ML20129H918
| ML20129H918 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 10/29/1996 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20129H909 | List: |
| References | |
| NUDOCS 9611060066 | |
| Download: ML20129H918 (390) | |
Text
{{#Wiki_filter:. -- . . - . . - - . - . . - Attachment to Enclosure 1A D ' it-Specific Information Relative to General Electric Licensing Topical Report NEDC-32410P-A , Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC-PRNM) Retrofit Plus Option III Stability Trip Function I i The information provided in this attachment adheres to the format presented in the l NUMAC Power Range Neutron Monitor (PRNM) Licensing Topical Report (LTR) i NEDC-32410P-A, with the exception of Section 8, which discusses the Technical l Specifications changes addressed in Enclosure I A. I 1 i I i 9611060066 961029 I DR ADOCK 050003 1
Attachment to Enclosure l A
- NUMAC PRNM Licensing Topical Report NEDC-32410P-A {
Plant-Specific Infonnation i 4 l LTR Section LTR Utility Action Required GPC Response I 2.3.4 " Confirm that the actual plant configuration
- The current Plant Hatch Units I and 2 I is included in the variations covered in this configuration and the replacement report, and the configuration alternative (s) PRNM configuration are described in ,
i~ being applied for the replacement PRNM are the following PRNM LTR sections: covered by this report. Document in the plant-specific licensing submittal for the Current Proposed PRNM project the actual current plant APRMs 2.3.3.1.1 2.3.3.1.2 configuration and the configuration of the RBMs 2.3.3.2.1 2.3.3.2.2 replacement PRNM, and document Flow Units 2.3.3.3.1 2.3.3.3.2 l confirmation that those are covered by the Rod Control 2.3.3.4.1 2.3.3.4.2 ; PRNM LTR. ARTS 2.3.3.5.1 2.3.3.5.2 PanelInterface 2.3.3.6.1 2.3.3.6.2 l e For any changes to the plant operator's . The replacement PRNM design j panel, document in the submittal the complies with the Hatch Human Factors ; ) human factors review actions that were Design Standard. This Standard is 1
- taken to confirm compatibility with based on NUREG 0700, " Guideline for ;
i existing plant commitments and Control Room Design Reviews" and j procedures " Instrument Society of America (ISA) RP i 60.3, " Human Engineering for Control
- Centers."
i j 3.4 "As part of the plant-specific licensing submittal, the utility should document the { following: i l e The pre-modification flow channel e The current Plant Hatch flow channel configuration, and any changes planned. configuration consists of four flow
- channels and eight transmitters. No NOTE: If transmitters are added, the process changes, other than modifying i requirements on the added transmitters the transmitters to have 4-20 mA output
{ should be: signals, are planned. (
- Non-safety related, but qualified
. environmentally and scismically to 2 operate in the application , environment
- Mounted with structures equivalent or
- better than those for the currently installed channels
+ Cabling routed to achieve separation j to the extent feasible using existing cableways and routes i
. HL-5054 A-1 i s
Attachment to Enclosure l A . NUMAC PRNM LTR NEDC-32410P-A Plant-Specific Information LTR Section LTR Utility Action Required GPC Response 3.4 (Cont.) e Document the APRM trips currently e Current APRM trips are the same as applied at the plant. If different from those identified in the LTR (Mode 2: those identified in this report [PRNM neutron flux - high; Mode 1: flow biased LTR], document the plan to change to simulated thermal power - high, fixed those defined in this report [the LTR]. neutron flux-high, downscale with the companion IRM upscale, and inop).
. Document the current status related to e ARTS has been implemented on both ARTS, and planned post modification Units 1 and 2 and will be retained with status as: the replacement PRNM system.
- ARTS currently implemented, and retained in the PRNM
- ARTS will be implemented concurrently with the PRNM (reference ARTS submittal)
- ARTS not implemented and will not be implemented with the PRNM
- ARTS not applicable" 4.4.1.11 "This section [the PRNM LTR] identifies The current Plant Hatch PRNM was supplied requirements that are expected to encompass by GE. As discussed in the LTR, the most specific plant commitments relative to replacement PRNM system is designed to the the PRNM rep!acement project, but may not same requirements as the original system.
be complete and some may not apply to all The Plant Hatch FSAR design and system plants. Therefore, the utility must confirm commitments are consistent with the that the requirements identified here [in the regulatory requirements described in PRNM LTR] address all of those identified section 4.4 of the LTR. in plant commitments. The plant-specific licensing submittal should identify the specific requirements applicable for the plant, confirm that any clarifications included here lin the PRNM LTR] apply to the plant, and document the specific l requirements that the replacement PRNM is intended to meet for the plant." l l HL-5054 A-2
A Attachment to Enclosure 1 A .
- NUMAC PRNM LTR NEDC-32410P-A j Plant-Specific Inf'ormation i
- LTR Section LTR Utility Action Required GPC Response
! 4.4.2.2.1.4 " Plant-specific action will confirm that the The Plant Hatch maximum control room j maximum control room temperatures plus temperature is 105'F. The heat load of the j mounting panel temperature rise, allowing replacement PRNM equipment is less than for heat load of the PRNMS equipment, does the existing equipment. Therefore, the not execed the above temperatures [ stated in replacement PRNM clectronics will not
; the PRNM LTRJ, and that control room exceed the qualified temperature of 5 to humidity is maintained within the above 50'C (41 to 122 F).
j, limits [ stated in the PRNM LTR). This } evaluation will normally be accomplished by The maximum humidity in the control room ] determining the operating temperature of the is < 75% relative humidity. This level does j current equipment which will be used as a not exceed the qualified 90% relative 3 bounding value because the heat load of the humidity of the PRNM clectronics. ! replacement system is less than the current j system u hile the panel structure, and thus 4 cooling, remains essentially the same. l Documentation of the above action, including the specific method used for the e required confirmation, should be included in l plant-specific licensing submittals." 4.4.2.2.2.4 " Plant-specific action will confirm that the The Plant Hatch maximum control room maximum control room pressure does not pressure is + 0.1 in. water gauge. This level exceed the above limits [ stated in the PRNM does not exceed the replacement PRNM LTR]. Any pressure differential from inside electronics qualified maximum pressure of to outside the mounting panel [is] assumed + 1.0 in. water gauge ambient. to be negligible since the panels are not scaled and there is no forced cooling or ventilation. Documentation of this action and the required confirmation should be included in plant-specific licensing submittals." 4.4.2.2.3.4 " Plant-specific-action will confirm that the The Plant Hatch control room maximum maximum control room radiation levels do design dose rate is 5.0 E-4 Rads (carbon)/hr, not exceed the above limits [ stated in the and the maximum total integrated dose is PRNM LTR). Documentation of this action 1.75 E+2 Rads (carbon) for normal and the required confirmation should be conditions and 3.7 E 1 Rads (carbon) for included in plant-specific submittals." accident conditions. These levels do not exceed the replacement PRNM electronics qualified dose rate of 0.036 Rads /hr [1 E-5 Rads (carbon)/s], and a total integrated dose of 1 E+4 Rads (carbon) for nonnal conditions and 3.0 E+1 Rads (carbon) for accident conditions. HL-5054 A-3 I
. - . _ _ __ _ _ . . _-. _.~._._ _ _ _ . _ _ _ _ m._ _ _
1 ! ' Attachment to Enclosure 1 A l } NUMAC PRNM LTR NEDC-32410P-A
- Plant-Specific Information ;
i ) LTR l l Section LTR Utility Action Required GPC Response i , 4.4.2.3.4 " Plant-specific action or analysis will As part of the Plant Hatch design process, j confirm that the maximum scismic the mounting configuration of the ; i accelerations at the mounting locations of replacement PRNM equipment was the equipment (control room floor evaluated to the qualification type test and l
- acceleration plus panel amplification) for subject to the floor response spectra of the l both OBE and SSE spectrums do not exceed Plant Hatch control room. This analysis ;
l the above limits [ stated in the PRNM LTR]. determined the installation does not exceed l- Documentation of this action and the the seismic qualification of the electronics.
- required confirmation should be included in !
] plant-specific licensing submittals." 4 4 j 4.4.2.4.4 "The utility should establish or document As part of the Plant Hatch design process for l j practices to control emissions sources, digital upgrades, EMI conditions are : t maintain good grounding practices and evaluated using EPRI Guideline TR-102323 J f maintain equipment and cable separation. as guidance. This process is complete for ; ] the replacement PRNM and determined not l to exceed EMI qualification of the ! f electronics. . { 1) Controlline Emissions ; i ! i a) Portable transceivers (walkie-talkies): a) The qualification levels used for the , j Establish practices to prevent NUMAC PRNM exceed the levels I j operation of portable transceivers in expected to result from portable j
- close proximity of equipment transceivers, even if such transceivers are j sensitive to such emissions. operated immediately adjacent to the l
! NUMAC equipment. In addition, Plant l l Hatch procedures restrict the use of l
- portable transceivers in the main control
{ room where the NUMAC equipment will ] be located. i ! b) ARC welding: Establish practices to b) The qualification levels used for the ! assure that ARC welding activities do NUMAC PRNM minimize the likelihood not occur in the vicinity of equipment of detrimental effects due to ARC sensitive to such emissions, welding as long as reasonable ARC particularly during times uhen the welding control and shielding practices i potentially sensitive equipment is are used. ! required to be operational for plant l safety. ! c) Limit emissions from new equipment: c) EMI qualification of the NUMAC PRNM Establish practices for ucw equipment included emissions testing, both and plant modifications to assure that conducted and radiated. Design they either do not produce procedures state that all digital systems to unacceptable levels of emissions. emissions of EMIin accordance with l
)
HL-5054 A-4 J l l
i Attachment to Enclosure 1 A ' NUMAC PRNM LTR NEDC-32410P-A Plant-Specific Information j l LTR I Section LTR Utility Action Required GPC Response 4.4.2.4.4 or installation of shiciding, filters, EPRI Guideline TR 102323. In lieu of EMI l (Cont.) grounding, or other methods prevent such testing for nonsafety equipment, high I' emissions from reaching other potentially frequency EMI emitters require RFI filtering sensitive equipment, particularly conducted on the AC power and shielded cables for all I emissions on power line; and power signal leads. distribution systems. Related to power distribution, both the effects of new equipment injecting noise on the power system and the power system conducting noise to the connected equipment should be l addressed. l
- 2) Groundine Practicpji a) Existing grounding system: The a) The PRNM retrofit will be installed to the specific details and effectiveness of same panel interfaces as the current PRM the original grounding system in system. No problems have been BWRs varied significantly. As part of identified with the current system. The the modification process, identify any PRNM hardware is less se.sitive to EMI known or likely problem areas based than the existing electconics. installation on previous experience and include in specifications for the new PRNM the modification program either an provides effective grounding measures.
cvaluation step to determine if problems actually exist, or include corrective action as part of the modification. b) Grounding practices for new b) NUMAC PRNM equipment qualification modifications: New plant was performed in a panel assembly I modifications process should include comparable to the plant panel. l a specific evaluation of grounding methods to be used to assure both that the new equipment is installed in a way equivalent to the conditions used in qualification.
- 3) Equipment and Cable Separation a) Cabling: Establish cabling practices a) The original PRM cable installation met ;
to assure that signal cables with the this criterion. The replacement PRNM l potential to be " receivers" are kept uses the same cable routes and paths as separate from cables that are sources the original installation. Since there are ; of noise. no specific problems identified with the existing system, no problems are enticipated. HL-5054 A-5
Attachment to Enclosure 1 A . l [' NUMAC PRNM LTR NEDC-32410P-A Plant-Specific Information l l 5 ~ LTR j Section LTR Utility Action Required GPC Response l j 4.4.2.4.4 b) Equipment: Establish equipment b) The replacement PRNM clectronics will j (Cont.) separation and shielding practices for be mounted in a panel similar to the i the installation of new equipment to mounting assembly the qualification was
, simulate that equipment's performed.
4 qualification condition, both relative < to susceptibility and emissions. The plant-specific submittal should identify the practices that are in place or will be applied for the PRNM modification to address each of the above items." 6.6 "The utility must confirm applicability of the above [ failure analysis] conclusions [ stated in the PRNM LTR] by: (1) confirming that the events in EPRI (1) The events in NEDC-30851P A i Report No. NP-2230 or Appendices F (Ref. I1 of the PRNM LTR), ; and G of Reference 1I [LTR Appendices F and G, encompass the l NEDC-30851P-A, " Technical events analyzed in the Unit I and l Specification Improvement Analysis for Unit 2 Final Safety Analysis Reports l BWR Reactor Protection System," dated and the Safe Shutdown Analysis ; March 1988] encompass the events that Reports. are analyzed for the plant, (2) confirming that the configuration (2) The replacement PRNM configmation implemented by the plant is within the is within the limits described in the limits described in this report [the LTR, as discussed in the response to PRNM LTR), and LTR Section 2.3.4. a (3) preparing a plant-specific 10CFR50.59 (3) As part of the normal Plant Hatch design evaluation of the modification per the process,10 CFR 50.59 safety evaluation applicable plant procedures." of the proposed PRNM replacement i modification has been documented in l accordance with applicable procedures. ! These confirmations and conclusions should be documented in the plant-specific licensing
]
submittal for the PRNM modification." l l 7.6 "The specific action [ required for FSAR As part of the HNP design process, the updates) will vary between plants. In all FSAR is reviewed and appropriate cases, however, existing FSAR documents revision (s) are made once the modification is should be reviewed to identify areas that complete. The revisions are submitted to the have descriptions specific to the current NRC in accordance with 10 CFR 50.71.(c). PRNM using the general guidance of HL-5054 A-6
- .m..
I Attachment to Enclosure 1 A . NUMAC PRNM LTR NEDC-32410P-A ! Plant-Specific Information 4 : , LTR . I 3 Section LTR Utility Action Required GPC Response j 7.6 (Cont.) Sections 7.2 through 7.5 to identify potential i 1 areas impacted. The utility should include in t the plant-specific licensing submittal a statement of the plans for updating the plant !
- FSAR for the PRNM project." i
- 9.1.3 "As part of the plant specific licensing The PRNM retrofit project activities comply submittal, the utility should document the with the applicable requirements of the
! established program that is applicable to the Hatch Quality Assurance Program based on 4 project modification. The subruittal should 10 CFR 50, Appendix B. j also document for the project what scope is j being performed by the utility and what For the PRNM retrofit project GE is under scope is being supplied by others. For scope contract to GPC to piovide engineering, { supplied by others, document the utility licensing, software development and testing,
- actions taken or planned to define or training, design documentation, field
- establish requirements for the project, to installation instructions, and hardware with l assure those requirements are compatible qualification reports. GE is a qualified !
{ with the plant specific configuration. supplier to GPC and their Appendix B j Actions taken or planned by the utility to QA program is periodically audited. j assure compatibility of the GE quality program with the utility program should also As part of the Plant Hatch design process, a be documented. design package integrating vendor design has been prepared, using applicable design Utility planned level of participation in the procedures, and provided to the Plant prior ] overall V&V [ verification and validation) to implementation. Site installation will be process for the project should be controlled by applicable Plant Hatch documented, along with utility plans for procedures. GPC has participated with GE software configuration management and in establishing design requirements and the provision to support any required changes V&V program for the Hatch PRNM project. after delivery should be documented." GPC representatives witnessed the initial V&V and participated with GE in the factory acceptance tests on hardware for both units. Firmware provided by GE will be controlled at the plant under applicable site procedures. GE will maintain control of any firmware changes required after initial delivery. , i l l HL-5054 A-7
Attachment to Enclosure 1 A NUMAC PRNM Licensing Topical Report NEDC-32410P-A Plant-Specific Information
.B f
k i l HL-5054 A-1
Enclosure IB Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications: Oscillation Power Range Monitor Basis for Change Reauest
Background
The following proposed Technical Specifications changes suppon the Oscillation Power Range hionitor (OPRhi) as the long-term stability solution. The proposed changes to the Unit 1 Technical Specifications are described. Unless otherwise noted, the same changes are proposed for the Unit 2 Technical Specifications. As discussed in the cover letter, the proposed OPRhi changes are requested to be made effective 6 months after the PRNhi changes are made effective. The proposed changes, which incorporate the OPRhi features as part of the Average l Power Range hionitor (APRhi) retrofit described in Enclosure I A, are consistent with l General Electric Licensing Topical Report (LTR) NEDC-32410P-A and Supplement 1, " Nuclear hieasurement Analysis and Control Power Range Neutron hionitor (NUMAC-PRNhi) Retrofit plus Option III Stability Trip Function." The Option III stability algorithms are described in BWR Owners' Group LTR NEDO-31960-A and Supplement 1, "Long-Term Stability Solutions Licensing hiethodology," and BWR Owners' Group LTR NEDO-32465-A, " Reactor Stability Detect and Suppress Solutions Licensing Basis hiethodology and Reload Applications." The proposed changes to Technical Specification 3.4.1, Recirculation Loops Operating, are consistent with NUREG 1433. s 1 Technical Specifications Proposed Chances P_ROPOSED CHANGF,_1 In Technical Specification 3.3.1.1, " and Function 2.f" is added to the Note in Required Action A.2 and to the Note in Condition B. Both Notes read:"Not applicable for Functions 2.a,2.b,2.c,2.d, and 2.f." Basis for Proposed Change 1 Required Action A.2," Place associated trip system in trip," is not applicable to Function 2.f, since each OPRhi provides signals to both RPS trip systems. Condition B is not applicable to Function 2.f. since each OPRhi provides signals to both RPS trip systems. l l HL-5054 ElB-1
l 4 Enclosure IB Oscillation Power Range hionitor i Basis for Change Request l 1 PROPOSED CHANGE 2 In Technical Specification 3.3.1.1, the following two Conditions, including Required Actions and Completion Times, are added: . A. Condition I, which reads: "As required by Required Action D.1 and referenced in Table 3.3.1.1-1." Required Action I.1 states:" Initiate alternate method to detect and suppress thermal-hydraulic instability oscillations," and has a Completion Time i of 12 hours. Required Action I.2 states:" Restore required channels to OPERABLE," and has a Completion Time of 120 days. B. Condition J, which reads: " Required Action and associated Completion Time of Condition I not met." Required Action J.1 states: "Be in hiODE 2", and has a
- Completion Time of 4 hours.
Ruis for Proposed Change 2 Condition I allows an alternate method to detect and suppress thermal-hydraulic instability. Recognition of the OPRh1 trip capability is a new Function; thus, it is remotely possible that experience may reveal some problem with the algorithm and/or . implementation. The contingent alternate method will meet the requirements of the BWR Owners' Group Interim Corrective Actions (ICAs) outlined in the letter to the NRC dated
- June 6,1994. The inclusion of the proposed Action Statement pre-plans for such a contingency with an established alternate method and requires OPRhi OPERABILITY restored within 120 days.
