ML20045E663
| ML20045E663 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 06/29/1993 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Marriott P GENERAL ELECTRIC CO. |
| References | |
| NUDOCS 9307020288 | |
| Download: ML20045E663 (35) | |
Text
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b-vfu.J sk _e a ntat, UNITED STATES.
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205 5 0001
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June 29,-1993
' Docket No.52-001 o
t Mr. Patrick W. Marriott, Manager 0
Licensing & Consulting Services GE Nuclear Energy 175 Curtner Avenue San Jose, California 95125 si
Dear Mr. Marriott:
SUBJECT:
ADVANCED BOILING WATER REACTOR (ABWR) TECHNICAL SPECIFICATIONS MILESTONES Enclosed are the proof and review ABWR technical specifications and their -
bases for the following sections:
2.0 Safety Limits 3.0 Applicability As we. discussed at our management meeting on June 10, 1993, the Nuclear Regulatory Commission (NRC) staff will be providing GE Nuclear Energy (GE),
between now and August 31, selected sections of proof.and-review ABWR techni-cal specifications. These sections are based on'the NRC staff review of the GE mark-up of the BWR-6 and BWR-4 Standard Technical Specifications;~ the sections, as provided, are acceptable to the NRC staff.
As discussed, we p-anticipate that GE will interface very closely with the staff to resolve the majority of any issues on these sections prior-to August 31, 1993.
Under this-arrangement, we anticipate that formal comments to proof and review ABWR technical specifications made by September 20, 1993,1will be few.
The electronic text of these sections is available:on the NRC Technical Specifications Branch electronic bulletin board. (OTSB-BBS) in Wordperfect 5.1
-format. The data telephone number for the OTSB-BBS is (301) 504-1778, and the
- system operator is Tom Dunning who is available for assistance at ~(301)
.504-1189.
Also, in accordance with our agreements,-GE will maintain these sections in Wordperfect 5.1 format and will produce subsequent issues of the'ABWR techni-cal specifications.in Wordperfect 5.1 format.
010006 EM 9307020288.930629' PDR. ADOCK 05200001
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Mr. Patrick W. Harriott June 29, 1993 If you have any questions about technical specifications please contact Mark
.Reinhart with the' Huclear Reactor Regulation Technical Specifications Branch.
He may'be reached at (301) 504-1185.
Sincerely, (Original signed by)
Dennis M. Crutchfield, Associat'e Director for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next'page DISTRIBUTION:
Docket File PDST R/F DORS R/F OTSB R/F PDR Central Files TEMurley/FJMiraglia WTRussell
~JGPartlow DMCrutchfield BKGrimes CIGrimes-BABoger JERichardson ACThadani FJCongel JSWermiel RBBarrett CEMcCracken RCJones CEBerlinger.
_AEChaffee GHMarcus WBHardin, RES-LCShao, RES JA0'Brien.
RWBorchardt JNWilson CPoslusny SNinh SMLHagruder GEGrant, 17G12 JEMoore, 15B18
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,s Mr. Patrick W. Harriott Docket No.52-001 l General Electric Company cc: Mr. Robert Mitchell Mr. Joseph Quirk General Electric Company GE Nuclear Energy 175 Curtner Avenue General Electric Company San Jose, California 95125 175 Curtner Avenue, Mail Code-782 San Jose, California 95125-Mr. L. Gifford, Program Manager R:gulat:ry Programs GE Nuclear Energy 12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852 Director, Criteria & Standards Division Office of Radiation Programs U. -S. Environmental Protection Agency 401 M Street, S.'W.
Washington, D.C.
20460 Mr. Sterling Franks U. S.' Department of Energy NE-42 Washington,.D.C. 20585 Marcus A. Rowden, Esq.
Fried, Frank, Harris, Shriver & Jacobson 1001 Pennsylvania Avenue, N.W.
Suite 800 Washington, D.C.
20004 Jay M. Gutierrez, Esq.
j' Newman & Holtzinger, P.C.
1615 L Strhet, N.W.
Suite 1000 t!ashington, D.C.
20036 4
T
- TABLE OF CONTENTS
' l.0 USE AND APPLICATION 1.1-1 1.1 Definitions 1.1-1
' l.2 Logical Connectors...................
1.2-1 1.3 Completion Times....................
1.3-1 1.4-1 1.4
_ Frequency 2.0 SAFETY. LIMITS (SLs) 2.0-1 2.1 SLs 2.0-1 2.2 SL Violations 2.0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY....
3.0-1; 3.0 SURVEILLANCE. REQUIREMENT (SR) APPLICABILITY 3.0-4 e
3.1 REACTIVITY CONTROL SYSTEMS 3.1-1 3.1.1 SHUTDOWN MARGIN (SDM) 3.1-1 3.1.2 Reactivi ty Anomalies................
3.1-5 3.1.3 Control Rod OPERABILITY 3.1-7 3.1.4 Control Rod Scram Times 3.1-13 3.1.5 Control Rod Scram Accumulators........... ' 3.1 3.1.6 Rod Pattern Control 3.1................
