ML022330315
| ML022330315 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 08/07/2002 |
| From: | Price J Dominion Nuclear Connecticut |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| B18691 | |
| Download: ML022330315 (17) | |
Text
Dominion Nuclear Connecticut, Inc. w'~JP"'D mnin Millstone Power Station Rope Ferry Road Waterford, CT 06385 Docket No. 50-336 B18691 Re: 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Millstone Nuclear Power Station, Unit No. 2 Changes to Technical Specifications Updating List of Documents Describing the Analytical Methods Specified in Technical Specification 6.9.1 .8b (LBDCR 2-11-02)
Pursuant to 10 CFR 50.90, Dominion Nuclear Connecticut, Inc. (DNC) hereby proposes to amend Operating License DPR-65 by incorporating the attached proposed changes into the Millstone Unit No. 2 Technical Specifications. The proposed changes will update the list of documents, describing the analytical methods used to determine the core operating limits, specified in Technical Specification 6.9.1.8b. The reason for these changes is to incorporate the Nuclear Regulatory Commission (NRC) approved, methodology documents in Millstone Unit No. 2 Technical Specifications. These changes will update the documents describing Steam Line Break (SLB) methodology, and Departure from Nucleate Boiling (DNB) correlation for High Performance Fuel (HPF). provides a discussion of the proposed changes and the Safety Summary. provides the Significant Hazards Consideration. Attachment 3 provides the marked-up version of the appropriate pages of the current Technical Specifications. provides the retyped pages of the Technical Specifications.
Environmental Considerations DNC has evaluated the proposed changes against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.22. DNC has determined that the proposed changes meet the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92(b). This determination is based on the fact that the changes are being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to use of a facility component located within the restricted area, as defined by
U.S. Nuclear Regulatory Commission B18691/Page 2 10 CFR 20, or that changes a surveillance requirement, and that the amendment request meets the following specific criteria.
(i) The proposed changes involve no Significant Hazards Consideration.
As demonstrated in Attachment 2, the proposed changes do not involve a Significant Hazards Consideration.
(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released off-site.
The proposed amendment will update the list of documents, describing the analytical methods used to determine the core operating limits, specified in Technical Specification 6.9.1.8b. However, the operability requirements for equipment associated with these Technical Specifications will remain the same.
The proposed changes are consistent with the design basis of the plant. The proposed changes will not result in an increase in power level, will not increase the production of radioactive waste and by-products, and will not alter the flowpath or method of disposal of radioactive waste or by-products. Therefore, the proposed changes will not increase the type and amounts of effluents that may be released off-site.
(iii) There is no significant increase in individual or cumulative occupational radiation exposure.
The proposed amendment will update the list of documents, describing the analytical methods used to determine the core operating limits, specified in Technical Specification 6.9.1.8b. However, the operability requirements for equipment associated with these Technical Specifications will remain the same.
The proposed changes will not result in changes in the configuration of the facility. There will be no change in the level of controls or methodology used for processing radioactive effluents or the handling of solid radioactive waste. There will be no change to the normal radiation levels within the plant. Therefore, there will be no increase in individual or cumulative occupational radiation exposure resulting from the proposed changes.
Conclusions The proposed changes were evaluated and we have concluded that they are safe. The proposed changes do not involve an impact on public health and safety (see the Safety Summary provided in Attachment 1) and do not involve a Significant Hazards Consideration pursuant to the provisions of 10 CFR 50.92 (see the Significant Hazards Consideration provided in Attachment 2).
i U.S. Nuclear Regulatory Commission B18691/Page 3 Site Operations Review Committee and Nuclear Safety Assessment Board The Site Operations Review Committee and Nuclear Safety Assessment Board have reviewed and concurred with the determinations.
Schedule We request issuance of this amendment prior to restart from refueling outage 15, which is currently scheduled for the Fall of 2003, with the amendment to be implemented prior to Mode 4 operation of Cycle 16.
State Notification In accordance with 10 CFR 50.91(b), a copy of this License Amendment Request is being provided to the State of Connecticut.
There are no regulatory commitments contained in this letter.
If you should have any questions regarding this submittal, please contact Mr. Ravi Joshi at (860) 440-2080.
Very truly yours, DOMINION NUCLEAR CONNECTICUT, INC.