If Condition I is not met, Condition J requires the plant to be in hiODE 2, a safe power level below the area ofinstability. PROPOSED CHANGE 3 In Table 3.3.1.1-1 (page 2 of 3), APRhi Function 2.f, OPRh1 Upscale, is added. The following information is included under the column headings: The Applicable hiodes or Other Specified Conditions is h10DE 1. The Number of Required Channels per Trip System is "3", with a footnote stating: "Each APRM channel provides inputs to both trip systems." The Condition referenced from Required Action D.1 is " Condition 1." The Surveillance Requirements are SR 3.3.1.1.1, SR 3.3.1.1.8, SR 3.3.1.1.10, SR 3.3.1.1.13, and SR 3.3.1.1.17. There is no Allowable Value for this Function. HL-5054 ElB-2
1 Enclosure IB . Oscillation Power Range Monitor i- Basis for Change Request ' Basis for Proposed Change 3 f. i Only the period based detect and suppress algorithm is used as the basis for the safety } analysis for the OPRM Upscale Function. The other two algorithms discussed in the j Technical Specifications Bases provide defense in depth, but are not required for OPRM j Upscale OPERABILITY. The consideration of the OPRM Upscale as a sub-function of ;
- the APRM has no effect on the actual trip function. ,
i i Due to the integrated nature of the OPRM Function in the APRM channel, the OPRM j Inop Function and the Two-out-of-Four Voter Function are included with the
- corresponding APRM Inop and APRM Two-out-of-Four Voter Function. The integration
- of the OPRM Inop with the APRM Inop reflects actual system design. Unlike the APRM i trips, the OPRM Upscale trip is voted separately from the Inop Trip Function in the Two- !
l out-of-Four Voter Function. That is, an APRM/OPRM Inop trip in one APRM channel t and an OPRM Upscale trip in another will result in two half-trips in each of the two-out- ! of-four voter channels, but no RPS trip. Conversely, an Inop trip in any two APRM , [ channels or an OPRM Upscale trip in any two channels will result in RPS trip outputs l l from all 4 two-out-of-four voter channels. i 1 i- For the APRM Flux Trip Function, an APRM/OPRM Inop in one APRM channel and an APRM Upscale trip in another channel will result in RPS trip outputs from all four voters. This reflects a somewhat more conservative APRM system design relative to response to channel failures when' compared to the OPRM design. The APRM design results in an immediate trip (in that channel) upon detection of a failure. However, the unavailability analysis supporting the Technical Specifications LCO times assumes the channel may be out of service for up to 12 hours. This additional conservatism is oflimited value in the OPRM design. Due to the uncertainties in the exact performance of the OPRM algorithms in normal plant conditions, the algorithms could, ifimplemented for the OPRM, result in unnecessary spurious trips. However, an automatic trip will occur upon an unexpected systematic failure of multiple APRM channels. This will result in an APRM/OPRM Inop in two or more non-bypassed channels, regardless of the OPRM Upscale (or APRM flux) trip status. Independent of the APRM/OPRM Inop logic, which originates in the APRM channel, a loss of communication from an APRM channel to a voter channel will result in both the MRM and the OPRM voting logic in the two-out-of-four voter channel declaring the inputs from that APRM channel inoperative. The condition is alarmed via the two-out-of- i four voter self-test diagnostics. A loss of communication may be the result of either a i hardware failure (may affect input to one or more voters) or a loss of power to the APRM channel (affects input to all voters). Loss of power to the two-out-of-four voter channel will result in immediate RPS trip outputs from that voter. l HL-5054 El B-3
-. . . . - . - - - . - . - - - - - - . - - - - - - . - - . . ~ ~_ .. . . .
l 1 l Enclosure IB . j Oscillation Power Range Monitor
- Basis for Change Request p
i l Combining the OPRM Function and the APRM Two-out-of-Four Voter Function into a j single Function simplifies overall operation and the decision-making process, because most 3 conditions affecting OPERABILITY of the voter channel will affect both the APRM trip
- voting and the OPRM trip voting logic. However, the final voting and output relays (from j the voter) for these two Functions are different. Furthermore, the output relays for both
, Functions are redundant;i.e., two relay outputs for the APRM trips and two additioaal j relay outputs for the OPRM trips. Even though there is only one voter channel for both
- the APRM and the OPRM trips, the LCO clock will start as soon as any portion of a voter i channel is determined to be inoperable Trip capability for both the APRM and the OPRM l Functions is still maintained. The Required Actions are in accordance with the Technical
! Specifications for each inoperable voter channel. Actions associated with the loss of trip l capability are not required. Consistent with the APRM Neutron Flux - High Function, the OPRM Upscale Function is ! required only when the plant is operating in the Run Mode (MODE 1). In addition, the i OPRM Upscale is bypassed automatically when THERMAL POWER is below 25% RTP l (as indicated by APRM Simulated Thermal Power) or with flow above 60% rated core I j flow (as indicated by drive flow). In the region below 25% RTP and above 60% rated
- core flow, theimal-hydraulic instabilities are not considered credible. The 25% RTP
)
. provides additional margin from the nominal 30% RTP OPRM scram enable setpoint j determined by the BWROG (ref BWROG letter to the NRC, " Guidelines for Stability l Option III ' Enabled Region' (TAC M92882)" dated September 17,1996). This additional ; i margin accounts for the previous Hatch power uprate and for any future uprates up to
- 10%. !
4 Identified events can take the plant from 100% power and flow to < 60% flow without ! operator action. Other events can take the plant from < 25% RTP to a power greater than { 1 25% RTP without operator action. Therefore, even though the OPRM Upscale Trip is !
- bypassed at > 60% flow and < 25% RTP, the Function must be OPERABLE so that if one 4
of the identified events occurs, the OPRM Upscale trip capability is immediately available { without operator intervention. Requiring OPRM OPERABILITY at MODE 1 provides l adequate margin to cover the operation region where oscillations may occur, and the l operation region from which the plant might enter the region where instabilities may occur j j without operator action. i l The outputs of the OPRM channels are shared by each RPS trip system via the independent two-out-of-four voter channels. Any two of the four OPRM channels and j one of the two-out-of-four voter channels in each RPS trip system are required to function for the OPRM Upscale Function to be accomplished. Therefore, a minimum of three i OPRM channels assures at least two OPRM channels can provide trip inputs to the two-l out-of-four voter channels, even in the event of a single OPRM channel failure. The j minimum of 2 two-out-of-four voter channels per RPS trip system assures at least one HL-5054 E1 B-4 i
Enclocure IB Oscillation Power Range Monitor I Basis for Change Request voter channel will be OPERABLE per RPS trip system, even in the event of a single voter channel failure. , The Two-out-of-Four Logic Module is designed for simplicity to assure high reliability and detect loss ofinput signals from the OPRM channels (dynamically encrypted to assure no passive fault at the interface will go undetected). This feature, combined with the highly reliable digital electronics implementing the OPRM Function and the on-line automatic self-test functions, assures the four-channel OPRM configuration will provide t reliability, relative to the safety trip functions, equal to or greater than the current APRM l system. This level of reliability is adequate for the OPRM Upscale Function. l l Since the OPRM Upscale Function is implemented in the same equipment as the APRM Function, equipment reliability is also the same. The OPRM Surveillance Requirements l CHANNEL CHECK, LPRM calibration, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION are the same as the APRM Functions proposed for the APRM retrofit. The expected demand for the OPRM trip is equal to or less than the demand for the APRM Functions. Therefore, the OPRM Surveillance Requirements are l adequate. l l l PROPOSED CHANGE 4 l SR 3.3.1.1.17," Verify OPRM is not bypassed when APRM Simulated Thermal Power is
> 25% and recirculation drive flow is < 60% of rated recirculation drive flow," is added.
The Surveillance Frequency is 18 months. l Basis for Proposed Change 4 i SR 3.3.1.1.17 is added to provide verification that the OPRM Upscale is enabled when ! APRM Simulated Thermal Power is ;t 25% and recirculation flow is < 60% rated flow. i The " auto-enable" region is determined by Simulated Thermal Power and drive flow setpoints in the APRM channels. Even though these setpoints are unlikely to change once set, periodic confirmation is appropriate. Other Required Actions verify the relationship l between THERMAL POWER and APRM Simulated Thermal Power, and core flow and l recirculation flow are within acceptable tolerances. The combined Surveillances ensure the OPRM Upscale trip is enabled in the intended region on the plant power / flow map. The 18-month Frequency is based on engineeringjudgment and the fact that the actual values are stored digitally, with no drift. Any hardware failures affecting the Simulated Thermal Power and recirculation drive flow setpoints will likely be detected by the automatic self-test function. HL-5054 ElB-5 t
l Enclosure IB Oscillation Power Range Monitor Basis for Change Request , Based on the above discussion, adding the OPRM Upscale Function to the Technical Specifications is reasonable and consistent with the instability detect and suppress i objectives. ER_ OPOSED CHANGE 5 Limiting Condition for Operation 3.4.1.b is removed to reflect deletion of Figure 3.4.1-1, )
" Power-Flow Operating Map with One Reactor Coolant System Recirculation Loop in Operation." The corresponding Action B of this LCO and SR 3.4.1.2 are deleted.
l The LCO Action A, which requires the mode switch to be placed in shutdown with no recirculation loops in operation is also deleted. The remaining LCO and Action ] Statements are renumbered as appropriate. Basis for Proposed Change 5 I Both the power-to-flow map for single loop operation (SLO) and the requirement for ) immediate shutdown with no recirculation pumps in operation were Technical l Specifications requirement pHor to the LaSalle thermal-hydraulic oscillation event in 1988.
]
r Both rcquirements are not contained in NUREG-1433 (the BWR/4 Improved Technical Specifications). Although exhaustive search with similar BWR plants was not performed, GPC believes these two Technical Specifications requirements are unique to Plant Hatch. The proposed changes described above are now identical to NUREG-1433. Approximately 10 years ago, Figure 3.4.1-1 was placed in the Technical Specifications when the NRC approved Unit 1 Amendment 141 and Unit 2 Amendment 77 to allow continuous SLO. The figure provided a restricted region of operation during SLO and was an alternative to placing certain recommendations of GE Service Information Letter (SIL) 380 in the Technical Specifications. The requirement to place the mode switch in shutdown immediately after a loss of both recirculation pumps has been in the Unit I and Unit 2 Technical Specifications for many years. The Technical Specifications of other plants similar to Plant Hatch provide several hours to restart a recirculation pump before entry into MODE 3 is required. The two requirements (i.e., SLO restricted region and immediate shutdown requirement l with no pumps operating) were retained in the Technical Specifications during the I conversion to the Improved Technical Specifications for the following reasons.
- l. The thermal-hydraulic stability issue, which arose from the LaSalle event in 1988, was not fully resolved.
i l 1 HL-5054 ElB-6 I
._ _ _ _ _ _ . _ - _ . _ . _ _ _ _ _ _ _ _ _ _ . __.. _ .._ _ .-_ _..__ _ ~ _ __
i i Enclosure IB . l Oscillation Power Range Monitor Basis for Change Request !
- 2. Both Unit I and Unit 2 operated (and continue to operate) under procedural Interim ;
Corrective Actions (ICAs) for thermal hydraulic stability. The ICAs restrict l operation even with hplh recirculation pumps in operation to a larger area of the power-to-flow map than Figure 3.4.1-1; i.e., the ICAs are more restrictive than the - Technical Specification figure).
- 3. The ICAs have procedurally required all BWRs similar to Plant Hatch to shut down !
immediately if a duel recirculation pump inp occurs. As stated in the cover letter, some or all of the ICAs will be removed once the Option III OPRM is installed and its Scram Function enabled. The OPRM system will detect and , automatically suppress any significant core wide or regional oscillations over a large l portion of the power-to-flow map. This automatic function provides more protection than ; these two requirements proposed for deletion. Having the ability to insert control rods and restart a recirculation pump (instead of shutting down immediately) following a recirculation pump trip is important for plant availability. This capability is one of the ; reasons GPC chose Option III over other less expensive long-term solutions for thermal- ! hydraulic stability. l HL-5054 ElB-7
Enclosure 2 Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications: Power Range Neutron Monitor Retrofit and Oscillation Power Range Monitor 10 CFR 50.92 Evaluation In 10 CFR 50.92(c), the NRC provides the following standards to be used in determining the existence of a significant hazards consideration:
..a proposed amendment to an operating license for a facility licensed under 50.21(b) or 50.22 for a testing facility involves no significant hazards i consideration if operation of the facility in accordance with the proposed )
amendment would not: (1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Crcate the possibility l of an accident of a new or different kind from any previously evaluated; or (3) Involve a significant reduction in the margin of safety. Georgia Power Company has reviewed the proposed license amendment request and , determined its adoption does not involve a significant hazards consideration based on the I I following discussion. Basis for no sienificant hazards considerine determination i
- 1. The proposed changes do not involve a sigmficant increase in the probability or consequences of an accident previously evahtated.
The purpose of the proposed amendment is to incorporate the Power Range Neutron , Monitoring (PRNM) retrofit and Oscillation Power Range Monitor (OPRM) l installation. The types of Average Power Range Monitor (APRM) Functions that are credited to mitigate accidents were previously evaluated. The proposed OPRM Upscale Function is implemented in the same hardware that implements the APRM Functions. The change to a two-out-of-four RPS logic was analyzed and determined to be equal to the originallogic. The modification involves equipment that is intended to detect the symptoms of some accidents and initiate mitigating action. The worst case failure of the equipment involved in the modification is a failure to initiate mitigating action (sc:am), but no failure can cause an accident. As discussed in the bases for proposed changes, the PRNM replacement system is designed to perform the same operations as the existing Power Range Monitoring (PRM) system and to meet or exceed all of its operational requirements. Therefore, it is concluded that the probability of an HL-5054 E2-1
Enclosure 2 10 CFR 50.92 Evaluation accident previously evaluated is not increased as a result of replacing the existing equipment with the PRNM equipment. The installation of the OPRM stability hardware provides the long term solution required by Generic Letter 94-02. This hardware incorporates the Option III detect and suppress solution reviewed and approved by the NRC in NEDO-31960, "BWROG Long Term Stability Solutions Licensing Methodology." The OPRM system is designed to meet all the requirements in GDC 10 and GDC 12 by automatically detecting and suppressing all design basis thermal-hydraulic instabilities prior to violating the MCPR Safety Limit. Since the OPRM system provides this protection in the region of the power / flow map where instabilities are most likely to occur, the current Interim Corrective Action (ICA) restrictions on plant operations 4 are deleted from the Technical Specifications, including region avoidance during ] single loop operation and the requirement for the operator to manually scram with no recirculation loops operating. Operation at high core powers with low core flows may cause a slight, but not a significant, increase in the probability an instability can occur.. This slight increase is acceptable, because, subsequent to the detection of a design basis instability, the OPRM system provides an automatic trip signal to the Reactor Protection System (RPS) which is faster protection than the operator- J initiated manual scram required by the current ICAs. Thus, as a result of this rapid automatic action, the consequences of an instability event are not be increased as a result of the instaliation of the OPRM system because it eliminates operator actions. 1 The PRNM reduces the need for tedious operator actions during normal conditions and allows the operator to focus more on overall plant conditions. The automatic self-test and increased operator information provided with the replacement system are likely to reduce th: burden during off-normal conditions. The replacement equipment qualifications fully envelope the environmental conditions, including electromagnetic interference (EMI), in the Plant Hatch control room. The replacement equipment was specifically designed to assure that it fully meets the response time requirements in the worst case. As a result, due to statistical variations resulting from the sampling and update cycles, the response time is typically faster than required to assure the required response time is always met. The replacement equipment includes up to five LPRM inputs on a single module compared to one per module on the current system. Up to 17 LPRM signals are processed through one preprocessor. The recirculation flow signals are processed in the same hardware as the LPRM processing. The net efTect of these architectural aspects is that there are some single failures that can cause a greater loss of"sub-functionality" than in the current system. Other architectural and functional aspects, however, have an offsetting effect. Redundant power supplies are used so that a HL-5054 E2-2
l l Enclosure 2 10 CFR 50.92 Evaluation l l l single failure of AC power has no effect on overall PRNM Functions, while still ' resulting in a half scram as does the current system. The continuous automatic self-test also assures that if a single failure does occur, it is much more likely to be l detected immediately. The net effect is that from a total system level, there is no increased risk ofloss of critical functionality and reduction in safety margins due to the architecture of the replacement system. 1 Common-cause failures in software controlled functions arejudged to not be a significant failure mode. However, in spite of that conclusion, means are provided within the system to mitigate the effects of such a failure and alert the operator. Therefore, such a failure, even ifit occurred, will not increase the consequences of a previously evaluated accident. To reduce the likelihood of common-cause failures of software controlled functions, thorough and careful verification and validation (V&V) activities were performed both for the requirements and the implementing software design. In addition, the software is designed to limit the loading that external systems or equipment can place on the system, thus significantly reducing the risk that some abnormal dynamic condition external to the system can cause an overload. For conservatism, however, despite these V&V activities, common-cause failures of software controlled functions due to residual software design faults are assumed to occur. Both the software and hardware are designed to manage the consequences of such failures (also cover potential common-cause hardware failures). Safety outputs are designed to be fail safe by requiring dynamic update of output modules or data signals, where failure to update the information is detected by simple receiving hardware, which in turn, forces a trip. This aspect covers all but rather complex failures where the software or hardware executes a portion of the overall logic but fails to process some portion of new information (inputs " freeze") or some portion of the logic (outputs " freeze"). To help reduce the likelihood of complex failures, a watchdog timer is used which is updated by a very simple software routine that in turn monitors the operational cycle time of all tasks in .he system. The software design is such that as long as all tasks are updating at the design rate, it is likely that software controlled functions are executing as intended. Conversely, if any task fails to update at the design rate, that is a strong indication of at least some unanticipated condition. If such a condition occurs, the watchdog timer will not be updated, the computer will be restarted, and the outputs will detect an abnormal condition and provide an alarm. The information available to the operator is the same as with the current system. No actions are required by the operator to obtain information normally used and equivalent to that available with the current equipment. However, the replacement HL-5054 E2-3 10 CFR 50.92 Evaluation system does provide more directly accessible information regarding the condition of the equipment, including automatic self-test, which can aid the operator in diagnosing unusual situations beyond those defined in the licensing basis. The replacement system has a significantly lower power requirement and as a result also reduces the heat load in the mounting cabinet. The replacement equipment is generally smaller and reduces somewhat the seismic loading on the panels. The equipment qualification also includes Ehil emissions which, combined with the fact that the replacement equipment is mounted in its own cabinet (replaces all of the current equipment), minimizes the likelihood of significant impact on other existing equipment. The replacement equipment makes increased use of optical methods to provide both safety and functional isolation between safety-related and nonsafety-related systems. Where fiber optic methods cannot be used, the isolation provided is comparable to or better than that provided in the current system. Software common-cause failures arejudged to not be a significant failure mode. However, it is very diflicult to quantify a sof1 ware common-cause failure rate. Analyses for the current system did consider common-cr e failures and assessed them to be at a rate of about 0.3 times the random faili s rate. The reference analysis uses a field basis for the random rates. The analysis for the replacement design uses conservative estimates for failure rates of equipment that are actually a little higher than those assumed for the current equipment. The methodology being applied concludes that the common-mode failure rate for the replacement system is somewhat higher than the current system. However, that is offset by more frequent surveillance tests performed by the self-test that result in an estimated slightly lower unavailability for the PRNhi Scram Function compared to the current PRhi system. The FSAR, in general, considers the failure rate of the Function, not that of sub-components. On that basis, there does not appear to be an increase, due to software common-cause failure, in the probability of a malfunction analyzed in the FS AR. The net electrical and thermal loads for the replacement system are less than those for the current system. The replacement system meets or exceeds all applicable requirements. The use of fiber optic connections between the APRh1 and RBhi improves the separation and reduces the dependence of the system on common grounds. However, the noise rejection, the equipment design, and manufacturing requireinents assure improved grounding of the actual equipment. No change in wiring or grounding external to the panels containing the replacement equipment is necessary for correct operation of the replacement equipment. HL-5054 E2-4
Enclosure 2 10 CFR 50.92 Evaluation The replacement design has a lower heat load than the current equipment, and is movated in the same panels. No forced cooling is required. The replacement design was specifically designed to have the same or more conservative " fail safe" failure modes as the current system. For example, in the case of a single power bus failure, the current system loses about one half of the LPRM information and an output trip occurs. For the replacement system, that failure still results in an output trip, but no LPRM information is lost. In the current system, a static failure in several areas in the system could result in a " fail-as-is" state of the outputs. In the replacement system, dynamic coupling starting in the main processor and going to the final output virtually eliminates " fail-as-is" failure modes and replaces them with " fail tripped" modes. Software common-cause failures arejudged to not be a significant failure mode. Due to design aspects that are intended to " protect" the digital system from excessive loading triggered by external events, it is unlikely that a software common-cause failure will occur concurrently with other events assumed in the safety analysis. However, even if that were to occur, the hardware and software aspects discussed previously are likely to mitigate any adverse consequences resulting in at most a safe failure of the PRNM. The replacement system has the same loss-of-power failure mode as the current system relative to the trip outputs and for loss of AC power. For loss of DC power, the replacement system in most cases continues to operate normally due to redundancy of the power supplies. Therefore, the consequences are at least as good or better than those considered in the FSAR. Both the current system and the replacement system automatically startup on application of power (or re-application). However, the replacement system may take 1 slightly longer to reach normal operation due to initializing activities. However, no FSAR evaluations take credit for rapid start of the PRM. Therefore, the slightly longer startup time from point of power application is bounded by the FSAR I analysis. Upon application of power, once the system is set up for the specific application (plant), it automatically returns to those settings upon application of power. All such setup parameters are stored in non-volatile memory. j Human-machine interface (HMI) failures in the current system could be related to l incorrectly adjusted settings, incorrect reading of meters, and failure to return the equipment to the normal operating configuration. There are comparable failure modes for some of these problems in the digital system where an erroneous ; potentiometer adjustment in the current system is equivalent to an erroneous digital ! HL-5054 E2-5
Enclosure 2 . 10 CFR 50.92 Evaluation l l entry in the replacement system. Certain potential" failure to reconfigure errors"in , the current system have no counterpart in the replacement system, because any !
