BVY 05-086, Technical Specifications Proposed Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information

Technical Specifications Proposed Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information

ML052660025
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	09/18/2005
From:	Rademacher N Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BVY 05-086, TAC MC0761
Download: ML052660025 (79)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

Vermont Yankee Ad P.O. Box 0500 185 Old Ferry Road E n vBrattleboro, VT 05302-0500 tf oy Tel 802 257 5271 September 18, 2005 Docket No. 50-271 BVY 05-086 TAC No. MC0761 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Vermont Yankee Nuclear Power Station Technical Specification Proposed Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information

References:

1) Entergy letter to U.S. Nuclear Regulatory Commission, 'Vermont Yankee Nuclear Power Station, License No. DPR-28 (Docket No. 50-271), Technical Specification Proposed Change No. 263, Extended Power Uprate," BVY 03-80, September 10, 2003

2) U.S. Nuclear Regulatory Commission (Richard B. Ennis) letter to Entergy Nuclear Operations, Inc. (Michael Kansler), 'Request for Additional Information -

Extended Power Uprate, Vermont Yankee Nuclear Power Station (TAC No. MC0761)," September 7, 2005

3) Entergy letter to U.S. Nuclear Regulatory Commission, 'Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263 - Supplement No. 32, Extended Power Uprate -

Additional Information," BVY 05-083, September 10, 2005

4) Entergy letter to U.S. Nuclear Regulatory Commission, 'Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263 - Supplement No. 33, Extended Power Uprate -

Response to Request for Additional Information," BVY 05-084, September 14, 2005 This letter provides additional information regarding the application by Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (Entergy) for a license amendment (Reference 1) to increase the maximum authorized power level of the Vermont Yankee Nuclear Power Station (VYNPS) from 1593 megawatts thermal (MWt) to 1912 MWt.

The attachments to this letter provide supplemental information in response to requests for additional information from the NRC staff (Reference 2) and other supplemental information to update the application for a license amendment. As a result of recent discussions with the NRC staff and its recent audit of analytical methodologies of General Electric (GE) that are used for AP I

BVY 05-086 Docket No. 50-271 Page 2 of 4 the design and evaluation of VYNPS' fuel, the NRC staff identified the need for additional information reflected in several of the requests for additional information (RAls) contained in Reference 2. Because of the recency of the requests, one (Reference 2) RAI remains to be addressed (i.e., NRC RAI SRXB-A-68); the remaining RAI will be addressed in a submittal that will be made by September 23, 2005. to this letter is a revision to Exhibit EMEB-B-1 8-1, Rev. 1, Attachment 4 (regarding the steam dryer acoustic load uncertainty evaluation) that was provided to the NRC staff in Reference 4. Inadvertently, several figures were not included in the original submittal. The omitted figures include comparisons of power spectral densities for certain transmitter locations. consists of thirty figures (EMEB-B-18-1-4-1 through EMEB-B-18-1-4-30) and supersedes, in its entirety, Exhibit EMEB-B-18-1, Rev. 1, Attachment 4 provided in Reference 4, (Proprietary Information) and Attachment 8 (Non-Proprietary Version). to this letter does not contain proprietary information.

In the response to RAI SRXB-A-66 (Reference 3), Entergy stated that certain tabulated data supporting the response to the RAI would be submitted to the NRC staff as Microsoft Excel spreadsheets. That information is included herein as Attachment 2 on a compact disk. The data contained on the compact disk is considered Proprietary Information to General Electric and is covered by the affidavit accompanying the response to SRXB-A-66 in Reference 3. An explanatory 'Read Me' file (non-proprietary) contained on the CD is included in hardcopy as part of Attachment 2.

As a result of discussions with the NRC staff, Entergy is providing in Attachment 3 a more extensive response to RAI SRXB-A-64. This response supplements the response that was originally provided in Reference 3. contains responses to NRC Reactor Systems Branch RAls SRXB-A-65 and SRXB-A-67 that were posed in Reference 2. These RAls and the responses thereto contain Proprietary Information as defined by 10CFR2.390 and should be handled in accordance with the provisions of that regulation. Attachment 4 is considered to be Proprietary Information in its entirety. Attachment 5 is a non-proprietary version of Attachment 4. An affidavit provided by General Electric Company, supporting the proprietary nature of the document, is provided as. provides a response to RAI SRXB-A-71 that was asked in Reference 2. of this letter provides a copy of the demonstrated shutdown margin (SDM) calculation for the current operating cycle (i.e., cycle 24). This SDM calculation is referenced in the response to RAI SRXB-A-67, part (b).

There are no new regulatory commitments contained in this submittal.

This supplement to the license amendment request provides additional information to clarify Entergy's application for a license amendment and does not change the scope or conclusions in the original application, nor does it change Entergy's determination of no significant hazards consideration.

BVY 05-086 Docket No. 50-271 Page 3 of 4 The following attachments are included in this submittal:

Attachment Title 1

Revised Exhibit EMEB-B-18-1, Rev. 1, Attachment 4 2

RAI SRXB-A-66 Data (Compact Disk)

(PROPRIETARY INFORMATION) 3 Supplemental Response to SRXB-A-64 4

Responses to RAls SRXB-A-65 and SRXB-A-67 (Proprietary Information) 5 Responses to RAls SRXB-A-65 and SRXB-A-67 (Non-Proprietary Version) 6 Response to RAI SRXB-A-71 7

General Electric Affidavit 8

Demonstrated Shutdown Margin Entergy stands ready to support the NRC staff's review of this submittal and suggests meetings at your earliest convenience to resolve any remaining issues. If you have any questions or require additional information, please contact Mr. James DeVincentis at (802) 258-4236.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on September 18, 2005.

Sincerely, Norman L. Rademacher Director, Nuclear Safety Assurance Vermont Yankee Nuclear Power Station Attachments (8) cc:

(see next page)

BVY 05-086 Docket No. 50-271 Page 4 of 4 cc:

Mr. Richard B. Ennis, Project Manager Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0 8 B1 Washington, DC 20555 Mr. Samuel J. Collins (w/o attachments)

Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 USNRC Resident Inspector (w/o attachments)

Entergy Nuclear Vermont Yankee, LLC P.O. Box 157 Vernon, Vermont 05354 Mr. David O'Brien, Commissioner (w/o proprietary information)

VT Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601

BVY 05-086 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information Revised Exhibit EMEB-B-18-1, Rev. 1, Attachment 4 Total number of pages in Attachment 1 (excludina this cover sheet) is 30.

to BVY 05-086 Docket No. 50-271 Page 1 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 M

g a-0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P1 QC2 P1-79OMWe FiueEM B1-1---

to BVY 05-086 Docket No. 50-271 Page 2 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 I

I I

I I

20 40 60 80 100 120 140 160 180 20 0.01 Il 0.001 0.0001 A

0.00001 0.000001 0.0000001 Frequency Hz I

Predict P2 QC2 P2-79OMWe Figure EMEB-B-18-1-4-2 to BVY 05-086 Docket No. 50-271 Page 3 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 NX E

tea.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P3 0C2 P3-79OMWe Figure EMEB-B-18-1-4-3 to BVY 05-086 Docket No. 50-271 Page 4 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

to 0.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P4 QC2 P4-79OMWe Figure EMEB-B-18-1-4-4 to BVY 05-086 Docket No. 50-271 Page 5 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

