ML050610676
| ML050610676 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 02/25/2005 |
| From: | Quennoz S Portland General Electric Co |
| To: | Document Control Desk, NRC/FSME |
| References | |
| PGE-8010, VPN-005-2005 PGE-1015-2004 | |
| Download: ML050610676 (10) | |
Text
/PGE Portland General Electric Company Trojan Nuclear Plant 71760 ColumbiaRiver Hwy Rainier OR 97048 (503) 556-3713 February 25, 2005 VPN-005-2005 Trojan Nuclear Plant Docket 50-344 License NPF-1 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 PGE- 1015-2004. "Annual Report of the Trojan Nuclear Plant for 2004" The enclosure to this letter is Portland General Electric Company's Annual Report of the Trojan Nuclear Plant for Calendar Year 2004. This report is submitted in accordance with the requirements of Appendix C, Section 1.5.1.1 of PGE-8010, "Trojan Nuclear Plant Nuclear Quality Assurance Program."
Any questions concerning this report may be directed to Mr. J. D. Reid, of my staff, at (503) 556-6474.
Sincerely, Stephen M. Quennoz Vice President, Generation Enclosure c: Director, NRC Region IV, DNMS J. T. Buckley, NRC, NMSS, DWM D. Stewart-Smith, ODOE A. Bless, ODOE msso/
Connecting People, Power and Possibilities
PGE-1015-2004 ANNUAL REPORT OF THE TROJAN NUCLEAR PLANT FOR 2004 Docket 50-344 License NPF-1 PORTLAND GENERAL ELECTRIC COMPANY 121 S. W. Salmon Street Portland, Oregon 97204
TABLE OF CONTENTS Pag INTRODUCTION........................................................................................................................... ii
SUMMARY
OF OPERATING EXPERIENCE IN 2004 ....................................... ii
- 1. Annual Personnel Exposure and Monitoring Report ............................................ 1
- 2. Changes, Tests, and Experiments ............................................ 2 i
INTRODUCTION The Annual Report for Trojan Nuclear Plant For Calendar Year 2004 is submitted in accordance with the requirements of PGE-8010, "Trojan Nuclear Plant Nuclear Quality Assurance Program."
SUMMARY
OF OPERATING EXPERIENCE IN 2003 Throughout the entire year, the plant remained permanently shut down. The spent fuel was transferred to an Independent Spent Fuel Storage Installation (ISFSI) licensed under the site-specific provisions of 10 CFR 72 during 2003.
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- 1. ANNUAL PERSONNEL EXPOSURE AND MONITORING REPORT Requirement PGE-8010, "Trojan Nuclear Plant Nuclear Quality Assurance Program," Appendix C, Section 1.5.1.1, requires an occupational radiation exposure report covering the activities of the facility for the previous calendar year be submitted that includes:
".. a tabulation on an annual basis of the number ofstation, utility, and otherpersonnel(including contractors)receiving exposures
> 100 mrem/yr and their associatedman-rem exposure according to work andjobfunctions (e.g., fuel handling, surveillance, maintenance and waste processing). This tabulationsupplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescentdosimeter(TLD), orfilm badge measurements. Small exposures totaling < 20% of the individual total dose need not be accountedfor. In the aggregate, at least 80% of the total whole body dose receivedfrom external sources should be assigned to specific major workfunctions."
Report There were no station, utility, or other personnel (including contractors) who received exposures greater than 100 mrem/year during 2004, based on TLD results.
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- 2. CHANGES. TESTS, AND EXPERIMENTS Requirement Federal Regulation 10 CFR 50.59 requires:
"(c)(1) A licensee may make changes in thefacility as described in the final safety analysis report (as updated), make changes in the proceduresas described in thefinal safety analysis report (as updated), and conduct tests or experiments not described in thefinal safety analysis report (as updated), without obtaininga license amendment pursuant to §50.90 only if:
(i) A change to the technicalspecifications incorporatedin the license is not required, and (ii) The change, test, or experiment does not meet any of the criteriain paragraph(c)(2) of this section.
"(d)(2) The licensee shall submit, as specified in §50.4, a report containinga briefdescription of any changes, tests, and experiments, including a summary of the evaluation of each. A report must be submitted at intervals not to exceed 24 months."
Report In accordance with 10 CFR 50.59 as cited above, the following provides a description of changes, tests, and experiments completed during 2003 and 2004 and a summary of the supporting evaluation.
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2004 10 CFR 50.59 Annual Report Evaluation Number 99-037 Subiect Spent Fuel Rack Removal per Defueled Plant Modification Request (DPMR)99-008, Detailed Construction Packages (DCPs) 1 and 2 Description of Change, Test, or Experiment DPMR 99-008, DCPs 1 and 2 removed all spent fuel racks from the Spent Fuel Pool (SFP) to support fuel loading activities for dry fuel storage at the Independent Spent Fuel Storage Installation (ISFSI) site and decommissioning of the SFP.