3.1.7 Standby Liquid Control (SLC) System........
3.1-21_
3.2 POWER DISTRIBUTION LIMITS...............
3.2-1 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)
'3.2-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) 3.2-2 3.2,3 LINEAR HEAT GENERATION RATE (LHGR)-(0ptional) 3.2...
3.2.4 Average Power Range Monitor (APRM) Gain and Setpoints (0ptional) 3.2-4 3.3
-INSTRUMENTATION....................
3.3-1 3.3.1.1 Reactor Protection System (RPS) Instrumentation 3.3-1 3.3.1.2 Source-Range Monitor (SRM) Instrumentation....-.
3.3-9 I
Control Rod Block Instrumentation 3.3-14 3.3.2.1 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation...
3.3-18' 3.3.3.2 Remote Shutdown System......._........
3.3-22 3.3.4.1 End.of Cycle Recirculation Pump Trip (E0C-RPT)
Instrumentation.................
3.3 3.3.4.2
_ Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation 3.3-29 3.3.5.1 Emergency Core Cooling _ System (ECCS)
Instrumentation.................
3.3.6.1 Primary Containment Isolation Instrumentation
. 3.3 3.3-48 3.3.6.2 Secondary Containment Isolation Instrumentation:
3.3-59' (continued)
ABWR TS v
Rev. O,11/4/92 1
~
Tf,BLE OF CONTENTS-3.3 INSTRUMENTATION (continued)
-l 3.3. 7. J Control Room Fresh Air (CRFA) System Instrumentation................. 3.3-73 3.3.8.1 Loss of Power (LOP).. Instrumentation 3.3-77 3.3.8.2 Load Shed and Sequence Instrumentation.
3.3-80 3.4 REACTOR COOLANT SYSTEM (RCS)...............
3.4-1
'3.4.1 Reactor Internal Pumps (RIPS) Operating 3.4-1 2.1.2 Safety / Relief Valves (S/RVs)..'...........
3.4-7 3.4.3 RCS Operational LEAKAGE 3.4.....
3.4.4 RCS Pressure ' Isolation Valve (PIV) Leakage.. -... 3.4-11 3.4.5 RCS Leakage Detection Instrumentation 3.4-14 3.4.6 RCS Specific. Activity 3.4-17 3.4.7 Residual _ Heat Removal (RHR) Shutdown Cooling System-410t Shutdown 3.4-20 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown..............
3.4-23 3.4. 9 RCS Pressure and Temperature (P/T) Limits..-...
3.4-25 3.4.10 Reactor Steam Dome Pressure
............ 3.4-29 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)
........-.- 3.5 3.5.1 ECCS-Operating 3.5..................
3.5.2 ECC S -Sh u t d own......... -..........
3.5 '
3.6 CONTAINMENT SYSTEMS..................
3.6 3.6.1.1 Primary Containment 3.6-1 3.6.1.2 Primary Containment Air Locks 3.6-3 ~
3.6.1.3 Primary Containment Isolation Valves (PCIVs)....
3.6-9 3.6.1.4 Primary Containment Pressure............ 3.6-19 3.6.1.5 Primary Containment Air Temperature.........
3.6 3.6.1.6 Suppression Chamber to Dry Well Vacuum Breakers 3.6-21' 3.6.2.1 Suppression Pool Average Temperature........ 3.6-29 g
3.6.2.2 Suppression Pool Water Level
........... 3.6-32 l
j (continued)
ABWR TS vi Rev.
O,11/4/92-j
TABLE OF CONTENTS
-3.6
. CONTAINMENT SYSTEMS. (continued) 3.6.2.3 Residual Heat Removal (RHR) Suppression Pool C o ol i n g.....................
3.6-33 3.6.2.4 Residual Heat Removal (RHR) Wetwell Spray, System..
3.6-35 3.6.3.1 Primary Containment Hydrogen Recombiners......
3.6-37 3.6.3.2 Primary Containment Oxygen Concentration System 3.6-39.
3.6.4.1 Secondary Containment 3.6-44
-3.6.4.2 Secondary Containment Isolation Valves (SCIVs)...
3.6-47 3.6.4.3 Standby Gas Treatment (SGT) System......... _3.6-S1 3.7 PLANT SYSTEMS 3.7-1 3.7.1.1 Reactor Cooling Water (RCW) System and Reactor Service Water.
(RSW) System................... 3.7-1 3.7.1.2 Ultimate Heat Sink (VHS)..............
3.7-1 3.7.2 Control Room Air Intake, Recirculation and Purge System.....................
3.7-7 3.7.3 Control Room Heating Ventilation and Air Conditioning (HVAC) System 3.7-11 3.7.4 Main Condenser Offgas 3.7...............
3.7.5 Main Turbine Bypass System............. 3.7-16 3.7.6 Fuel Pool Water Level 3.7-18 3.8 ELECTRICAL POWER SYSTEMS................
3.8-1 3.8.1 AC Sources-0perating 3.8-1 3.8.2 AC Sources --Shutdown................ 3.8-19 3.8.3 Diesel Fuel Oil, Lube Oil, and Starting Air Subsystem.....................~3.8-23 3.8.4 DC Sources-Operating 3.8-26.