U, 0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P5 ---

QC2 P5-79OMWel Figure EMEB-B-18-1-4-5 to BVY 05-086 Docket No. 50-271 Page 6 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

07E I-co 0.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P6 -

-QC2 P6-79OMWel Figure EMEB-B-18-1-4-6 to BVY 05-086 Docket No. 50-271 Page 7 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

(I) 0.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P7 QC2 P7-790MWe Figure EMEB-B-18-1-4-7 to BVY 05-086 Docket No. 50-271 Page 8 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

i;I coE en 0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

FPredict P8

-M-B-QC2 P8-79MWe Fi-gure EMEB-B-18-1-4-8 to BVY 05-086 Docket No. 50-271 Page 9 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

L 0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P9 QC2 P9-79OMWe Figure EMEB-B-18-1-4-9 to BVY 05-086 Docket No. 50-271 Page 10 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 M

E a-0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P10

- - -QC2 P1 0-790MWe Fi-qure EMEB-B-18-1-4-10 to BVY 05-086 Docket No. 50-271 Page 11 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

0.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P11 QC2 P11 -790MWe Figure EMEB-B-18-1-4-11 to BVY 05-086 Docket No. 50-271 Page 12 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 a-V4 0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P12

- -QC2 P12-79OMWe Figure EMEB-B-18-1-4-12 to BVY 05-086 Docket No. 50-271 Page 13 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

Z4 0.E U) 0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P13 QC2 P13-79OMWe Figure EMEB-B-18-1-4-13 to BVY 05-086 Docket No. 50-271 Page 14 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

E en a.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P14

- -QC2 P14-79OMWe Figure EMEB-B-18-1-4-14 to BVY 05-086 Docket No. 50-271 Page 15 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

0 0.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P15 QC2 P15-790MWe Figure EMEB-B-18-1-4-15 to BVY 05-086 Docket No. 50-271 Page 16 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 Nzi~I 0.E W

IL 0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P16 QC2 P16-79OMWe Figure EMEB-B-1 8-1-4-16 to BVY 05-086 Docket No. 50-271 Page 17 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

X 0.

a.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P17 QC2 P17-79OMWe I

Figure EMEB-B-18-1-4-17 to BVY 05-086 Docket No. 50-271 Page 18 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

0.

E U) 0.0001 0.00001 0.000001 0.0000001 Frequency Hz l

Predict P18 QC2 P1 8-79OMWe I

Figure EMEB-B-18-1-4-18 to BVY 05-086 Docket No. 50-271 Page 19 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

X a.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P19 QC2 P19-79OMWe Figure EMEB-B-18-1-4-19 to BVY 05-086 Docket No. 50-271 Page 20 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

8 iq 0.E

=1 V) 0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P20 0C2 P20-79OMWe I

Figure EMEB-B-18-1-4-20 to BVY 05-086 Docket No. 50-271 Page 21 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

X 0

Eci CL 0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P21 QC2 P21-790MWe Figure EMEB-B-18-1-4-21 to BVY 05-086 Docket No. 50-271 Page 22 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

0.E rnaM 0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P22 QC2 P22-79OMWe I

Figure EMEB-B-1 8-1-4-22 to BVY 05-086 Docket No. 50-271 Page 23 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

(I)a-0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P23

- -QC2 P23-79OMWe I

Figure EMEB-B-18-1-4-23 to BVY 05-086 Docket No. 50-271 Page 24 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

Z4 a-E 0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P24 QC2 P24-79OMWe I

Figure EMEB-B-18-1-4-24 to BVY 05-086 Docket No. 50-271 Page 25 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

E 0.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P25 QC2 P25-790MWe I

Figure EMEB-B-18-1-4-25 to BVY 05-086 Docket No. 50-271 Page 26 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

2 a-CL 0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P26

- -QC2 P26-79OMWe Figure EMEB-B-18-1-4-26 to BVY 05-086 Docket No. 50-271 Page 27 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

0.E CO) 0.0001 0.00001 0.000001 0.0000001 Frequency Hz Predict P27 QC2 P27-79OMWe Figure EMEB-B-18-1-4-27 to BVY 05-086 Docket No. 50-271 Page 28 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

~0 E

a-0.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P22 - Predict P23 Figure EMEB-13-18-1-4-28 P22 - P23 to BVY 05-086 Docket No. 50-271 Page 29 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 M

E a.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P20 - Predict P14 P20 - P14 Figure EMEB-B-18-1-4-29 to BW 05-086 Docket No. 50-271 Page 30 of 30 Exhibit EMEB-B-18-1, Rev. 1, VYNPS Steam Dryer Load Uncertainty (Rev 1)

PSD Comparison, QC2 Data vs. ACA Predictions plus Uncertainty 0.1 0.01 0.001 N

(n 0.

0.0001 0.00001 0.000001 0.0000001 Frequency Hz I

Predict P3 - Predict P13 P3 - P13 I

Figure EMEB-B-18-1-4-30

BVY 05-086 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information RAI SRXB-A-66 Data Total number of pages in Attachment 2 (excluding this cover sheet) is 1. includes a compact disk (The CD contains Proprietary Information).

to BVY 05-086 Docket No. 50-271 Page 1 of 1 Vermont Yankee Nuclear Power Station Data Supporting Response to RAI SRXB-A-66 Entergy's letter of September 10, 2005, responded to NRC RAI SRXB-A-66. Microsoft Excel spreadsheet files designated C4-TGBLA6_diff.Rev1.xIs and Lat_7009_T6_C4_FD_Data.xIs contain data supporting that response.

The file designated C4-TGBLA6_diff.Rev1.xIs contains the basis of the RAI SRXB-A-66 plots of CASMO-4 and TGBLA06 data comparisons (K-inf, local peaking, plutonium isotopes, void coefficient, etc.) for a number of lattice designs, void history depletions, and instantaneous void cases. Void fraction definitions used in both methods are consistent and are based upon the heated channel area only with bypass and water rods at zero void. As noted in the response to RAI SRXB-A-66, these comparisons were within expectations for comparisons of different calculational methods. Larger differences are noted for 90% void history depletion cases, but consideration that only a small portion of any BWR core obtains these values for a small exposure window near beginning of life minimizes the impact. These cases are not currently used in any production or licensing basis applications. For the lower void fraction cases, the differences seen in K-inf and local peaking would be expected to be insignificant when incorporated into the homogenized nodal models in which the data are used. This has been demonstrated via comparisons of PANAC1 1 and SIMULATE-3 core-follow results.

The file designated Lat_7009_T6_C4_FDData.xls contains detailed pin-by-pin power data from CASMO-4 and TGBLA06 for a single lattice design for various void history depletions. These data were requested at the September 7, 2005, NRC audit to aid in the staff's evaluation of GNF's methods via an independent method.

