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Category:Technical Specification
MONTHYEARML24103A2042024-04-12012 April 2024 Constellation Energy Generation, LLC, Application to Revise Technical Specifications to Adopt TSTF-591-A, Revise Risk Informed Completion Time (RICT) Program Revision 0 and Revise 10 CFR 50.69 License Condition ML22298A0112022-10-25025 October 2022 License Amendment Request to Adopt Technical Specification Task Force TSTF-295-A, Modify Note 2 to Actions of (Post-Accident Monitoring) PAM Table To. RS-22-086, R. E. Ginna Nuclear Power Plant - Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam.2022-08-10010 August 2022 R. E. Ginna Nuclear Power Plant - Application to Revise Technical Specifications to Adopt TSTF-577, Revised Frequencies for Steam. RS-22-008, Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report to Add References to NRC-Approved Topical Report with Admin2022-01-24024 January 2022 Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report to Add References to NRC-Approved Topical Report with Admin ML21278A2972021-09-0707 September 2021 6 to Technical Specification Bases ML21165A4062021-06-14014 June 2021 Spent Fuel Pool Cooling - Shutdown Cooling Systems Licensing Design Basis License Amendment Request ML19346E5362019-12-11011 December 2019 License Amendment Request: 10 CFR 50.90 Proposed Changes to Technical Specification (TS) 3.8.1 Emergency Diesel Generator Surveillance Requirements for Frequency and Voltage Tolerances ML19325C1282019-11-21021 November 2019 License Amendment Request - Revision to Technical Specification 5.5.7, Reactor Coolant Pump Flywheel Inspection Program RS-19-039, Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls2019-06-26026 June 2019 Fleet License Amendment Request - Common Language for Technical Specification High Radiation Area Administrative Controls ML19127A0762019-05-0606 May 2019 Application to Revise Technical Specifications to Adopt TSTF-490, Deletion of E-Bar Definition and Revision to RCS Specific Activity Tech Spec ML18239A2602018-08-27027 August 2018 Supplement to Response to Request for Additional Information License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 1, Provide Risk .. ML18235A1992018-08-23023 August 2018 License Amendment Request: 10 CFR 50.90 Proposed Changes to Technical Specification (TS) 3.8.1 Actions A.3 and D.3 to Extend the Offsite Circuit Inoperable Completion Times from 72 Hours to ... ML18172A1452018-06-21021 June 2018 Response to Request for Additional Information License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 1, Provide Risk-Informed ... ML18262A0962018-03-0606 March 2018 Technical Specification Bases, Revisions 62 Through 63 ML17261A2342017-07-19019 July 2017 Bases 3.7, Plant Systems, Revisions; B 3.7.1-1 Through B 3.7.18-3 ML17261A2252017-07-19019 July 2017 Bases 3.2, Power Distribution Limits, Revisions; B 3.2.1-1 Through B 3.2.5-5 ML17261A2352017-07-19019 July 2017 Bases 3.8, Electric Power Systems, Revisions; B 3.8.1-1 Through B 3.8.10-5 ML17261A2272017-07-19019 July 2017 Bases 3.3, Instrumentation, Revisions; B 3.3.1-18 Through B 3.3.12-4 ML17261A2212017-07-19019 July 2017 Bases 2.0, Safety Limits, Revision 2; B 2.1.1-1 Through B 2.1.2-4 ML17261A2192017-07-19019 July 2017 Technical Specification Bases ML17261A2312017-07-19019 July 2017 Bases 3.5, Emergency Core Cooling System (Eccs), Revisions; B 3.5.1-1 Through B 3.5.5-5 ML17261A2232017-07-19019 July 2017 Bases 3.1, Reactivity Control Systems, Revisions; B 3.1.1-1 Through B 3.1.8-5 ML17261A2322017-07-19019 July 2017 Bases 3.0, Containment System, Revisions; B 3.0.1-1 Through B 3.0.8-4 ML17261A2292017-07-19019 July 2017 Bases 3.4, Reactor Coolant System(Rcs), Revisions; B 3.4.1-1 Through B 3.4318-8 ML17261A2382017-07-19019 July 2017 Bases 3.9, Refueling Operations, Revisions; B 3.9.1-1 Through B 3.9.6-2 ML17261A2222017-07-19019 July 2017 Basis 3.0, Limiting Condition for Operation (LCO) Applicability, Revisions; B 3.0-1 Through B 3.0-25 ML16266A0862016-09-22022 September 2016 License Amendment Request - Control Room Emergency Ventilation System ML16258A0762016-09-0808 September 2016 Technical Specification Bases B 3.2, Power Distribution Limits - Linear Heat Rate (Lhr) ML16258A0732016-09-0808 September 2016 Technical Specification Bases B 2.1.1-1, Safety Limits ML16258A0772016-09-0808 September 2016 Technical Specification Bases B 3.3.1, Reactor Protective System (RPS) Instrumentation-Operating ML16258A0722016-09-0808 September 2016 Inc. - Technical Specification Bases, Unit Nos. 1 and 2 ML16258A0742016-09-0808 September 2016 Technical Specification Bases B 3.0-1, Limiting Condition for Operation (LCO) Applicability ML16258A0842016-09-0808 September 2016 Technical Specification Bases B 3.9.1, Refueling Operations, Boron Concentration ML16258A0802016-09-0808 September 2016 Technical Specification Bases B 3.5.1, Emergency Core Cooling System (Eccs), Safety Injection Tanks (Sits) ML16258A0812016-09-0808 September 2016 Technical Specification Bases B 3.6, Containment Systems ML16258A0752016-09-0808 September 2016 Technical Specification Bases B 3.1, Reactivity Control Systems ML16258A0832016-09-0808 September 2016 Technical Specification Bases B 3.8.1, Electrical Power Systems, AC Sources-Operating ML16258A0822016-09-0808 September 2016 Technical Specification Bases B 3.7.1, Plant Systems, Main Steam Safety Valves (Mssvs) ML16258A0782016-09-0808 September 2016 Technical Specification Bases B 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ML16209A2182016-07-26026 July 2016 Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing. ML15257A2002015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 8 of 12 ML15257A2032015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 11 of 12 ML15257A1992015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 7 of 12 ML15257A1982015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 6 of 12 ML15257A1972015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 5 of 12 ML15257A1962015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 4 of 12 ML15257A1952015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 3 of 12 ML15257A1942015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 2 of 12 ML15257A1932015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 1 of 12 ML15257A2022015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 10 of 12 2024-04-12
[Table view] Category:Bases Change
MONTHYEARRS-22-008, Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report to Add References to NRC-Approved Topical Report with Admin2022-01-24024 January 2022 Application to Revise Technical Specifications 5.6.5, Core Operating Limits Report to Add References to NRC-Approved Topical Report with Admin ML21278A2972021-09-0707 September 2021 6 to Technical Specification Bases ML21165A4062021-06-14014 June 2021 Spent Fuel Pool Cooling - Shutdown Cooling Systems Licensing Design Basis License Amendment Request ML19346E5362019-12-11011 December 2019 License Amendment Request: 10 CFR 50.90 Proposed Changes to Technical Specification (TS) 3.8.1 Emergency Diesel Generator Surveillance Requirements for Frequency and Voltage Tolerances ML18262A0962018-03-0606 March 2018 Technical Specification Bases, Revisions 62 Through 63 ML17261A2342017-07-19019 July 2017 Bases 3.7, Plant Systems, Revisions; B 3.7.1-1 Through B 3.7.18-3 ML17261A2222017-07-19019 July 2017 Basis 3.0, Limiting Condition for Operation (LCO) Applicability, Revisions; B 3.0-1 Through B 3.0-25 ML17261A2382017-07-19019 July 2017 Bases 3.9, Refueling Operations, Revisions; B 3.9.1-1 Through B 3.9.6-2 ML17261A2292017-07-19019 July 2017 Bases 3.4, Reactor Coolant System(Rcs), Revisions; B 3.4.1-1 Through B 3.4318-8 ML17261A2232017-07-19019 July 2017 Bases 3.1, Reactivity Control Systems, Revisions; B 3.1.1-1 Through B 3.1.8-5 ML17261A2252017-07-19019 July 2017 Bases 3.2, Power Distribution Limits, Revisions; B 3.2.1-1 Through B 3.2.5-5 ML17261A2352017-07-19019 July 2017 Bases 3.8, Electric Power Systems, Revisions; B 3.8.1-1 Through B 3.8.10-5 ML17261A2272017-07-19019 July 2017 Bases 3.3, Instrumentation, Revisions; B 3.3.1-18 Through B 3.3.12-4 ML17261A2212017-07-19019 July 2017 Bases 2.0, Safety Limits, Revision 2; B 2.1.1-1 Through B 2.1.2-4 ML17261A2192017-07-19019 July 2017 Technical Specification Bases ML17261A2322017-07-19019 July 2017 Bases 3.0, Containment System, Revisions; B 3.0.1-1 