Affairs Tennessee Valley Authority Tennessee Valley Authority 6A Lookout Place 4K Lookout Place 1101 Market Street 1101 Market Street Chattanooga, TN 37402-2801 Chattanooga, TN 37402-2801 Vice President Mr. Masoud Bajestani, Vice President Nuclear Support Watts Bar Unit 2 Tennessee Valley Authority Watts Bar Nuclear Plant 3R Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Spring City, TN 37381 Mr. Michael J. Lorek Mr. Michael K. Brandon, Manager Vice President Licensing and Industry Affairs Nuclear Engineering & Technical Services Watts Bar Nuclear Plant Tennessee Valley Authority Tennessee Valley Authority 3R Lookout Place P.O. Box 2000 1101 Market Street Spring City, TN 37381 Chattanooga, TN 37402-2801 Mr. Gregory A. Boerschig, Plant Manager General Counsel Watts Bar Nuclear Plant Tennessee Valley Authority Tennessee Valley Authority 6A West Tower P.O. Box 2000 400 West Summit Hill Drive Spring City, TN 37381 Knoxville, TN 37902 Senior Resident Inspector Mr. John C. Fornicola, Manager Watts Bar Nuclear Plant Nuclear Assurance U.S. Nuclear Regulatory Commission Tennessee Valley Authority 1260 Nuclear Plant Road 3R Lookout Place Spring City, TN 37381 1101 Market Street Chattanooga, TN 37402-2801 County Executive 375 Church Street Mr. Larry E. Nicholson, General Manager Suite 215 Performance Improvement Dayton, TN 37321 Tennessee Valley Authority 3R Lookout Place County Mayor 1101 Market Street P. O. Box 156 Chattanooga, TN 37402-2801 Decatur, TN 37322 Mr. Michael D. Skaggs Mr. Lawrence E. Nanney, Director Site Vice President Division of Radiological Health Watts Bar Nuclear Plant Dept. of Environment & Conservation Tennessee Valley Authority Third Floor, L and C Annex P. O. Box 2000 401 Church Street Spring City, TN 37381 Nashville, TN 37243-1532

Request for Additional Information Regarding Referencing Westinghouse Topical Report WCAP-16932 Watts Bar Nuclear Plant, Unit 1 Tennessee Valley Authority Docket No. 50-390 By letter dated August 1, 2008 (Agencywide Documents Access and Management System Accession No. ML082180093), Tennessee Valley Authority (TVA) proposed a change to the Watts Bar Nuclear Plant Unit 1 (WBN-1) Technical Specifications (TSs). The proposed change would revise TS 5.9.5 for the core operating limit report. The application contained Westinghouse Topical Report WCAP-16932, Control Rod Insertion Following a Cold Leg LOCA

[Loss-of-Coolant Accident] for Watts Bar Unit 1.

After reviewing the information provided by TVA, the U.S. Nuclear Regulatory Commission (NRC) staff requests the following information regarding WCAP-16932:

Section 2.2 Seismic Response Spectra

1. WCAP-16932 includes the following:

a. Section 2.2 states that plant specific safe shutdown earthquake (SSE) response spectra were used to generate synthesized time history acceleration input for the seismic analysis of the reactor vessel/internals/fuel model.

b. Section 2.2 states that the enveloped horizontal and vertical response spectra used in the Watts Bar Unit 1 analyses are shown in Figures 1 and 2. Figures 1 and 2 are titled 4-loop 17x17 plants implying that the envelope spectra shown as solid lines are bounding plant group spectra.

c. Section 4.3 and 4.4 states that the bounding plant group seismic spectra were used in the analysis.

Item (1a) appears to be in conflict with Items (1b) and (1c). Please confirm whether the synthetic time histories were generated from the site specific Watts Bar Unit 1 SSE or from the envelope spectra shown in Figures 1 and 2.