This information was taken from existing calculation files generated to support the response to RAI SRSB-A-66. Due to differences in the units of depletion (MWD/MT vs. MWD/ST), only a limited number of depletion points are close enough to the same exposure to provide meaningful comparisons at a pin level. Pin power differences (RMS for the entire lattice) are shown graphically and further detailed study is possible from the data available. The review performed of this information indicates that the difference in gadolinium treatment drives the difference in pin powers calculated by the two methods over the whole lattice. Peak pin agreement, as shown in the response to RAI SRXB-A-66, is generally good.

BVY 05-086 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information Supplemental Response to SRXB-A-64 l Total number of pages in Attachment 3

( (excludina this cover sheet) is 2.

l to BVY 05-086 Docket No. 50-271 Page 1 of 2 RAI SRXB-A-64 Provide the values for maximum bundle power and average power densities at VYNPS before and after the EPU.

Supplemental Response to RAI SRXB-A-64 Core thermal power information for VYNPS is provided in Table SRXB-A-64-1.

The table provides the average power densities before and after EPU. The table also provides channel (bundle) power information requested by the RAI.

Table SRXB-A-64-1 Vermont Yankee Nuclear Power Station Power Information Parameter Pre-EPU Post-EPU

% Change Total Core Thermal Power (MWt) 1593 1912 20 Power Density (kW/liter) 48.9 58.7 20 Channel Average Power (MWt) 4.33 5.20 20 Maximum Channel Power (MWt)

-7

-7 N/A The channel average power is the total core thermal power divided by the number of fuel channels (368).

The maximum channel powers shown in Table SRXB-A-64-1 are essentially unchanged by EPU operation. The values are presented as approximately 7 MWt in order to emphasize this point. The reason for this is that high power channels are limited by thermal limits. In other words, the peak LHGR and/or OLMCPR limits effectively put a ceiling on the maximum allowable bundle power. These limits are associated with the fuel and core designs, and are not a direct function of EPU. The actual pre-and post-EPU maximum bundle powers are 7.02 and 7.37 MWt, respectively. Again, the maximum values will likely change in the future depending on the particular reload core and bundle design. The maximum bundle power could also (potentially) be impacted by other design constraints, for example, the margin to the OLMCPR limit (i.e., how the peak bundles are projected to operate relative to the limit).

The NRC safety evaluation (SE) for constant pressure EPU documented in NEDC-33004P-A summarizes key elements related to the power uprate, including a discussion of power density.

Section 1.3.3 of the SE contains the statements: 'The CPPU approach achieves the power uprate by increasing the core average power density proportional to the core thermal power increase.

This affects the reload core design and operating flexibility, the reactivity characteristics and the cycle energy requirements. No changes in fuel mechanical designs or fuel design limits are required to implement the CPPU process." From a core designer's point of view, the power uprate is effectively achieved by flattening the core radial power shape. More channels operate at or above the pre-uprate average bundle power level.

to BVY 05-086 Docket No. 50-271 Page 2 of 2 The next VYNPS operating cycle (i.e., cycle 25) core was designed to support operation under constant pressure power uprate (CPPU) conditions. The additional reactivity necessary to achieve the target power and cycle length is provided through the reload core design (i.e., the selection of bundle enrichments and the reload batch fraction).

BVY 05-086 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information Responses to RAls SRXB-A-65 and SRXB-A-67 NON-PROPRIETARY VERSION l Total number of pages in Attachment 5 (excludina this cover sheet) is 28.

l to BVY 05-086 Docket No. 50-271 Page 1 of 28 NON-PROPRIETARY VERSION RAI SRXB-A-65 Linear Heat Generation Rate (LHGR)

The NRC staff had previously asked whether any uncertainties were applied to the LHGR limit (curve) and the actual operating nodal steady state kilowatt/foot (kw/ft). The response to RAI SRXB-A-41 took credit for a reduced value in the gradient uncertainty.

However, the power allocation and the pin power uncertainty values were increased to accommodate the lack of gamma scans of the current GE14 fuel designs as operated.

The RAI response states that a local uncertainty of ((

)) in LHGR is assumed in the development of the LHGR, implying that the ((

)) kw/ft uncertainty addressed in the response to the staff RAI 5, associated with the NRC-approved safety limit minimum critical power ratio (SLMCPR) topical report NEDC-32694P-A, was intended for the generation of the LHGR limit.

However, it is the staff's understanding that the uncertainty analyses provided in the RAI 5 response was addressing the uncertainty to be applied to the kw/ft calculated by the core monitoring system (e.g., 3D MONICORE) as opposed to a ((

)) uncertainty assumed during the development of the LHGR curve.

The RAI 5 to NEDC-32694P-A stated that the process computer monitors peak kw/ft and maximum average planar linear heat generation rate (MAPLHGR). The peak kw/ft and the MAPLHGR depend on the bundle axial power distribution and, consequently, are significantly more sensitive to the 3-D MONICORE replacement of the traversing incore probe (TIP)/local power range montior (LPRM) axial power distribution. The RAI asked for uncertainty analysis for the 3-D MONICORE prediction of peak kw/ft and MAPLHGR.

In the response, GE provided the following uncertainty analyses, which specified the uncertainty that would be applied to the peak kw/ft calculations:

Nodal Power Uncertainty:

The nodal power uncertainty for 3D MONICORE is a combination of: 1) the uncertainty in the four bundle power at axial node k; 2) the uncertainty in the power allocation factor at node k; 3) the LPRM update uncertainty; and

4) the uncertainty in the TIP axial power distribution at node k. ((

)) The total nodal power uncertainty is, therefore, equal to:

1))

Pin Power Peaking Uncertainty: The pin power peaking uncertainty can be determined from the factors outlined for the R-factor uncertainty summarized in Section 3 of NEDC-32601. Specifically, the pin power peaking uncertainty is a combination of 1) the model uncertainty, 2) the manufacturing uncertainty, and 3) the channel bow uncertainty.

As in Section 3 of NEDC-32601 P, the model uncertainty is a combination of the pin to BVY 05-086 Docket No. 50-271 Page 2 of 28 NON-PROPRIETARY VERSION peaking uncertainty determined from Monte Carlo comparisons (1.44%) and an uncertainty due to flux gradients from neighboring bundles. ((

)) All of these pin power uncertainties have been combined in NEDC-32601 P as:

((:

1))

The total LHGR uncertainty is the combination of nodal and pin power uncertainties:

((

1]

Staff Position As shown in the NRC-approved SLMCPR methodology specified in NEDC 2694P-A, aLHGR changes with apa, and amc.

Accepting the reduction in the gradient uncertainty, a GLHGR of ((

D] should be applied to the calculated kwlft as discussed and specified in the NRC-approved licensing topical report. Because a ((

)) uncertainty is assumed in the generation of the LHGR limit, this does not mean that the uncertainties due to the impact of modeling uncertainties on the operating kw/ft can be traded off with the ((

))

uncertainty assumed in the development of the limit. The limit is developed based on the accuracy of the thermal-mechanical analytical models, methods and code systems.

Therefore, any uncertainty currently applied in the development of the LHGR limit, can only be taken credit for or changed if it is demonstrated that for the current fuel designs and operating conditions additional nonconservatisims would not offset the 'no cause"

((

)) uncertainty.