3.8.5 DC Sources-Shutdown.....
3.8-30 3.8.6 Battery Electrolyte..
3.8-32 3.8.7 Inverters-0perating................
3.8-36 3.8.8 Inverters-Shutdown 3.8-38 g.
3.8.9 Distribution Systems-Operating 3.8...........
3.8.10 Distribution Systems-Shutdown...........
3.8-42 3.9 REFUELING OPERATIONS..................
3.9-1 3.9.1 Refueling-- Equipment Interlocks...........
3.9-1 3.9.2 Refuel Position One-Rod-Out Interlock 3.9-2 3.9.3 Control Rod Position................ l 3.9-4 3.9.4 Control Rod Position Indication ~
3.9-5 3.9.5 Control Rod OPERABILITY-Refueling.........
3.9 )
(continued)~
i ABWR TS vii Rev.
0, 11/4/92 i
TABLE OF CONTENTS 3.9 HEFUELING OPERATIONS (continued) 2.o.f Reactor Pressure Vessel (RPV) Water Level.....
3.9-8 3.9.7 Residual Heat Removal (RHR)--High Water Level.... 3.9-10 3.9.8 Residual Heat Removal (RHR)--Low Water Level....- 3.9-12 3.10 SPECIAL OPERATIONS...................
3.10-1 3.10.1 Inservice Leak and Hydrostatic Testing Operation..
3.10-1 3.10.2 Reactor Mode Switch Interlock Testing 3.10.......
? 10.3 Single Control Rod Withdrawal--Hot Shutdown 3.10-6 3.10.4 Single Control Rod Withdrawal--Cold Shutdown....
3.10-9 3.10.5 Single Control Rod Drive (CRD) Removal-Refueling 3.10-13 3.10.6 Multiple Control Rod Withdrawal--Refueling..... 3.10-16 3.10.7 Control Rod Testing--Operating........... 3.10-18 3.10.8 SHUTDOWN MARGIN (SDM) Test--Refueling 3.10.......
3.10.9 R I P s -- Te s t i n g...................
3.10-22 3.10.10 Training Startups 3.10-24 4.0 DESIGN FEATURES 4.0-1 4.1 Site........................
4.0-1 4.2 Reactor Core....................
4.0-1 4.3 Fuel Storage....................
4.0-2 5.0 ADMINISTRATIVE CONTROLS 5.0-1 5.1 Responsibility...................
- 5. 0 5.2 Organization........
5.0 5.3 Unit Staff Qualifications 5.0-8 '
5.4 Training......................
5.0 5.5 Reviews and Audits.................
5.0-10 5.6 Technical Specifications (TS) Bases Control 5.0....
5.7 Procedures, Programs, and Manuals 5 0-17 5.8 Safety Function Determination Program (SFDP)....
5.0-30 5.9 Reporting Requirements...............
5.0-32 p
5.10 Record Retention..........-........
5.0 [5.11 High Radiation Area 5.0-41]
ABWR TS viii Rev.
O, 11/4/92 i
SLs 2.0 2.0 SAFETY LIMITS (SLs)-
2.1 SLs 2.1.1 Reactor Core SL1 2.1.1,1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:
THERMAL POWER shall ba s 25% RTP.
2.1.1.2 With the' reactor steam dome pressure 2 785 psig and core flow 210% rated core flow:
MCPR shall be 2 [1.07] for operation with 9 or 10 RIPS.
2.1.1.3 Reactor vessel. water level shall be greater than the top -
of active irradiated fuel.
2.1.2 Beactor Coolant System Pressure SL Reactor steam dome pressure shall be maintained 51325 psig.
2.2 SL Violations With any SL violation, the following actiors shall be completed:
2.2.1 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance with 10 CFR 50.72.
9 2.2.2 Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods.
2.2.3 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the [ General Manager-Nuclear Plant and.
Vice President-Nuclear Operations) and the [offsite reviewers specified in Specification 5.5.2, "[0ffsite] Review and Audit"].
l (continued) l R
ABWR TS 2.0-1 Rev.
O',11/4/92 d
i.
SLs.
2.0 L
2.0 SLs 2.2 SL Violations (continued) 2.2.4 Within 30 days, a Licensee Event Report (LER) shall be prepared pursuant to 10 CFR 50.73. The LER shall be submitted to the NRC, the [offsite reviewers specified in Specification 5.5.2], and the
[ General fianager--Nuclear Plant and Vice President--Nuclear-Operations).
2.2.5 Operation of the unit shall not be resumed until authorized by the NRC.
4 4
i ABWR-TS 2.0-2 Rev.
O, 11/4/92 l
LCO Applicability 3.0
,3.0 _ LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LC0 3.0.7.
LC0 3.0.2 Upon discovery of a failure to meet an'LCO, the Required Actions of the associated Conditions shall be met, except as provided in LC0 3.0.6.
If the LC0 is met or. is no longer applicable prior to expiration of the specified completion Time (s), completion of the Required Action (s) is not -required, unless otherwise stated.