2. Section 3.7 of Watts Bar Unit 1 Updated Final Safety Analysis Report (UFSAR) discusses three sets of seismic criteria: (1) Set A for original seismic analysis, (2) Set B for evaluation seismic analysis using site specific response spectra, and (3) Set C for future design and modification of existing design. Please confirm that the seismic criteria used in the analysis is in compliance with the Watts Bar Unit 1 current licensing basis.

3. Figures 1 and 2 show 3-percent damping Watts Bar spectra, shown as dotted and dash lines, plus 2-percent damping envelope spectra shown as solid line. Considering that the damping for Watts Bar spectra is higher than the envelope spectra, please provide justification relative to difference in damping and validity of Figures 1 and 2 spectra comparison, specifically in Figure 2 where the Watts Bar spectra nearly exceed the envelope spectra. Furthermore, Enclosure

please confirm the compatibility of the seismic criteria used in the analysis and the Watts Bar spectra (dotted and dash lines) shown in Figures 1 and 2.

4. Watts Bar Unit 1 UFSAR, Section 3.7 states that the Standard Review Plan (SRP) 1981, Revision 1 formed the basis for Set B and Set C seismic analyses, updated to the provisions of SRP 1989, Revision 2. WCAP-16932 references NUREG 75/087, Section 3.7.1 dated 1975 for acceptability of response spectra corresponding to the synthetic time history accelerations. Considering Item 2 above, please provide justification that the SRP version used in WCAP-16932 is in compliance with the current licensing basis of Watts Bar Unit 1.

5. Watts Bar Unit 1 UFSAR includes the following:

a. Section 3.7.1.2.1 states that for Set A and Set C time history analyses, four synthesized acceleration time histories were developed so that the response spectra produced by the arithmetic average of the response spectra of each individual record envelope the site specific seismic design response spectra.

b. Section 3.7.1.2.2 states that Set B analyses utilize three statistically independent acceleration time histories. The power spectral density function enveloping criteria of NUREG/CR-5347 were used to ensure adequate energy content of the synthetic time histories.

Please provide justification that the number of synthesized time history accelerations and criteria for acceptability of synthesized time history accelerations used in the seismic analysis included in WCAP-16932 are in compliance with the current licensing basis of Watts Bar Unit 1.

6. Considering Item No. 5 above, please provide the following:

a. A detailed description of the method(s) and criteria (e.g., time increment, total duration, amplitude, basis for selecting Taft earthquake, etc.) used to generate the synthetic time history accelerations. Provide the curves of the horizontal and vertical synthetic seismic time histories used in the analysis.

b. A detailed description of method(s) used to demonstrate acceptability of the response spectra corresponding to the synthesized time history accelerations.

7. As stated in Section 3.7.3.14, Seismic Analysis for Fuel Elements, Control Rod assemblies, Control Rod Drives, and Reactor Internals of the Watts Bar Unit 1 UFSAR, the horizontal and vertical seismic analysis was based on the modal response spectrum method. For the combination of LOCA and SSE, the most unfavorable sign convention for the SSE was assumed. WCAP-16932 is based on a time history analysis using a synthesized time history acceleration generated based on response spectra shown in Figures 1 and 2. This method of analysis differs from what is presently described in Watts Bar Unit 1 UFSAR. Also, Enclosure 4, proposed UFSAR change to TVA letter dated August 1, 2008, does not include the affected UFSAR pages relative to this issue. Please provide justification.

Section 4.1 Mathematical System Model of Vessel/Internals/Fuel

1. Provide diagrams of the actual vessel/internals/fuel to clarify the correspondence with the models shown in Figures 4 through 8, or provide references where the diagrams may be found.

2. Provide the locations on the models where the LOCA hydraulic loads are applied.

Section 4.2 Analysis Methodology

1. Provide a description of the non-linear superposition method used in the WECAN/PLUS computer code, and any NRC conditions associated with the staff review of this method.

2. State if all conditions resulting from the NRC review and approval of the MULTIFLEX computer code have been satisfied.

