ML023180500
| ML023180500 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 11/07/2002 |
| From: | Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML023180500 (38) | |
Text
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07, 2002 Page 1
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TSB APPROVED AMENDMENT TO THE UNIT 2 TECHNICAL SPECIFICATIONS BASES MANUAL REVISION 33 Replace the following pages of the Technical Specifications Bases Manual with the enclosed pages. The revised pages are identified by Revision Number and contain vertical lines indicating the area of change.
REMOVE PAGES REV. #
INSERT PAGES REV. #
TS / LOES -1 through 4 32 TS / B LOES -1 through 4 33 TS B 3.3-104 1
TS I B 3.3-104 2
TS I B 3.3-109 1
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TS /B 3.3-117 1
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SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS-BASES)
Section Title Revision B 2.0 SAFETY LIMITS BASES Page B 2.0-1 1
Pages TS / B 2.0-2 through TS I B 2.0-4 2
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B 3.0 LCO AND SR APPLICABILITY BASES Pages B 3.0-1 through B 3.0-15 0
B 3.1 REACTIVITY CONTROL BASES Pages B 3.1-1 through B 3.1-51 0
B 3.2 POWER DISTRIBUTION LIMITS BASES Pages B 3.2-1 through B 3.2-4 1
Pages TS / B 3.2-5 and TS B 3.2-6 2
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B 3.3 INSTRUMENTATION Pages TS / B 3.3-1 throaigohTSi B 3.3-10 1
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SUSQUEHANNA-UNIT2 TS / B LOES-1I Revision 33
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS-BASES)
Section Title Revision Pages TS / B 3.3-116 and TS / B 3.3-117 2
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B 3.4 REACTOR COOLANT SYSTEM BASES Pages B 3.4-1 through B 3.4-9 1
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B 3.5 ECCS AND RCIC BASES Pages TS / B 3.5-1 and TS / B 3.5-2 1
Page TS / B 3.5-3 r2 Pages TS / B 3.5-4 through TS / B 3.5-10 1
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SUSQUEHANNA - UNIT 2 TS / B LOES-2 Revision 33
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS-BASES)
-Section Title Revision Page TS / B 3.18 1
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B 3.6 CONTAINMENT SYSTEMS BASES Page TS / B 3.6-1 2
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SUSQUEHANNA - UNIT 2 TSIB LOES-3 Revision 33 SUSQUEHANNA - UNIT 2 TS / B LOES-3 Revision 33
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS-BASES)
Section Titte Revision B 3.7 PLANT SYSTEMS BASES Pages TS / B 3.7-1 through TS / B 3.7-6 1
Page TS / B 3.7-6a 1
Pages B 3.7-7 through B 3.7-11 0
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B 3.8 ELECTRICAL POWER SYSTEMS BASES Pages B 3.8-1 through B 3.8-4 0
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B 3.9 REFUELING OPERATIONS BASES Pages TS / B 3.9-1 and TS / B 3.9-2 1
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Pages TS / B 3.9-3 and TS / B 3.9-4 1
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B 3.10 SPECIAL OPERATIONS BASES Page TS / B 3.10-1 1
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SUSQUEHANNA - UNIT 2 TS / B LOES-4 Revision 33
ECCS Instrumentation B 3.3.5.1 BASES BACKGROUND High Pressure Coolant Iniection System (continued)
The HPCI System may be initiated by either automatic or manual means.
Automatic initiation occurs for conditions of Reactor Vessel Water Level--Low Low, Level 2 or Drywell Pressure-High. Each of these variables is monitored by four redundant instruments. The instrument outputs are connected to relays whose contacts are arranged in a one-out-of-two taken twice logic for each Function.
The HPCI System also monitors the water level in the condensate storage tank (CST). HPCI suction is normally maintained on the CST until it transfers to the suppression pool on low CST level or is manually transferred by the operator. Reactor grade water in the CST is the normal source. Upon receipt of a HPCI initiation signal, the CST suction valve is automatically signaled to open (it is normally in the open position) unless the suppression pool suction valve is open. If the water level in the CST falls below a preselected level, first the suppression pool suction valve automatically opens, and then the CST suction valve automatically closes. Two level switches are used to detect low water level in the CST.