LCO 3.0.3 When an LC0 is not met and the associated ACTIONS are not-met or an associated ACTION is not provided, the unit shall l
be placed in a MODE or other specified condition in which.
the LC0 is not applicable. Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:
a.
MODE 2 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />; b.
MODE 3 within'13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and c.
MODE 4 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.
Exceptions to this Specification are stated in the-p individual Specifications.
Where corrective measures are completed that' permit operation in accordance with the LC0 or ACTIONS, completion -
of the actions required by LCO 3.0.3 is not. required.
LC0 3.0.3 is. applicable' in MODES-1, c2, and 3.
f
'LC0 3.0.4 When an LCO is not met, entry into a MODE or_other specified-condition in the-Applicability shall not be made.except when the associated ACTIONS to be entered permit continued operation in the' MODE or other specified condition in the Applicability for an unlimited period of time.
This Specification'shall not prevent changes _in-MODES or other.
(continued)
ABWR TS 3.0-1
-Rev.
O, 11/4/92 I
LC0 Applicability 3.0 a.v LLv M PLICABILITY LC0 3.0.4 specified conditions in the Applicability that are required (continued) to comply with ACTIONS.
Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered allow unit cperation in the MODE or other specified condition in the Applicability only for a limited period of time.
LCO 3.0.5 Equipment removed fron. service or declared inoperable to comply with ACTIONS may be returned to service under-administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LC0 3.0.2 for the system returned to service under administrative controlto perform the testing required to demonstrate OPERABILITY.
LC0 3.0.6 When a supported system LC0 is not met solely due to a support system LC0 not being met, the Conditions.and:
Required Act. ions associated with this supported system are.
not required to be entered._ Only the support system LC0 ACTIONS are required to be entered.
This is an exception to LC0 3.0.2 for the supported-system.
In this. event, additional evaluations and limitations may be required in g
accordance with Specification 5.8, " Safety Function Determination Program (SFDP)."
If a-loss of safety functicn is determined to exist by this program, the appropriate.
Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.
When a support system's Required' Action' directs a' supported.
system _to be declared inoperable or directs entry into Conditions and Required-Actions for a supported system,-the applicable Conditions and Required Actions-shall be entered in accordance with LC0 3.0.2.
(continued).
ABWR TS 3.0-2 Rev.
O,11/4/92
LCO Applicability 3.0 3.0 LCO APPLICABILITY (continued)
LCO 3.0.7 Special Operations LCOs in Section 3.10 allow specified Technical Specifications (TS) requirements to be changed _ to permit performance of special tests and operations.
Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Special Operations LCOs is optional. When a Special Operations _LC0 is desired to be met but is not met, the ACTIONS of the Special Operations LC0 shall be met. When a Special Operations LC0 is not desired to be met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with the other applicable Specifications.
d ABWR TS 3.0-3 Rev.
O,11/4/92-
SR Applicability-3.0 5.0 50RVULLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other soecified conditions in the Applicability for individu'al LCOs, unless otherwise stated in the SR.
Failure to meet a Surveillance, whether such failure is experienced during.the performance of. the Surveillance or between ' performances of the Surveillance, shall be failure to meet the LCO.
Failure to perform a Surveil _ lance within the'specified Frequency shall be failure to meet the LCO except as provided in SR:3.0.3.
Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.
SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time.a specified condition of the Frequency is met.
For Frequencies specified as "once," the above interval extension does not apply.
If a Completion Time requires periodic performance-on a "once per..." basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
I SR 3.0.3 If it-is discovered.that a Surveillance was'not performed within its specified Frequency, then compliance.with the requirement to declare the LCO not met may be delayed,-from the time of discovery, 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,-whichever is less. This delay period is permitted to allow performance of the Surveillance.
If the Surveillance is not. performed within the delay.
period, the LC0 must immediately be declared not met, and.
the applicable Condition (s) must be entered..'The_ Completion Times of the Required Actions begin immediately upon expiration of the delay period.
(continued)
ABWR TS' 3.0-4 Rev. O,'11/4/92
SR Applicability 3.0 3.0 SR APPLICABILITY
'SR 3.0.3 When the Surveillance is performed within the delay period (continued) and the Surveillance is not met, the LC0 must immediately be declared not met, and the applicable Condition (s) must be entered.
The Completion Times of the Required Actions begin immediately upon failure to meet the Surveillance.
SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LC0 shall not be mada unless the LC0's Surveillances have been met within their specified Frequency. This provision shall not prevent passage through or to MODES or other 'specified conditions in compliance with Required Actions, e
O, 11/4/92
Safety Limits (SLS)
Reactor Core SIA BASES BACKGROUND GDC 10 (Reference B2.1.1-1) requires, and SLs ensure, that specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).
The fuel cladding integrity SL is set, such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, stepback approach is used to establish an SL, such that the MCPR is not less than the limit specified in Specification 2.1.1.2. MCPR greater than the specified limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforatioas or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.