Through B 3.0.8-4 ML17261A2312017-07-19019 July 2017 Bases 3.5, Emergency Core Cooling System (Eccs), Revisions; B 3.5.1-1 Through B 3.5.5-5 ML16258A0742016-09-0808 September 2016 Technical Specification Bases B 3.0-1, Limiting Condition for Operation (LCO) Applicability ML16258A0732016-09-0808 September 2016 Technical Specification Bases B 2.1.1-1, Safety Limits ML16258A0722016-09-0808 September 2016 Inc. - Technical Specification Bases, Unit Nos. 1 and 2 ML16258A0772016-09-0808 September 2016 Technical Specification Bases B 3.3.1, Reactor Protective System (RPS) Instrumentation-Operating ML16258A0782016-09-0808 September 2016 Technical Specification Bases B 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits ML16258A0832016-09-0808 September 2016 Technical Specification Bases B 3.8.1, Electrical Power Systems, AC Sources-Operating ML16258A0762016-09-0808 September 2016 Technical Specification Bases B 3.2, Power Distribution Limits - Linear Heat Rate (Lhr) ML16258A0842016-09-0808 September 2016 Technical Specification Bases B 3.9.1, Refueling Operations, Boron Concentration ML16258A0802016-09-0808 September 2016 Technical Specification Bases B 3.5.1, Emergency Core Cooling System (Eccs), Safety Injection Tanks (Sits) ML16258A0812016-09-0808 September 2016 Technical Specification Bases B 3.6, Containment Systems ML16258A0822016-09-0808 September 2016 Technical Specification Bases B 3.7.1, Plant Systems, Main Steam Safety Valves (Mssvs) ML16258A0752016-09-0808 September 2016 Technical Specification Bases B 3.1, Reactivity Control Systems ML16209A2182016-07-26026 July 2016 Application to Revise Technical Specifications to Adopt TSTF-545, Revision 3, TS Lnservice Testing Program Removal & Clarify SR Usage Rule Application to Section 5.5 Testing. ML15257A2012015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 9 of 12 ML15257A2042015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 12 of 12 ML15257A2032015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 11 of 12 ML15257A1932015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 1 of 12 ML15257A1942015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 2 of 12 ML15257A1962015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 4 of 12 ML15257A2022015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 10 of 12 ML15257A2002015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 8 of 12 ML15257A1992015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 7 of 12 ML15257A1982015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 6 of 12 ML15257A1972015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 5 of 12 ML15257A1952015-09-0404 September 2015 Technical Specification Bases, Revisions 49 Through 55. Part 3 of 12 ML14267A2322014-09-19019 September 2014 B 3.7.1-1, Mssvs, Plant Systems Through B 3.7.17-2, SFP Storage ML14267A2312014-09-19019 September 2014 B 3.6.1-1, Containment Systems Though B 3.6.8-4, Irs ML14267A2302014-09-19019 September 2014 B 3.5.1-1, Sits, Emergency Core Cooling System Through B 3.5.5-5, Stb ML14267A2292014-09-19019 September 2014 B 3.4.1-1, RCS Pressure, Temperature and Flow DNB Limits Through B 3.4.18-8, SG Tube Integrity ML14267A2282014-09-19019 September 2014 B 3.3.1-1, Rs Instrumentation-Operating Through B 3.3.12-5, Wide Range Logarithmic Neutron Flux Monitor Channels ML14267A2272014-09-19019 September 2014 B 3.2.1-1, Power Distribution Limits Through B 3.2.5-0, Asi ML14267A2262014-09-19019 September 2014 B 3.1.1-1, Reactivity Control Systems Through B 3.1.8-5, STE-Modes 1 and 2 ML14267A2252014-09-19019 September 2014 B 3.0-1, Limiting Condition for Operation (LCO) Applicability Through B 3.0-26, SR Applicability, Bases 2022-01-24
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Reactor Core SLs B 2.1.1 B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs BASES BACKGROUND Reference 1, Appendix 1C, Criterion 6 requires, and Safety Limits (SLs) ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs). This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95% probability at a 95% confidence level (95/95 DNB criterion) that DNB will not occur, and by requiring that fuel centerline temperature stays below the melting temperature.