Either switch can cause the suppression pool suction valve to open and the CST suction valve to close. To prevent losing suction to the pump, the suction valves are interlocked so that one suction path must be open before the other automatically closes.
The HPCI provides makeup water to the reactor until the reactor vessel water level reaches the Reactor Vessel Water Level--High, Level 8 trip, at which time the HPCI turbine trips, which causes the turbine's stop valve, minimum flow valve, the cooling water isolation valve, and the injection valve to close. The logic is two-out-of-two to provide high reliability of the HPCI System. The HPCI System automatically restarts if a Reactor Vessel Water Level--Low Low, Level 2 signal is subsequently received.
(continued)
SUSQUEHANNA-UNIT 2 TS / B 3.3-104 Revision 2
ECCS Instrumentation B 3.3.5.1
,BASES APPLICABLE provide adequate protection because instrumentation uncertainties, SAFETY ANALYSES process effects, calibration tolerances, instrument drift, and severe LCO, and environment errors (for channels that must function in harsh APPLICABILITY environments as defined by 10 CFR 50.49) are accounted for.
(continued)
An exception to the methodology described to derive the Allowable Value is the methodology used to determine the Allowable Values for the ECCS pump start time delays and HPCI CST Level 1 - Low. These Allowable Values are based on system calculations and/or engineering judgement which establishes a conservative limit at which the function should occur.
In general, the individual Functions are required to be OPERABLE in the MODES or other specified conditions that may require ECCS (or DG) initiation to mitigate the consequences of a design basis transient or accident. To ensure reliable ECCS and DG function, a combination of Functions is required to provide primary and secondary initiation signals.
The specific Applicable Safety Analyses, LCO, and Applicability discussions are listed below on a Function by Function basis.
Core Spray and Low Pressure Coolant Iniection Systems l.a. 2.a.
Reactor Vessel Water Level--Low Low Low, Level 1 Low reactor pressure vessel (RPV) water level indicates that the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. The low pressure ECCS and associated DGs are initiated at Level 1 to ensure that core spray and flooding functions are available to prevent or minimize fuel damage. The Reactor Vessel Water Level-Low Low Low, Level 1 is one of the Functions assumed to be OPERABLE and capable of initiating the ECCS during the transients analyzed in References 2. In addition, the Reactor Vessel Water Level-Low Low Low, Level 1 Function is directly assumed in the analysis of the recirculation line break (Ref. 1). The core cooling function of the ECCS, along with the scram action of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.3-109 Revision 2
ECCS Instrumentation B 3.3.5.1 BASES 3.e. Manual Initiation The Manual Initiation push button channel introduces signals into the HPCI logic to provide manual initiation capability and is redundant to the automatic protective instrumentation. There is one push button for the HPCI System.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.3-117 Revision 2
ECCS Instrumentation B 3.3.5.1 BASES ACTIONS C.1 and C.2 (continued)
Function 3.c (which also requires entry into this Condition if a channel in this Function is inoperable), since the loss of one channel results in a loss of the Function (two-out-of-two logic). This loss was considered during the development of Reference 3 and considered acceptable for the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowed by Required Action C.2.
The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. This Completion Time also allows for an exception to the normal "time zero" for beginning the allowed outage time "clock." For Required Action C.1, the Completion Time only begins upon discovery that the same feature in both subsystems (e.g., both CS subsystems) cannot be automatically initiated due to inoperable channels within the same Function as described in the paragraph above. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time from discovery of loss of initiation capability is acceptable because it minimizes risk while allowing time for restoration of channels.
Because of the diversity of sensors available to provide initiation signals and the redundancy of the ECCS design, an allowable out of service time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> has been shown to be acceptable (Ref. 3) to permit "restoration of any inoperable channel to OPERABLE status. If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, Condition G must be entered and its Required Action taken. The Required Actions do not allow placing the channel in trip since this action would either cause the initiation or it would not necessarily result in a safe state for the channel in all events.