Summary of Evaluation The spent fuel racks were removed using the guidance of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," and included safety factors in accordance with ANSI B30.9-1971, "Slings." Rack removal did not reduce the Spent Fuel Pool water level below the Technical Specification limit. Fuel movements were not performed concurrently with the rack removal. Administrative controls were in place to ensure the safe completion of this activity. Therefore, these changes did not (1) increase the frequency of, or consequences from, an accident previously evaluated in the Defueled Safety Analysis Report (DSAR), (2) create a different accident scenario than those previously evaluated in the DSAR, (3) increase the likelihood of, or consequences from, a malfunction of an SSC important to safety previously evaluated in the DSAR, (4) create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the DSAR, (5) result in a design basis limit for a fission product barrier as described in the DSAR being exceeded or altered, or (6) result in a departure from a method of evaluation described in the DSAR used in establishing the design bases or in the safety analyses.
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2004 10 CFR 50.59 Annual Report Evaluation Number 2002-021 Subject DSAR Table 3.5-1, Revision 18, and Nuclear Plant Engineering Procedure (NPEP) 200-11, "Verification of Design," Revision 7 Description of Change. Test, or Experiment DSAR Table 3.5-1 was revised to clarify compliance with Regulatory Guide 1.64, "Quality Assurance Requirements for the Design of Nuclear Power Plants. Specifically, the DSAR revision modifies the restrictions for supervisors performing design verifications without prior general manager approval. NPEP 200-11 was revised to document this change.
Summary of Evaluation In certain cases, due to staffing levels and decommissioning activities, supervisory and management personnel are the most qualified individuals to perform reviews and independent verifications. This position remains consistent with NUREG-0800, "Standard Review Plan," as design changes will be specifically approved by the facility general manager and verified by Nuclear Oversight audit. Therefore, these changes do not (1) increase the frequency of, or consequences from, an accident previously evaluated in the DSAR, (2) create a different accident scenario than those previously evaluated in the DSAR, (3) increase the likelihood of, or consequences from, a malfunction of an SSC important to safety previously evaluated in the DSAR, (4) create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the DSAR, (5) result in a design basis limit for a fission product barrier as described in the DSAR being exceeded or altered, or (6) result in a departure from a method of evaluation described in the DSAR used in establishing the design bases or in the safety analyses.
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2004 10 CFR 50.59 Annual Report Evaluation Number 2003-001 Subiect License Change Application (LCA) 237, "Spent Fuel Cask Loading in the Fuel Building,"
Revision 6 Description of Change, Test, or Experiment LCA 237 was revised to change the sequence for filling a Multi-Purpose Canister (MPC) with borated water while suspended in the Cask Loading Pit.
Summary of Evaluation NOTE: Because of its effect on fuel loading activities for the Independent Spent Fuel Storage Installation (ISFSI), LCA 237 was also evaluated and reported separately in accordance 10 CFR 72.48. This summary addresses only changes associated with 10 CFR 50.59.
Filling the MPC with borated water as previously described in LCA 237 was intended to prevent the PWR Basket from floating out of the Transfer Cask during fuel loading activities. However, the Trojan ISFSI MPC design was changed to accommodate a heavier MPC that would not float due to its density relative to the density of water and because the annulus between the MPC and Transfer Cask was sealed at the top. The requirements for lifting heavy loads as discussed in the ISFSI Safety Analysis Report, LCA 237, and NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants," continued to be enforced. Since this change reduced the initial weight carried by the Transfer Cask lifting equipment, the change did not (1) increase the frequency of, or consequences from, an accident previously evaluated in the DSAR, (2) create a different accident scenario than those previously evaluated in the DSAR, (3) increase the likelihood of, or consequences from, a malfunction of an SSC important to safety previously evaluated in the DSAR, (4) create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the DSAR, (5) result in a design basis limit for a fission product barrier as described in the DSAR being exceeded or altered, or (6) result in a departure from a method of evaluation described in the DSAR used in establishing the design bases or in the safety analyses.
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2004 10 CFR 50.59 Annual Report Evaluation Number 2003-012 Subject Changes to PGE-1061, "DSAR and License Termination Plan (PGE-1078)," Revision 19; PGE-80 10, "Nuclear Quality Assurance (QA) Program;" PGE-1063, "Environmental Report Supplement;" and PGE-1021, "Offsite Dose Calculation Manual (ODCM)," Amendment 24, associated with DPMR 2003-01, DCP 4 Description of Change. Test, or Experiment DPMR 2003-01, DCP 4, deactivated PERM-9 along with its associated instrumentation and alarm/trips. A temporary Liquid Radioactive Waste System was also installed. This evaluation addressed the changes to licensing documents made to accommodate results of this work.
PGE-1061 was revised to reflect these changes and the current stage of decommissioning.
PGE-8010 and PGE-1063 were revised to clarify intent for using the methodology and parameters contained in the ODCM if monitoring instruments are used.
Changes were made to the ODCM to delete references to PERM-9 and it's associated instrumentation and alarmn/trips. Clarification was made to ensure that calculation methodology remain in the ODCM if setpoints are used in the future.
Summary of Evaluation The Liquid Radioactive Waste System no longer performs a safety function. Grab samples are used to ensure compliance with liquid radioactive waste effluent limits. Therefore, the changes do not (1) increase the frequency of, or consequences from, an accident previously evaluated in the DSAR, (2) create a different accident scenario than those previously evaluated in the DSAR, (3) increase the likelihood of, or consequences from, a malfunction of an SSC important to safety previously evaluated in the DSAR, (4) create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in the DSAR, (5) result in a design basis limit for a fission product barrier as described in the DSAR being exceeded or altered, or (6) result in a departure from a method of evaluation described in the DSAR used in establishing the design bases or in the safety analyses.
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