JAonVrce" Sitei Ve President - Millstone Sworn to and subscribed before me this day of k.,- 2002 Notary Pub&C WM. E. BROWN My Commission expires NOTARYPUBLIC MyCQMMISSION EXPIRE .~t31,2M0 Attachments cc: Next Page -
U.S. Nuclear Regulatory Commission B18691/Page 4 cc: H. J. Miller, Region I Administrator R. B. Ennis, NRC Project Manager, Millstone Unit No. 2 NRC Senior Resident Inspector, Millstone Unit No. 2 Director Bureau of Air Management Monitoring and Radiation Division Department of Environmental Protection 79 Elm Street Hartford, CT 06106-5127
I Docket No. 50-336 B18691 Attachment 1 Millstone Nuclear Power Station, Unit No. 2 Change to Technical Specifications Updating List of Documents Describing the Analytical Methods Specified in Technical Specification 6.9.1.8b (LBDCR 2-11-02)
Discussion of Changes and Safety Summary
U.S. Nuclear Regulatory Commission B18691/Attachment 1/Page 1 Change to Technical Specifications Updating List of Documents Describing the Analytical Methods Specified in Technical Specification 6.9.1.8b (LBDCR 2-11-02)
Discussion of Changes and Safety Summary Introduction Dominion Nuclear Connecticut, Inc. (DNC) hereby proposes to amend Operating License DPR-65 by incorporating the attached proposed changes into the Millstone Unit No. 2 Technical Specifications. The proposed changes will update the list of documents, describing the analytical methods used to determine the core operating limits, specified in Technical Specification 6.9.1.8b.
Description of Proposed Chanqie The proposed changes will update the documents describing the Nuclear Fuel and Safety Analysis vendor's methodology given in Technical Specification 6.9.1.8b. The following are the proposed changes:
- 1. The document contained in section 6.9.1.8b.4, describing Steam Line Break methodology for PWRs, is deleted and documents 6.9.1.8b.5 through 6.9.1.8b.15 are renumbered. This document is deleted since it has been superceded by the most recent methodology as described in the document contained in section 6.9.1.8b.15 (renumbered 6.9.1.8b.14).
- 2. A new document is added as 6.9.1.8b.15. This change is required to identify the most recent methodology description for the HTP: Departure from Nucleate Boiling Correlation for High Performance Fuel.0') This document is referenced in the document currently in section 6.9.1.8b.15 (renumbered 6.9.1.8b.14). This document has been reviewed and approved by the Nuclear Regulatory Commission (NRC) and the associated Safety Evaluation Report (SER) is contained in a letter dated December 28, 1993. 2 )
Safety Summary The proposed changes Will update the list of documents, describing the analytical methods used to determine the core operating limits, specified in Technical Specification 6.9.1.8b.
(1) EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation, March 1994.
(2) Acceptance For Referencing of Siemens Power Corporation Topical Report EMF-92 153(P), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," dated December 28, 1993.
U.S. Nuclear Regulatory Commission B18691/Attachment 1/Page 2 Deleting the document contained in section 6.9.1.8b.4 is required since it has been superceded by the most recent methodology as described in the document contained in section 6.9.1.8b.15 (renumbered 6.9.1.8b.14). The use of the most recent methodology constitutes an improvement over the previous methodology and is in accordance with the NRC's SER. Therefore, this proposed change will have no adverse effect on plant safety.
Adding the new document associated with the new section 6.9.1.8b.15 to the list of references is required for completeness. This document is referenced in the document currently in section 6.9.1.8b.15 (renumbered 6.9.1.8b.14) and describes the HTP:
Departure from Nucleate Boiling Correlation for High Performance Fuel. Since this new document complements the document currently in section 6.9.1.8b.15 (renumbered 6.9.1.8b.14) and provides completeness and clarity, this proposed change will have no adverse effect on plant safety.