D.1, D.2.1, and D.2.2 Required Action D.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in a complete loss of automatic component initiation capability for the HPCI System. Automatic component initiation capability is lost if two Function 3.d channels are inoperable and untripped. In this situation (loss of automatic suction swap), the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> allowance of Required Actions D.2.1 and D.2.2 is not appropriate and the HPCI
__)*
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.3-127 Revision 2
ECCS--Operating B 3.5.1 BASES BACKGROUND Full flow test lines are provided for each LPCI subsystem to route water from (continued) the suppression pool, to allow testing of the LPCI pumps without injecting water into the RPV. These test lines also provide suppression pool cooling capability, as described in LCO 3.6.2.3, "RHR Suppression Pool Cooling."
The HPCI System (Ref. 3) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping for the system is provided from the CST and the suppression pool. Pump suction for HPCI is normally aligned to the CST source to minimize injection of suppression pool water into the RPV.
Whenever the CST water supply is low, an automatic transfer to the suppression pool water source ensures an adequate suction head for the pump and an uninterrupted water supply for continuous operation of the HPCI System. The steam supply to the HPCI turbine is piped from a main steam line upstream of the associated inboard main steam isolation valve.
The HPCI System is designed to provide core cooling for a wide range of reactor pressures (165 psia to 1225 psia). Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open and the turbine accelerates to a specified speed. As the HPCI flow increases, the turbine control valve is automatically adjusted to maintain design flow.
Exhaust steam from the HPCI turbine is discharged to the suppression pool.
A full flow test line is provided to route water to the CST to allow testing of the HPCI System during normal operation without injecting water into the RPV.
The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in these lines automatically open to prevent pump damage due to overheating when other discharge line valyes are closed. To ensure rapid delivery of water to the RPV and to minimize water hammer effects, all ECCS pump discharge lines are filled with water. The HPCI, LPCI and CS System discharge lines are kept full of water using a "keep fill" system that is supplied using the condensate transfer system.
"(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-3 Revision 2
TSB APPROVED AMENDMENT TO THE UNIT 2 TECHNICAL SPECIFICATIONS BASES MANUAL REVISION 34 Replace the following pages of the Technical Specifications Bases Manual with the enclosed pages. The revised pages are identified by Revision Number and contain vertical lines indicating the area of change.
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Section Titre Revision B 2.0 SAFETY LIMITS BASES Page TS / B 2.0-1 1
Pages TS / B 2.0-2 through TS I B 2.0-4 2
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B 3.0 LCO AND SR APPLICABILITY BASES Pages B 3.0-1 through B 3.0-15 0
B 3.1 REACTIVITY CONTROL BASES Pages B 3.1-1 through B 3.1-51
- 2) 0 B 3.2 POWER DISTRIBUTION LIMITS BASES Pages TS / B 3.2-1 through TS / B 3.2-4 1
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SUSQUEHANNA -UNIT 2 TS / B LOES-1 Revision 34
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS-BASES)
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B 3.5 ECCS AND RCIC BASES Pages TS / B 3.5-1 and TS / B 3.5-2 1
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SUSQUEHANNA - UNIT 2 TS / B LOES-2 Revision 34
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS-BASES)
Section Titre Revision B 3.6 CONTAINMENT SYSTEMS BASES Page TS / B 3.6-1 2
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SUSQUEHANNA - UNIT 2 TS / B LOES-3 Revision 34
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS-BASES)
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SUSQUEHANNA - UNIT 2 TS / B LOES-4 Revision 34
RCS P/T Limits B 3.4.10 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.10 RCS Pressure and Temperature (P/T) Limits BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design assumptions and the stress limits for cyclic operation.
This Specification contains PIT limit curves for heatup, cooldown, and inservice leakage and hydrostatic testing, and limits for the maximum rate of change of reactor coolant temperature. The heatup curve provides limits for both heatup and criticality.
Each P/T limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region.
The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure.
Therefore, the LCO limits apply mainly to the vessel.
10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials.
Reference 1 requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests.
It mandates the use of the ASME Code, Section Xl, Appendix G (Ref. 2).
The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.4-49 Revision 2
RCS P/T Limits B 3.4.10 BASES BACKGROUND (continued)
The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions.
