1CAN051901, License Amendment Request Application to Revise Technical Specifications to Adopt TSTF 563, Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program.
| ML19149A290 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/29/2019 |
| From: | Gaston R Entergy Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 1CAN051901 | |
| Download: ML19149A290 (19) | |
Text
Entergy Operations, Inc.
1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 1CAN051901 May 29, 2019 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555
SUBJECT:
License Amendment Request Application to Revise Technical Specifications to Adopt TSTF 563, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program" Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51 Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) is submitting a request for an amendment to the Technical Specifications (TSs) for Arkansas Nuclear One, Unit 1 (ANO-1).
Entergy requests adoption of TSTF-563, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program." TSTF-563 revises the TS definitions of Channel Calibration and Channel Functional Test. The definitions of Channel Calibration and Channel Functional Test currently permit performance by any series of sequential, overlapping, or total channel steps. Both definitions are revised to allow the required frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program.
The enclosure provides a description and assessment of the proposed changes. Attachment 1 of the enclosure provides the existing TS pages marked to show the proposed changes. of the enclosure provides revised (clean) TS pages.
No new regulatory commitments are included in this amendment request.
Approval of the proposed amendment is requested by June 1, 2019. Once approved, the amendment shall be implemented within 60 days.
In accordance with 10 CFR 50.91, Entergy is notifying the State of Arkansas of this amendment request by transmitting a copy of this letter and enclosure to the designated State Official.
1CAN051901 Page 2 of 2 If there are any questions or if additional information is needed, please contact Tim Arnold at 479-858-7826.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on May 29, 2019.
Sincerely, ORIGINAL SIGNED BY RON GASTON Ron Gaston RG/dbb
Enclosure:
Description and Assessment cc: NRC Region IV Regional Administrator NRC Senior Resident Inspector - Arkansas Nuclear One NRC Project Manager - Arkansas Nuclear One Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205
Enclosure to 1CAN051901 Description and Assessment
Enclosure to 1CAN051901 Page 1 of 4
1.0 DESCRIPTION
Entergy Operations, Inc. (Entergy) requests adoption of TSTF-563, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program." TSTF-563 revises the Technical Specification (TS) definitions of Channel Calibration and Channel Functional Test.
The definitions of Channel Calibration and Channel Functional Test currently permit performance by any series of sequential, overlapping, or total channel steps. Both definitions are revised to allow the required frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program (SFCP).
2.0 ASSESSMENT 2.1 Applicability of Safety Evaluation Entergy has reviewed the safety evaluation for TSTF-563 provided to the Technical Specifications Task Force in a letter dated December 4, 2018. This review included a review of the NRC staffs evaluation, as well as the information provided in TSTF-563. As described herein, Entergy has concluded that the justifications presented in TSTF-563 and the safety evaluation prepared by the NRC staff are applicable to Arkansas Nuclear One, Unit 1 (ANO-1) and justify this amendment for the incorporation of the changes to the ANO-1 TS.
A SFCP was incorporated into the ANO-1 TS in a license amendment dated May 22, 2019 (ML19098A955).
2.2 Optional Changes and Variations Entergy is not proposing any variations from the TS changes described in the TSTF-563 or the applicable parts of the NRC staffs safety evaluation dated December 4, 2018. Note, however, that the additional wording adopted for the two affected TS definitions required information to be moved from page to page in order to accommodate needed space. Therefore, several additional pages from the ANO-1 TS Definitions section are included in Attachments 1 and 2 of this enclosure. Information relocated to another TS page is not highlighted with revision bars where no change to the wording occurred. Entergy considers this movement of information to be administrative in nature and has no impact on the applicability of TSTF-563 to ANO-1.
The traveler and Safety Evaluation discuss the applicable regulatory requirements and guidance, including the 10 CFR 50, Appendix A, General Design Criteria (GDC). ANO-1 was not licensed to the 10 CFR 50, Appendix A, GDC. ANO-1 was originally designed to comply with the 70 "Proposed General Design Criteria for Nuclear Power Plant Construction Permits,"
published in July 1967. However, the ANO-1 Safety Analysis Report (SAR) provides a comparison with the Atomic Energy Commission (AEC) GDC published as Appendix A to 10 CFR 50 in 1971. The applicable AEC GDC were compared to 10 CFR 50, Appendix A, GDC as discussed below.
Enclosure to 1CAN051901 Page 2 of 4 TSTF-563 references 10 CFR 50, Appendix A, GDC 13, "Instrumentation and Control,"
which states:
Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.
Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.
GDC 13 is discussed in ANO-1 SAR Section 1.4.9, "Instrumentation and Control," which states, in part:
Adequate instrumentation and controls are provided to maintain operating variables within prescribed ranges for normal operation and monitor accident conditions as appropriate to assure adequate safety.