" reconfiguration" is automatically returned to normal by the system. Also, since l parameters are available for review at any time, even if an error, such as a digital entry error occurs, it is more likely that the error would be almost immediately detected by recognition that the displayed value is not the correct one.
The failure analysis of the current system assumes certain rates of human error. The l rates for the replacement system will be lower and, hence, are bounded by the FSAR analysis. 1 I Therefore, GPC concludes the proposed changes do not involve a significant increase I in the probability or consequences of an accident previously evaluated.
- 2. The proposed changes do not create the possibility of a new or different kind of accidentfrom any accidentpreviously evaluated.
The APRM Trip Functions credited in the accident analyses are retained in the PRNM retrofit. The response time of the new electronics meets or exceeds the required response criteria. No new interfaces or interactions with other equipment c will introduce any new failure modes. The modification involves estuipment that is intended to detect the symptoms of some accidents and initiate mitigating action. The worst-case failure of the equipment involved in the modification is a failure to initiate mitigating action (scram), but no failure can cause an accident. This is unchanged from the current I system. Sonware common-cause failures can at most cause the system to fail to perform its safety function. In that case, it could fail to initiate action to mitigate the consequences of an accident, but would not cause one. The new system is a digital system with sonware (firmware) control. As such, it has
" central" processing points and sonware controlled digital processing where the current system had analog and discrete component processing. The result is that the specific failures of hardware and potentially common-cause sonware failures are different from the current system. Also, automatic self-test results in some cases in a direct trip as a result of a hardware failure where the current system may have remained "as-is" However, when these are evaluated at the system level, there are no new effects. In general, FSARs assume simplistic failure modes (relays for ; example) but do not specifically evaluate such effects as self-test detection and automatic trip or alarm.
HL-5054 E2-6
- l Enclosure 2 !
10 CFR 50.92 Evaluation l The effects of software common-cause failure are mitigated by hard vare design and
- system architecture. The replacement equipment is fully qualified to operate in its (
- installed location and will not affect other equipment.
l Therefore, GPC concludes the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. l 1
- 3. The proposed changes do not irwoh e a sigmjicant reduction in a margin ofsafety. ,
! The replacement equipment provides the same function as the original electronics. Response time and operator information are either maintained or improved. The equipment was qualified, where appropriate, to assure its intended safety function is ! performed. i The replacement system has improved channel trip accuracy compared to the current i
- system and meets or exceeds system requirements assumed in setpoint analysis. The ,
i channel response time exceeds the requirements. The channel indicated accuracy is j improved over the current system, and meets or exceeds system requirements. The -
- replacement system meets or exceeds all system requirements.
i j The BWROG Stability Option III was developed to meet the requirements of
- GDC 10 and GDC 12 by providing a hardware system that detects the presence of j thermal-hydraulic instabilities and automatically initiates the necessary actions to suppress the oscillations prior to violating the MCPR Safety Limit. The NRC has ,
4 reviewed and accepted the Option III methodology described in Licensing Topical i Report NEDO-31960 and concluded this solution will provide the intended
- protection. Therefore, it is concluded that there will be no reduction in the margin of safety as defined in the Technical Specifications as a result of the installation of the
- OPRM system and the simultaneous removal of the operating restrictions imposed by
- the ICAs.
i ! Therefore, GPC concludes the proposed changes do not involve a significant i reduction in a margin of safety. s l Conclusion i I Based on the preceding analysis, GPC has determined the proposed changes to the Technical Specifications will not significantly increase the probability or consequences of i an accident previously evaluated, create the possibility of a new or different kind of l j accident from any accident previously evaluated, or involve a significant reduction in a l
' I margin of safety. Therefore, GPC has determined the proposed changes meet the requirements of 10 CFR 50.92(c) and thus, do not involve a significant hazards consideration.
4 HL-5054 E2-7 1
.. _ - = . - _ . - - . . . - . - . . - . . . . . . . . . - . - .- . _ .~ ~ - - . .
Enclosure 3A . 4 Edwin I. Hatch Nuclear Plant !
- Request to Revise Technical Specifications
l Power Range Neutron Monitor Retrofit 1 Pane Chance Instructions and Revised Pages j Unit I f Remove Page Insert Page ! 3.3-1 3.3-1 3.3-2 3.3-2 j 3.3-3 3.3-3 1 3.3-4 3.3-4 1 3.3-5 3.3-5 i
- 3.3-6 3.3-6 3.3-7 3.3-7 i 3.3-16 3.3-16 3.3-19 3.3-19 4 3.4-1 3.4-1 3 3.10-20 3.10-20 i 3.3-22 3.3-22 Unit 2 l 1,
j Remove Page Insert Page , 3.3-1 3.3-1
- 3.3-2 3.3-2 1 3.3-3 3.3-3 3.3-4 3.3-4 I 3.3-5 3.3-5 3.3-7 3.3-7 3.3-8 3.3-8 3.3-17 3.3-17 3.3-20 3.3-20 3.4 1 3.4-1 3.10-20 3.10-20 3.3-22 3.3-22 I
l 1 l HL-5054 E3A-1
l RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTAT'ON , 3.3.1.1. Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1. 1 , i ACTIONS ! i
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required. A.1 Place channel in 12 hours channels inoperable. trip. . I QE A.2 --------NOTE--------- I Not applicable for i Functions 2.a, 2.b, ! 2.c, and 2.d. i Place associated trip 12 hours ; system in trip. ! i B. --------NOTE---------- B.1 Place channel in one 6 hours ! Not applicable for trip system in trip. ! Functions 2.a, 2.b, : 2.c, and 2.d. QB ! B.2 Place one trip system 6 hours ! One or more Functions in trip. . with one or more ! required channels > inoperable in both , i trip systems. (continued) ! HATCH UNIT 1 3.3-1 Proposed Amendment No. 7/16/96 l l l _
l a RPS Instrumentation 3.3.1.1 ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME I 4 C. One or more Functions C.1 Restore RPS trip 1 hour i with RPS trip capability. , capability not I , maintained. l 'I , D. Required Action and D.1 Enter the Condition Immediately l associated Completion referenced in l Time of Condition A, Table 3.3.1.1-1 for l B, or C not met. the channel. ! J E. As required by E.1 Reduce THERMAL POWER 4 hours Required Action D.1 to < 30% RTP. j and referenced in Table 3.3.1.1-1. l i F. As required by F.1 Be in MODE 2. 6 hours ! , Required Action D.1 and referenced in Table 3.3.1.1-1.
)
As required by G. G.1 Be in MODE 3. 12 hours Required Action D.1
! and referenced in i
Table 3.3.1.1-1. H. As required by H.1 Initiate action to Immediately Required Action D.1 fully insert all and referenced in insertable control Table 3.3.1.1-1. rods in core cells containing one or more fuel assemblies. HATCH UNIT 1 3.3-2 Proposed Amendment No. 7/16/96 l
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS A
-------------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function. l 4
- 2. When a channel is placed in an inoperable status solely for performance of d
required Surveillances, entry into associated Conditions and Required . Actions may be delayed for up to 6 hours provided the associated Function ' maintains RPS trip capability. ! SURVEILLANCE FREQUENCY 3 i SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.1.2 ------------------NOTE------------------- Not required to be performed until 12 hours after THERMAL POWER 2: 25% RTP. i Verify the absolute difference between 7 days the average power range monitor (APRM) , channels and the calculated power is s 2% RTP while operating at 2: 25% RTP.
- SR 3.3.1.1.3 (Not used.) l i
SR 3.3.1.1.4 ------------------NOTE------------------- Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. ; Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. 7 days (continued) HATCH UNIT 1 3.3-3 Proposed Amendment No. 7/16/96
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.7 ------------------NOTE------------------- Only required to be met during entry into MODE 2 from MODE 1. l Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.8 Calibrate the local power range monitors. 1000 effective full power hours I l SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days i SR 3.3.1.1.10 ------------------NOTE------------------- For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 184 days l (continued) HATCH UNIT 1 3.3-4 Proposed Amendment No. 7/16/96
_. . . . = _ _ _- . - _ - _. - . . . .- RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 184 days Turbine Control Valve Fast Closure, Trip Oil Pressure -- Low Functions are not bypassed when THERMAL POWER is 2: 30% RTP. ) 1 I , SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months a SR 3.3.1.1.13 ------------------NOTES------------------ l 1. Neutron detectors are excluded.
- 2. For Function 1, not required to be l performed when entering MODE 2 from '
. MODE 1 until 12 hours after entering MODE 2. l j Perform CHANNEL CALIBRATION. 18 months
]
SR 3.3.1.1.14 (Not used.) l I SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months a SR 3.3.1.1.16 ------------------NOTE------------------- Neutron detectors are excluded. Verify the RPS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST BASIS HATCH UNIT 1 3.3-5 Proposed Amendment No. 7/16/96
RPS Instrumentation I
.3.3.1.1 Tabte 3.3.1.1-1 (pose 1 of 3)
Reactor Protection system Instrumentation APPLICA8LE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FR(M SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE. FUNCT!0N CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 1. Intermodlate Range Monitor
- a. Neutron Flux - High 2 3 G SR 3.3.1.1.1 s 120/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.6 futt scale sa 3.3.1.1.7 sa 3.3.1.1.13 st 3.3.1.1.15 5(a) 3 H SR 3.3.1.1.1 s 120/125 sa 3.3.1.1.5 divisions of SR 3.3.1.1.13 futi seale sa 3.3.1.1.15
- b. Inop' 2 3 G sa 3.3.1.1.4 NA sa 3.3.1.1.15 5(a) 3 N SR 3.3.1.1.5 NA sa 3.3.1.1.15
- 2. Aversee Power Renee Monitor
- a. Neutron Flux - High 2 3(c) G SR 3.3.1.1.1 s 20% RTP l (setdown) l SR' 3.3.1.1.7 SR 3.3.1.1.8 sa 3.3.1.1.10 SR 3.3.1.1.13 1 1.
- b. simulated Thermal 1 3(C) F SR 3.3.1.1.1 s 0.58 W +
Power - High l SR 3.3.1.1.2 62% RTP and 5 115.5% l SR 3.3.1.1.8 RTP(b) sa 3.3.1.1.10 sa 3.3.1.1.13 (continued) (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) 0.58 W + 62% - 0.58 AW RTP when reset for single loop operation per LCO 3.4.1, " Recirculation Loops operating." , (c) Each APRM channel provides inputs to both trip systems. l HATCH UNIT 1 3.3-6 Proposed Amendment No. 7/16/96
RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (page 2 of 3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVE!LLANCE ALLOWASLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE ;
- 2. Average Power Range Monitor (continued) )
- c. Neutron Flux - High 1 3(c) F SR 3.3.1.1.1 5 120% RTP l SR 3.3.1.1.2 j SR 3.3.1.1.8 l
SR 3.3.1.1.10 i SR 3.3.1.1.13
- d. Inop 1,2 3(C) G SR 3.3.1.1.10 NA
- e. Two-out of Four 1,2 2 .G SR 3.3.1.1.1 NA Voter SR 3.3.1.1.10 SR 3.3.1.1.15 SR 3.3.1.1.16
- 3. Reactor Vessel Stsam 1,2 2 G SR 3.3.1.1.1 5 1085 psig Dome Pressure - High SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 1
- 4. Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 2 0 inches Level - Lew, Level 3 SR 3.3.1.1.9 '
SR 3.3.1.1.13 l SR 3.3.1.1.15 1
- 5. Main Steam Isolation 1 8 F SR 3.3.1.1.9 5 10% closed Valve - Closure SR 3.3.1.1.13 SR 3.3.1.1.15
- 6. Drywetl Pressure - High 1,2 2 G SR 3.3.1.1.1 5 1.92 psis SR 3.3.1.1.9 i SR 3.3.1.1.13 I SR 3.3.1.1.15 '
(continued) (c) Each APRM channet provides inputs to both trip systems. l HATCH UNIT 1 3.3-7 Proposed Amendment No. 7/16/96
Control Rod Block Instrumentation 3.3.2.1 C DITI REQUIRED ACTION COMPLETION TIME
- E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch -- Shutdown withdrawal.
Position channels inoperable. ANQ l l E.2 Initiate action to Immediately , fully insert all ! insertable control j' rods in core cells containing one or
- more fuel assemblies.
1 SURVEILLANCE REQUIREMENTS
- -------------------------------------NOTES------------------------------------
! 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
- 2. When an RBM channel is placed in an inoperable status solely for l performance of required Surveillances, entry into associated Conditions i and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability.
l I SURVEILLANCE FREQUENCY l SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. 184 days l l (continued) ? 1 i 1 I 4 HATCH UNIT 1 3.3-16 Proposed Amendment No. 7/16/96
7 4 k Control Rod Block Instrumentation 3.3.2.1 j Table 3.3.2.1-1 (page 1 of 1) l i Control Rod Block Instrumentation ) i 1 ! APPLICABLE I NODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE , FUNCTION CONDITIONS CHANNELS REQUIREMENTS W.LUE
- 1. Rod Btock Monitor
- a. Low Power Range-Upscale (a) 2 st 3.3.2.1.1 s 115.5/125 st 3.3.2.1.4- divisions of SR 3.3.2.1.7 full scate
- b. Intermediate Power (b) 2 st 3.3.2.1.1 s 109.7/125 Range - Upecate st 3.3.2.1.4 divisions of SR 3.3.2.1.7 futt scale
- c. High Power Range-Upecate (c),(d) 2 sR 3.3.2.1.1 s 105.9/125 sR 3.3.2.1.4 divisions of st 3.3.2.1.7 futt scale
- d. Inop (d),(e) 2 st 3.3.2.1.1 NA
- e. Downscale (d),(e) 2 st 3.3.2.1.1 t 93/125 st 3.3.2.1.7 divisions of j full scale
- 2. Rod Worth Minimizer 1(I) 2(f)
, 1 SR 3.3.2.1.2 NA st 3.3.2.1.3 st 3.3.2.1.5 sa 3.3.2.1.8
- 3. Reactor Mode switch - shutdown (g) 2 st 3.3.2.1.6 NA Position (a) THERMAL POWER t 29% and < 64% RTP and MCPR < 1.70.
(b) THERMAL POWER R 64% and < 84% RTP and MCPR < 1.70. (c) THERMAL POWER t 84% and < 90% RTP and MCPR < 1.70. (d) THERMAL POWER R 90% RTP and MCPR < 1.40. (e) THERMAL POWER t 29% and < 90% RTP and MCPR < 1.70. j (f) With THERNAL POWER < 101 RTP. (g) Reactor mode switch in the shutdoun position. HATCH UNIT 1 3.3-19 Proposed Amendment No. 7/16/96
Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation, DE One recirculation loop shall be in operation with:
- a. The following limits applied when the associated LCO is applicable:
- 1. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits spe.cified in the COLR;
- 2. LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR),"
single loop operation limits specified in the COLR;
- 3. LCO 3.3.1.1, " Reactor Protection System (RPS)
Instrumentation,"' Function 2.b (Average Power Range Monitor Simulated Thermal Power - High), Allowable l Value of Table 3.3.1.1-1 is reset for single loop operation; and
- b. Core flow as a function of core thermal power in the
" Operation Allowed Region" of Figure 3.4.1-1.
APPLICABILITY: MODES I and 2. 4 ! ACTIONS i CONDITION REQUIRED ACTION COMPLETION TIME i A. No recirculation loops A.1 Place the reactor Immerii _tely in operation. mode switch in the shutdown position. (continued) HATCH UNIT.1 3.4-1 Proposed Amendment No. 7/16/96
SDM Test - Refueling 4. 3.10.8 i i 3.10 SPECIAL OPERATIONS 4
- 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling LC0 3.10.8 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/ hot standby position, and operation considered not to be in MODE 2, to allow SDM testing, provided the following requirements are met
- a. LCO 3.3.1.1, " Reactor Protection System Instrumentation," MODE 2 requirements for Functions 2.a, 2.d, and 2.e of Table 3.3.1.1-1;
- b. 1. LCO 3.3.2.1, " Control Rod Block Instrumentation,"
MODE 2 requirements for Function 2 of Table 3.3.2.1-1, with the banked position withdrawal sequence requirements of SR 3.3.2.1.8 changed to require the control rod sequence to conform to the SDM test sequence, 08
- 2. Conformance to the approved control rod sequence for the SDM test is verified by a second licensed operator or other qualified member of the technical staff;
- c. Each withdrawn control rod shall be coupled to the associated CRD;
- d. All control rod withdrawals during out of sequence control rod moves shall be made in notch out mode;
- e. No other CORE ALTERATIONS are in progress; and
- f. CR0 charging water header pressure a: 940 psig.
APPLICABILITY: MODE 5 with the reactor mode switch in startup/ hot standby position. HATCH UNIT 1 3.10-20 Proposed Amendment No. 7/16/96
i SDM Test -- Refueling 3.10.8 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ; l 1 SR 3.10.8.1 Perform the MODE 2 applicable SRs for According to LC0 3.3.1.1, Functions 2.a, 2.d, and 2.e of the applicable l Table 3.3.1.1-1. SRs SR 3.10.8.2 -------------------NOTE-------------------- i Not required to be met if SR 3.10.8.3 satisfied.
- Perform the MODE 2 applicable SRs for According to LC0 3.3.2.1, Function 2 of Table 3.3.2.1-1. the applicable
- SRs i
- I l
l l ' SR 3.10.8.3 -------------------NOTE-------------------- Not required to be met if SR 3.10.8.2 l { satisfied. I , Verify movement of control rods is in During control
- compliance with the approved control rod rod movement
- sequence for the SDM test by a second
- licensed operator or other qualified member 3
of the technical staff. i j SR 3.10.8.4 Verify no other CORE ALTERATIONS are in 12 hours progress. (continued) HATCH UNIT 1 3.10-22 Proposed Amendment No. 7/16/96
_ . . _ - . _ - _ _ . _ . _ . ~ ~ . . _ . _ _ - _._ _..._ _ . _ . _ _ _ _ . _ _ . . - E l j RPS Instrumentation 3.3.1.1 i 3.3 INSTRUMENTATION f 3.3.1.1 Reactor Protection System (RPS) Instrumentation 9 1~ LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. 1 i APPLICABILITY: According to Table 3.3.1.1-1. $ l l ! ACTIONS ! -------------------------------------NOTE------------------------------------- 1 Separate Condition entry is allowed for each channel.