While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, the fuel cladding SL is defined with a.
margin to the conditions that would produce onset of transition boiling (i.e., MCPR = 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL ensures that during normal operation and during AOOs, at least 99.9% of the fuel rods in the core do not ' experience transition boiling.
Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of transition boiling 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, resulting in an uncontrolled release of activity to the reactor coolant.
APPLICAllLE The fuel cladding must not sustain damage as a result of normal SAFLTY ANALYSES operation and AOOs. The reactor core SLs are established to preclude violation of the fuel design criterion that an htCPR is to be established, such that at least 99.9% of the fuel rods in the core would not be expected to experience the onset of transition boiling.
The Reactor Protection System setpoints (LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation"), in combination with all the LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and TIIER5fAL POWER level that would result in reaching the h1CPR SL.
2.1.1.la Fuel Cladding Integrity (General Electric Corporation (GE)
Fuel)
GE critical power correlations are applicable for all critical power calculations at pressures 2 785 psig or core flows 210% of rated flow. For operation at low pressures and low flows, another basis is used, as follows:
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows.
will always be > 4.5 psi. Analyses (Referente B2.1.1-2) show that with a bundle flow of 28 x 103 lb/hr, bundle pressure drop 3
is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be
> 28 x 103 lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 51wt. With the design peaking factors, this corresponds to a TIIERh!AL POWER.
> 50% RTP. Thus, a TIIERh!AL POWER limit of 25% RTP -
for reactor pressure < 785 psig is conservative.
2.1.1.11 h1CPR (GE Fuel)
- The fuel cladding integrity SL is set, such that no significant fuel damage is calculated to occur if the limit is not violated. Since the parameters that i
result in fuel damage are not directly observable during reactor operation,.
the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the beginning of the region in which fuel damage could occur. Although it is recognized that the onset of transition boiling would not result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as c convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity SL is defined as the critical power ratio in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.
The MCPR SL is determined using a statistical model that combines all the uncertainties in operating parameters and the procedures used to calculate critical power. 'nie probability of the occurrence of boiling.
transition is determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL calculation -
are given in Reference B2.1.1-2 Reference B2.1.1-2 also includes a tabulation of the uncertainties used in the determination of the MCPR SL -
and of the nominal values of the parameters used in the MCPR SL' statistical analysis, d-2.1.1.2 Reactor Vessel Water Level During MODES I and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability. With fuel in the reactor vessel during periods when the reactor is shut down, consideration must be given to water level requirements due to the effect of decay heat. If the water level should drop below the top of the active irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level becomes <
of the. core height. The reactor vessel water level SL has -
been established at the top of the active irradiated fuel to provide a point that can be monitored and to also provide adequate margin for effective -
action.
The reactor core SLs are established to protect the integrity of the fuel '
clad barrier to the release of radioactive materials to the environs.
SL 2.1.1.1 and SL 2.1.1.2 ensure that the core operates within the fuel design criteria. SL 2.1.1.3 ensures that the reactor vessel water level is greater than the top of the active irradiated fuel, thus maintaining a coolable geometry.
Siis 2.1.1.1,2.1.1.2 and 2.1.1.3 are applicable in all MODES. However, in MODES 3, 4, and 5, with the reactor shut down, it is unlikely that fuel -
cladding integrity SLs would be violated.
2.2.1 If any SL is violated, the NRC Operations Center must be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance with 10 CFR 50.72 (Reference B2.1.1-4).-
2.2.2 Exceeding an SL may cause fuel damage and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria,"
limits (Reference B2.1.1-5). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action and also ensures that the probability of an accident occurring during this period is minimal.
2.2.3 If any SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified _within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period provides time for plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the senior management..
N-2.2.4 If any SL is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC, the senior management of the
]
nuclear plant, and the utility Vice President-Nuclear Operations.
This requirement is in accordance with 10 CFR 50.73 (Reference B2.1.1-6).
2.2.5 If any SL is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all
]
necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.
REFERENCES 10 CFR 50, Appendix A, GDC 10.
2 1
NEDE-24011-P-A, (latest approved revision).
1 4
2 I
1 I
i
.l
)
t.
Safety Limits (SLS)
Reactor Coolant System (RCS) Pressure SL BASES The SL on reactor steam dome pressure protects 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. Establishing an upper limit on reactor steam dome pressure ensures continued RCS integrity. According to 10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and GDC 15, " Reactor Coolant Systerr. Design" (Reference B2.1.2-1), the reactor coolant pressure boundary (RCPB) shall be designed with '
sufficient margin to ensure that the design conditions are not exceeded during normal operation and anticipated operational occurrences (AOOs). During MODES I and 2, the reactor vessel water level is required to be above the top of the active fuel to provide core cooling capability.
During normal operation and AOOs, RCS pressure is limited from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code (Reference B2.1.2-2). To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure, in accordance with ASME Code requirements, prior to initial operation when there is no fuel in the core. Any further hydrostatic testing with fuel in the core is done under LCO 3.10.1,
" Inservice Leak and Ilydrostatic Testing Operation." Following p
inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code, Section XI-(Reference B2.1.2-3).
Overpressurization of the RCS could result in a breach of the RCPB.