TSTF-563 references 10 CFR 50, Appendix A, GDC 21, "Protection System Reliability and Testability," which states:
The protection system shall be designed for high functional reliability and inservice testability commensurate with the safety functions to be performed. Redundancy and independence designed into the protection system shall be sufficient to assure that (1) no single failure results in loss of the protection function and (2) removal from service of any component or channel does not result in loss of the required minimum redundancy unless the acceptable reliability of operation of the protection system can be otherwise demonstrated. The protection system shall be designed to permit periodic testing of its functioning when the reactor is in operation, including a capability to test channels independently to determine failures and losses of redundancy that may have occurred.
GDC 21 is discussed in ANO-1 SAR Section 1.4.17, "Protection System Reliability and Testability," which states, in part:
The design of protection systems meets this criterion by specific instrument location, component redundancy, and in-service testing capability. The major design criteria as stated have been applied to the design of the instrumentation. In addition, the protection systems meet the single failure criterion of IEEE 279-1968.
Following implementation of the proposed change, ANO-1 will remain in compliance with AEC GDC as discussed in the SAR. Therefore, this difference does not alter the conclusion that the proposed change is applicable to ANO-1.
Enclosure to 1CAN051901 Page 3 of 4
3.0 REGULATORY ANALYSIS
3.1 No Significant Hazards Consideration Analysis Entergy Operations, Inc. (Entergy) requests adoption of TSTF-563, "Revise Instrument Testing Definitions to Incorporate the Surveillance Frequency Control Program." TSTF-563 revises the Technical Specification (TS) definitions of Channel Calibration and Channel Functional Test.
The definitions of Channel Calibration and Channel Functional Test currently permit performance by any series of sequential, overlapping, or total channel steps. Both definitions are revised to allow the required frequency for testing the components or devices in each step to be determined in accordance with the TS Surveillance Frequency Control Program (SFCP).
Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
- 1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed change revises the TS definitions of Channel Calibration and Channel Functional Test to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS SFCP. All components in the channel continue to be tested. The frequency at which a channel test is performed is not an initiator of any accident previously evaluated; therefore, the probability of an accident is not affected by the proposed change. The channels surveilled in accordance with the affected definitions continue to be required to be operable and the acceptance criteria of the surveillances are unchanged. As a result, any mitigating functions assumed in the accident analysis will continue to be performed.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
- 2. Does the proposed amendment create the possibility of a new or different kind of accident from any previously evaluated?
Response: No The proposed change revises the TS definitions of Channel Calibration and Channel Functional Test to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS SFCP. The design function or operation of the components involved are not affected and there is no physical alteration of the plant (i.e.,
no new or different type of equipment will be installed). No credible new failure mechanisms, malfunctions, or accident initiators not considered in the design and licensing bases are introduced. The changes do not alter assumptions made in the safety analysis.
The proposed changes are consistent with the safety analysis assumptions.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
Enclosure to 1CAN051901 Page 4 of 4
- 3. Does the proposed amendment involve a significant reduction in a margin of safety?
Response: No The proposed change revises the TS definitions of Channel Calibration and Channel Functional Test to allow the frequency for testing the components or devices in each step to be determined in accordance with the TS SFCP. The SFCP assures sufficient safety margins are maintained, and that the design, operation, surveillance methods, and acceptance criteria specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plants' licensing basis. The proposed change does not adversely affect existing plant safety margins, or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits, or limiting safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by method of determining surveillance test intervals under an NRC-approved licensee-controlled program.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Entergy concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
ATTACHMENTS
- 1. Proposed Technical Specification Changes (Mark-Up)
- 2. Revised Technical Specification Pages
Enclosure Attachment 1 to 1CAN051901 Proposed Technical Specification Page Markups (6 pages)
Definitions 1.1 1.1 Definition CHANNEL CALIBRATION The CHANNEL CALIBRATION may be performed by (continued) means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
CONTROL RODS CONTROL RODS shall be all full length safety and regulating rods that are used to shutdown the reactor and control power level during maneuvering operations.
CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS The COLR is the ANO-1 specific document that provides REPORT (COLR) cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
Moved DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of to I-131 (microcuries per gram) that alone would produce Next the same committed effective dose equivalent (CEDE) as Page the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The CEDE dose conversion factors used to determine the DOSE ANO-1 1.1-2 Amendment No. 215,243,
Definitions 1.1 EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion.