-t Docket No. 50-336 B18691 Attachment 2 Millstone Nuclear Power Station, Unit No. 2 Change to Technical Specifications Updating List of Documents Describing the Analytical Methods Specified in Technical Specification 6.9.1.8b (LBDCR 2-11-02)
Sianificant Hazards Consideration
1V U.S. Nuclear Regulatory Commission B18691/Attachment 2/Page 1 Change to Technical Specifications Updating List of Documents Describing the Analytical Methods Specified in Technical Specification 6.9.1.8b (LBDCR 2-11-02)
Significant Hazards Consideration Significant Hazards Consideration In accordance with 10 CFR 50.92, Dominion Nuclear Connecticut, Inc. (DNC) has reviewed the proposed changes and has concluded that they do not involve a Significant Hazards Consideration (SHC). The basis for this conclusion is that the three criteria of 10 CFR 50.92(c) are not compromised. The proposed changes do not involve a SHC because the changes would not:
,1. Involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed change to delete the document contained in section 6.9.1.8b.4 is required since it has been superceded by the most recent methodology as described in the document contained in section 6.9.1.8b.15 (renumbered 6.9.1.8b.14). Adding the new document associated with the new section 6.9.1.8b.15 to the list of references is required for completeness. This change has no impact on plant equipment operation. Since the changes only affect description of the safety analysis methodology and do not revise any setpoints assumed in the accident analyses, they cannot affect the likelihood or consequences of accidents. Therefore, this change will not increase the probability or consequences of an accident previously evaluated.
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes will not alter the plant configuration (no new or different type of equipment will be installed) or require any new or unusual operator actions.
They do not alter the way any structure, system, or component functions and do not alter the manner in which the plant is operated. These changes do not introduce any new failure modes. Therefore, the proposed changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.
- 3. Involve a significant reduction in a margin of safety.
The proposed changes have no impact on plant equipment operation. The proposed changes do not revise any setpoints assumed in the analyses and do not affect the acceptance criteria for the Steam Line Break accident. Therefore, the proposed changes will not result in a reduction in a margin of safety.
Docket No. 50-336 B18691 Attachment 3 Millstone Nuclear Power Station, Unit No. 2 Change to Technical Specifications Updating List of Documents Describing the Analytical Methods Specified in Technical Specification 6.9.1.8b (LBDCR 2-11-02)
Marked-Up Pages
1,l List of Affected Paaes Technical Specification Title of Section Affected Page with Section Number Amendment Number 6 9.1.8b List of References used to determine core 6-18a, Amend. No. 260 operating limits 6-19, Amend. No. 260
4'-
ADMINISTRATIVE CONTROLS 0Decenbei,9 00t' MONTHLY OPERATING REPORT (Con't)
( Administrator, Region I, and one copy to the NRC Resident Inspector, no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.8 a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.
3/4. 1.1 .1 SHUTDOWN MARGIN - T > 200°F 3/4. 1. 1.2 SHUTDOWN MARGIN - Ta < 200°F 3/4.1.1.4 Moderator Temperature Coefficient 3/4.1.3.6 Regulating CEA Insertion Limits 3/4.2.1 Linear Heat Rate 3/4.2.3 Total Integrated Radial Peaking Factor - FT r 3/4.2.6 DNB Margin
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1) EMF-96-029(P)(A) Volumes I and 2, "Reactor Analysis System for PWRs Volume 1 - Methodology Description, Volume 2 Benchmarking Results," Siemens Power Corporation.
( 2) ANF-84-73 Appendix B (P)(A), "Advanced Nuclear Fuels Methodology, for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels.
- 3) XN-NF-82-21(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"
Exxon Nuclear Company.
Ai raw na nn*,n, ,a, .... , .- - -
4 .. ., i Z b-caMiinc I"* Krc-aK -m1Ctfl~fl4iOO.' fnr I1* 'II bifln i~ewr Enlanwnrtiann I ý,
I q1 XN-75-32(P)(A) Supplements I through 4, "Computational Procedure for' Evaluating Fuel Rod Bowing," Exxon Nuclear Company.
&') EMF-2328(P)(A), "PWR Small Break LOCA Evaluation Model S RELAP5 Based," Framatome ANP.
6,75 EMF-2087(P)(A), "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Siemens Power Corporation.
MILLSTONE - UNIT 2 .6-18a Amendment No. ,1 0706 77 , 79 ,
77y, 77, 7n
CORE OPERATING LIMITS REPORT (CONT.)
7 A XN-NF-78-44(NP)(A),
Ejection Transient "A Generic Analysis of the Control rod Nuclear Company.
for Pressurized water reactors," Exxon
(
XN-NF-621(P)(A), "Exxon Nuclear DNB Correlation for PWR Fuel Designs," Exxon Nuclear Company.