If this occurred in conjunction with a fuel cladding failure, fission products could enter the containment atmosphere, raising concerns relative to limits on radioactive releases specified in 10 CFR 100,
~
" Reactor Site Criteria" (Reference B2.1.2-4).
APPLICABLE The RCS safety / relief valves and the Reactor Protection System SAFETY ANALYSES Reactor Vessel Steam Dome Pressure-Iligh Function have settings -
established to ensure that the RCS pressure SL will not be exceeded.
The RCS pressure SL has been selected such that it is at a pressure-below which it can be shown that the integrity of the system is not endangered. The reactor pressure vesselis designed to ASME, Boiler
and Pressure Vessel Code,Section III, [1971 Edittor.], including Addenda through the [ winter of 1972] (Reference B2.1.2-5), which permits a maximum pressure transient of 110%,1375 psig, of design.
pressure 1250 psig. The SL of 1325 psig, as measured by the reactor steam dome pressure Indicator, is equivalent to 1375 psig at the lowest elevation of the RCS. The RCS is designed to ASME Code,Section III,1974 Edition (Reference B2.1.2-6), for the reactor recirculation piping, which permits a maximum pressure transient of 110% of design pressures of 1250 psig for suction piping and 1650 psig for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.
The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III,is 110% of design pressure. The; maximum transient pressure allowable in the RCS piping, valves, and fittings is 110% of design pressures of 1250 psig for suction piping and 1500 psig for discharge piping. The most limiting of these two allowances is the 110% of design pressure; therefore, the SL on maximum allowable RCS pressure is established at 1375 psig.
SL 2.1.2 applies in all MODES; however, in MODE S, because the reactor vessel head closure bolts are not fully tightened, it is unlikely the RCS would be pressurized.
2.2.1 if any SL is violated, the NRC Operations Center must be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, in accordance with 10 CFR 50.72 (Reference B2.1.2-7).
l' 2.2.2 Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100,
" Reactor Site Criteria," limits (Reference B2.1.2-4). Therefore,- it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time ensures that the operators take prompt remedial action.
1 2.2.3 If any SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period
i, provides time for plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the senior management.
2.2.4 If any SL is violated, a Licensee Event Report shall be prepared and.
submitted within 30 days to the NRC, the senior management of the nuclear plant, and the utility Vice President-Nuclear Operations. This requirement is in accordance with 10 CFR 50.73 (Reference B2.1.2-8).
2.2.5 If any SL is violated, restart of the unit shall not commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.
REFERENCES 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.
ASME, Boiler and Pressure Vessel Code,Section III, Art'icle~ NB-7000.
ASME, Boiler and Pressure Vessel Code,Section XI, Article IW-5000 10 CFR 100.
ASME, Boiler and Pressure Vessel Code, [1971 Edition], Addenda,
[ winter of 1972].
ASME, Boiler and Pressure Vessel Code, (1974 Edition].
h i
imi ng Condition for Operation (LCO) Applicability LCOs LCO 3.0.1 through LCO 3.0.7 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.
LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or ott:er specified conditions of the Applicability statement of each Specification).
LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure 'o meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The -
Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:
a.
Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification; and b.
Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.
There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met.
Y.
This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. OVhether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering LCO 3.0.2 -
ACTIONS.) The second type,
of Required Action specifies the (continued) remedial measures that. permit continued operation of the unit that is not-further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for; continued operation.
-i
Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.
The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the ca.se. An example of this is in LCO 3.8.1, "AC Sourem-Operating."
The Completion Times of the Required Actions are also applicable' when a system or component is removed from service intentionally.
The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not he made for operational convenience.
Alternatives that would not result in redundant equipment being -
Inoperable should be used instead. Doing so limits the time both --
subsystems / trains of a safety function are Inoperable and limits the time other conditions exist which result in LCO 3.0.3 being entered.
Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing.
In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains b
removed from service or bypassed.
When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter _a MODE or other specified condition in which another Specification becomes applicable, in this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition (s) are entered.
= LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an -
LCO is not met and:
a.
An associated Required Action and Completion Time is not met -
and no_other. Condition applies; or b.
The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of -
Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS
. specifically state a Condition corresponding to such combinations and n!so that LCO 3.0.3 be entered immediately..
This Specification delineates the time limits for placing the unit in a safe MODE or other specified condition when operation cannot be male.lained within the limits for safe operation as defined by the LCO and its ACTIONS. It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.
Upon entering LCO 3.0.3,1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcizer to ensure the stability and availability of the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE.
This reduces thermal stresses on components of the Reactor Coolant -
System and the potential Ice a plant upset that could challenge safety -
N systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, j
Completion Times.
A unit shutdown required in accordance with LCO 3.0.3 may be l
terminated and LCO 3.0.3 exited if any of the following occurs:
a.
The LCO is now met.-
h.
A Condition exists for which the Required Actions have now been performed.
c.
ACTIONS exist that do not have expired Completion Times..
These Completion Times are applicable from the point in time.
that the Condition is initially entered and not from the time LCO 3.0.3 is exited.