Section 4.3 Core Plate Motions

1. State the scope and a description of the Nuclear Fuel Business Unit evaluation.

2. State whether Reference 4.2 (WCAP-12828, Reactor Pressure Vessel and Internals System Evaluation for D. C. Cook Unit 2 Vantage 5 Fuel Upgrade with IMFs December 1990) was reviewed and approved by the NRC.

3. Since the structural evaluation is based on non-linear modal superposition analysis, provide justification for calculating the effects of the LOCA and seismic core plate motions separately, and then combining these by using the SRSS method.

Section 5.1 Rod Cluster Control Assembly (RCCA) Upper Internals Guide Tube Loads

1. Provide a detailed description of the proprietary scale model and plant test measurements of guide tube strains that form the basis for estimating these loads. Provide the magnitude of these strains.

Section 5.3 Allowable Loads for RCCA Upper Internals Guide Tubes

1. Guide tube scram tests were reported in WCAP-15704 and WCAP-15245. State whether these reports were reviewed and approved by the NRC. Conversely, provide a detailed description of these tests.

2. Provide the deflections corresponding to the allowable loads listed in Table 2.

3. Provide the limiting guide tube deflection that will prevent control rod insertion.

Section 6.2 Allowable Loads for Fuel Grids

1. Section 6.2 references NUREG-800, Appendix A, Section C.2. The proper reference should be NUREG-0800, Section 4.2, Appendix A, Section C.2.

2. Provide a description of the tests performed on the grids.

Section 6.3 Analysis Results

1. Table 4 on page 27 shows the Mid Grid and the intermediate flow mixer (IFM) Grid impact loads resulting from LOCA and SSE bounding analyses. These loads are combined by SRSS to obtain the combined loads from both analyses. In the column labeled Mid Grid,

the LOCA load is listed as 1358 pounds force (lbf) and the SSE load is listed as 2,004 lbf.

The square root sum of the squares (SRSS) of the two loads is 2421 lbf, as compared with the listed value of 2004 lbf. In addition, the design margin for this combination of loads is 45-percent, as compared with the listed value of 54.6-percent. Likewise, in the column labeled IFM Grid, the LOCA load value is listed as 371 lbf and the SSE load value is listed as 236 lbf. The SRSS of these two loads is 440 lbf, as compared to the listed value 372 lbf.

In addition, the design margin is 73.5-percent, as compared to the listed value of 77.5-percent. Therefore, the listed values for SRSS and design margin in both columns are in error and should be corrected.

2. Provide the maximum membrane plus bending stresses corresponding to the allowable loads for ZIRLO mid-grid and IFM grid, and show that these are below the yield stress.

Section 7 Control Rod Driveline Integrity

1. Provide a diagram showing the various components referred to in this section, or provide a reference where it can be found.

2. Expand the discussion in the last paragraph on page 30, and provide a description of the method for showing that the driveline misalignment at the limiting guide tube location is less than the allowable value for a period in excess of 5 minutes.

3. Show the basis for the conclusion that sufficient drive rod clearances are available for post-LOCA control rod insertion based upon the worst assumptions for upper package displacement during the combined LOCA and seismic event and for the maximum cool down rate of the upper support package during the post-LOCA period.

4. Show that the largest calculated relative deflections of the guide tubes and the control rods are within the limiting guide tube deflections.

Section 8.2 Analysis Results

1. Provide clarification that the application of the non-linear modal superposition method to the analysis of the control rod drive mechanism (CRDM) support platform assembly did not include plastic deformation.

2. Table 5 shows the allowable stress intensity of 73.5 kilopounds per square inch for platform welding. Provide the basis for this allowable.

3. The last paragraph of this section refers to another plant without identifying this plant.

a. Identify the other plant.

b. Provide the seismic spectral comparisons with the other plant that support the acceptability of the CRDM head adapter and CRDM pressure housing loads.

c. Show that the CRDM head adapter and CRDM pressure housing loads met the allowable loads.