XN-NF-82-06(P)(A) and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company.
/0j.{) ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation.
1/
A XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"
Lxxon Nuclear Company.
ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized I Water Reactors: Analysis of Non-LOCA Chapter 15 Events,"
Advanced Nuclear Fuels Corporation.
t3)4A EMF-1961(P)(A), "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation.
EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology Pressurized Water Reactors," Framatome ANP.
for (
'2XSERT A /
- c. The core operating limits shall be determined so that all applicable limits (e.g., .fuel thermal-mechanical limits, thermal-hydraulic limits, ECCS limits, nuclear limits core such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
- d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be. provided issuance, for each reload cycle, to the NRC Document Control upon with copies to the Regional Administrator and Resident Inspector. Desk SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, copy to the Regional Administrator, one Region "I, and one copy to the NRC Resident Inspector within the time period specified each report. for These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
- a. Deleted MILLSTONE - UNIT 2 6-19 Amendment No. 7gy7, y, 0706 .... ?P g......
Page 6-19, Insert A
- 15) EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation.
Docket No. 50-336 B18691 Attachment 4 Millstone Nuclear Power Station, Unit No. 2 Change to Technical Specifications Updating List of Documents Describing the Analytical Methods Specified in Technical Specification 6.9.1.8b (LBDCR 2-11-02)
Retyped Pages
ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORT (Con't)
Administrator, Region I, and one copy to the NRC Resident Inspector, no later than the 15th of each month following the calendar month covered by the report.
CORE OPERATING LIMITS REPORT 6.9.1.8 a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle.
3/4.1.1.1 SHUTDOWN MARGIN - T > 200°F 3/4.1.1.2 SHUTDOWN MARGIN - Tavg < 200°F 3/4.1.1.4 Moderator Temperature Coefficient 3/4.1.3.6 Regulating CEA Insertion Limits 3/4.2.1 Linear Heat Rate 3/4.2.3 Total Integrated Radial Peaking Factor - FTr 3/4.2.6 DNB Margin
- b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
- 1) EMF-96-029(P)(A) Volumes I and 2, "Reactor Analysis System for PWRs Volume 1 - Methodology Description, Volume 2 Benchmarking Results," Siemens Power Corporation.
- 2) ANF-84-73 Appendix B (P)(A), "Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels.
- 3) XN-NF-82-21(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations,"
Exxon Nuclear Company.
- 4) XN-75-32(P)(A) Supplements I through 4, "Computational I Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company.
- 6) EMF-2087(P)(A), "SEM/PWR-98: ECCS Evaluation Model for PWR LBLOCA Applications," Siemens Power Corporation.
- 7) XN-NF-78-44(NP)(A), "A Generic Analysis of the Control rod I Ejection Transient for Pressurized water reactors," Exxon Nuclear Company.
MILLSTONE - UNIT 2 6-18a Amendment No. 7, Y, 7/9, 777, 0706 777, 77w, i ,
"'NADMINISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT (CONT.)
- 9) XN-NF-82-06(P)(A) and Supplements 2, 4, and 5, "Qualification of Exxon Nuclear Fuel for Extended Burnup," Exxon Nuclear Company.
- 10) ANF-88-133(P)(A) and Supplement 1, "Qualification of Advanced Nuclear Fuels PWR Design Methodology for Rod Burnups of 62 GWd/MTU," Advanced Nuclear Fuels Corporation.
- 11) XN-NF-85-92(P)(A), "Exxon Nuclear Uranium Dioxide/Gadolinia Irradiation Examination and Thermal Conductivity Results,"
Exxon Nuclear Company.
- 12) ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events,"
Advanced Nuclear Fuels Corporation.
- 13) EMF-1961(P)(A), "Statistical Setpoint/Transient Methodology for Combustion Engineering Type Reactors," Siemens Power Corporation.
- 14) EMF-2310(P)(A), "SRP Chapter 15 Non-LOCA Methodology for Pressurized Water Reactors," Framatome ANP.
- 15) EMF-92-153(P)(A) and Supplement 1, "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation.
- c. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.
- d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, one copy to the Regional Administrator, Region I, and one copy to the NRC Resident Inspector within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
- a. Deleted MILLSTONE - UNIT 2 6-19 Amendment No. 7*, 7*, ,
0706 M,