~
i
-g l
m The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> for the unit to be in MODE 4 when a shutdown is required during MODE 1 operation. If the unit is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies. If a lower MODE is reached in less time than allowed, however, the total allowable time to reach MODE 4, or other applicable MODE, is not reduced. For example, if MODE 2 is reached in 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, then the time allowed for reaching MODE 3 is the next 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />, because the total time for reaching MODE 3 is not reduced from the allowable limit of 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. Therefore, if remedial measures g
are completed that would permit a return to MODE 1, a penalty is' not incurred by having to reach a lower MODE of operation in less than the total time allowed.
In MODES 1,2, and 3, LCO 3.03 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 4 and 5 because the unit is already in the most restrictive Condition required by LCO 3.03. The requirements of LCO 3.03 do not apply 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.
Exceptions to LCO 3.03 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the as'sociated condition of the unit.
An example of this is in LCO 3.7.7, " Fuel Pool Water Ixvel."
LCO 3.7.7 has an Applicability of "During movement of irradiated fuel LCO 3.03 assemblies in the associated fuel storage pool." Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.7 are not met while in MODE 1,2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.7 of " Suspend movement ofirradiated fuel assemblies in the associated fuel storage pool (s)" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3. These exceptions are addrased in the individual Specifications.
LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It
-t
'a precludes placing the unit in a different h10DE or other specified condition when the following exist:
a.
.The requirements of an LCO, in the h10DE or other specified condition to be entered, are not met; and b.
Continued noncompliance with these LCO requirements would result in the unit being required (c-be placed in a h10DE or othe specified condition in which the LCO does not apply to comply with the Required Actions.
Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued-operation. This is without regard to the status of the unit before or after the 510DE change. Therefore, in such cases, entry into a h10DE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions. The provisions of this Specification should not be interpreted as ~ endorsing the failure to exercise the good practice of restoring systems or I
components to OPERABLE status before unit'startup.
The provisions of LCO 3.0.4 si all not prevent changes in MODES or other specilled 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 A10 DES or other specified conditions in the Applicability that result from a normal shutdown.
Exceptions to LCO 3.0.4 are stated in the individual Specifications.
&ceptions may apply to all the ACTIONS or to a specific Required E
Action of a Specification.
Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, changing h10 DES or other specified conditions while in an ACTIONS Condition, either in compliance with LCO 3.0.4, or where an exception to LCO 3.0.4 is stated, is not a violation of SR 3.0.1 or SR 3.0.4 for those Surveillances that do not have to be performed due to the associated inoperable equipment. IIowever, SRs must he met to ensure -
OPERABILITY prior to declaring the associated equipment.
OPERABLE (or variable within limits) and restoring compliance with -
the affected LCO.
LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service -
under administrative control' when it has been removed from service s
or declared inoperable to comply.with ACTIONS. The sole purpose.of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the-performance of SRs to demonstrate:
a.
The OPERABILITY of the equipment being returned to service; or 3
b.
The OPERABILITY of other equipment.
The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS '
is limited to the time absolutely necessary to perform the allo ved SRs.
This Specification does not provide time to perform any other preventive or corrective maintenance.
An example of demonstrating the OPERABILITY of th' equipment e
being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions, and must be-reopened to perform tiae SRs.
An examp!e of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of an SR on another channel in the other trip system, A similar example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the -
appropriate response during the performance of an SR on another channel in the same trip system.
LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for support systems.-
that have an LCO specified in the Technical Specifications (TS). This exception is provided because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system LCO be entered solely due to the inoperability of the support system. This exception is justified because the actions that are ;
required to ensure the plant is maintained in a safe condition are specified in the support system LCO's Required Actions. These Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.
When a suppo'rt system is inoperable and there is an LCO specified f
n t'
E
r -
for it in the TS, the supported system (s) are required to be decleed inoperable if determined to be inoperable as a result of the support system inoperability. However, it is not necessary to enter into the y
supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsisteracy of requirements related to the entry into~
multiple support and supported systems' LCOs' Conditions and -
Required Actions are eliminated by providing all the actions that are :
necessary to ensure the plant is maintained in a safe condition in the support system's Required Actions.
However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry imo Conditions and Required Actions'for the supported system. This may occur immediately or after some specified delay to perform some other Required Act!on. Regardless of whether it is Imraediate or after some delay, when a support system's Required '
Action directs a supported system to be dec'ared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.
Specification 5.8, " Safety Function Determination Program" (SFDP),.
ensures loss of safety function is detected and appropriate actions are taken. 'Upon failure to meet two or more LCOs concurrently, an evaluation shall be made to determine if loss of safety function exists.
Additionally, other limitations, remedial actions, or compensatory I~
actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. The SFDP implements the requirements of LCO 3.0.6.
Cross division checks to identify a loss of safety function for those support systems that support safety systems are required. The cross
}~
division check verifies that the supported systems of the redundant OPERABLE support nystera are OPERABLE, thereby ensuring safety function is retained. If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the bss of safety function _ exists are required to be entered.
LCO 3.0.7 There are certain special tests and operations required to be k.