The increase in the power allocation and pin power uncertainty applied to the SLMCPR does not directly lead to a proactive increase in the predicted steady state kw/ft.

Therefore, potential underestimation in the nodal powers (bundle and peak pin) need to be accounted for. As evident in the RAI responses, the core-wide axial and nodal uncertainties determined through the TIP comparisons are not applied to the transient or accident analyses. The core-wide radial (e.g., bundle uncertainty aP4B) uncertainty is limited to the SLMCPR calculations. Therefore, there are no nodal or pin uncertainties that are applied to the predicted kw/ft.

It is the staff's position that a ((

)) kw/ft uncertainty be applied to the operating kw/ft calculated in the core simulator code, because of the following reasons:

1. Since there are no measurement data to validate the bundle and pin axial power, the uncertainties in the cross-sections and the pin powers are based on the TIP four bundle readings and the MCNP/TGBLA code-to-code comparisons. The four radial bundle uncertainty 0 P4B nodal is derived from TIP comparisons and is applied to the SLMCPR.

The power allocation between the four bundles OPAL nodal derived from measurement data is also applied to the SLMCPR. The predicted operating kw/ft to BVY 05-086 Docket No. 50-271 Page 3 of 28 NON-PROPRIETARY VERSION relies on the predicted axial bundle power and the pin powers. Although the 3D MONICORE adjusts the four bundle axial power peaking to the TIP reading, the adjusted axial power peaking is based on at least four bundle TIP response.

Therefore, the power allocation in each bundle must be incorporated in the predicted kwtft.

Similarly, the uncertainty in the pin power needs to be included in the calculation of the peak kw/ft. Therefore, the calculated ((

)) uncertainty needs to be applied to the predicted kw/ft, to account for the uncertainties in the cross-sections and the pin powers.

2. The ((

)) power uncertainty bias, applied in the fuel rod internal pressure cited in the Alternative Approach (Supplement 30, Attachment 1), accounts for the differences between the design conditions the rod internal pressure calculations are based on and the rod internal pressures that would be obtained if actual operating history conditions were simulated. In other words, the ((

)) uncertainty accounts for the difference between the as-designed and as-operated conditions.

3. The Alternative Approach cites an additional power uncertainty of ((

)) power that is not specifically assigned to any cause. The Alternative Approach also states that separate experimental benchmarking information confirms that the model uncertainties remain valid. However, it is the NRC staff's understanding that, for the current fuel designs (GE14) as operated, no benchmarking of the fission gas inventory was performed.

It is also the understanding that the ((

)) Uno cause' uncertainty is based on the original NRC-approval of the thermal-mechanical methodology and models. Therefore, it is not evident if a conservatism of ((

))

would actually be available, if the operating and core design changes implemented since the initial development of the fuel thermal-mechanical models are evaluated.

Neither the RAI response nor the Alternative Approach demonstrated this. The RAI response also did not discuss what uncertainties are assumed in the transient overpower kw/ft and if there is sufficient margin available.

4. The application of ((

)) margin to the calculated kw/ft values would ensure that there are sufficient margins to the pellet exposure limits.

The ((

)) additional margin in the peak kw/ft would require a decrease in the nodal (bundle-wise) operating kw/ft, which would provide additional margin in bundle averaged accumulated exposure.

Response to RAI SRXB-A-65 The 3D Monicore surveillance system discussed in the RAI is intended to be ((

))

The following points are provided in response to items 1 - 4 under the staff position heading in the RAI.

1. As stated above, the GE objective is for the core monitoring methods to provide the most accurate ((

)) quantification of the actual operating state.

The to BVY 05-086 Docket No. 50-271 Page 4 of 28 NON-PROPRIETARY VERSION uncertainty in that operating state calculation is addressed ff

)), even when uprated conditions are considered, as discussed further below.

2. The ((

)) bias applied to fuel rod internal pressure calculations is an allowance ((

))

Variations between the analyzed power history and actual power histories are addressed through the analysis assumption ((

)) Figure SRXB-A-65-1 presents the ((

)) (LHGR Operating Limit), as compared to an actual projected operating history for Bundle JLC505 Rod K4 Node 5 both under power uprate conditions and without power uprate. JLC505 experiences the highest bundle nodal exposure (Node 5) for any bundle in the VYNPS Cycle 25 core both with and without power uprate conditions. Rod K4 of JLC505 experiences the highest local exposure within that peak exposure bundle node. It is noted from Figure SRXB-A-65-1 that (1) the difference between the non-uprated and uprated nodal operating histories is relatively small, and (2) both operating histories are well bounded by ((

)) the LHGR Operating Limit. It should be noted that at any point in time the local fuel rod power level could potentially momentarily approach or even be at the LHGR Operating Limit[f

]. The presented power history for JLC505 Rod K4 Node 5 provides a characterization of a typical operating history for a fuel rod node that operated at highest power, on the average over lifetime, of all fuel rods in the third cycle reload batch present in VYNPS Cycle 25. In this case, JLC505 Rod K4 Node 5 did not approach the LHGR Operating Limit prior to Cycle 25 and is not projected to approach the LHGR Operating Limit during VYNPS Cycle 25, although, again, it is recognized that any individual fuel rod, either JLC505 Rod K4 during actual Cycle 25 operation or any other fuel rod, could briefly operate at the LHGR Operating Limit.

((

to BVY 05-086 Docket No. 50-271 Page 5 of 28 NON-PROPRIETARY VERSION

3. The basic fuel rod thermal-mechanical design analysis methodology currently used by GNF was implemented with GESTAR Amendment 7 with corresponding NRC approval as documented in Reference 65-1. Subsequent to the initial methodology approval, the NRC, in conjunction with NRC consultant and fuel rod thermal-mechanical analysis expert Dr. Carl Beyer (PNL), again reviewed the fuel rod thermal-mechanical design analysis methodology as documented in Reference 65-2.

At the time of the original NRC review and approval of the fuel rod thermal-mechanical design and analysis methodology, the uncertainty in the fuel rod operating power level was addressed (1) directly through explicit consideration of the local power level variations that could develop ((

to BVY 05-086 Docket No. 50-271 Page 6 of 28 NON-PROPRIETARY VERSION B].

In NEDC-32694-P-A, the response to RAI-5 (page A-10) identified an uncertainty in local LHGR of ((

)). Later revisions to the uncertainty treatment described in RAI-5 resulted in a slight increase to ((

)) (page B-3 in the same topical report). Applying the adjusted uncertainty driven by lack of gamma scan data from RAI SRXB-A-41 for VYNPS would result in an uncertainty of [f 11 For the fuel rod thermal-mechanical transient overpower analyses, again, the fuel rod is assumed ((

)). This approach introduces considerable conservatism relative to the conditions that would be calculated for an actual operating history with a randomly placed transient event.

4. ((

]3 exposure limits are established for each product line. These limits are conservatively established with approved methods, including appropriate provisions for uncertainties. The limit established for GE14 fuel is applicable under the proposed CPPU conditions for VYNPS.