The heatup curve used to develop the P/T limit curve composite represents a different set of restrictions than the cooldown curve used to develop the
- PIT limit curve composite because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner walls.
The criticality limits include the Reference 1 requirement that they be at least 40oF above the heatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leakage and hydrostatic testing.
The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components.
ASME Code, Section Xl, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits.
APPLICABLE SAETY ANALYSES The P/T limits are not derived from Design Basis Accident (DBA) analyses.
They are prescribed during normal operation to avoid 6ncountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed. Reference 7 establishes the methodology for determining the P/T limits. Since the P/T limits are not derived from any DBA, there are no acceptance limits related to the P/T limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.
RCS P/T limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 8).
(continued)
SUSQUEHANNA - UNIT 2 I
TS / B 3.4-50 Revision 1
RCS P/T Limits B 3.4.10 BASES LCO The elements of this LCO are:
- a. RCS pressure and temperature are to the right of the applicable curves specified in Figures 3.4.10-1 through 3.4.10-3 and within the applicable heat-up or cool down rate specified in SR 3.4.10.1 during RCS heatup, cooldown, and inservice leak and hydrostatic testing;
- b. The temperature difference between the reactor vessel bottom head coolant and the reactor pressure vessel (RPV) coolant is < 1451F during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- c. The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel is <50OF during recirculation pump startup, and during increases in THERMAL POWER or loop flow while operating at low THERMAL POWER or loop flow;
- d. RCS pressure and temperature are to the right of the criticality limits specified in Figure 3.4.10-3 prior to achieving criticality; and N
- e. The reactor vessel flange and the head flange temperatures are> 70OF when tensioning the reactor vessel head bolting studs.
These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure.
The P/T limit composite curves are calculated using the worst case of material properties, stresses, and temperature change rates anticipated under all heatup and cooldown conditions. The design calculations account for the reactor coolant fluid temperature impact on the inner wall of the vessel and the temperature gradients through the vessel wall. Because these fluid temperatures drive the vessel wall temperature gradient, monitoring reactor coolant temperature provides a conservative method of ensuring the P/T limits are not exceeded. Proper monitoring of vessel temperatures to assure compliance with brittle fracture temperature limits and vessel thermal stress limits during normal heatup and cooldown, and during inservice leakage and hydrostatic testing, is established in PPL Calculation EC 062-0573 (Ref. 9). For P/T curves A, B, and C, the bottom head drain line coolant temperature should be monitored and maintained to the right of the most limiting curve.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.4-51 Revision 2
RCS P/T Limits B 3.4.10 BASES LCO Curve A must be used for inservice leakage and hydrostatic testing.
(continued)
Curve A is based on pressure stresses only. During inservice leakage and hydrostatic testing, thermal stresses are assumed to be insignificant.
Therefore, heatup and cooldown rates are limited to 20oF in a one-hour period when using Curve A to ensure minimal thermal stresses. The recirculation loop suction line temperatures should be monitored to determine the temperature change rate.
Curves B and C are to be used for non-nuclear and nuclear heatup and cooldown, respectively. Heatup and cooldown rates are limited to 100OF in a one-hour period when using Curves B and C. This limits the thermal gradient through the vessel wall, which is used to calculate the thermal stresses in the vessel wall. Thus, the LCO for the rate of coolant temperature change limits the thermal stresses and ensures the validity of the P/T curves. The vessel beltline fracture analysis assumes a 1 00°F/hr coolant heatup or cooldown rate in the beltline area. The 1 OOoF limit in a one-hour period applies to the coolant in the beltline region, and takes into account the thermal inertia of the vessel wall. Steam dome saturation temperature (TsAT), as derived from steam dome pressure, should be monitored to determine the beltline temperature change rate at temperatures above 212oF. At temperatures below 212oF, the recirculation loop suction line temperatures should be monitored.
Violation of the limits places the reactor vessel outside of the bounds of the stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:
- a. The severity of the departure from the allowabli operating pressure temperature regime or the severity of the rate of change of temperature;
- b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced); and
- c. The existences, sizes, and orientations of flaws in the vessel material.