= ==----_____________________________________________________________________
i CONDITION REQUIRED ACTION COMPLETION TIME i A. One or more required A.1 Place channel in 12 hours channels inoperable. trip. QB A.2 --------NOTE--------- Not applicable for j Functions 2.a, 2.b, l 2.c, and 2.d. j j Place associated trip 12 hours ! system in trip. j l i B. --------NOTE--------- B.1 Place channel in one 6 hours I Not applicable for trip system in trip. Functions 2.a, 2.b, 2.c, and 2.d. QB B.2 Place one trip system 6 hours One or more Functions in trip, with one or more required channels inoperable in both trip systems. l (continued) HATCH UNIT 2 3.3-1 Proposed Amendment No. 7/16/96
i RPS Instrumentation
- 3.3.1.1 1
l ACTIONS (continued) ; 4 1 ! CONDITION REQUIRED ACTION COMPLETION TIME !
- 1 4 i i- C. One or more Functions C.1 Restore RPS trip 1 hour i with RPS trip capability. ,
- capability not l maintained.
I ! D. Required Action and D.1 Enter the Condition Immediately
- associated Completion referenced in
- Time of Condition A, Table 3.3.1.1-1 for B, or C not met. the channel.
l l E. As required by E.1 Reduce THERMAL POWER 4 hours l Required Action D.1 to < 30% RTP. and referenced in
- Table 3.3.1.1-1.
j F. As required by F.1 Be in MODE 2. 6 hours
- Required Action D.1 l and referenced in
! Table 3.3.1.1-1. i l G. As. required by G.1 Be in MODE 3. 12 hours Required Action D.1 l and referenced in j Table 3.3.1.1-1. i 4 H. As required by H.1 Initiate action to Immediately j Required Action D.1 fully insert all 4 and referenced in insertable control i l Table 3.3.1.1-1. rods in core cells ! I containing one or l
- more fuel assemblies. j i
t 1 i l 4 i HATCH UNIT 2 3.3-2 Proposed Amendment No. 7/16/96 l
i l RPS Instrumentation > 3.3.1.1 SURVEILLANCE REQUIREMENTS
NOTES------------------------------------
- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
- 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
i SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.1.2 ------------------NOTE------------------- Not required to be performed until 12 hours after THERMAL POWER 2: 25% RTP. Verify the absolute difference between 7 days the average power range monitor (APRM) channels and the calculated power is s 2% RTP while operating at 2 25% RTP. SR 3.3.1.1.3 (Not used.) l SR 3.3.1.1.4 -------------
----NOTE-------------------
Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. 7 days (continued) HATCH UNIT 2 3.3-3 Proposed Amendment No. 7/16/96
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY 1 l SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overl ap. SRMs from the fully inserted position SR 3.3,1.1.7 ------------------NOTE------------------- Only required to be met during entry into MODE 2 from MODE 1. j Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.8 Calibrate the local power range monitors. 1000 effective full power hours SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.10 ------------------NOTE------------------- For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 184 days l (continued) 1 HATCH UNIT 2 3.3-4 Proposed Amendment No. 7/16/96 l
t , RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 18 months Turbine Control Valve Fast Closure, Trip 011 Pressure -- Low Functions are not bypassed when THERMAL POWER is 2 30% RTP. SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months SR 3.3.1.1.13 ------------------NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 1, not recuired to be' performed when enteri..g MODE 2 from MODE 1 until 12 hours after entering MODE 2.
Perform CHANNEL CALIBRATION. 18 months SR 3.3.1.1.14 (Not used.) l SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months (continued) HATCH UNIT 2 3.3-5 Proposed Amendment No. 7/16/96
l RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (page 1 of 3) Reactor Protection system Instrumentation i I APPLICA8LE CONDITIONS l MODES OR REQUIRED REFERENCED l OTHER CHANNELS FROM i SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS system ACTION D.1 REQUIREMENTS VALUE
- 1. Intermodlate Range Monitor l a. Neutron Flux - High 2 3 G st 3.3.1.1.1 s 120/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.6 full scale I st 3.3.1.1.7 l SR 3.3.1.1.13 st 3.3.1.1.15 5(*) 3 N SR 3.3.1.1.1 s 120/125 st 3.3.1.1.5 divisions of SR 3.3.1.1.13 full scale l
l SR 3.3.1.1.15 l b. Inap 2 3 G sa 3.3.1.1.4 NA j st 3.3.1.1.15 5(a) 3 N st 3.3.1.1.5 NA st 3.3.1.1.15
- 2. Average Power Range Monitor
- a. Neutron Flux - Nigh 2 3(C) G sa 3.3.1.1.1 5 20% RTP (Setdown)
I st 3.3.1.1.7 I sa 3.3.1.1.8 i SR 3.3.1.1.10 st 3.3.1.1.13 1
- b. simulated Thermal 1 3(c) F sa 3.3.1.1.1 s 0.58 W + l Power - Nigh st 3.3.1.1.2 62% RTP and i
5 115.5% l l SR 3.3.1.1.8 RTP(b) st 3.3.1.1.10 st 3.3.1.1.13 l I (continued) (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) 0.58 W + 62% - 0.58 &W RTP when reset for single loop operation per LCO 3.4.1, " Recirculation Loops operating." (c) Each APRM channel provides inputs to both trip systems. l 1 l I ! HATCH UNIT 2 3.3-7 Proposed Amendment No. 7/16/96 i
l l RPS Instrumentation ) 3.3.1.1 1 Table 3.3.1.1-1 (page 2 of 3) Reactor Protection System Instrumentation 1 APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM I SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE I l FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 2. Average Power Range !
Monitor (continued)
- c. Neutron Flux - High 1 3(C) F SR 3.3.1.1.1 s 120% RTP l SR 3.3.1.1.2 SR 3.3.1.1.8 i
SR 3.3.1.1.10 SR 3.3.1.1.13
- d. Inop 1,2 3(C) G SR 3.3.1.1.10 NA
- e. Two out-of-Four 1,2 2 G SR 3.3.1.1.1 NA Voter SR 3.3.1.1.10 SR 3.3.1.1.15 i SR 3.3.1.1.16
- 3. Reactor vessel Steam 1,2 2 G SR 3.3.1.1.1 s 1085 psig Dome Pressure - High SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 l SR 3.3.1.1.16 4 Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 t 0 inches Level - Low, Level 3 SR 3.3.1.1.9 SR 3.3.1.1.13 l SR 3.3.1.1.15 SR 3.3.1.1.16
- 5. Main Steam Isolation 1 8 F SR 3.3.1.1.9 s 10% closed Valve - Closure SR 3.3.1.1.13 l
SR 3.3.1.1.15 SR 3.3.1.1.16
- 6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.1 s 1.92 psig SR 3.3.1.1.9 l
SR 3.3.1.1.13 SR 3.3.1.1.15 l l (continued) (c) Each APRM channel provides inputs to both trip systems. l l l 1 1 HATCH UNIT 2 3.3-8 Proposed Amendment No. 7/16/96 1 i 1
i Control Rod Block Instrumentation 3.3.2.1 ACTIONS (continued) i CONDITION REQUIRED ACTION COMPLETION TIME I i E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch -- Shutdown withdrawal. ' Position channels inoperable. AND l E.2 Initiate action to Immediately fully insert all insertable control ! rods in core cells : i containing one or
- more fuel assemblies.
! i l > SURVEILLANCE REQUIREMENTS ___________________------------------NOTES------------------------------------
- 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function. ,
- 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions '
l and Required Actions may be delayed for up to 6 hours provided the ; associated Function maintains control rod block capability. l l l SURVEILLANCE FREQUENCY l SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. 184 days l ! l ! (continued) ^ , t i i HATCH UNIT 2 3.3-17 Proposed Amendment No. 7/16/96
I Centrol Rod Bleck Instrumentation 3.3.2.1 l Table 3.3.2.1 1 (pose 1 of 1) Control Rod Block Instrumentation , 4'PLICASLE l MODES OR ! OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE l l 1. Rod Block Monitor ! a. Low Power Range-Upsca1e (a) 2 st 3.3.2.1.1 s 115.5/125 st 3.3.2.1.4 divisions of l sa 3.3.2.1.7 futt scale
- b. Intermediate Power (b) 2 st 3.3.2.1.1 5 109.7/125 Range - Upscale st 3.3.2.1.4 divisions of st 3.3.2.1.7 full scale
- c. High Power Range-Upscale (c),(d) 2 sR 3.3.2.1.1 s 105.9/125 SR 3.3.2.1.4 divisions of SR 3.3.2.1.7 full scate
- d. Inop (d),(e) 2 st 3.3.2.1.1 NA
, e. Downscale (d),(e) 2 SR 3.3.2.1.1 2 93/125 i SR 3.3.2.1.7 divisions of futt scale 1
- 2. Rod Worth Minimiter 1(f) 2(f)
, 1 SR 3.3.2.1.2 NA sR 3.3.2.1.3 SR 3.3.2.1.5
! sR 3.3.2.1.8 l
- 3. Reactor Mode switch - Shutdown (g) 2 SR 3.3.2.1.6 NA j Position l
l (a) THERMAL POWER t 29% and < 64% RTP and MCPR < 1.70. ! (b) THERMAL POWER t 64% and < 84% RTP and MCPR < 1.70. l l (c) THERMAL POWER t 84% and < 90% RTP and MCPR < 1.70. (d) THERMAL POWER t 90% RTP and MCPR < 1.40. l (e) THERMAL POWER t 29% and < 90% RTP and MCPR < 1.70. l l (f) With THERMAL POWER < 10% RTP. (g) Reactor mode switch in the shutdown position. t i HATCH UNIT 2 3.3-20 Proposed Amendment No. 7/16/96
l l R:cfrculetion Lonps Operating 3.4.1 l 3.4 REACTOR COOLANT SYSTEM (RCS) l 3.4.1 Recirculation Loops Operating LC0 3.4.1 Two recirculation loops with matched flows shall be in operation, i E One recirculation loop shall be in operation with:
- a. The following limits applied when the associated LC0 is applicable:
- 1. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;
- 2. LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR),"
single loop operation limits specified in the COLR;
- 3. LC0 3.3.1.1, " Reactor Protection System (RPS)
Instrumentation," Function 2.b (Average Power Range Monitor Simulated Thermal Power - High), Allowable l Value of Table 3.3.1.1-1 is reset for single loop operation; and
- b. Core flow as a function of core thermal power in the
, " Operation Allowed Region" of Figure 3.4.1-1. I APPLICABILITY: MODES I and 2. I ACTIONS 1 CONDITION REQUIRED ACTION COMPLETION TIME A. No recirculation loops A.1 Place the reactor Immediately l in operation. mode switch in the , shutdown position. (continued) 4 f HATCH UNIT 2 3.4-1 Proposed Amendment No. 7/16/96 t
SDM Test - Refueling 3.10.8 3.10 SPECIAL OPERATIONS ! 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling LC0 3.10.8 The reactor mode switch position specified in Table 1.1-1 '
- for MODE 5 may be changed to include the startup/ hot standby .
position, and operation considered not to be in MODE 2, to ' allow SDM testing, provided the following requirements are l met: l
- a. LC0 3.3.1.1, " Reactor Protection System
- l Instrumentation," MODE 2 requirements for Functions 2.a, 2.d, and 2.e of Table 3.3.1.1-1;
- b. 1. LC0 3.3.2.1, " Control Rod Block Instrumentation,"
MODE 2 requirements for Function 2 of Table 3.3.2.1-1, with the banked position withdrawal sequence requirements of SR 3.3.2.1.8 changed to require the control rod sequence to conform to the SDM test sequence, M
- 2. Conformance to the approved control rod sequence for the SDM test is verified by a second licensed operator or other qualified member of the technical staff;
- c. Each withdrawn control rod shall be coupled to the associated CRD;
- d. All control rod withdrawals during out of sequence control rod moves shall be made in notch out mode;
- e. No other CORE ALTERATIONS are in progress; and
- f. CRD charging water header pressure a 940 psig.
APPLICABILITY: MODE 5 with the reactor mode switch in startup/ hot standby position. l i HATCH UNIT 2 3.10-20 Proposed Amendment No. 7/16/96 l
i l SDM Test -- Refueling 3.10.8 SURVEILLANCE REQUIREMENTS ! l SURVEILLANCE FREQUENCY SR 3.10.8.1 Perform the MODE 2 applicable SRs for According to LCO 3.3.1.1, Functions 2.a, 2.d, and 2.e of the applicable l Table 3.3.1.1-1. SRs SR 3.10.8.2 -------------------NOTE-------------------- Not required to be met if SR 3.10.8.3 satisfied. l Perform the MODE 2 applicable SRs for According to LC0 3.3.2.1, Function 2 of Table 3.3.2.1-1. the applicable SRs bn 3.10.8.3 -------------------NOTE-------------------- Not required to be met if SR 3.10.8.2 satisfied. l l l Verify movement of control rods is in During control ! t compliance with the approved control rod rod movement ' i sequence for the SDM test by a second i licensed operator or other qualified member l of the technical staff. l l ! SR 3.10.8.4 Verify no other CORE ALTERATIONS are in 12 hours 1 progress. ; 1 l l i (continued) i I i HATCH UNIT 2 3.10-22 Proposed Amendment No. 7/16/96
i RPS Instrumentation 3.3.1.1 '
- 3.3 INSTRUMENTATION I 3.3.1.1 I Reactor Protection System (RPS) Instrumentation '
1 ! LC0 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1
- shall be OPERABLE.
APPLICABILITY: According to Table 3.3.1.1-1. i ACTIONS
-------------------------------------NOTE-------------------------------------
] Separate Condition entry is allowed for each channel. g i CONDITION REQUIRED ACTION COMPLETION TIME 4 A. One or more required A.1 Place channel in 12 hours channels inoperable. trip. OR v ln ser+'A* l A.2 ciated trip 12 hours j system in trip.
,m losed *A e Functions B.1 Place channel in one 6 hours with one or more trip system in trip.
required channels inoperable in both 0R trip systems. B.2 Place one trip system 6 hours in trip. C. One or more Functions C.1 Restore RPS trip I hour with RPS trip capability. capability not ! maintained. (continued) HATCH UNIT 1 3.3-1 Amendment No. 195 i k-
. ._. _ . ._ . _ . _ _ _ . - . . . _ _ _ . . _ - . . . _ . _ _ _ _ . _ . _ .__m._._._... .. .__. . . _ . .. ~ . .
Insert 'A' . Technical Specification 3.3.1.1 Reactor Protection System (RPS) ! Instrumentation LCO A & B ! I
-..---NOTE..~. :
Not applicable for Functions 2.a,2.b, 2.c, and 2.d. i 6 i l 4 i l l I l L i i I l 4 t i ! 1 I
-, - , . . . . - < - 7 m-
RPS Instrumentation- ; 3.3.1.1 SURVEILLANCE REQUIREMENTS
-------------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
r i 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains RPS trip capability. 3 l l SURVEILLANCE FREQUENCY l SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.1.2 ------------------NOTE------------------- i Not required to be performed until l 12 hours after THERMAL POWER a 25% RTP. l' Verify the absolute difference between 7 days the average power range monitor (APRM) channels and the calculated power is s 2% RTP while operating at a 25% RTP. \ (Ho+ usec\.) V l 3.3.1.1. SR Vp" ".-+ ^..- y "y- '.,-' y" y ",,, ,>,,, ,d ' - , . ,#,
-s#. -
...,....n v, - . ,,n,.
A yk f- , l SR 3.3.1.1.4 ------------------NOTE------------------- I Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. 1 Perform CHANNEL FUNCTIONAL TEST. 7 days l
- (continued) t HATCH UNIT 1 3.3-3 Amendment No. 195
RPS Instrumentation 3.3.1.1 l SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l l SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. 7 days i SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.7 ------------------NOTE------------------- Only required to be met during entry into MODE 2 from MODE 1. Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.8 Calibrate the local power range monitors. 10M effective fui power houra l l SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days l SR 3.3.1.1.10 ---------------
------------------NOTE
'v.')$tf;M;t6d;p'% //j-2:)2;~W
\
AA A A[r'ForFunction2.a,notrequiredtobe performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. _______________ D _________ ____ FUNCTlONAL TEST Perform CHANNE I" '@tdr. 184 days V l (continued) t HATCH UNIT 1 3.3-4 Amendment No. 195
RPS Instrumentatien 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve - Closure and 184 days Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are not bypassed when THERMAL POWER is 2: 30% RTP. l SR 3.3.1.1.12 Perfonn CHANNEL FUNCTIONAL TEST. 18 months l j SR 3.3.1.1.13 ------------------NOTES------------------ ! 1. Neutron detectors are excluded. I
- 2. For Function 1, not required to be performed when entering MODE 2 from l
MODE 1 until 12 hours after entering MODE 2. 1 Perform CHANNEL CALIBRATION. 18 months I m n - l SR 3.3.1.1.14 y"(:Hof use d. )
- -ify th: ^^a", F;;; ai;;;d Si, ;;;;ted la ;,;,,th;
'h:' :1 .";= r ."igh ti;; c;r,;te;;t i; l
l M Wir, th; lieit; ;pecified ir, the COL". I l l SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months l SR 3.3.1.1.16 Verify the RPS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST BASIS l
- - - - - - - - - - - N O T E' - - - - - - - - - - - -
t4eutron defec+ ors are, excluded. { l HATCH UNIT 1 3.3-5 Amendment No. 195 l t
l RPS Instrumentation 3.3.1.1 ' Table ~. i.1.1 1 (page 1 of 3) Reactor P.otes tion System Instrunentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED FUNCTION SURVEILLANCE ALLOWABLE CON 0lfl0Ns SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 1. Intermediate Range Monitors
- a. Neutron Flux - High 2 3
\
G sR 3.3.1.1.1 s 120/125 SR 3.3.1.1.4 divisions of ) SR 3.3.1.1.6 full scale j SR 3.3.1.1.7 SR 3.3.1.1.13 } sR 3.3.1.1.15 i 5(a) 3 i H SR 3.3.1.1.1 5 120/125
$R 3.3.1.1.5 divisions of l SR 3.3.1.1.13 full scale SR 3.3.1.1.15
- b. Inop 2
( 3 C SR 3.3.1.1.4 NA SR 3.3.1.1.15 l 5(*) 3 N SR 3.3.1.1.5 NA SR 3.3.1.1.15
- 2. Average Power Range a.
g Neutron Flux - High, 2 3 -4 G SR 3.3.1.1.1 5 20% RTP i C (Set down) 'lj;j;;;;;) sR 3.3.1.1.8 SR {-{*}-}*'04 .S*R 3. 3. l. I.13
- b. " ^ " - -
' 1;imul a ted 1 Thermal Power - Higt) 3 F
$1L._LJ3.1.1. 2 +62% RTP and
= :.:. . .. s 115.5%
l SR 3.3.1.1.8 RTP(b)
= !.!. f
- ..t.1 l.N. .?< sR 3. 3.1. /.13
= !.:.-
(continued) 1 i (a) With any control rod withdrawn f rom a core cell containing one or more fuel assemclies. l (b) 0.58 W + e - 4 ery 4nT "3 :0.58 N AWARJ eset for sin a-loop _ operation per LCO 3.4.1,
- Recirculation Loops sf ~g (c) Eack APRM channel provides inputs fo both trip sys+em s.