The fuel rod thermal-mechanical performance consideration of greatest interest at exposures near the peak pellet exposure limit is the fuel rod internal pressure. As discussed above, a significant conservatism, most especially for the fuel rod internal pressure calculation, is ((

)). Therefore, no additional conservatism in local exposure monitoring is required to maintain fuel integrity.

The discussion below supports items 1 - 4 above and contains additional information regarding the first paragraph of RAI SRXB-A-65.

As a point of clarification to the first paragraph of the RAI, the response to RAI 11.5 in NEDC-32694P-A applies to uncertainties and core monitoring considerations. The LTR covers these topics, as well as their relevance to the SLMCPR methodology. The original RAI response provided a derivation of the uncertainty in the predicted peak LHGR.

As discussed in the topical report, the same component uncertainties are incorporated into the SLMCPR.

However, the LTR did not directly address how the uncertainties were incorporated ((

)). The responses documented in to BVY 05-086 Docket No. 50-271 Page 7 of 28 NON-PROPRIETARY VERSION NEDC-32694P-A accurately describe the uncertainties, but only in terms of their application in the SLMCPR.

The response to SRXB-A-41 indicated a slight increase in the predicted peak LHGR uncertainty. The response also indicated that power uncertainty is considered ((

)). The response included the statement 'A local uncertainty of ((

))." This statement is accurate.

This ((

)) local power uncertainty is utilized with the application of the GESTR thermal-mechanical model ((

)) for each fuel product line. Additional discussion concerning determination of the exposure-dependent LHGR Operating Limit is given below.

For each GNF fuel design, including GE14 as applied to VYNPS, LHGR operating limits are determined and specified in the form of allowable ((

)) LHGR as a function of ((

)) exposure. These fuel rod thermal-mechanical performance based operating limits are specified for each fuel rod type (UO2 or (U,Gd)02 for various gadolinia concentrations) so that if each fuel rod type is operated within its respective exposure-dependent LHGR limit, all thermal-mechanical design and licensing criteria, including those which address response to anticipated operational occurrences, are explicitly satisfied.

The exposure-dependent LHGR operating limits are determined through the performance of a number of fuel rod thermal-mechanical analyses. As shown to the NRC staff during the GE Methods audit, an important assumption with these analyses is

)). This assumption represents a significant conservatism; ((

))-

With this conservative ((

]l assumption, the thermal-mechanical analyses are performed either on a worst tolerance basis or statistically. For those analyses performed statistically, such as the fuel rod internal pressure analysis, the uncertainty in each fuel rod fabrication parameter is determined and specifically addressed.

The fuel rod thermal-mechanical model prediction uncertainty is also determined and addressed.

((

to BVY 05-086 Docket No. 50-271 Page 8 of 28 NON-PROPRIETARY VERSION

))

For the GE14 fuel rod thermal-mechanical design and the preceding component uncertainties are: ((

licensing analyses, the values of

)).

The LHGR Operating Limit is derived for an individual fuel design using the following basic procedure.

to BVY 05-086 Docket No. 50-271 Page 9 of 28 NON-PROPRIETARY VERSION to BVY 05-086 Docket No. 50-271 Page 10 of 28 NON-PROPRIETARY VERSION

))

Figure SRXB-A-65-2 is a chart presented to the USNRC in recent discussions to describe the results of the GE14 fuel rod thermal-mechanical design and licensing analyses, and is included here for documentation purposes. The primary result of the fuel rod thermal-mechanical design and licensing analyses is development of the LHGR Operating Limit. The analyses that contribute directly to the development of that limit are the analyses for ((

to BVY 05-086 Docket No. 50-271 Page 11 of 28 NON-PROPRIETARY VERSION

))

In summary, with this methodology, the exposure-dependent LHGR Operating Limit is determined to ensure that the fuel rod thermal-mechanical design and licensing limits, such as the fuel rod internal pressure limit, will not be exceeded ((

B].

References 65-1.

Letter from C. 0. Thomas (NRC) to J. S. Chamley (GE), 'Acceptance for Referencing of Licensing Topical Report NEDE-2401 1-P-A Amendment 7 to Revision 6, GE Standard Application for Reactor Fuel," March 1, 1985 65-2. Letter from Robert M. Gallo (NRC) to C. P. Kipp (GE), 'NRC Inspection Report No. 99900003/96-01," September 10, 1996 to BVY 05-086 Docket No. 50-271 Page 12 of 28 NON-PROPRIETARY VERSION 1]

Figure SRXB-A-65-1 VYNPS Cycle 25 Projected Actual Operating History for JLC505 Rod K4 Node 5 - Comparison Between Uprated and Non-Uprated Conditions (JLC505 Node 5 is the highest projected bundle nodal exposure in VYNPS Cycle 25; rod K4 is the highest exposure rod node in bundle JLC505 Node 5. See further description in Item 2 text above on page 4 of 28.)

to BVY 05-086 Docket No. 50-271 Page 13 of 28 NON-PROPRIETARY VERSION 11 to BVY 05-086 Docket No. 50-271 Page 14 of 28 NON-PROPRIETARY VERSION RAI SRXB-A-67 Shutdown Margin (SDM)

In the Alternative Approach and in the RAI responses, VYNPS SDM data was not provided as discussed in the July 12, 2005, telephone conference. As the NRC staff pointed out in the June 30, 2005 meeting, Figure 25-18, "Cold Critical Eigenvalues-AII Cycles Studies," of the MFN-05-029 shows that the actual cold eigenvalue tracking of different plants show a scatter of the bias of each plant.

However, the uncertainty applied to each plant is obtained by RMS averaging of bias from all plants. Thus, it seems that a bias of 0.38% Ak/k is applied to the calculated core-wide critical keff (in-sequence cold eigenvalue) although the bias from critical (keff = 1.0) may be larger for a given plant.

Also, presenting the calculated cold critical eigenvalue alone does not indicate if the critical control rod positions were predicted.

a) Provide the VYNPS cold critical eigenvalues for at least two cycles. Include the recent mid-cycle startup cold critical eigenvalue. Include tables of the predicted keff with the CR withdrawals and indicate predicted critical eigenvalue and the calculated cold critical eigenvalue corresponding to when the core became critical. Evaluate the bias in the VYNPS cold critical eigenvalue data.

b) Provide the actual calculated SDM, with the correction for the period, temperature and peak reactivity.

c) The alternative approach states that for VYNPS "the standard design SDM is 1.1%

Ak/k to provide additional flexibility in cycle length and operations."

Clarify this statement. Is this an additional margin included to meet the cycle energy needs or is this additional conservatism that ensures SDM for any point in the cycle?

d) The Alternative Approach did not include impact of potential underprediction in reactivity and bundle and pin powers on the SLC system cold shutdown capability.

Provide an evaluation of the SLC system shutdown capability and rod withdrawal error analysis.

e) Demonstrate that the ((

)) would not have an important impact when the

)) void fraction and extrapolation to higher voids are used.

Also, provide a discussion on what such an under-prediction would have on the accuracy of the local reactivity predictions and what impact, if any, it would have on the SDM, SLC system cold shutdown and rod withdrawal error calculations.

f) The RAI responses stated that the objective is for the eigenvalue trendline to remain constant and consistent from cycle to cycle for a given plant, unless significant change in core loading design results in some change in the trendline. However, the trendline is not a licensing parameter and can be adjusted according to a new trendline fitting a change in the data.