APPLICABILITY The potential for violating a P/T limit exists at all times. For example, P/T limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup. Therefore, this LCO is applicable even when fuel is not loaded in the core.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.4-52 Revision 1
RCS P/T Limits B 3.4.10 BASES (continued)
ACTIONS A.1 and A.2 Operation outside the P/T limits while in MODES 1, 2, and 3 must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses.
The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner.
Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed if continued operation is desired. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components.
ASME Code,Section XI, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is reasonable to accomplish the evaluation of amild violation. More severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed if continued operation is desired.
Condition A is modified by a Note requiring Required Adtion A.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the-excursion outside the allowable limits. Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity.
B.1 and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.4-53 Revision 1
RCS P/T Limits B 3.4.10 BASES (continued)
ACTIONS B.1 and B.2 (continued)
Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored.
Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > 2000F. Several methods may be used, including comparison with pre-analyzed transients, new analyses, or inspection of the components. ASME Code, Section Xl, Appendix E (Ref. 6),
may be used to support the evaluation; however, its use is restricted to evaluation of the beltiine.
SURVEILLANCE SR 3.4.10.1 REQUIREMENTS Verification that operation is within limits (i.e., to the right of the applicable curves in Figures 3.4.10-1 through 3.4.10-3) is reqbired every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations.
Surveillance for heatup, cooldown, or inservice leakage and hydrostatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied.
This SR has been modified with a Note that requires this Surveillance to be performed only during system heatup and cooldown operations and inservice leakage and hydrostatic testing.
Notes to the acceptance criteria for heatup and cooldown rates ensure that more restrictive limits are applicable when the P/T limits associated with hydrostatic and inservice testing are being applied.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.4-54 Revision 2
RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.2 REQUIREMENTS (continued)
A separate limit is used when the reactor is approaching criticality.
Consequently, the RCS pressure and temperature must be verified within the appropriate limits (i.e., to the right of the criticality curve in Figure 3.4.10-3) before withdrawing control rods that will make the reactor critical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of reactor criticality. Although no Surveillance Frequency is specified, the requirements of SR 3.4.10.2 must be met at all times when the reactor is critical.
SR 3.4.10.3 and SR 3.4.10.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation "K.i loop (Ref. 10) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.4 is to compare the temperatures of the operating recirculation loop and the idle loop. If both loops are idle, compare the temperature difference between the reactor coolant within.the idle loop to be started and coolant in the'reactor vessel.
SR 3.4.10.3 has been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4. In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required. The Note also states the SR is only required to be met during a recirculation pump start-up, because this is when the stresses occur.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.4-55 Revision 2
RCS P/T Limits B 3.4.10 BASES SURVEILLANCE SR 3.4.10.5 and SR 3.4.10.6 REQUIREMENTS (continued)
Differential temperatures within the applicable limits ensure that thermal stresses resulting from increases in THERMAL POWER or recirculation loop flow during single recirculation loop operation will not exceed design allowances. Performing the Surveillance within 15 minutes before beginning such an increase in power or flow rate provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the change in operation.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.10.6 is to compare the temperatures of the operating recirculation loop and the idle loop.
Plant specific startup test data has determined that the bottom head is not subject to temperature stratification at power levels > 30% of RTP and with single loop flow rate _> 21,320 gpm (50% of rated loop flow). Therefore, SR 3.4.10.5 and SR 3.4.10.6 have been modified by a Note that requires the Surveillance to be met only under these conditions. The Note for SR 3.4.10.6 further limits the requirement for this Surveillance to exclude comparison of the idle loop temperature if the idle loop is isolated from the RPV since the water in the loop can not be introduced into the remainder of the Reactor Coolant System.
SR 3.4.10.7, SR 3.4.10.8. and SR 3.4.10.9 Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits.
The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature < 80oF, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature _ 1 OOF, monitoring of the flange temperature is required every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to ensure the temperature is within the specified limits.
The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency is reasonable based on the rate of temperature change possible at these temperatures.
(continued)
SUSQUEHANNA-UNIT 2 TS / B 3.4-56 Revision I
RCS P/T Limits B 3.4.10 BASES (continue(
REFERENCES
- 2. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix G.