ANO-1 1.1-2 Amendment No. 215,243,
Definitions 1.1 1.1 Definition (continued)
DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of Moved from I-131 (microcuries per gram) that alone would produce Previous the same committed effective dose equivalent (CEDE) as Page the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The CEDE dose conversion factors used to determine the DOSE EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion.
DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."
INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).
LEAKAGE LEAKAGE shall be:
- a. Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
- 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
ANO-1 1.1-3 Amendment No. 215,224,243,257,
Definitions 1.1
- b. Unidentified LEAKAGE Moved to All LEAKAGE (except RCP seal water injection and Next leakoff) that is not identified LEAKAGE; Page
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
ANO-1 1.1-3 Amendment No. 215,224,243,257,
Definitions 1.1 1.1 Definition (continued)
LEAKAGE (continued) b. Unidentified LEAKAGE Moved from All LEAKAGE (except RCP seal water injection and Previous leakoff) that is not identified LEAKAGE; Page
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
OPERABLE-OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.
These tests are:
- a. Described in the SAR;
- b. Authorized under the provisions of 10 CFR 50.59; or
- c. Otherwise approved by the Nuclear Regulatory Commission.
QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.
Power in any Core Quadrant QPT 100 1 Average Power in all Quadrants Moved to RATED THERMAL POWER RTP shall be a total steady state reactor core heat Next (RTP) transfer rate to the reactor coolant of 2568 MWt.
Page ANO-1 1.1-4 Amendment No. 215,243,
Definitions 1.1 1.1 Definition (continued)
Moved from RATED THERMAL POWER RTP shall be a total steady state reactor core heat Previous (RTP) transfer rate to the reactor coolant of 2568 MWt.
Page SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. All full length CONTROL RODS (safety and regulating) are fully inserted except for the single CONTROL ROD of highest reactivity worth, which is assumed to be fully withdrawn. With any CONTROL ROD not capable of being fully inserted, the reactivity worth of these CONTROL RODS must be accounted for in the determination of SDM;
- b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level; and
- c. There is no change in APSR position.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
ANO-1 1.1-5 Amendment No. 215,218,264,
Enclosure Attachment 2 to 1CAN051901 Revised Technical Specification Pages (4 pages)
Definitions 1.1 1.1 Definition CHANNEL CALIBRATION The CHANNEL CALIBRATION may be performed by (continued) means of any series of sequential, overlapping, or total channel steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total steps, and each step must be performed within the Frequency in the Surveillance Frequency Control Program for the devices included in the step.
CONTROL RODS CONTROL RODS shall be all full length safety and regulating rods that are used to shutdown the reactor and control power level during maneuvering operations.
CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS The COLR is the ANO-1 specific document that provides REPORT (COLR) cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
ANO-1 1.1-2 Amendment No. 215,243,
Definitions 1.1 1.1 Definition (continued)
DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same committed effective dose equivalent (CEDE) as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The CEDE dose conversion factors used to determine the DOSE EQUIVALENT I-131 shall be performed using Table 2.1 of EPA Federal Guidance Report No. 11, 1988, Limiting Values of Radionuclide Intake and Air Concentration and Dose conversion Factors for Inhalation, Submersion, and Ingestion.
DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil."
INSERVICE TESTING PROGRAM The INSERVICE TESTING PROGRAM is the licensee program that fulfills the requirements of 10 CFR 50.55a(f).
LEAKAGE LEAKAGE shall be:
- a. Identified LEAKAGE
- 1. LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
- 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
- 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
ANO-1 1.1-3 Amendment No. 215,224,243,257,
Definitions 1.1 1.1 Definition (continued)
LEAKAGE (continued) b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection and leakoff) that is not identified LEAKAGE;
- c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
OPERABLE-OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.
These tests are:
- a. Described in the SAR;
- b. Authorized under the provisions of 10 CFR 50.59; or
- c. Otherwise approved by the Nuclear Regulatory Commission.
QUADRANT POWER TILT QPT shall be defined by the following equation and (QPT) is expressed as a percentage.
Power in any Core Quadrant QPT 100 1 Average Power in all Quadrants ANO-1 1.1-4 Amendment No. 215,243,
Definitions 1.1 1.1 Definition (continued)
RATED THERMAL POWER RTP shall be a total steady state reactor core heat (RTP) transfer rate to the reactor coolant of 2568 MWt.
SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
- a. All full length CONTROL RODS (safety and regulating) are fully inserted except for the single CONTROL ROD of highest reactivity worth, which is assumed to be fully withdrawn. With any CONTROL ROD not capable of being fully inserted, the reactivity worth of these CONTROL RODS must be accounted for in the determination of SDM;
- b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level; and
- c. There is no change in APSR position.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
ANO-1 1.1-5 Amendment No. 215,218,264,