The licensing parameter is the SDM.

to BVY 05-086 Docket No. 50-271 Page 15 of 28 NON-PROPRIETARY VERSION Therefore, from a licensing and safety perspective, the difference between the calculated keff for a critical reactor and the deviation from 1.0 is the most important parameter.

Explain why it is not desirable for the keff bias and uncertainty to be derived on plant-specific bases. Thus, ensuring a better adjustment applied to the keff bias assumed in the SDM calculations would be based on individual plant's characteristic response and the accuracy of the neutronic methods.

Response to RAI SRXB-A-67 Response to Part (a)

The cold critical eigenvalues for the Vermont Yankee Nuclear Power Station (VYNPS)

Cycles 23 and 24 are presented below.

The results shown below are for the TGBLA06/PANAC1 1 set of methods. Because Cycle 24 was the first cycle at VYNPS to be designed and licensed with PANACI 1, no predicted eigenvalues had been established for earlier cycles.

The previous cycle cold criticals were analyzed with PANAC1 1 however in order to establish a data base from which the Cycle 24 predicted eigenvalues were developed.

The mid-cycle Cycle 23 predicted eigenvalue was established by taking the actual beginning of cycle (BOC) eigenvalue and adjusting it by the standard reduction in cold eigenvalue with cycle exposure (used when sufficient mid-cycle information is not available for a plant). The process for determining predicted cold critical eigenvalues is discussed in the response to part (f) of this request.

The cold eigenvalues shown are very typical for other BWRs operating with GE fuel and analyzed with PANAC1 1 methods (Reference 67a-1).

((

1]

Cyle Cycle Exposure Predicted Critical Difference yc(MWdIST)

Eigenvalue Elgenvalue (Ak) 23 BOC

((

7417 24 BOC 961

))

to BVY 05-086 Docket No. 50-271 Page 16 of 28 NON-PROPRIETARY VERSION ResDonse to Part (b)

The VYNPS Technical Specification (TS) Shutdown Margin (SDM) is determined following a core reload, at the beginning of each cycle during plant startup. A copy of the demonstrated SDM calculation for the current operating cycle (cycle 24) is attached (see ). The calculation involves correcting the SDM for the effects of temperature and period present at the critical measurement. As calculated, the SDM also includes a correction for any difference in peak reactivity at any point in the cycle, R. The period and temperature correction is obtained from the Cycle Management Report, as is the correction for the difference in peak reactivity, R. It should be noted that the temperature correction is a translation to the most reactive condition.

As shown in the accompanying worksheet, Cycle 24 SDM was demonstrated by test to be 1.291. The cycle was designed for Extended Power Uprate (115% CLTP) and a SDM of 1.1, which indicates that the SDM design criterion was easily met for this cycle.

During discussions related to this subject, some other issues were identified by the reviewers, and are addressed below:

The VYNPS TS require that the SDM, at any time there is fuel in the core, shall be greater than or equal to 0.38% AK/K with the analytically determined highest worth rod fully withdrawn. The 0.38% AK/K was determined based upon a statistical combination of allowed manufacturing tolerances and calculational uncertainties. The calculational uncertainties were determined from a statistical analysis of measured and calculated criticals performed at an operating BWR.

Procedurally, if the demonstrated SDM is less than 0.38% AK/K, then the shift manager is immediately notified and SDM must be restored within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or the reactor must be in Hot Shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If the corrected critical eigenvalue is different from the expected critical eigenvalue by more than 1% AK/K, then the shift manager is immediately notified and the reactor must be shut down until the cause is determined.

Additionally, if the corrected critical eigenvalue is different from the expected critical eigenvalue by greater than 0.75% AK/K, then the reactor engineering superintendent is notified and a Condition Report is initiated.

Typically, the SDM demonstration is performed during the beginning-of-cycle (BOC) startup. Within the calculation of the demonstrated in-sequence SDM, there is a factor, R, that accounts for a decrease in SDM during the most reactive point in the cycle. This factor is zero when SDM is determined at the most reactive point in the cycle. For those situations when the SDM is not determined at the most reactive point in the cycle, the R factor is subtracted from the demonstrated SDM.

With regard to the effect of the assumed critical eigenvalue and its uncertainty on the demonstrated SDM, the following discussion is offered:

Per the Cycle Management Report (CMR), the equation for SDM is as follows:

to BVY 05-086 Docket No. 50-271 Page 17 of 28 NON-PROPRIETARY VERSION SDM = Kclt - KsRo + KTemp -

KPer-R

where, Kcnt = Eigenvalue when critical is achieved, KSRO = Eigenvalue with the strongest rod out (SRO),

KTemp = AK temperature correction, Kper = AK period correction, and R = Maximum decrease in SDM throughout the cycle.

But since, Kcrit

= Keft with all rods in (ARI) + AK of the critical rod pattern (CRP)

= KARI + AKCRP, and KsRO

= Keff with all rods in + AK of the strongest rod out

= KARI + AKSRo, then SDM = (KARI + AKCRP) - (KARI + AKSRO) + KTemp -

Kper - R which simplifies to:

SDM = AKCRP - AKSRo + KTemp -

Kper - R KARI is subject to the influence of the assumed critical eigenvalue and its uncertainty. It can be seen from the final equation that KARI cancels out and the demonstrated SDM is not influenced by the assumed critical eigenvalue or its uncertainty. However, it should be noted that AKSRO includes a 0.003 AK/K adjustment to account for the methods bias which occurs when normalizing shutdown margin calculations to a cold eigenvalue derived from in-sequence critical benchmarking data.

to BVY 05-086 Docket No. 50-271 Page 18 of 28 NON-PROPRIETARY VERSION Response to Part (c)

The VYNPS TS 3.3.A.1 requires that any time fuel is the in the core, the core loading shall be limited to that which may be made subcritical in the most reactive condition during the operating cycle with the highest worth, operable control blade fully withdrawn and all other operable rods inserted.

The shutdown margin shall be:

(a) Greater than or equal to 0.38% Ak/k with the highest worth rod analytically determined; or (b) Greater than or equal to 0.28% Ak/k with the highest worth rod determined by test.

Entergy confirms sufficient SDM for VYNPS at the BOC based upon greater than or equal to 0.38% Ak/k.

A failure to meet the Technical Specification SDM requirement is severe in that a redesign of the core loading and/or fuel design would be required to restart the plant. To ensure that 2 0.38% Ak/k is always satisfied, a design margin of 1% SDM has been used by GE for many years. The additional margin between the Technical Specification SDM and 1% allows for the following factors to impact the prediction capability of the simulator:

1. Operation of the plant different than that projected

2. Fuel manufacturing tolerances

3. Control rod worth reduction due to depletion of control rod absorber material

4. Methodology approximations

5. Inexact tracking of actual plant parameters

6. Other unidentified factors In all of these factors, the most significant factor is allowance for operation different from that projected. VYNPS must maintain sufficient operational flexibility to protect the core and fuel while maintaining acceptable economic objectives. Factors affecting the GE application methodology are quantified through the uncertainty in cold critical eigenvalue and deviation from expectations. These data are provided in the responses to RAls SRXB-A-67 part (a) and SRXB-A-67 part (b).