The restrictions of this SL prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate below the level at which fuel centerline melting occurs. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Fuel centerline melting occurs when the local linear heat rate or power peaking in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel. Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.
Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB and the resultant sharp reduction in heat transfer coefficient. Inside the steam film, high cladding temperatures are reached and a cladding-water (zirconium-water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, CALVERT CLIFFS - UNITS 1 & 2 B 2.1.1-1 Revision 2
Reactor Core SLs B 2.1.1 BASES resulting in an uncontrolled release of activity to the reactor coolant.
The Reactor Protective System (RPS), in combination with the Limiting Conditions for Operation (LCOs), is designed to prevent any anticipated combination of transient conditions for Reactor Coolant System (RCS) temperature, pressure, and THERMAL POWER level that would result in a violation of the reactor core SLs.
APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES normal operation and AOOs. The reactor core SLs are established to preclude violation of the following fuel design criteria:
- a. There must be at least 95% probability at a 95%
confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB; and
- b. The hot fuel pellet in the core must not experience fuel centerline melting.
The RPS setpoints, LCO 3.3.1, in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for RCS temperature, pressure, and THERMAL POWER level that would result in a departure from nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.
Automatic enforcement of these reactor core SLs is provided by the following functions:
- a. Pressurizer Pressure-High trip;
- b. Power Level-High trip;
- c. Rate of Change of Power-High trip;
- d. Reactor Coolant Flow-Low trip;
- e. Steam Generator Pressure-Low trip;
- f. Steam Generator Level-Low trip;
- g. Axial Power Distribution-High trip;
- h. Thermal Margin/Low Pressure trip; CALVERT CLIFFS - UNITS 1 & 2 B 2.1.1-2 Revision 2
Reactor Core SLs B 2.1.1 BASES
- i. Steam Generator Pressure Difference trip; and
- j. Steam Generator Safety Valves.
The SL represents a design requirement for establishing the RPS trip setpoints identified previously. Limiting Condition for Operation (LCO) 3.2.1, or the assumed initial conditions of the safety analyses (as indicated in Reference 1, Section 14.1), provide more restrictive limits to ensure that the SLs are not exceeded.
SAFETY LIMITS The curves provided in Figure 2.1.1-1 show the loci of points of THERMAL POWER, pressurizer pressure, and highest operating loop cold leg temperature for which the minimum DNBR is not less than the safety analysis limit. Safety Limit 2.1.1.2 ensures that fuel centerline temperature remains below melting.
APPLICABILITY Safety Limit 2.1.1 only applies in MODEs 1 and 2 because these are the only MODEs in which the reactor is critical.
Automatic protection functions are required to be OPERABLE during MODEs 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1.
In MODEs 3, 4, 5, and 6, Applicability is not required, since the reactor is not generating significant THERMAL POWER.
SAFETY LIMIT The following SL violation responses are applicable to the VIOLATIONS reactor core SLs.
2.2.1 If SL 2.1.1 is violated, the requirement to go to MODE 3 places the unit in a MODE in which this SL is not applicable.
CALVERT CLIFFS - UNITS 1 & 2 B 2.1.1-3 Revision 2
Reactor Core SLs B 2.1.1 BASES The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable and reduces the probability of fuel damage.
REFERENCES 1. UFSAR CALVERT CLIFFS - UNITS 1 & 2 B 2.1.1-4 Revision 2
RCS Pressure SL B 2.1.2 B 2.0 SAFETY LIMITS (SLs)
B 2.1.2 Reactor Coolant System (RCS) Pressure SL BASES BACKGROUND The SL on RCS pressure protects the integrity of the RCS against overpressurization. In the event of fuel cladding failure, fission products are released into the reactor coolant. The RCS then serves as the primary barrier in preventing the release of fission products into the atmosphere. By establishing an upper limit on RCS pressure, continued RCS integrity is ensured. According to Reference 1, Appendix 1C, Criteria 9 and 33, the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal operation and AOOs. Also, according to Reference 1, Appendix 1C, Criterion 32, reactivity accidents do not result in rupturing the RCPB.