\ y Q .. N - N A - /C" i
HATCH UNIT 1 3.3-6 Amendment No. 195
RPS Instrumentaticn 3.3.1.1 4 Table 3.3.1.1 1 (page 2 of 3) 1 Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLom BLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REGUIREMENTS VALUE
- 2. Aversee Pouer Renee i Men (continued)
! (G
- c. Jfsfliis(soutron 1 3 4- F st 3.3.1.1.1 s 120% RrP ,
- F Ltat - Nigh at 3.3.1.1.2 '
'W st 3.3,.1. 1. 10-i C: r ::i: 1 '.3 ::--
- 3.3.1.1.3 44 3.3.1.' t x d, -e, Insp 1, G 3.3.0.1.C
- 3(c) -a-4,g, m E i.+...:-:... i..< 3. 3. i. uo 1,2 2 G SM 5 1085 psis C d , " Reactor- vessel
" " ~ ' - steen ll!:!:i:i:i>
st 3.3.1.1.15 l
- 4. Reactor vessel Water 1,2 2 G st 3.3.1.1.1 R 0 inches Levet -Low, Levet 3 ca 3.3.1.1.9 st 3.3.1.1.13 st 3.3.1.1.15
- 5. Main steen Isolation 1 8 F st 3.3.1.1.9 s 10% closed Valve - Closure st 3.3.1.1.13 st 3.3.1.1.15
- 6. Drywell Pressure-Nigh 1,2 2 G st 3.3.1.1.1 5 1.92 psig st 3.3.1.1.9 SR 3.3.1.1.13 st 3.3.1.1.15 1
(continued) s [ N(K (c) Eock APRm channe.l prwide.c inputs -10 bofk hQ systerns. N -D A A C i l HATCH UNIT I 3.3-7 Amendment No. 197
Insert 'B' - Technical Specification 3.1.1.1, Table 3.3.1.1-1 (page 2 of 3) Beactor Protection System Instrumentation
- e. 2-out-of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.10 SR 3.3.1.1.15 SR 3.3.1.1.16
Control Rod Block Instrumentation 3.3.2.1 ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME E. One or more Reactor E.1 Suspend control rod Immediately 1 Mode Switch -- Shutdown withdrawal. Position channels 1 inoperable. AND E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies. SURVEILLANCE REQUIREMENTS
-------------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Block Function.
- 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capabilit
_____________________________________________________________y. _________________ SURVEILLANCE FREQUENCY n, IB 4 SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. /95/ days (continued) HATCH UNIT 1 3.3-16 Amendment No. 195
Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1) Control Rod Block Instrunentation i i l APPLICABLE ! MODES OR I OTHER ! SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE l l
- 1. Rod Block Monitor
- a. Low Power Range -Upscale (a) SR 3.3.2.1.1 2
SR 3.3.2.1.4 s 115.5/125 divisions of
)
SR 3.3.2.1.7 full scale l l
- b. Intermediate Power (b) 2 SR 3.3.2.1.1 s 109.7/125 Range - Upscale SR 3.3.2.1.4 divisions of i SR 3.3.2.1.7 full scale '
- c. High Power Range -Upscale (c),(d) 2 SR 3.3.2.1.1 s 105.9/125 SR 3.3.2.1.4 divisions of SR 3.3.2.1.7 full scale
- d. Inop (d),(e) 2 SR 3.3.2.1.1 NA l
- e. Downscale (d),(e) 2 SR 3.3.2.1.1 2 93/125 SR 3.3.2.1.7 divisions of full scale i
'%f e ^[ MD i,7--- n :: ; -a- = 3.3.2.'.' : 2.^ --
' W d'
- 2. Rod Worth Minimizer 1(f3,2(f) 1 SR 3.3.2.1.2 NA SR 3.3.2.1.3 SR 3.3.2.1.5 SR 3.3.2.1.8
- 3. Reactor Mode Switch - Shutdown (g) 2 SR 3.3.2.1.6 NA l Position I
(a) THERMAL POWER t 29% and < 64% RTP and MCPR < 1.70. (b) THERMAL POWER 164% and < 84% RTP and MCPR < 1.70. (c) THERMAL POWER 184% and < 90% RTP and MCPR < 1.70. (d) THERMAL POWER R 90% RTP and MCPR < 1.40. (e) THERMAL POWER t 29% and < 90% RTP and MCPR < 1.70. (f) With THERMAL POWER < 10% RTP. i (g) Reactor mode switch in the shutdown position. HATCH UNIT 1 3.3-19 Amendment No. 195 l l
-.= - . . - . - - . . . . . - _ . - - .-. . _... - . .- ..~ . .__. -
Recirculation loops Operating , l 3.4.1-i { 3.4 REACTOR COOLANT SYSTEM (RCS) t 3.4.1 Recirculation Loops Operating
.LCO 3.4.1 -Two recirculation loops with matched flows shall be in operation, QB One recirculation loop shall be in operation with: !
- a. The following limits applied when the associated LCO,is applicable:'
l
- 1. LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION '
RATE (APLHGR)," single loop operation limits specified in the COLR;
- 2. LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," ,
single loop operstion. limits specified in the COLR; !
- 3. LCO 3.3.1.1, "neactor Protection System (RPS)
Instru tion 2.b (Average Power Range j Monito imulated Thermal 1 Power - ble' Value of Table 3.3.1.1-1 is - J reset for single. loop operation; and
- b. Core flow as a function of core thermal power in the
" Operation Allowed Region" of Figure 3.4.1-1.
l
)
i APPLICABILITY: MODES 1-and 2. 1 i ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME l l I A. No- recirculation loops A.1 Place the reactor Immediately l in operation, mode switch in the shutdown position. (continued) ! HATCH UNIT 1 3.4-1 Amendment No. 195
SDM Test - Refueling 3.10.8 3.10 SPECIAL OPERATIONS 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling LCO 3.10.8 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/ hot standby position, and operation considered not to be in MODE 2, to allow SDM testing, provided the following requirements are met:
- a. LCO 3.3.1.1, " Reactor Protection System Instrumentation," MODE 2 requirements for Functions 2 3 2. . d ;
and 2.e of Table 3.3.1.1-1;
- b. 1. LC0 3.3.2.1, " Control Rod Block Instrumentation,"
MODE 2 requirements for Function 2 of Table 3.3.2.1-1, with the banked position withdrawal sequence requirements of SR 3.3.2.1.8 changed to require the control rod sequence to conform to the SDM test sequence, QB
- 2. Conformance to the approved control rod sequence for the SDM test is verified by a second licensed operator or other qualified member of the technical staff;
- c. Each withdrawn control rod shall be coupled to the associated CRD;
- d. All control rod withdrawals during out of sequence control rod moves shall be made in notch out mode;
- e. No other CORE ALTERATIONS are in progress; and
- f. CRD charging water header pressure 2: 940 psig.
APPLICABILITY: MODE 5 with the reactor mode switch in startup/ hot standby position. HATCH UNIT 1 3.10-20 Amendment No. 195
- .- - - - . - . _ _ - . _ . = - - _ - _ = _ - . - . . . - . . . . . . . -
J I SDM Test -- Refueling 4 3.10.8 .i SURVEILLANCE REQUIREMENTS ,
- SURVEILLANCE FREQUENCY 4
SR 3.10.8.1 Perform the MODE 2 applica Rs for According to i LC0 3.3.1.1, Functions .e of the applicable Table 3.3.1.1-1. . #^and SRs , f b> SR 3.10.8.2 -------------------NOTE-------------------- Not required to be met if SR 3.10.8.3 satisfied. Perform the MODE 2 applicable SRs for According to LC0 3.3.2.1, Function 2 of Table 3.3.2.1-1. the applicable , SRs SR 3.10.8.3 -------------------NOTE-------------------- Not required to be met if SR 3.10.8.2 I satisfied. ' Verify movement of control rods is in During control compliance with the approved control rod rod movement ; sequence for the SDM test by a second licensed operator or other qualified member of the technical staff. SR 3.10.8.4 Verify no other CORE ALTERATIONS are in 12 hours progress. (continued) HATCH UNIT 1 3.10-22 Amendment No. 195
4 : i RPS Instrumentation J 3.3.1.1 i l 3.3 INSTRUMENTATION i i ! 3.3.1.1 Reactor Protection System (RPS) Instrumentation ' l ~ LC0 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. I ' i APPLICABILITY: According to Table 3.3.1.1-1. ' I l ACTIONS ' l l ' I
-------------------------------------NOTE------------------------------------- '
Separate Condition entry is allowed for each channel. l CONDITION REQUIRED ACTION COMPLETION TIME l A. One or more required A.1 Place channel in 12 hours channels inoperable. tri . Ob in ser + "A I V A.2 associated trip 12 hours system in trip. _O [ I n s er +*A" l VOn re Functions B.1 Place channel in one 6 hours ith one or more trip system in trip. ' required channels l inoperable in both QR trip systems. l B.2 Place one trip system 6 hours in trip. C. One or more Functions C.1 Restore RPS trip 1 hour with RPS trip capability. capability not i maintained. 4 k a (continued) e HATCH UNIT 2 3.3-1 Amendment No. 135 t
i 1 1 l l Insert 'A' - Technical Specification 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO A & B
---------NOTE-----
Not applicable for Functions 2.a 2.b, 2.c, and 2.d. I l 1 i 1 e l l i i
r < 3 i RPS 2nstrumentation ! 3.3.1.1 ; ! SURVEILLANCE REQUIREMENTS ' 4
-------------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS 4 Function.
! 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function i maintains RPS trip capability. i SURVEILLANCE FREQUENCY i'
- SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours i i
? , i l ? i SR 3.3.1.1.2 ------------------NOTE------------------- ! Not required to be performed until 12 hours after THERMAL POWER 2: 25% RTP. _________________________________________ ) v I i Verify the absolute difference between 7 days
, j the average power range monitor (APRM) channels and the calculated power is s 2% RTP while operating at 2: 25% RTP.
m i (Ho t U.S e d.) Y Sg 3.3.3.}.3 Yms,3 y+ a ,o a_y _ _,ou,... A. +. J_
,A
- 3, _ . _ wn . . ,. , - , - . ,, . ..
3 3 i SR 3.3.1.1.4 ------------------NOTE------------------- l Not required to be performed when j entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 7 days (continued) i
)
i HATCH UNIT 2 3.3-3 Amendment No. 135 l 1 j
i i RPS Instrumentation l 3.3.1.1 ! l SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE ! FREQUENCY SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. i 7 days l l SR 3.3.1.1.6 Verify the source range monitor (SRM) and
- Prior to intermediate range monitor (IRM) channels withdrawing overlap.
SRMs from the fully inserted ; J, position I 1 i SR 3.3.1.1.7 ------------------NOTE------------------- Only required to be met during entry into MODE 2 from MODE 1.
=-- =_ ------------
Verify the IRM and APRM channels overlap. 7 days i SR 3.3.1.1.8 Calibrate the local power range monitors. I 1000 effective l full power l hours 1 SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days j SR 3.3.1.1.10
- 7. ,. 7~.,. . <
, ., .. . x.
7
, . , . ,. -,~r . .
or Function 2.a. not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2
-_-- . _---_-- -- C O -+-----
UNCTIO^tAL. TEST Perform CHANNE ; F / " #,J M. 184 days (continued) HATCH UNIT 2 3.3-4 Amendment No. 135
RPS fnstrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve - Closure and 18 months Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are not bypassed when THERMAL POWER is 2 30% RTP. SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months SR 3.3.1.1.13 ------------------NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
Perform CHANNEL CALIBRATION. 18 months N_ (
~
SR 3
- 3
- 1 RZ
- 1' f,* '14l!
#- + " * " ' ' ^ -
E :I ' /u V"("No+ "--d c"used.)
; 2 ";c1.' ."K'7;
'^+^d
~Z ' # L '# "
rrT 7" lA.. . . 7. ." ',C., . .., . . T,,, _ L. " T. .' . ,'G, . ;. .I.a" Z, . . ' I. . . .' ,M_ _ /o
\
SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months (continued) HATCH UNIT 2 3.3-5 Amendment No. 135
i 5
- RPS Instrumentation
? 3.3.1.1 Table 3.3.1.1 1 (page 1 of 3) Reactor Protection system Instrtmentation ' ) 1 APPLICA8LE CON 0lfl0Ns { MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM : SPECIFIED PER TRIP REQUIRED FUNCTION SURVE!LLANCE ALLOWABLE ! CONDlflCNs SYSTEM ACTION D.1 REQUIREMENis I VALUE
- 1. Intermediate Range I Monitors '
1 , 4
- a. Neutron Flux = High 2 3 G SR 3.3.1.1.1 5 120/125 i
- sR 3.3.1.1.4 divisions of '
J sR 3.3.1.1.6 full scale 2 SR 3.3.1.1.7 5 ! SR 3.3.1.1.13 . sa 3.3.1.1.15 ! 5(a) ! j 3 H $R 3.3.1.1.1 s 120/125 ; i SR 3.3.1.1.5 divisions of d sa 3.3.1.1.13 full scale SR 3.3.1.1.15
- b. Inop 2 .
3 G SR 3.3.1.1.4 NA SR 3.3.1.1.15 ' 5(a) 3 H sR 3.3.1.1.5 NA sa 3.3.1.1.15
- 2. Average Power Range .
Monito >
- a. Flux = High 2 3 G ... s 20% RTP Setdoan) ;;
; ; '.1.7 M SR 3.3.1 ?
SR 3.3.1.1.8 ? SR } 3*}*}*}?< s J. J./. /. /J
- b. '!:u !!::-r imulated 1 3 (C) 2- F SR 3. . . s 0.58 W hermal Po r - H igh
- 3. . +62% RTP and !
- ! . ; . * . '. . ! s 115.5% i st 3.3.1.1.8 * "
R I i se ?.I.' SR
., l-l }-}'.}.94 SR 3,3, l. l. l3
- !.!.'.".'! /
00 !.:.M (continued) (a) With any control rod withdrawn from a core cell containing one or more fuel asseniblies. l j (b) AW RTP when_ reset for single i o operation per LCO 3. r*Recirculat on i Loops D.58 7'"~'5"W + 621 =-A5(V N (c) Each APRtd channel provi es input.s 46 ho+h hip systems.
-GA s( -- N q HATCH UNIT 2 3.3-7 Amendment No. 135
i RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (pose 2 of 3) Reactor Protection system Instrumentation APPLICA8LE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 2. Average Power Range Monitor (continued) -
-54aed ~ ron 1 3(c)e- F st 3.3.1.1.1 s 120s RTP i Flux - Igh d .1.2
[=st 3.3.).}.}g< gg y,3, /, /, / y i
)
e :.:.'.'.ts l l 1 r =s + _e_ + = :. . .'.: 2 c.= n-
-= :. .'.1.0
" 3 3 * ' I5 d *. Inop 1,2 3(c)e- G = :.3.'.'.: NA g, lngeg $3 W 'd 3 3-]-} ? A S R -J. 3.1.3.I
- 3. A b teen 1,2 2 C sR . 1085 psig l Dome Pressure - Hign SR 3.3.1.1.9 st 3.3.1.1.13 i sa 3.3.1.1.15 l SR 3.3.1.1.16 )
- 4. Reactor vessel Water 1,2 2 C SR 3.3.1.1.1 t 0 Inches Level - Low, Level 3 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 st 3.3.1.1.16
- 5. Main steem isolation 1 8 F st 3.3.1.1.9 s 10% closed Vatve - CL osure SR 3.3.1.1.13 st 3.3.1.1.15 SR 3.3.1.1.16
- 6. Drywell Pressure - High 1,2 2 G st 3.3.1.1.1 s 1.92 psig SR 3.3.1.1.9 sa 3.3.1.1.13 st 3.3.1.1.15
~s ~S Qcontinued)
(c) Cach APPM ckonel provides inputs to Qh -frap sy stem s . jy j% /% j" HATCH UNIT 2 3.3-8 Amendment No. 138
l Insert 'B' - Technical Specification 3.1.1.1 Table 3.3.1.1-1 (page 2 of 3) Reactor Protection System Instrumentation l
- e. 2-out-of-4 Voter 1,2 2 G SR 3.3.1.1.1 NA SR 3.3.1.1.10 SR 3.3.1.1.15 SR 3.3.1.1.16
)
l 1 1
Control Rod Block Instrumentation 3.3.2.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLET10N TIME E. One or more Reactor E.1 Suspend control rod Immediately Mode Switch - Shutdown withdrawal. Position channels inoperable. AND E.2 Initiate action to Immediately fully insert all insertable control rods in core cells containing one or more fuel assemblies. SURVEILLANCE REQUIREMENTS
NOTES------------------------------------
- 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control Rod Bicek Function.
- 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours provided the associated Function maintains control rod block capability. l i
SURVEILLANCE f _ FREQUENCY iS + SR 3.3.2.1.1 Perform CHANNEL FUNCTIONAL TEST. E days (continued) HATCH UNIT 2 3.3-17 Amendment No. 135
- .. _- . ~ . . . -.... - . - .. - - . . - - - . - ... .. . . - - . .- -.-
Control Rod Block ~ Instrumentation l 3.3.2.1 i Table 3.3.2.1 1 (page 1 of 1) f Control Rod Block Instrumentation i APPLICABLE MODES OR OTHER ' l SPECIFIED REQUIRED SURVEILL.ANCE ALLOWABLE l FUNCTION CONDITIONS CHANNELS REQUIREHENTS VALUE , i i
- 1. Rod Block Monitor !
- a. Low Power Range -Upscale (a) 2 SR 3.3.2.1.1 s 115.5/125 SR 3.3.2.1.4 divisions of j SR 3.3.2.1.7 full scale ;
- b. Intermediate Power (b) 2 SR 3.3.2.1.1 s 109.7/125 :
Range - Upscale SR 3.3.2.1.4 divisions of f l SR 3.3.2.1.7 full scale ; l
- c. High Power Range -Upscale (c),(d) 2 SR 3.3.2.1.1 s 105.9/125 !
SR 3.3.2.1.4 divisions of SR 3.3.2.1.7 full scale l t I d. Inop (d),(e) 2 SR 3.3.2.1.1 NA ' I i
- e. Downscale (d),(e) 2 SR 3.3.2.1.1 a 93/125 ;
SR 3.3.2.1.7 divisions of i _,,___ _ _ . -fdpffet- -&- - ...... .. _ ___ ___- _ . i
<v\ - 3.3.2.' 7
- 2. Rod Worth Minimizer 1(f) 2(f)
, 1 SR 3.3.2.1.2 NA j SR 3.3.2.1.3 :
SR 3.3.2.1.5 ! i SR 3.3.2.1.8 i i l 3. Reactor Mode Switch - Shutdown (g) 2 SR 3.3.2.1.6 NA ! l Position i l l (a) THERMAL POWER t 29% and < 64% RTP and MCPR < 1.70. i j (b) THERMAL POWER R 64% and < 84% RTP and MCPR < 1.70. l l (c) THERMAL POWER t 84% and < 90% RTP and MCPR < 1.70. 1 (d) THERMAL POWER t 90% RTP and MCPR < 1.40. (e) THERMAL PC.WER t 29% and < 90% RTP and MCPR < 1.70. , (f) With THERMAL POWER < 10% RTP. f
- (g) Reactor mode switch in the shutdown position.
- r 1
f I HATCH UNIT 2 3.3-20 Amendment No. 135
. . - _ . . . _ _~ _ . _ . -
d Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation, DE One recirculation loop shall be in operation with:
- a. The following limits applied when the associated LCO is applicable:
- 1. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specifieyintheCOLR;
- 2. LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR),"
single loop operation limits specified in the COLR;
- 3. LCO 3.3.1.1, " Reactor Protection System (RPS)
Instrum ion, tion 2.b (Ave. rage Power Range Monitor -- 9 =: Sh:::' imulated Thermal Power - le Value of Table 3.3.1.1-1 is reset for single loop operation; and
- b. Core flow as a function of core thermal power in the
" Operation Allowed Region" of Figure 3.4.1-1.
APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. No recirculation loops A.1 Place the reactor Immediately in operation. mode switch in the shutdown position. (continued) HATCH UNIT 2 3.4-1 Amendment No. 135
l SDM Test -- Refueling ! 3.10.8 l 1 l 3.10 SPECIAL OPERATIONS 3.10.8 SHUTDOWN MARGIN (SDM) Test - Refueling LC0 3.10.8 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/ hot standby position, and operation considered not to be in MODE 2, to allow SDM testing, provided the following requirements are met:
- a. LCO 3.3.1.1, " Reactor Protection System Instrumentation," MODE 2 requirements for Functions #
2.d; and 2.e of Table 3.3.1.1-1; i
- b. 1. LC0 3.3.2.1, " Control Rod Block Instrumentation," l MODE 2 requirements for Function 2 of Table i 3.3.2.1-1, with the banked position withdrawal l sequence requirements of SR 3.3.2.1.8 changed to '
require the control rod sequence to conform to the SDM test sequence, Q8
- 2. Conformance to the approved control rod sequence for the SDM test is verified by a second licensed operator or other qualified member of the technical staff;
- c. Each withdrawn control rod shall be coupled to the associated CRD; j
- d. All control rod withdrawals during out of sequence l control rod moves shall be made in notch out mode;
- e. No other CORE ALTERATIONS are in progress; and j
- f. CRD charging water header pressure 2: 940 psig.
APPLICABILITY: MODE 5 with the reactor mode switch in startup/ hot standby position, i HATCH UNIT 2 3.10-20 Amendment No. 13"
i SDM Test -- Refueling 3.10.8 [ SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l l SR 3.10.8.1 Perform the MODE 2 applic e S s for According to LCO 3.3.1.1, Functions .a and e of the applicable l j^ Table 3.3.1.1-1. SRs i 2.d
- SR 3.10.8.2 -------------------NOTE--------------------
i Not required to be met if SR 3.10.8.3 satisfied. Perform the MODE 2 applicable SRs for According to LC0 3.3.2.1, Function 2 of Table 3.3.2.1-1. the applicable SRs l l SR 3.10.8.3 -------------------NOTE-------------------- l Not required to be met if SR 3.10.8.2 l satisfied. l l Verify movement of control rods is in During control compliance with the approved control rod rod movement sequence for the SDM test by a second licensed operator or other qualified member , of the technical staff. ' SR 3.10.8.4 Verify no other CORE ALTERATIONS are in 12 hours progress. (continued) l I i. i i i HATCH UNIT 2 3.10-22 Amendment No. 135
Enclosure 3B . . l Edwin I. Hatch Nuclear Plant . Request to Revise Technical Specifications: l
^
Oscillation Power Range Monitor Page Change Instructions and Revised Pages , Unit 1 Remove Pagg Insert Page l 3.3-1 3.3-1 3.3-2 3.3-2 3.3-3 3.3-3 3.3-4 3.3-4 3.3-5 3.3-5 ! 3.3-6 3.3-6 l l 3.3-7 3.3-7 l l 3.3-8 3.3-8 - 3.3-8a 3.3-8b 3.4-1 3.4-1
- 3.4-2 3.4-2 j 3.4-3 3.4-3 3.4-4 3.4-4 t
Unit 2 ) i , ! ! Remove Pace Insert Pace' l l l 3.3-1 3.3-1 3.3-2 3.3-2 l 3.3-3 3.3-3 l 3.3-4 3.3-4 l 3.3-5 3.3-5 l 3.3-6 3.3-6 3.3-7 3.3-7 3.3-8 3.3-8 3.3-8a 3.3-8b 3.4 1 3.4-1 3.4-2 3.4-2 i 3.4-3 3.4-3 l 3.4-4 3.4-4 d HL-5054 E3B-1
RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation l LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. ; APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS
NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours , channels inoperable. trip. ; 1 M A.2 --------NOTE--------- l Not applicable for Functions 2.a, 2.b, 2.c, 2.d and 2.f. l i Place associated trip 12 hours I system in trip. B. --------NOTE---------- B.1 Place channel in one 6 hours Not applicable for trip system in trip. Functions 2.a, 2.b, 2.c, 2.d, and 2.f. E l B.2 Place one trip system 6 hours One or more Functions in trip, with one or more required channels inoperable in both trip systems. (continued) HATCH UNIT 1 3.3-1 Proposed OPRM 7/31/96
RPS InstrumentatiCn 3.3.1.1 i
. ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME l l
C. One or more Functions C.1 Restore RPS trip I hour ; with RPS trip capability. .
. capability not maintained.
D. Required Action and D.1 Enter the Condition Immediately ! associated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for B, or C not met. the channel, f E. As required by E.1 Reduce THERMAL POWER 4 hours l Required Action D.1 to < 30% RTP. i and referenced in Table 3.3.1.1-1. ! F. As required by F.1 Be in MODE 2. 6 hours Required Action D.1 ; and referenced in Table 3.3.1.1-1. l G. As required by G.1 Be in MODE 3. 12 hours Required Action D.1 i and referenced in Table 3.3.1.1-1. ! H. As required by H.1 Initiate action to Immediately l Required Action D.1 fully insert all l and referenced in insertable control Table 3.3.1.1-1. rods in core cells containing one or , more fuel assemblies. (continued) ) l HATCH UNIT 1 3.3-2 Proposed OPRM 7/31/96 l
-._._.-.._._.._..._m ,.
RPS Instrumentation 3.3.1.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME I. As required by I.1 Initiate alternate 12 hours ! Required Action D.1 method to detect and ; and referenced in suppress thermal- ; Table 3.3.1.1-1. hydraulic instability
-> oscillations. ,
8.ND I.2 Restore required 120 days channels to OPERABLE. ; J. Requi;ted Action and J.1 Be in MODE 2. 4 hours , associated Completion Time of Condition I not met. SURVEILLANCE REQUIREMENTS , 1
-------------------------------------NOTES------------------------------------
- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.
- 2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions way be delayed for up to 6 hours provided the associated Function maintains RPS trip capability.
SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours (continued) HATCH UNIT 1 3.3-3 Proposed OPRM 7/31/96
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.2 ------------------NOTE-----=== =------ Not required to be performed until 12 hours after THERMAL POWER at 25% RTP. Verify the absolute difference between 7 days the average power range monitor (APRM) channels and the calculated power is s 2% RTP while operating at 2: 25% RTP. SR 3.3.1.1.3 (Not used.) l SR 3.3.1.1.4 ------------------NOTE------------------- Not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position (continued) j HATCH UNIT 1 3.3-4 Proposed OPRM 7/31/96 l
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.7 ------------------NOTE------------------- Only required to be met during entry into MODE 2 from MODE 1. _________________________________________ i Verify the IRM and APRM channels overlap. 7 days I SR 3.3.1.1.8 Calibrate the local power range monitors. 1000 effective I full power i hours SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days l I SR 3.3.1.1.10 ------------------NOTE------------------- For Function 2.a not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering l MODE 2. Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 184 days Turbine Control Valve Fast Closure, Trip Oil Pressure -- Low Functions are not bypassed when THERMAL POWER is 2: 30% RTP. SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months l (continued) l l \ f HATCH UNIT 1 3.3-5 Proposed OPRM 7/31/96 l l I
\
RPS Instrumentation i 3.3.1.1 i SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.13 ------------------NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
Perform CHANNEL CALIBRATION. 18 months SR 3.3.1.1.14 (Not used.)- SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months i SR 3.3.1.1.16 ------------------NOTE------------------- Neutron detectors are excluded. Verify the RPS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST BASIS SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 18 months Simulated Thermal Power is 2 25% and recirculation drive flow is < 60% of rated recirculation drive flow, i I HATCH UNIT 1 3.3-6 Proposed OPRM 7/31/96
RPS II:strs:,nentation 3.3.1.1 ; Table 3.3.1.1 1 (pose 1 of 3) , Reactor Protection system Instrumentation I i APPLICAsLE CONDITIONS MODES OR REGUIRED REFERENCED OTHER CHANNELS FROM ! SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE ! FUNCTION CONDITIONS STsTEM ACTION D.1 REQUIREMENTS VALUE j
- 1. Intermodlate Range Monitor !
- s. Neutron Flux - High 2 3 G sa 3.3.1.1.1 s 120/125 l SR 3.3.1.1.4 divisions of ;
st 3.3.1.1.6 full scale SR 3.3.1.1.7 ' SR 3.3.1.1.13 ' SR 3.3.1.1.15 5(*) 3 N SR 3.3.1.1.1 s 120/125 st 3.3.1.1.5 divisions of SR 3.3.1.1.13 futi scale [ SR 3.3.1.1.15 l
- b. Inop 2 3 G SR 3.3.1.1.4 NA st 3.3.1.1.15 SI ") - ' ' ' 3 N SR 3.3.1.1.5 NA sa 3.3.1.1.15 ,
- 2. Average Power Range !
Nonitor j
- a. Neutron Flux - High 2 3(C) G SR 3.3.1.1.1 s 20% RTP (setdown) sa 3.3.1.1.7 ,
sa 3.3.1.1.8 i SR 3.3.1.1.10 i st 3.3.1.1.13 I
- b. simulated Thermal 1 3(*) F SR 3.3.1.1.1 5 0.58 W + i Power - Migh SR 3.3.1.1.2 62% RTP and SR 3.3.1.1.8 s 115.5%
sa 3.3.1.1.10 Ryp(b) st 3.3.1.1.13
- c. Neutron Flux - High 1 3ICI F st 3.3.1.1.1 's 120% RTP sa 3.3.1.1.2 st 3.3.1.1.8 SR 3.3.1.1.10 sa 3.3.1.1.13
- d. Inop 1,2 3(c) G SR 3.3.1.1.10 NA (continued)
(a) With any control rod withdrawn from a core cett containing one or more fuel assemblies. (b) 0.58 W + 62% - 0.58 AW RTP when reset for single loop operation per LCO 3.4.1, " Recirculation Loops operating." (c) Each APRM channel provides inputs to both trip systems. HATCH UNIT 1 3.3-7 Proposed OPRM 7/31/96 l
RPS Instrumentation 3.3.1.1 Table 3.3.1.1 1 (page 2 of 3) Reactor Protection System Instrumentation APPLICA8LE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 2. Average Power Range Monitor (continued)
- e. Two-out of Four 1,2 2 G SR 3.3.1.1.1 NA voter SR 3.3.1.1.1D SR 3.3.1.1.15 SR 3.3.1.1.16
- f. OPRM Upscale 1 3(C) 1 SR 3.3.1.1.1 NA SR 3.3.1.1.8 SR 3.3.1.1.1D SR 3.3.1.1.13 SR 3.3.1.1.17
- 3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.1 s 1085 pois Dome Pressure - High SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15
- 4. Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 t 0 inches Level - Low, Level 3 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15
- 5. Main Steam isolation 1 8 F SR 3.3.1.1.9 s 1D% closed Valve -. Closure SR 3.3.1.1.13 SR 3.3.1.1.15
- 6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.1 5 1.92 psig SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15
- 7. Scram Discharge Volume Water Level - High
- e. Resistance Temperature 1,2 2 G SR 3.3.1.1.9 s 71 gallons Detector SR 3.3.1.1.13 SR 3.3.1.1.15 5(a) 2 H SR 3.3.1.1.9 s 71 gallons SR 3.3.1.1.13 SR 3.3.1.1.15
- b. Float Switch 1,2 2 G SR 3.3.1.1.13 5 71 gallons SR 3.3.1.1.15 5(a) 2 H SR 3.3.1.1.13 5 71 gallons SR 3.3.1.1.15 (continued) s (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.
(c) Each APRM channel provides inputs to both trip systems. HATCH UNIT 1 3.3-8 Proposed OPRM 7/31/96
RPS Instrumentation 3.3.1.1
'able 3.3.1.1 1 (page 3 of 3)
Reactor Protection system Instrtamentation APPLICAilLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOW 48LE FUNCTION CONDIT10Ns SYSTEM ACTION 0.1 REQUIREMENTS VALUE
- 8. Turbine stop t 30% RTP 4 E SR 3.3.1.1.9 5 10% closed velve - Closure st 3.3.1.1.11 SR 3.3.1.1.13 SR 3.3.1.1.15
- 9. Turbine Control Velve a 30% RTP 2 E st 3.3.1.1.9 2 600 psig Fast Closure, Trip 011 st 3.3.1.1.11 Pressure - Low SR 3.3.1.1.13 st 3.3.1.1.15 sa 3.3.1.1.16
- 10. Reactor Mode switch - 1,2 1 G st 3.3.1.1.12 NA shutdown Position SR 3.3.1.1.15 5(*) 1 M SR 3.3.1.1.12 NA l SR 3.3.1.1.15
- 11. Manuet scram 1,2 1 0 SR 3.3.1.1.5 NA st 3.3.1.1.15 5(*) 1 N st 3.3.1.1.5 NA l SR 3.3.1.1.15 l i
1 (a) With any control rod withdrawn from a core cell contelning one or more fuel assenbtles. l l i l l 1 HATCH UNIT 1 3.3-8a Proposed OPRM 7/31/96 l
RPS Instrumentation 3.3.1.1 i l l 1 l l This page intentionally left blank. l HATCH UNIT 1 3.3-8b Proposed OPRM 7/31/96 l
Recirculation Loops Operating 3.4.1 , 3.4 REACTOR COOLANT SYSTEM (RCS) ; 3.4.1 Recirculation Loops Operating i LC0 3.4.1 Two recirculation loops with matched flows shall be in operation, QB One recirculation loop shall be in operation with the l following limits applied when the associated LCO is ! applicable: a, LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE l (APLHGR)," single loop operation limits specified in the COLR;
- b. LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and
- c. LCO 3.3.1.1, " Reactor Protection System (RPS) l Instrumentation," Function 2.b (Average Power Range l Monitor Simulated Thermal Power - High), Allowable Value '
of Tabla 3.3.1.1-1 is reset for single loop operation. APPLICABILITY: MODES 1 and 2. ! l HATCH UNIT 1 3.4-1 Proposed OPRM 7/0 /96
Recirculation Loops Operating 3.4.1 i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 24 hours LCO not met. requirements of the LCO. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. , l E I No recirculation loops in operation. I l l i I l l HATCH UNIT 1 3.4-2 Proposed OPRM 7/31/96
Recirculation Loops Operating e 3.4.1 , I l SURVEILLANCE REQUIREMENTS
- SURVEILLANCE FREQUENCY
, SR 3.4.1.1 --------------------NOTE------------------- ! Not required to be performed until 24 hours after both recirculation loops are in a operation. i Verify recirculation loop jet pump flow 24 hours , mismatch with both recirculation loops in ! operation is: 1 l ' a. s 10% of rated core flow when operating at < 70% of rated core flow; and
- b. s 5% of rated core flow when operating at h 70% of rated core flow.
SF, 3.4.1.2 (Not used.) i 1 i HATCH UNIT 1 3.4-3 Proposed OPRM 7/31/96
3 Recirculation Leops Operating i 3.4.1 1 l !i s 1 i i 4 5 i 4 e i . } l i 1 1 f
, . I d }
t 1 l 1 1 1 .i 4 i i i : i i' i 4 ; i ! ) i 4 t 1 { i 1 1 i i i 1 4
's 1
4 i j -HATCH UNIT 1 3.4-4 Proposed OPRM 7/31/96 i
l RPS Instrumentation l 3.3.1.1 l 1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation l LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours channels inoperable. trip. QB A.2 --------NOTE--------- Not applicable for Functions 2.a 2.b, 2.c, 2.d, and 2.f. l Place associated trip 12 hours system in trip. B. --------NOTE--------- B.1 Place channel in one 6 hours Not applicable for trip system in trip. Functions 2.a, 2.b, 2.c, 2.d, and 2.f. QR l B.2 Place one trip system 6 hours l One or more Functions in trip. with one or more required channels i inoperable in both trip systems. i (continued) l HATCH UNIT 2 3.3-1 Proposed OPRM 7/31/96
i RPS Instrumentation 1 3.3.1.1 I ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME i C. One or more Functions C.1 Restore RPS trip 1 hour
- with RPS trip capability.
capability not maintained. D. Required Action and D.1 Enter the Condition Immediately
- associated Completion referenced in i Time of Condition A, Table 3.3.1.1-1 for i B, or C not met. _
the channel. 4 E. As required by E.1 Reduce THERMAL POWER 4 hours Required Action D.1 to < 30% RTP. J and referenced in Table 3.3.1.1-1. t i F. As required by F.1 Be in MODE 2. 6 hours 2 Required Action D.1 and referenced in Table 3.3.1.1-1. l G. As required by G.1 Be in MODE 3. 12 hours Required Action D.1 and referenced in Table 3.3.1.1-1. 1 i H. As required by H.1 Initiate action to Immediately Required Action D.1 fully insert all 1 and referenced in insertable control Table 3.3.1.1-1. rods in core cells containing one or
; more fuel assemblies.
i (continued) i HATCH UNIT 2 3.3-2 Proposed OPRM 7/31/96 l
4 , I RPS Instrumentation i j 3.3.1.1 i l 2 ACTIONS (continued) t i; CONDITION REQUIRED ACTION COMPLETION TIME ! j l l I. As required by I.1 Initiate alternate 12 hours ! i Required Action D.1 method to detect and i j and referenced in suppress thermal- !
- Table 3.3.1.1-1. hydraulic instability !
! oscillations. l' 1 E j I.2 Restore required- 120 days ; i channels to OPERABLE. I f J. Required Action and J.1 Be in MODE 2. 4 hours ; i associated Completion ' ! Time of Condition I , I not met.
- i
! I i !
) i l
SURVEILLANCE REQUIREMENTS l
! -------------------------------------NOTES------------------------------------
j 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS l Function.
- 2. When a channel is placed in an inoperable status solely for performance of
! required Surveillances, entry into associated Conditions and Required j Actions may be delayed for up to 6 hours provided the associated Function i maintains RPS trip capability. 1 4 I 1 i SURVEILLANCE FREQUENCY l-i i-SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours l i
- (continued) 1 l HATCH UNIT 2 3.3-3 Proposed OPRM 7/31/96 l
; 1 i
t
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.2 ------------------NOTE------------------- Not required to be performed until ; 12 hours after THERMAL POWER 2 25% RTP. l Verify the absolute difference between 7 days l the average power range monitor (APRM) channels and the calculated power is ! s 2% RTP while operating at 2 25% RTP. SR 3.3.1.1.3 (Not used.) SR 3.3.1.1.4 ------------------NOTE------------------- Not required to be performed when entering MODE 2 from NODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 7 days SR 3.3.1.1.5 Perform CHANNEL FUNCTIONAL TEST. 7 days i SR 3.3.1.1.6 Verify the source range monitor (SRM) and Prior to l intermediate range monitor (IRM) channels withdrawing I overlap. SRMs from the fully inserted position (continued) l HATCH UNIT 2 3.3-4 Proposed OPRM 7/31/96 l
1 RPS Instrumentation 3.3.1.1 SURVEILLANCE REOUIREMENTS (continued) SURVEILLANCE FREQUENCY l SR 3.3.1.1.7 ------------------NOTE------------------- Only required to be met during entry into MODE 2 from MODE 1. l Verify the IRM and APRM channels overlap. 7 days i l SR 3.3.1.1.8 Calibrate the local power range monitors. 1000 effective j full power i hours SR 3.3.1.1.9 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.10 ------------------NOTE------------------- For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 18 months Turbine Control Valve Fast Closure, Trip 011 Pressure -- Low Functions are not bypassed when THERMAL POWER is 1 30% RTP. SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months (continued) HATCH UNIT 2 3.3-5 Proposed OPRM 7/31/96 l
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.13 ------------------NOTES------------------
- 1. Neutron detectors are excluded.