The additional 0.1% Ak/k that VYNPS requires results from consideration of inverted B4C tubes in the core. Based upon a total of 82 inverted B4C tubes in 44 control rods in 1975, a 0.07% Ak/k SDM adder was required to compensate for the inverted B4C tubes.

[Reference 67c-1] While there are only 30 inverted B4C tubes in 13 peripherally located control blades, the 0.07% Ak/k SDM adder is still being applied until all affected control rods are discharged.

to BVY 05-086 Docket No. 50-271 Page 19 of 28 NON-PROPRIETARY VERSION If the SDM demonstration at VYNPS results in a SDM less than Technical Specification requirement, the plant will take actions as specified in the Technical Specifications.

Response to Part (d)

The standby liquid control system (SLCS) calculation is performed on a cycle specific basis to assure that the plant will remain subcritical in the most reactive condition when the Technical Specification (Tech Spec) minimum requirement for soluble boron is introduced into the core.

The calculation is performed as a function of exposure throughout the cycle to determine the minimum SLCS shutdown margin during the cycle.

This is an analytical determination, and no actual demonstration of this shutdown capability is performed as is done in the one-rod-out shutdown margin.

In order to provide a high degree of assurance that the analytically determined shutdown margin will indeed result in a subcritical condition, a SLCS shutdown margin criteria is established, requiring that the analytically determined shutdown margin be greater than or equal to this shutdown margin criteria. The criteria accounts for all of the biases and uncertainties inherent in the various components of the SLCS methodology.

It should be noted that unlike the one-rod-out shutdown margin, which must be demonstrated subsequent to any reconfiguration of the core, and which is highly sensitive to the local conditions in the four bundles surrounding the withdrawn blade, the SLCS shutdown margin is driven more by core-wide reactivity effects. This makes the calculation less sensitive to nodal uncertainties in exposure and isotopic content, and more dependent on the average exposure and reactivity behavior of the various fuel batches loaded in each cycle.

to BVY 05-086 Docket No. 50-271 Page 20 of 28 NON-PROPRIETARY VERSION

))

The severity of the RWE transient is largely dependent on the worth of the rod being withdrawn. The limiting bundle for the RWE for the VYNPS Cycle 25 analysis shows a controlled to uncontrolled AK-of approximately ((

)) Of the four bundles face-adjacent to the error rod, two bundles are approximately ((

)) including the limiting bundle. The other two bundles are approximately ((

)). The higher exposure bundles show a smaller AK-, ((

)), and a corresponding lower change in power and CPR during the RWE. The trend of reduction in AK-, and corresponding lower change in power and CPR during the RWE, continues at exposures greater than Response to Part (e)

((:

)) as the impact on the 0, 40, and 70% void data is minimal. Consequently, this effect does not significantly impact the extrapolation using the 0, 40 and 70% void data to voids higher than 70%.

The above discussion indicates that there is potential for a change in the lattice reactivity of ((

to BVY 05-086 Docket No. 50-271 Page 21 of 28 NON-PROPRIETARY VERSION 1]

To demonstrate the reactivity impacts of this modification to the ((

evaluation, a cycle of plant performance tracking using GE14 fuel in a high power density core was performed using both the current TGBLA production engineering computer program (ECP) and a version of TGBLA that was modified to correct this issue.

The hot core reactivity impact on the core tracking is shown in Figure SRXB-A-67-1 and the impact to Shutdown Margin (SDM) as a function of cycle exposure is shown in Figure SRXB-A-67-2.

Table SRXB-A-67-1 and Table SRXB-A-67-2 provide the core reactivity and SDM detailed results comparisons, respectively.

As shown in the figures and tables, ((

)) These levels of impact are not significant compared to the historical uncertainty of these calculated parameters.

((

The response to NRC RAI SRXB-A-67 part (d) provides a discussion of the impact of this potential reactivity uncertainty on the SLCS SDM and Rod Withdrawal Error (RWE) analyses.

The ((

)) (see response to NRC RAI SRXB-67d).

Response to Part (f)

The current process is consistent with the expressed concern ("Explain why it is not desirable for the k-eff bias and uncertainty be derived on plant-specific bases."). ((

to BVY 05-086 Docket No. 50-271 Page 22 of 28 NON-PROPRIETARY VERSION 1]

Cycle Exposure Critical ycle (MWd/ST)

Eigenvalue 21 BOC O

22 BOC 23 BOC 7417 24 BOC 961 11 to BVY 05-086 Docket No. 50-271 Page 23 of 28 NON-PROPRIETARY VERSION

References:

67a-1 MFN 05-029, TAC No. MC5780 67c-1 Letter, Dennis L. Ziemann (NRC) to G. Carl Andognini (YAEC), Docket No. 50-271, June 6, 1975 to BVY 05-086 Docket No. 50-271 Page 24 of 28 NON-PROPRIETARY VERSION 1[

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to BVY 05-086 Docket No. 50-271 Page 25 of 28 NON-PROPRIETARY VERSION 4

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4 4-IL to BVY 05-086 Docket No. 50-271 Page 26 of 28 NON-PROPRIETARY VERSION I

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to BVY 05-086 Docket No. 50-271 Page 27 of 28 NON-PROPRIETARY VERSION

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to BVY 05-086 Docket No. 50-271 Page 28 of 28 NON-PROPRIETARY VERSION

((

]1

BVY 05-086 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information Response to RAI SRXB-A-71 Total number of pages in Attachment 6 (excludina this cover sheet) is 1.

to BVY 05-086 Docket No. 50-271 Page 1 of 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING APPLICATION FOR EXTENDED POWER UPRATE LICENSE AMENDMENT VERMONT YANKEE NUCLEAR POWER STATION PREFACE This attachment provides a response to the NRC Reactor Systems Branch's (SRXB) request for additional information (RAI) SRXB-A-71 in NRC's letter dated September 7, 2005.1 Upon receipt of the RAI, discussions were held with the NRC staff to further clarify the RAI. The intent of individual RAI is addressed based on clarifications reached during these discussions. The information provided herein is consistent with those clarifications.

The RAI is re-stated as provided in NRC's letter of September 7, 2005.

RAI SRXB-A-71 In the response to RAI SRXB-A-6, the licensee stated "the reactivity events are analyzed with the steady state tools and the results presented regarding steady-state methods in this response are directly applicable. There are some increases in power, which are significant but remain within the comparisons between the above plants for corresponding events." This RAI response does not provide sufficient detail.

The response to RAI SRXB-A-57 requested clarification to the above quoted statement. The generic event sequence was described, rather than explaining the statement in the initial RAI response.

Please explain the intent of the statement in the initial submittal.