The design pressure of the RCS is 2500 psia. During normal operation and AOOs, the RCS pressure is kept from exceeding the design pressure by more than 10%, in accordance with Reference 2,Section III, Article NB-7000. To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, according to the American Society of Mechanical Engineers (ASME) Code requirements, prior to initial operation, when there is no fuel in the core.
Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of Reference 2,Section XI, Article IWX-5000.
Overpressurization of the RCS could result in a breach of the RCPB. If this occurs in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in Reference 1, Chapter 14.
APPLICABLE The RCS pressurizer safety valves, the main steam safety SAFETY ANALYSES valves, and the Reactor Pressure-High trip have settings established to ensure that the RCS pressure SL will not be exceeded.
The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, in accordance with Reference 2,Section III, CALVERT CLIFFS - UNITS 1 & 2 B 2.1.2-1 Revision 2
RCS Pressure SL B 2.1.2 BASES Article NB-7000. The transient that establishes the required relief capacity, and hence the valve size requirements and lift settings, is a complete loss of external load without a direct reactor trip. During the transient, no control actions are assumed except that the safety valves on the secondary plant are assumed to open when the steam pressure reaches the secondary plant safety valve settings, and nominal feedwater supply is maintained.
The Reactor Protective System trip setpoints (LCO 3.3.1),
together with the settings of the main steam safety valves (LCO 3.7.1) and the pressurizer safety valves, provide pressure protection for normal operation and AOOs. In particular, the Pressurizer Pressure-High trip setpoint is specifically set to provide protection against overpressurization (Reference 1, Section 14.1). Safety analyses for both the Pressure-High trip and the RCS pressurizer safety valves are performed, using conservative assumptions relative to pressure control devices.
More specifically, no credit is taken for operation of the following:
- a. Pressurizer power-operated relief valves;
- b. Steam Bypass Control System;
- c. Pressurizer Level Control System; or
- d. Pressurizer Pressure Control System.
SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel in Reference 2,Section III, Article NB-7000 is 110%
of design pressure. The maximum transient pressure allowable in the RCS piping, valves, and fittings under Reference 3, is 110% of design pressure. The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is established at 2750 psia.
APPLICABILITY Safety Limit 2.1.2 applies in MODEs 1, 2, 3, 4, and 5 because this SL could be approached or exceeded in these MODEs due to overpressurization events. The SL is not applicable in MODE 6 because the reactor vessel head is CALVERT CLIFFS - UNITS 1 & 2 B 2.1.2-2 Revision 2
RCS Pressure SL B 2.1.2 BASES unbolted, making it unlikely that the RCS can be pressurized.
SAFETY LIMIT The following SL violation responses are applicable to the VIOLATIONS RCS pressure SL.
2.2.2.1 If the RCS pressure SL is violated when the reactor is in MODE 1 or 2, the requirement is to restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
With RCS pressure greater than the value specified in SL 2.1.2 in MODE 1 or 2, the pressure must be reduced below this value. A pressure greater than the value specified in SL 2.1.2 exceeds 110% of the RCS design pressure and may challenge system integrity.
The allowed Completion Time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provides the operator time to complete the necessary actions to reduce RCS pressure by terminating the cause of the pressure increase, removing mass or energy from the RCS, or a combination of these actions, and to establish MODE 3 conditions.
2.2.2.2 If the RCS pressure SL is exceeded in MODE 3, 4, or 5, RCS pressure must be restored to within the SL value within 5 minutes.
Exceeding the RCS pressure SL in MODE 3, 4, or 5 is potentially more severe than exceeding this SL in MODE 1 or 2, since the reactor vessel temperature may be lower and the vessel material, consequently, less ductile. As such, pressure must be reduced to less than the SL within 5 minutes. This action does not require reducing MODEs, since this would require reducing temperature, which would compound the problem by adding thermal gradient stresses to the existing pressure stress.
CALVERT CLIFFS - UNITS 1 & 2 B 2.1.2-3 Revision 2
RCS Pressure SL B 2.1.2 BASES REFERENCES 1. UFSAR
- 2. ASME, Boiler and Pressure Vessel Code
- 3. ASME, USAS B31.7, Standard Code for Pressure Piping, 1967 CALVERT CLIFFS - UNITS 1 & 2 B 2.1.2-4 Revision 2