- 3. ASTM E 185-73
- 5. Regulatory Guide 1.99, Revision 2, May 1988.
- 6. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.
- 7. NEDO-21778-A, December 1978.
- 8. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
- 9. PPL Calculation EC-062-0573, "Study to Support the Bases Section of Technical Specification 3.4.10."
- 10. FSAR, Section 15.4.4.
SUSQUEHANNA-UNIT 2 TS / B 3.4-57 Revision 1 I
I
Reactor Steam Dome Pressure B 3.4.11 B 3.4 REACTOR COOLANT SYSTEM (RCS)
B 3.4.11 Reactor Steam Dome Pressure BASES BACKGROUND APPLICABLE SAFETY ANALYSES LCO APPLICABILITY The reactor steam dome pressure is an assumed initial condition of design basis accidents and transients and is also an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria.
The reactor steam dome pressure of < 1050 psig is an initial condition of the vessel overpressure protection analysis of Reference 1. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved.
Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statement (Ref. 3).
The specified reactor steam dome pressure limit of < 1050 psig ensures the plant is operated within the assumptions of the transient analyses.
Operation above the limit may result in a transient response more severe than analyzed.
In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required (continued)
In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant steam and the design basis accidents and transients are bounding.
SUSQUEHANNA-UNIT 2 TS / B 3.4-58 Revision 1
Reactor Steam Dome Pressure B 3.4.11 BASES APPLICABILITY (continued)
ACTIONS limit, and no anticipated events will challenge the overpressure limits.
A.1 With the reactor steam dome pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the bounds of the analyses. The 15 minute Completion Time is reasonable considering the importance of maintainirng the pressure within limits. This Completion Time also ensures that the probability of an accident occurring while pressure is greater than the limit is minimized. If the operator is unable to restore the reactor steam dome pressure to below the limit, then the reactor should be placed in MODE 3 to be operating within the assumptions of the transient analyses.
B.1 If the reactor steam dome pressure cannot be restored to within the limit within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE REQUIREMENTS SR 3.4.11.1 Verification that reactor steam dome pressure is < 1050 psig ensures that the initial conditions of the over-pressurization analysis are met. Operating experience has shown the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency to be sufficient for identifying trends and verifying operation within safety analyses assumptions.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.4-59 Revision 1
Reactor Steam Dome Pressure B 3.4.11 BASES (continued REFERENCES
- 1. FSAR, Section 5.2.2.1.
- 2. FSAR, Section 15.
- 3. Final Policy Statement on Technical Specifications Improvements, July 22, 1993 (58 FR 39132).
SUSQUEHANNA - UNIT 2 TS / B 3.4-60 Revision 1
)
ECCS-Operating B 3.5.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SR 3.5.1.4 Verification every 31 days that at least one RHR System cross tie valve is closed and power to its operator is disconnected ensures that each LPCI subsystem remains independent and a failure of the flow path in one subsystem will not affect the flow path of the other LPCI subsystem.
Acceptable methods of removing power to the operator include opening the breaker, or racking out the breaker, or removing the breaker. If both RHR System cross tie valves are open or power has not been removed from at least one closed valve operator, both LPCI subsystems must be considered inoperable. The 31 day Frequency has been found acceptable, considering that these valves are under strict administrative controls that will ensure the valves continue to remain closed with motive power removed.
SR 3.5.1.5 Verification every 31 days that each 480 volt AC swing bus transfers automatically from the normal source to the alternate source on loss of power while supplying its respective bus demonstrates that electrical power is available to ensure proper operation of the associated LPCI inboard injection and minimum flow valves and the recirculation pump discharge and bypass valves. Therefore, each 480 volt AC swing bus must be OPERABLE for the associated LPCI subsystem to be OPERABLE. The test is performed by actuating the load test switch or by disconnecting the preferred power source to the transfer switch and verifying that swing bus automatic transfer is accomplished. The 31 day Frequency has been found to be acceptable through operating experience.