- 2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
Perform CHANNEL CALIBRATION. 18 months SR 3.3.1.1.14 (Not used.) SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months 1 SR 3.3.1.1.16 ------------------NOTES------------------ j
- 1. Neutron detectors are excluded. 1
- 2. For Functions 3 and 4, channel sensors are excluded.
- 3. For Function 5, "n" equals 4 channels for the purpose of determining the l STAGGERED TEST BASIS Frequency.
Verify the RPS RESPONSE TIME is within 18 months on a l' limits. STAGGERED TEST BASIS i i SR 3.3.1.1.17 Verify 0PRM is not bypassed when APRM 18 months l Simulated Thermal Power is 2: 25% and recirculation drive flow is < 60% of rated recirculation drive flow. HATCH UNIT 2 3.3-6 Proposed OPRM 7/31/96
RPS Instrumentation I 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3) Reactor Protection system Instrumentation F APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED
- OTHER CHANNELS FROM
, SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE i l 1. Intermediate Range Monitor I a. Neutron Flux - Nigh 2 3 G SR 3.3.1.1.1 5 120/125 st 3.3.1.1.4 divisions of SR 3.3.1.1.6 futt scate i i st 3.3.1.1.7 ' SR 3.3.1.1.13 j sa 3.3.1.1.15
- 5(a) 3 H SR 3.3.1.1.1 s 120/125 j
- st 3.3.1.1.5 divisions of i
, st 3.3.1.1.13 futI scale sa 3.3.1.1.15 i b. Inop 2 3 G SR 3.3.1.1.4 NA j SR 3.3.1.1.15 i 5(*) 3 H SR 3.3.1.1.5 NA I sa 3.3.1.1.15 i
- 2. Average Power Range j Monitor
- a. Neutron Flux - High 2 3(c) G SR 3.3.1.1.1 s 20% RTP
, (setdown) SR 3.3.1.1.7 1 SR 3.3.1.1.8 a sa 3.3.1.1.10 i SR 3.3.1.1.13 f b. Simulated Theramt 1 3(c) F SR 3.3.1.1.1 s 0.58 W + 1 Power - High SR 3.3.1.1.2 62% RTP and , SR 3.3.1.1.8 s 115.5% i SR 3.3.1.1.10 nyp(b) sa 3.3.1.1.13 ! c. Neutron Flux - High 1 3(c) F st 3.3.1.1.1 5 120% RTP ! SR 3.3.1.1.2 st 3.3.1.1.8 ! SR 3.3.1.1.10 . st 3.3.1.1.13
- d. Inop 1,2 3(c) G SR 3.3.1.1.10 NA l
l 1 j (continued) a
- (a)- With any control rod withdrawn from a core cell containing one or more fuel assenbtles.
. (b) 0.58 W + 62% - 0.58 AW RTP when reset for single loop operation per LCO 3.4.1, " Recirculation Loops l Operating." (c) Each APRM channel provides inputs to both trip systems. 4 a 1 a 4 HATCH UNIT 2 3.3-7 Proposed OPRM 7/31/96 l 1
.. .. .. .. -. .. - - ~ _ . - - .. - . -.~_ . - _ -
l RPS Instrumentation 3.3.1.1 1 .
- Table 3.3.1.1 1 (page 2 of 3)
Reactor Protection System Instrumentation i-1 APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SilRVEILLANCE ALLOWASLE
- FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE I 2. Average Power Range i Monitor (continued)
J
- e. Two-out-of Four 1,2 2 G SR 3.3.1.1.1 NA ;
e voter SR 3.3.1.1.10 4 SR 3.3.1.1.15 SR 3.3.1.1.16 ] f. OPRM Upscale 1 3(c) I SR 3.3.1.1.1 NA 4 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.17
- 3. Reactor vesset Steam 1,2 2 G SR 3.3.1.1.1 s 1085 psig Dome Pressure - High SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16 4 Reactor Vessel Water 1,2 2 C SR 3.3.1.1.1 a 0 inches Level - Low, Level 3 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16
- 5. Main Steam Isolation 1 8 F SR 3.3.1.1.9 s 10% closed Valve - Closure SR 3.3.1.1.13 ,
SR 3.3.1.1.15 l SR 3.3.1.1.16
- 6. Drywell Pressure - High 1,2 2 G SR 3.3.1.1.1 s 1.92 pelo SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15
- 7. Scram Discharge Volume Water Level - High
- a. Resistance 1,2 2 G SR 3.3.1.1.9 s 57.15 l Temperature SR 3.3.1.1.13 gallons j Detector SR 3.3.1.1.15 5(a) 2 H SR 3.3.1.1.9 s 57.15 SR 3.3.1.1.13 gallons SR 3.3.1.1.15
- b. Float Switch 1,2 2 G SR 3.3.1.1.13 s 57.15 SR 3.3.1.1.15 gallons 5(*) 2 H SR 3.3.1.1.13 s 57.15 SR 3.3.1.1.15 gal lons (continued)
(a) With any control rod withdrawn from a core cell containing one or more fuel assenblies. (c) Each APRM channel provides inputs to both trip systems. HATCH UNIT 2 3.3-8 Proposed OPRM 7/31/96
i RPS InstrumentatiCn 3.3.1.1 l Table 3.3.1.1 1 (psee 3 of 3) Reactor Protection system Instrtamentation l l l [ APPLICAsLE CONDITIONS ' l 40 des OR REQUIRED REFERENCED OTHER CHANNELS FROM i l SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWAsLE I FUNCTION CONDITIONS STsTEM ACTION D.1 REQUIREMENTS VALUE l 8. Turbine stop t 30% RTP 4 E sa 3.3.1.1.9 s 10% closed l valve - Closure st 3.3.1.1.11 st 3.3.1.1.13 st 3.3.1.1.15 i st 3.3.1.1.16
- 9. Turbine Control Velve t 30% RTP 2 E SR 3.3.1.1.9 t 600 psig Fast Closure, Trip Oil SR 3.3.1.1.11 Pressure - Low st 3.3.1.1.13 SR 3.3.1.1.15 SR 3.3.1.1.16
- 10. Reactor Mode switch - 1,2 2 G SR 3.3.1.1.12 NA shutdown Position SR 3.3.1.1.15 5(*) 2 H $R 3.3.1.1.12 NA sa 3.3.1.1.15
! 11. Manual scram 1,2 2 G st 3.3.1.1.5 NA , SR 3.3.1.1.15 I 5(*) 2 H st 3.3.1.1.5 NA l SR 3.3.1.1.15 l l l (a) With any control rod with& sun from a core cell containing one or more fuel essemblies. l l l i i i i i i l l l l l i 1 HATCH UNIT 2 3.3-8a Proposed OPRM 7/31/96 l l
I l l I 1 l RPS Instrumentation 3.3.1.1
.i
- )
This page intentionally left blank. I e i i e I i l HATCH UNIT 2 3.3-8b Proposed OPRM 7/31/96 l l l i
l Recirculation Lesps Operating , 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating , LCO 3.4.1 Two recirculation loops with matched flows shall be in ! operation, 08 J One recirculation loop shall be in operation with the l following limits applied when the associated LCO is applicable:
- a. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE l (APLHGR)," single loop operation limits specified in the COLR; i
- b. LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and
- c. LC0 3.3.1.1, " Reactor Protection System (RPS) l Instrumentation," Function 2.b (Average Power Range .
Monitor Simulated Thermal Power -- High), Allowable Value ! of Table 3.3.1.1-1 is reset for single loop operation, j APPLICABILITY: MODES I and 2. I I l l 4 i HATCH UNIT 2 3.4-1 Proposed OPRM 7/31/96 l
Recirculation Loops Operating 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 24 hours LCO not met. requirements of the LCO. B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. DE No recirculation loops in operation. HATCH UNIT 2 3.4-2 Proposed OPRM 7/31/96
b i Recirculation loops Operating ; , 3.4.1 i SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
! SR 3.4.1.1 --------------------NOTE-------------------
!* Not required to be performed until 24 hours after both recirculation loops are in operation.
! Verify recirculation loop jet pump flow 24 hours
- mismatch with both recirculation loops in J
operation is: I
- a. s 10% of rated core flow when operating at < 70% of rated core flow; and
- b. s 5% of rated core flow when operating at 2: 70% of rated core flow.
SR 3.4.1.2 (Not used.) HATCH UNIT 2 3.4-3 Proposed OPRM 7/31/96
_ _ ~ . _ . . _ . .- _.__.. _._. . . - _ . . . _ _ _ - _ _ . 1 l Recirculation Lo:ps Operating f 3.4.1 . 1 i l l i i l l i l ! i i ,i 1 HATCH UNIT 2 3.4-4 Proposed OPRM 7/31/96
RPS Instrumentatien 3.3.1.1 3.3 INSTRUMENTATION i 3.3.1.1 Reactor Protection System (RPS) Instrumentation ' LC0 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. i APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS l
-------------------------------------NOTE-------------------------------------
Separate Condition entry is allowed for each channel. ! 1 i CONDITION REQUIRED ACTION COMPLETION TIME ; A. One or more required A.1 Place channel in 12 hours channels inoperable. trip.
.QB A.2 --------NOTE---------
Not appTI'dable C' l Fu ions 2.a, 2.b, l
- 2. , M 2.d, and 6.k
/
Place associated trip 12 hours system in trip. B. --------NOTE---------- B.1 Place channel in one 6 hours Not g 4caJb d r N trip system in trip. l N i Fu
- 2. ,ctions f# 2.d,2.a, 2.b, and 2. E _B / [
One or more Functions in trip. with one or more required channels ! inoperable in both trip systems. (continued) HATCH UNIT 1 3.3-1 Proposed Amendment No. 7/16/96
RPS Instrumentation > 3.3.1.1 i ACTIONS (continued) t . CONDITION REQUIRED ACTION COMPLETION TIME i ! ! C. One or more Functions C.1 Restore RPS trip 1 hour [ with RPS trip capability.
- capability not l maintained.
l D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for B, or C not met. the channel. l E. As required by E.1 Reduce THERMAL POWER 4 hours l Required Action D.1 to < 30% RTP. i and referenced in Table 3.3.1.1-1. F. As required by F.1 Be in MODE 2. 6 hours Required Action D.1 and referenced in Table 3.3.1.1-1. l G. As required by G.1 Be in MODE 3. 12 hours Required Action D.1 and referenced in Table 3.3.1.1-1. H. As required by H.1 Initiate action to Immediately Required Action D.1 fully insert all and referenced in insertable control l Table 3.3.1.1-1. rods in core cells containing one or more fuel assemblies.
- . m_
[/GSERT " A" A HATCH UNIT 1 3.3-2 Proposed Amendment No. 7/16/96 %
1 1 Insert 'A' - Technical Specifications 3.3.1.1 Reactor Protection System (RPS) . Instrumentation LCO I & J I. As required by Required I.1 Initiate alternate method to 12 hours Action D.1 and detect and suppress thermal-referenced in Table hydraulic instability ; 3.3.1.1-1. oscillations. AND l I.2 Restore required channels to 120 days OPERABLE. J. Required Action and 11 Be in MODE 2. 4 hours associated Completion i Time of Condition I not met. l l l l l
. , - ...-.._..-. .- -... _- - - _-. - .. - - . . _ . ~ . . . - _ -..
i RPS Instrumentation 3.3.1.1 i SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve -- Closure and 184 days Turbine Control Valve Fast Closure, Trip Oil Pressure -- Low Functions are not bypassed when THERMAL POWER is 2: 30% RTP. . SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months ! t l SR 3.3.1.1.13 ------------------NOTES------------------ !
- 1. Neutron detectors are excluded.
- 2. For Function 1, not required to be !
performed when entering MODE 2 from , MODE 1 until 12 hours after entering ' MODE 2. l Perform CHANNEL CALIBRATION. 18 months l l l l SR 3.3.1.1.14 (Not used.) ;d([ . ! SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months l l l , SR 3.3.1.1.16 ------------------NOTE------------------- OS$".!b!I!S$.-*$ISS$---_------- l Verify the RPS RESPONSE TIME is within 18 months on a l limits. STAGGERED TEST BASIS
-A lRf5ERT "6 i.
HATCH UNIT 1 3.3-5 Proposed Amendment No. 7/16/96
}
Insen 'B' - Technical Specifications 3.3.1.1 Reactor Protection System (RPS) Surveillance Requirements 3.3.1.1.17 SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 18 months Simulated Thermal Power is 2 25% and recirculation drive flow is <60% of rated recirculation drive flow. ; I I l l i i
4 I i RPS Instrumentation-i 3.3.1.1 4 1
- Table 3.3.1.1 1 (page 2 of 3) 4 Reactor Protection System Instrumentation i' j
j APPL!CAgLE . CONDITIONS j MODES OR REQUIRED REFERENCED
- OTHER CHANNELS FROM j SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWASLE i
FUNCTION CONDITIONS SYSTEN ACTION D.1 REQUIREMENTS VALUE I
- 2. Average Power Range
)
- i. Monitor (continued) 1 4
- c. Neutron Flux - High 3(c) 1 y $g 3,3,3,j,3 SR 3.3.1.1.2 s 120K RTP
[ l i !, SR 3.3.1.1.8 ' SR 3.3.1.1.10 3 SR 3.3.1.1.13 f d. Inop 1,2 3(C) G SR 3.3.1.1.10 NA
- e. Two-out of-Four 1,2 2 G SR 3.3.1.1.1 NA i voter SR 3.3.1.1.10 '
SR 3.3.1.1.15 l SR 3.3.1.1.16
- 3. % ess 5 1,2 2 G SR 3.3.1.1.1 s 1085 pois Don 6 Pressure - High SR 3.3.1.1.9
, y e it SR 3.3.1.1.13 i t SR 3.3.1.1.15
- 4. React -#er set-11I.er 1,2 2 G SR 3.3.1.1.1 t 0 inches Level - Low, Level 3 - SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15
- 5. Main Steam Isolation 1 8 F SR 3.3.1.1.9 s 10E closed valve - Closure SR 3.3.1.1.13 SR 3.3.1.1.15
- 6. Drywell Pressure- High 1,2 2 G SR 3.3.1.1.1 s 1.92 pois SR 3.3.1.1.9-SR 3.3.1.1.13 SR 3.3.1.1.15 (continued)
(c) Each APRM channel provides inputs to both trip systems. , 1 l HATCH UNIT 1 3.3-7 Proposed Amendment No. 7/16/96 1 1
Insert 'C' - Technical Specifications 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3) Reactor Protection Instrumentation
- f. OPRM Upscale 1 3 M I SR 3.3.1.1.1 NA SR 3.3.1.1.8 l SR 3.3.1.1.10 i SR 3.3.1.1.13 i SR 3.3.1.1.17 l
i
1 l l Recirculation loops Operating i 3.4.1 ' l 3.4 REACTOR COOLANT SYSTEM (RCS) 4 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in ' . operation, M One recirculation loop shall be in operation withg
-ar hhe following limits applied when the associated LC0 is applicable: !
a I. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR; b/. LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; c f. LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation," Function 2.b (Average Power Range l Monitor Simulated Thermal Power - High), Allowable l l
- Value of Table 3.3.1.1-1 is reset for single loop l operation; and i w/' A. s c, .c , s , r.. . . r.. /. c s. - . w. s . <.. . A . A J \
^ N
'Z' . X. . '. L.. m,
<r- . .'E. "X.y"X. ". :,/~ ~c' m '.z Z"'3. . X. ..o.
/T"77 7 l APPLICABILITY: MODES 1 and 2.
1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME u- us a. a z . . , ,A . ./ , .11 nA / / n,1. J. L . L / s/A/Af
' 7"" ' N ' 'Z '"~7'"'" N ' " /7. .M Z.r- ! " fZ. . . '.[' 7 T't, 7. '! % L' Z ' 1.. /
r .. I (continued) HATCH UNIT 1 3.4-I Proposed Amendment No. 7/16/96
Recirculation Loops 0 pirating 3.4.1 i ACTIONS (continued) j ! CONDITION REQUIRED ACTION COMPLETION TIME i l l ! d u!./-- s....r.+ u. ,... ./ / L. L /a.z- J , t ,JD i //TI LZ".Zi17,.;'.7 '/ / ',~Z"Z.I'J .:"3 '.1C 777 i
/ 4V.7:.'I'i::_ II.:~21 A Z;i'Z:J.~7C IC'/
I 1 ' r 'Z'_1sZ ' '"YI' 'Zi- I2: L 'Z T7
, 21' /"I' TA. 'Z '"
CIZ"; '"2 "'7-~f
- T,~,L 3 " Z '
- "7';
- ny_7- y ,..y. -. ';'Z. -.,
- .'. " A.". ,' . .,e/, /
i ! A / A l /. Requirements of the g.1 Satisfy the 24 hours ! h- s requirements of the
- LCO
.....,u_not met.Zul- '
LCO* b.Z31.L. . ..
. . . . _I'"Tc .,1",.'.l j ~ n 6 B i
- g. Required Action and 5.1 Be in MODE 3. 12 hours i associated Completion ;
- Time of Condition K+e- l
-C- not met. A l
l \ i OR t j do l'eCirCVh N W lMP jn operrd W . O f 4 J 4 i i a J I I HATCH UNIT 1 3.4-2 Amendment No. 195 3 l 4
Recirculation loops Operating 3.4.1 J-SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l 1 SR 3.4.1.1 --------------------NOTE------------------- Not required to be performed until 24 hours after both recirculation loops are in operation. l Verify recirculation loop jet pump flow 24 hours l mismatch with both recirculation loops in operation is:
- a. s; 10% of rated core flow when i operating at < 70% of rated core flow; I and
- b. s; 5% of rated core flow when operating at 2 70% of rated core flow.
(Nd usecL) SR 3.4.1.2 g- -g--7--,-- p ---- g--- j-
<"cI.C. 2X. . . ',,"._. . "Z, . '. 4. ." ' .' L. . . .."'"
.'4." .,
Z' ' 4. : 77..' 7 j.
,.per.ier../._____
A// YE n 5 z ifi M I5E':.i!cht i l 2 e rv-
._ .Z, ', ".':. ~"J.
7, . ' ;. .',.. _. ..'.. A. . ,;".,.?. , T ' " ' ' "'Y " ' ' 7 '" i l HATCH UNIT 1 3.4-3 Amendment No. 195
Recirculation Lo:ps Operating 3.4.1 i 1 l l l
\
n N / eo e _ oramnou mor Auowen s REGON 1 m j E E . .. l =~ \ I - oram non aumso
=__, Rosen l
l to
/
o- ,
/
CORE RDW p Mg e u. . . . w i_i A .,.. i X c is / /
'n'. ,%.%.. . s 1.
u.. y Ej. ,y y. . 2, ,' X.,.. 'y;fZ L:n, fr .07r
','.X, y :ti":. .
C 32 ...t 0= P.,,y .., *7.
. . rech 0 7 HATCH UNIT I 3.4-4 Amendment No. 197
RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LC0 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE. APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS
-------------------------------------NOTE-------------------------------------
- Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in 12 hours channels inoperable. trip. QB A.2 --------NOTE--------- I Not a lica fo \ Fun ons b / l
- 2. , 3st[ 2.d.a,dQ l
-g ,
Place associated trip 12 hours i system in trip. ; 1 l B. --------NOTE--------- B.1 Place channel in one 6 hours trip system in trip. y
'._ _ .7_. _
B.2 Place one trip system 6 hours One or more Tunctions in trip. with one or more required channels inoperable in both trip systems. (continued) HATCH UNIT 2 3.3-1 Proposed Amendment No. 7/16/96
. . --- -. . _ - . - . - = . - _ . . . . .. . .. . - - _ ..