Response to RAI SRXB-A-71 The intent of the statement in quotations was that the VYNPS events analyzed with the 3D core thermal-hydraulic PANACEA model, such as the Rod Withdrawal Error and Fuel Loading Error, started from conditions within the range in other analyses as shown in Figures 6-1 through 6-6 of the response to RAI SRXB-A-6 2. No comparison was made against the events analyzed with the steady state methods for the other plants of Figures 6-1 through 6-6 because of differences in the plant size, core design and loading, rod block monitor setup, power distribution and control rod patterns, which result in inconsistent comparisons.

1 U.S. Nuclear Regulatory Commission (Richard B. Ennis) letter to Entergy Nuclear Operations, Inc.

(Michael Kansler), "Request for Additional Information - Extended Power Uprate, Vermont Yankee Nuclear Power Station (TAC No. MC0761)," September 7, 2005 2Entergy letter to U.S. Nuclear Regulatory Commission, 'Vermont Yankee Nuclear Power Station, Technical Specification Proposed Change No. 263 - Supplement No. 24, Extended Power Uprate -

Response to Request for Additional Information," BVY 05-024, March 10, 2005

BVY 05-086 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information General Electric Affidavit Total number of pages in Attachment 7 l {excludina this cover sheet) is 3.

l

General Electric Company AFFIDAVIT I, George B. Stramback, state as follows:

(1)

I am Manager, Regulatory Services, General Electric Company ("GE"), have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2)

The information sought to be withheld is contained in Enclosure 2 of GE letter, GE-VYNPS-AEP-403, Responses to NRC RAls SRXB-64, 65, 67, and 71, dated September 16, 2005. The proprietary information in Enclosure 2, Responses to NRC RAIs SRXB-64, 65, 67, and 71, is delineated by a double underline inside double square brackets. Figures and large equation objects are identified with double square brackets before and after the object. In each case, the superscript notation refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3)

In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission. 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir. 1983).

(4)

Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, resulting in potential products to General Electric; GBS-05 af VYNPS-AEP-403 Methods RAIs 9-1605.doc Affidavit page I

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a., and (4)b, above.

(5)

To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6)

Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7)

The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation.

Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8)

The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results and conclusions from analyses supporting the extended power uprate of the Vermont Yankee Power Station utilizing analytical models and methods including computer codes and methods of applying these for safety analyses, which GE has developed. The development of these models and computer codes and methods was achieved at a significant cost to GE, on the order of several million dollars.

The development of the analytical methods and evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

GBS-05 af VYNPS-AEP403 Methods RAIs 9-16-05.doc Affidavit page 2

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 16 A of September 2005.

4eorgeB t General Electric GBS-05 af VYNPS-AEP403 Methods RAIs 9-16-05.doc Affidavit page 3

BVY 05-086 Docket No. 50-271 Vermont Yankee Nuclear Power Station Proposed Technical Specification Change No. 263 - Supplement No. 34 Extended Power Uprate - Additional Information Demonstrated Shutdown Margin Total number of pages in Attachment 8 (excludina this cover sheet) is 15.

l

.1 IN-SEQUENCE CRMTICAJJSDM WORKSBEET

1. Prerequisites Met I

~Prted Na=eSgatz

2. Conditions at Criticality Datefime of Citicality I

Mld

'IF

.° Computer point B023 prefenred Reactor Period seconds CriticalRodLacation t¢-s_

Notch Position 3 (P

3. Calculation of Control Rod Density Total Rod Positions Witbdrawn, N (e.g. IrodW-402 is 2positions)

I_____________

Contro Rod Density (CRD) =Can be verified via Computer (4M2-N)14272

Pon V

4. Calculation of Uncorrected Critical Eigenvalue kif of critical rod at analyzed withdraw Per c design do O limit Tndified s

@n bstiate Ftactio, of analyzd step widrawn at criticality o

k4ff of rod prior to ciicality at its Per cum &sign vendor or withdraw limit

q.

oof s

i UncorreCte ritical Eigenvalue k~=Fr * (ki - k2) + k S. Calculation of Corrected Critical Eigenvahue Period Correction Per core design vendor or k_

qualified substitte Temperature Correction 6 O I

Per core design vendor or k41 substitute Corrected Critical Egenvalue k.=

k,-

kp,, + kwp

.I I!

S VYOPF 4430.03 OP 4430 Rev. 22 Page 1of//S

I h

IN-SEQUENCE CRMTICAIJSDM WORKSBF~ (Continued) 6 Calculation of Demonstrated SDM Eigenvalue with strongest rod out Per core design vendor or qualified substitute OA9113 Ak.

Decrease in SDM with exposure over Per core desip vendor or qualified cycle substiMe. The value of RI is equal to zero when the SDM test is peffomed at the most reactive point dauing the cycle (in other words, when BOC is the most reactive point Ak in thecycle).

Potential SDM loss from inverted CR tubes R2 0.0007 Ak Per Reference 3.a Demonstrated SDM SDM"

= 100* (k= -ksRo Ri -4Ykcr

7. Calcuation of Difference BetweenCorrecedand ErpecedCiitical Eigenvalue~, Nf Expected Critical Eigenvalue 1.00 Difference Between Corrected and Expected Critical Eige!lUe Diff = 100IkapkW

& Comments, Remarks, or Discrepancies:

,q. &S Ete4 iY bt, dSin /1

9. Calculations Performed By l

Plm Nae Sinpature Datlme

10. Calculations Verified By Prnted Name fSignane 7 Date VYOPF4430.03 OP 4430 Rev. 22 Page 2 o/

/

IN-SEQUENCE CRMICAIISDM WORKSHEET (Continued)

VERIFICATION OF TEC[INICAL SPECIFMCAT1ON ACCEPTANCE CRfTERIA

11. Verification of Tech. Spec. Acceptance Criteria, SDM&,

2038 SDMA 1

0ŽD38 Yes No (Check applicable box)

VerifiedBy 5.6, s

-7

/

(Printed Name and Siature)

Pied Namc sigature Mae.

Verified By cdw41.-c 4-M4t f

7Q X

(Printed Name and Signature)

Pid Nm Signa IF SDM4

<0.38, THEN notify Shift Supervisor immediately that SPM must be restored within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or the reactor must be in hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Tech. Spec. 3.3.-A1 Verification of Tech-Spec. Acceptance Criteria, 1%

• Diffl %

IF Tech Spec Acceptance Criteria of -1% z DM ff +1% is not met, TIEN noty Shift Supervisor immediately that reactor must be shut down until cause has been determined and corrective actions have been taken if such actions are appropiate, per Tech Spec 3.3.E VER[CATION OF ADMISTRAXTIVE LIMT ACCEPTANCE CRMERI.

13 Verificatim of Adnin Acceptance Criteria, 47.5% SDIff 0.75%

-o.75%5Diff 0.75%

Yes (Check applicabIe box)

Verified By t& SesA 4A

-f.jM^

sh4 (Printed Name and Siate)

PndName signamr De Verified.By (Printed Name and Signature)

Pit ame iSignatw Da=e IF Administrative Acceptance Criteria of -0.75 iDff* +0.75% is not met, TfE notify Superintendcn4t Reactor Engineering immediately and initiate an ER.

VYOPF 4430.03 OP 4430 Rev. 22 Page 30/AS