SR 3.5.1.6 Cycling the recirculation pump discharge and bypass valves through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and provides assurance that the valves will close when required to ensure the proper LPCI flow path is established. Upon initiation of an automatic LPCI subsystem injection signal, these valves are required to be closed to ensure full LPCI subsystem flow injection in the reactor via the recirculation jet pumps. De-energizing the valve in the closed position will also ensure the proper flow path for the LPCI subsystem. Acceptable methods of de-energizing the valve include opening the breaker, or racking out the breaker, or removing the breaker.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.5-11 Revision 2
Secondary Containment B 3.6.4.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.4.1 Secondary Containment BASES BACKGROUND The function of the secondary containment is to contain, dilute, and hold up fission products that may leak from primary containment following a Design Basis Accident (DBA). In conjunction with operation of the Standby Gas.
Treatment (SGT) System and closure of certain valves whose lines penetrate the secondary containment, the secondary containment is designed to reduce the activity level of the fission products prior to release to the environment and to isolate and contain fission products that are released during certain operations that take place inside primary containment, when primary containment is not required to be OPERABLE, or that take place outside primary containment (Ref. 1).
The secondary containment is a structure that completely encloses the primary containment and those components that may be postulated to contain primary system fluid. This structure forms a control volume that serves to hold up and dilute the fission products. It is possible for the pressure in the control volume to rise relative to the environmental pressure (e.g., due to pump and motor heat load additions).
-The secondary containment boundary consists of the reactor building structure and associated removable walls and panels, hatches, doors, dampers, sealed penetrations and valves. The secondary containment is divided into Zone I, Zone II and Zone III, each of which must be OPERABLE depending on plant status and the alignment of the secondary containment boundary. Specifically, the Unit 1 secondary containment boundary can be modified to exclude Zone II. Similarly, the Unit 2 secondary containment boundary can be modified to exclude Zone I. Secondary containment may consist of only Zone III when in MODE 4 or 5 during CORE ALTERATIONS, or during handling of irradiated fuel within the Zone III secondary containment boundary.
(continued)
SUSQUEHANNA-UNIT2 TS / B 3.6-83 Revision 2
TSB APPROVED AMENDMENT TO THE UNIT 2 TECHNICAL SPECIFICATIONS BASES MANUAL REVISION 35 Replace the following pages of the Technical Specifications Bases Manual with the enclosed pages. The revised pages are identified by Revision Number and contain vertical lines indicating the area of change.
REMOVE PAGES REV. #
INSERT PAGES REV. #
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TS I B 3.0-13 through TS / B 3.0-15 1
SUSQUEHANNA STEAM ELECTRIC STATION LIST OF EFFECTIVE SECTIONS (TECHNICAL SPECIFICATIONS-BASES)
Section TitFe Revision B 2.0 SAFETY LIMITS BASES Page TS / B 2.0-1 1
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SUSQUEHANNA-UNIT2 TS/BLOES-4 Revision 35 SUSQUEHANNA -UNIT 2 TS / B LOES-4 Revision 35
SR Applicability B 3.0 BASES SR 3.0.3 period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, (continued) whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.
This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.
The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.
When a Surveillance with a Frequency based not on the time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix j, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance. However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.
SR 3.0.3 provides a time limit for, and allowances for the performance of Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.
Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals. While up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning,-availability of personnel, and the time required to perform the Surveillance.
(continued)
SUSQUEHANNA - UNIT 2 TS / B 3.0-13 Revision 1
SR Applicability B 3.0 BASES SR 3.0.3 (continued)
The risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, "Assessing and Managing Risk before Maintenance Activities at Nuclear Power Plants." This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods..The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the Corrective Action Program.
If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.
Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.
SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.
This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit.
The provisions of this Specification should not, be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.
(continued)
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BASES SR 3.0.4-.
(continued)
However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified condition change.
When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed per SR 3.0.1, which states that surveillances do not have to be performed on inoperable equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability. However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes.
The provisions of SR 3.0.4 shall not prevent changes in MODES or other.
specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown.
The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO Applicability would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately; the Surveillance may be stated in the form of a Note as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.
SR 3.0.4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3 or 4, or MODE 1 from MODE 2. Furthermore, SR 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3. The requirements of SR 3.0.4 do not apply in MODES 4 and 5, or in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.
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