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y
]qj performed at.various times over the life of the unit. These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities,'and to perform special evolutions. Special Operations LCOs in Section 3.10 allow specified TS requirements to be changed to permit performances of these special tests and operations, which otherwise could not be t-performed if required to comply with the requirements of these TS.
Unless otherwise specified, all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will remain in iffect.
The Apphenbility of a Special Operations LCO represents a condition not necessarily in compliance with the normal requireuents of the TS.
Compliance with Special Operations LCOs is optional. A special operation may be performed either under the provisions of the appropriate Special Operations LCO or under the other applicable TS.
requirements. If it is desired to perform the special operation under the provisions of the Special Operations LCO, the requirements of the Special Operations LCO shall be followed. When a Special Operations LCO requires another LCO to be met, only the requirements of the LCO statement are required to be met regardless of that LCO's Applicability (i.e., should the requirements of this other LCO not be met, the ACTIONS of the Special Operations LCO apply, not the ACTIONS of the other LCO). Ilowever, there are instances where the E
Special Operations LCO ACTIONS may direct the other LCOs' ACTIONS be met. The Surveillances of the other LCO are not required to be met, unless specified in the Special Operations LCO. If conditions exist such that the Applicability of any other LCO is met, all the other LCO's requirements (ACTIONS and SRs) are required to be met concurrent with the requirements of the Special Operations LCO.
T
Surveillame Requirement (SR) Applicabihty 13AShS S Rs --
SR 3.0.1 through SR 3.0.4 estab!!sh the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.
SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the I
individual SRs. This Specification is to" ensure that Surveillances are -
performed to verify the OPERABILITY of systems and components,.and.
that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.
Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is -
to be construed as implying that systems or components are OPERABLE when:
a.
The systems or components are known to be inoperable, although still meeting the SRs; or b.
The requirements of the Surveillance (s) are known to be not met between required Surveillance performances.
Surveillances do not have to be performed when the unit is in a MODE.
or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated-with a Special Operations LCO are only applicable when the Special Operations LCO is used as an allowable exception to the requirements of a
(
Specification.
Surveillances, including Surveillances invoked by Required Actions, do not have to be perfc,rmed on inoperable equipment because the ACTIONS.
define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.
Upon completion of maintenance,' app.op.. ate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring
- i i
- applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2; Post maintenance testing may not be possible in the current-MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been -
j established. In these situations, the equipment may be considered -
OPERABLE provided testing has b'een satisfactorily completed to the -
a s
n
?
4 extent possible and f ue equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a m
MODE or other specified condition where other necessary post maintenance tests can be completed. Some examples of this process are:-
a.
Control rod drive maintenance during refueling that requires scram testing at > 800 psi. However, if other appropriate testing is satisfactorily completed and the scram time testing of SR 3.1.3.4 is satisfied, the control rod can be considered OPERABLE. This allows startup to proceed to reach 800 psi to perform other.
d' necessary testing.
b.
High pressure core spray (HPCS) maintenance during shutdown that -
requires system fonctional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can" proceed with HPCF considerea OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.
SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency _
for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a."once per..." interval.
SR 3.0.2 permits a 25% extension of the interval specified in the.
Frequency. This extension facilitates Surveillance scheduling and considers.
plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).
,-b The 25% extension does not significantly degrade the reliability that results fram performing the Surveillance at its specified Frequency. This:
is based on the recognition that the most probable result of any panicular Surveillance being performed is the verification of conformance with the -
SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.
These exceptions are stated in the individual Specifications.. An example of where SR 3.0.2 does not apply is a Surveillance with a Frequency of "in accordance with'10 CFR 50, Appendix J, as modified by approved exemptions." The requirements of regulations take precedence over the -
TS. The TS cannot in and of themselves extend a test interval specified in' the regulations; Therefore, there is a Note in the Frequency stating,.
"SR 3.0.2 is not applicable."
r L
As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on -
a "once per..." basis. The 25% extension applies to each performance
~
after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial:
action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by -
checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.
The provisions of SR 3.0.2 are not intended to be used repeatedly '
merely as an operational convenience to extend Surveillance intervals or periodic Completion Time intervals beyond those specified.
SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A~
delay period of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 perfo'rm I'
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 time intervals, but upon specified unit conditions or operational situations, is discovered not to have been performed when specified, SR 3.0.3 allows the full delay perio' of 24 hcurs to perform the Surveillance.
SR 3 ' 3 also provides a time limit for completion 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 E
an infrequent occurrence. Use of the delay period established by.SR 3.0.3
4 M
H is a flexibility which is not intended to be used as an operational m
convenience to extend Surveillance intervals.
If a Surveillance is not completed within the allowed delay period, then.
the equipment is considered inoperable or the variable then is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon 1
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 Conditions begin immediately upon the failure of the Surveillance.
Completion of the Surveillance within the delay period allowed by this t
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;
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other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit. This Specification applies to-changes in' MODES or other specified conditions in the' Applicability
' ssociated with unit shutdown as well as startup.
a 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.
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 -
rnet. 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, k
e i
or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.
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