RPS Instrumentation I 3.3.1.1 : j ACTIONS (continued) j CONDITION REQUIRED ACTION COMPLETION TIME
- C. One or more Functions C.1 Restore RPS trip 1 hour with RPS trip capability.
- capability not
- maintained.
l } D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for
; B, or C not met. the channel.
i i E. As required by E.1 Reduce THERMAL POWER 4 hours Required Action D.1 to < 30% RTP. and referenced in Table 3.3.1.1-1. F. As required by F.1 Be in MODE 2. 6 hours Required Action 0.1 and referenced in Table 3.3.1.1-1. G. As required by G.1 Be in MODE 3. 12 hours Required Action D.1 and referenced in Table 3.3.1.1-1. H. As required by H.1 Initiate action to Immediately Required Action 0.1 fully insert all and referenced in insertable control Table 3.3.1.1-1. rods in core cells containing one or more fuel assemblies. anserA HATCH UNIT 2 3.3-2 Proposed Amendment No. 7/16/96 l
h Insert 'A' - Technical Specifications 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO I & J l I. As required by Required I.1 Initiate alternate method to 12 hours ) Action D.1 and detect and suppress thermal referenced in Table hydraulic instability 3.3.1.1-1, oscillations. AND I.2 Restore required channels to 120 days OPERABLE. J. Required Action and J.1 Be in MODE 2. 4 hours associated Completion l Time ofCondition I not met. l l
RPS Instrumentation 3.3.1.1 ) 1 SURVEILLANCi,[ludIREMENTS (continued) l SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve - Closure and 18 months Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are not bypassed when THEr1 MAL POWER is 130% RTP. 1 l SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 18 months { SR 3.3.1.1.13 ------------------NOTES------------------ l
- 1. Neutron detectors are excluded.
- 2. For Function 1,'not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
Perform CHANNEL CALIBRATION. 18 months SR 3.3.1.1.14 (Not used.) I SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months l 7# (continued) _1 A % C F 3 j , HATCH UNIT 2 3.3-5 Proposed Amendment No. 7/16/96
Insert 'B' - Technical Specifications 3.3.1.1 Reactor Protection System (RPS) Surveillance Reauirements 3.3.1.1.17 SR 3.3.1.1.17 Verify OPRM is not bypassed when APRM 18 months Simulated Thermal Power is 2 25% and recirculation drive flow is <60% of rated l recirculation drive flow. i l l i l 1 j i i l l
4 RPS Instrumentation ! 3.3.1.1 t i
\
Table 3.3.1.1 1 (pose 2 of 3) l 1, Reactor Protection system Instrumentation i i APPLICABLE CONDITIONS MODES OR REeutRED REFERENCED OTNER CNANNELS FRtBI SPECIFIED PER TRIP REQUIRED FUNCTION SURVEILLANCE ALLOWABLE t CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALLE i 2. Averese Power Renee Monitor (continued) i c. Neutron Flum - Nigh 1 3(c) F sa 3.3.1.1.1 s 120E STP i- sa 3.3.1.1.2 4 sa 3.3.1.1.8 i sa 3.3.1.1.10 \ sa 3.3.1.1.13 ' (
- d. Insp 1,2
' 3(C) a sa 3.3.1.1.10 NA i d 5
. Two- -
1,2 2 ( Veter G sa 3.3.1.1.1 NA > i sa 3.3.1.1.10 u rasy.r c
- j. stees Dame Pressure - Nigh 1,2 2 G
- i:! 1:11 sa 3.3.1.1.1
(( ' 5 7005 peig sa 3.3.1.1.9 sa 3.3.1.1.13
- sa 3.3.1.1.15 sa 3.3.1.1.16 4 Reactor vessel Water 1,2 2 Level - Low, Level 3 G sa 3.3.1.1.1 t 0 inches sa 3.3.1.1.9 i SA 3.3.1.1.13 SR 3.3.1.1.15 d
- 5. sa 3.3.1.1.16 Mein steem isolation 1 8 4
Vaive - Ctoeure F sa 3.3.1.1.9 s 10% closed SR 3.3.1.1.13 4 st 3.3.1.1.15
- 6. sa 3.3.1.1.16 Drywell Pressure - High 1,2 2 G SR 3.3.1.1.1 s 1.92 pele 4
SR 3.3.1.1.9 Sa 3.3.1.1.13 sa 3.3.1.1.15 (continued) (c) Each APRM channet provides inputs to both trip systems. ?, HATCH UNIT 2 3.3-8 Proposed Amendment No. 7/16/96
. . . _ _ _ _ _ . - _ . . _ . . . _ _ _ _ . _ . _ _ _ . ~ - . _ . ~ . _ . . . _ _ _ . __ _ _ _. ~ _- _ .. _ _ . _ ___ .. _ -__
i 1 1 1 l Insert 'C' - Technical Specifications 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3) Reactor i Protection Instrumentation ! i l
\
- f. OPRM Upscale 1 3M I SR 3.3.1.1.1 NA !
SR 3.3.1.1.8 l SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.17 I
Recirculation Lo:ps Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation loops Operating LCO S.4.1 Two recirculation loops with matched flows shall be in , optration, ! QB One recirculation loop shall be in operation withy 6 hhe following limits applied when the associated LCO is applicable: 0/. LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR; b g. LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; C I. LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation," Function 2.b (Average Power Range Monitor Simulated Thermal Power - High), Allowable l Value of Table 3.3.1.1-1 is reset for single loop operation; and v .>.z a. - r, ..a.. w s uL 7' / ;" A-X.JerL .. A rYY1."',7"c' Z..T.".,v'L,M'..
.1 - .
M::'X 77T r 7' y n m. n.3,vn ,v. v.v... . , .g . ./ APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME
^ N Z.^ .Ac M u.41^- 4;;(;
A[,MPhc-4h-(cr+6
.. J. .X
'dddlf
,. . .Z. . . L. . . 7
.7..... y .. u....
.._., c.
Z.. g+ v! (continued) HATCH UNIT 2 3.4-1 Proposed Amendment No. 7/16/96
1 l Recirculation Lceps Operating 3.4.1 I ACTIONS (continued) i i COWITION REQUIRED ACTION COMPLETION TIME ! /u z , - . , . . . .. , _ _ _ 4/ 4. . ._ __ u . _ _ __ 2., m e u.... 1 W T'X'*7;I
~~
Il7.2:9 "* ' "' T i'.l;'.'II "r/ Z Z
' ~ ~ ' ~
i ' __ ;G"_Zi 7 1 K L' T :T_ IL 'U~ l .- V. ' '7 'I'.' R '
. 'E' 'J: &T s' , I'_17' X' ' L L: z,1x.:Y '" E iZ';'"11
"'J'~~
! /J'. '1". '_.' /. . ,' . i. /,.. T_g,7/.T Z ' ' '_2...'Z. '.
- , . ;;m .., .7. .., j
- n
' J A A
. equirements of the Satisfy the 24 hours
? LCO not met. {.1 requirements of the
- 4... m . . d. - LCO*
b kDi.,li L '_-"7-..:Y4 7 l 6 8 l k equired Action and K.1 Be in MODE 3. 12 hours J associated Completion j Time of Condition K er-E'not met. A J f 3 i i l 0R i i p}o recircvlGSi n l"#lS i l jn oper Gh 5 - i i .I i t I i HATCH UNIT 2 3.4-2 Amendment No. 135 i 4
t Recirculation Loops. Operating 3.4.1 i 1 ! SURVEILLANCE REQUIREMENTS } SURVEILLANCE FREQUENCY j SR 3.4.1.1 --------------------NOTE---- - ------------ l Not required to be performed until 24 hours l after both recirculation loops are in t operation. Verify recirculation loop jet pump flow 24 hours , mismatch with both recirculation loops in i operation is: i
- a. s 10% of rated core flow when
] operating at < 70% of rated core flow; and i b. s 5% of rated core flow when operating j at a: 70% of rated core flow. 1 l Wo+ used.) i SR 3.4.1.2 --------------------NOT6------------- u A -- d , i .1 A f. LJ.- r A .. x11 A L-i 2"J g_'r..'.y" 'Z...y:"u. ..,. z.. '.,"r,. . ..,I.,. .',:7, ., : I. '/ U LJ A RA A 1 MP tth d ash AA RZT, T-.' Z: Z ' %:'Z' .E_X;I U "7~ l 'g,.?f.
., ."". ,', G_ ,"I.
. , ,. _,l Z. . ".,'
, '. '7' ' ' ' "7 " '7 " -'
. .,'X.
1 1 4 i i i 4 1 i Y h ) HATCH UNIT 2 3.4 3 Amendment No. 135
Recirculatien Lorps Operating 3.4.1 ( l l n \ l so r oPBu mowmo RmGnoN < l
/ !
a 7 x 1 .- . cremnou Remou 30 ( 10 0- ,ii . . i g . . 4, i . . i g, i i . . g, consnowmmm Figure 3.1.1-1 (P:ge 1 ef 1) I:wcr Fl;; Oper: ting ":p with One R:::t:r C01 t Sy:t:= P.::ircul:tten L::p in Oper:tien HATCH UNIT 2 3.4-4 Amendment No. 138
Enclosure 4A . Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications: PRNhi Bases Changes and Associated Afarkups i l 1 HL-5054 E4A-1
RPS Instrumentatien B 3.3.1.1 BASES i APPLICABLE Averaae Power Ranae Monitor (APRM) SAFETY ANALYSES, LCO, and The APRM channels provide the primary indication of neutron APPLICABILITY flux within the core and respond almost instantaneously to (continued) neutron flux increases. The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP. The APRM System is divided into 4 APRM channels and 4 two-out-of-four voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The APRM System is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one unbypassed APRM will result in a " half-trip" in all four voter channels, but no trip inputs to either RPS trip system. A trip from any two unbypassed APRM channels will result in a full-trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip logic channel (A1, A2, B1, and B2). Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the entire core, consistent with the design bases for APRM Functions 2.a 2.b, and 2.c, at least 17 LPRM inputs, with at least three LPRM inputs from each of the 1 four axial levels at which the LPRMs are located, must be OPERABLE for each APRM channel. j 2.a. Averaae Power Ranae Monitor Neutron Flux - Hiah (Setdown) For operation at low power (i.e., MODE 2), the Average Power Range Monitor Neutron Flux - High (Setdown) Function is capable of generating a trip signal that prevents fuel i damage resulting from abnormal operating transients in this ! power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux - High (Setdown) l j Function will provide a secondary scram to the Intermediate Range Monitor Neutron Flux - High Function because of the relative setpoints. With the IRMs at Range 9 or 10, it is (continued) HATCH UNIT 1 B 3.3-7 PROPOSED REVISION 7/16/96
RPS Instrumentation j B 3.3.1.1 l BASES i 5 APPLICABLE 2.a. Averaae Power Ranae Monitor Neutron Flux - Hiah SAFETY ANALYSES, (Setdown) (continued) LCO, and l ; APPLICABILITY possible that the Average Power Range Monitor Neutron i Flux -High (Setdown) Function will provide the primary trip l signal for a corewide increase in power. No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux - High (Setdown) l Function. However, this Function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 25% RTP (SL 2.1.1.1) when operating at low reactor pressura and low core flow.
- Therefore, it indirectly prevents fuel damage during
; significant reactivity increases with THERMAL POWER 3
< 25% RTP.
I The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 25% RTP. The Average Power Range Monitor Neutron Flux - High (Setdown 'i control r)ods may be withdrawn since the potential forFunctionl must criticality exists. In MODE 1, the Average Power Range Monitor Neutron l Flux - High Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events. 2.b. Averaae Power Ranae Monitor Simulated Thermal l t Power - Hiah l The Average Power Range Monitor Simulated Thermal Power - l High Function monitors neutron flux to approximate the l THERMAL POWER being transferred to the reactor coolant. The l APRM neutron flux is electronically filtered with a time ' constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. Changes to fuel design include an evaluation of the time constant to determine if the electronic filter requires replacement. The trip level is varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed (continued) HATCH UNIT 1 B 3.3-8 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 ~ BASES d APPLICABLE 2.b. Averaae Power RADge Monitor Simulated Thermal l SAFETY ANALYSES, Power - Hiah (continued) LCO, and APPLICABILITY control rod pattern) but is clamped at an upper limit that is always lower than the Average Power Range Monitor Neutron l Flux - High Function Allowable Value. The Average Power Range Monitor Simulated Thermal Power - l 1 High Function provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating events and protects the fuel cladding integritybyensuringtbattheMCPRSLisnotexceeded. During these events, the THERMAL POWER increase does not . significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram before the high neutron flux scram. For rapid neutron flux increase events, 4 the THERMAL POWER lags the neutron flux and the Average Power Range Monitor Neutron Flux - High Function will l provide a scram signal before the Average Power Range Monitor Simulated Thermal Power - High Function setpoint and l associated time delay are exceeded. Each APRM channel uses one total drive flow signal representative of total core flow. The total drive flow signal is generated by the flow processing logic, which is part of the APRM channel. The flow is calculated by summing j two flow transmitter signals, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel OPERABILITY requirements for this Function. The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heating event. The time constant is based on the 1 fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER. ~ The Average Power Range Monitor Simulated Thermal Power - l High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high . pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for
- fuel cladding integrity.
(continued) HATCH UNIT 1 B 3.3-9 PROPOSED REVISION 7/16/96 i
- RPS Instrumentation 4 B 3.3.1.1 BASES
.i 1 APPLICABLE 2.c. Averaae Power Ranae Monitor Neutron Flux -- Hiah SAFETY ANALYSES, LCO, and The Average Power Range Monitor N utron Flux - High Function APPLICABILITY is capable of generating a trip signal to prevent fuel (continued) damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux - High Function is l assumed to terminate the main steam isolation valve (MSIV) l closure event and, along with the safety / relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 7) takes credit for the
- Average Power Range Monitor Neutron Flux - High Function to terminate the CRDA. l The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses.
4 The Average Power Range Monitor Neutron Flux - High Function l 1s required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux - High Function l is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High - (Setdown) Function conservatively bounds the assumed trip l
- and, together with the assumed IRM trips, provides adequate 4
protection. Therefore, the Average Power Range Monitor l Neutron Flux - High Function is not required in MODE 2. j 2.d. Averaae Power Ranoe Monitor - Inoo This Function (Inop) provides assurance that the minimum number of APRM channels is OPERABLE. i For any APRM channel, any time: 1) its mode switch is in any position other than " Operate," 2) an APRM module is urolugged, or 3) the automatic self-test system detects a critical fault with the APRM channel, an Inop trip signal is i sent to all four voter channels. Inop trips from two or ) i more enbypassed APRM channels result in a trip output from l l all four voter channels to their associated trip system. l l I (continued) HATCH UNIT 1 B 3.3-10 PROPOSED REVISION 7/16/96 I 4 l
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.d. Averaae Power Ranae Monitor-Inoo (continued) SAFETY ANALYSES, LCO, and This Function was not specifically credited in the accident APPLICABILITY analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis. . I i There is no Allowable Value for this Function. i i This Function is required to be OPERABLE in the MODES where the APRM Functions are required. l i 2.e. Two-out-of-Four Voter The Two-out-of-Four Voter Function provides the interface ; between the APRM Functions and the final RPS trip system ; logic. As such, it is required to be OPERABLE in the MODES ! where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions. Therefore, the Two-out-of-Four Voter Function is l required to be OPERABLE in MODES 1 and 2. l 1 All four voter channels are required to be OPERABLE. Each voter channel also incudes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, an Inop trip is issued from that voter channel to the associated trip system. There is no Allowable Value for this Function.
- 3. Reactor Vessel Steam Dome Pressure - Hiah An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 4, reactor scram (the analyses conservatively (continued)
HATCH UNIT 1 B 3.3-11 PROPOSED REVISION 7/16/96
) j RPS Instrumentation
- B 3.3.1.1 :
BASES l i l ! APPLICABLE 3. 2 h (continued) Reactor Vessel Steam Dome Pressure - H1 l l SAFETY ANALYSES, i LCO, and assume scram on the Average Power Range Monitor Neutron l 4 APPLICABILITY Flux - High signal, not the Reactor Vessel Steam Dome j Pressure - High signal), along with the S/RVs, limits the a peak RPV pressure to less than the ASME Section III Code ! limits.
- High reactor pressure signals are initiated from four 3
^ pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure - High Allowable Value is e chosen to provide a sufficient margin to the ASME Section III Code limits during the event. i Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in j ! a one-out-of-two logic, are required to be OPERABLE to '
- ensure that no single instrument failure will preclude a i
scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES I and 2 when the RCS is pressurized and the potential for pressure increase exists. 4 $ 4. Reactor Vessel Water Level - Low. Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel i Water Level - Low, Level 3 Function is assumed in the ' analysis of the recirculation line break (Ref. 3). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core : Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. i Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level - Low, Level 3 i Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. (continued) HATCH UNIT 1 B 3.3-12 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 4. Reactor Vessel Water Level - Low. Level 3 (continued) SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level - Low, Level 3 Allowable APPLICABILITY Value is selected to ensure that (a) during normal operation the steam dryer skirt is not uncovered (this protects available recirculation pump net positive suction head (NPSH) from significant carryunder) and, (b) for transients involving loss of all normal feedwater flow, initiation of the low pressure ECCS subsystems at Reactor Vessel Water - Low Low Low, Level I will not be required. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS resulting in the limiting transients and accidents. ECCS initiations at Reactor Vessel Water Level - Low Low, Level 2 and Low Low Low, Level 1 provide sufficient protection for level transients in all other MODES.
- 5. Main Steam Isolation Valve - Closure MSIV closure results in loss of the main turbine and the condenser as a heat sink for the nuclear steam supply system and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a '
Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss of the normal heat sink and subsequent overpressurization ' transient. However, for the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux - High Function, along with the S/RVs, limits l the peak RPV pranure to less than the ASME Code limits. That is, the direct scram on position switches for MSIV closure events is not assumed in the overpressurization analysis. Additionally, MSIV closure is assumed in the transients analyzed in Reference 2 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. (continued) HATCH UNIT 1 B 3.3-13 PROPOSED REVISION 7/16/96
i RPS Instrumentation B 3.3.1.1 BASES j-i APPLICABLE 5. Main Steam Isolation Valve - Closure (continued) L SAFETY ANALYSES, i LCO, and MSIV closure signals are initiated from >osition switches i APPLICABILITY located on each of the eight MSIVs. Eact MSIV has two i position switches; one inputs to RPS trip system A while the , other inputs to RPS trip system B. Thus, each kPS trip system receives an input from eight Main Steam Isolation Valve - Closure channels, each consisting of one position L switch. The logic for the Main Steam Isolation Valve - Closure Function is arranged such that either the ' inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur. In addition, certain combinations of valves closed in two lines j will result in a half-scram. I The Main Steam Isolation Valve - Closure Allowable Value is i specified to ensure that a scram occurs prior to a ! significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient. Sixteen channels of the Main Steam Isolation Valve -- Closure
- Function, with eight channels in each trip system, are i required to be OPERABLE to ensure that no single instrument
, failure will preclude the scram from this Function on a
, valid signal. This Function.is only required in MODE 1
! since, with_the MSIVs open and the heat generation rate j high, a pressurization transient can occur if the MSIVs close. In MODE 2, the heat generation rate is low enough so ! that the other diverse RPS functions provide sufficient 1 i protection. I l j 6. Drywell Pressure - Hiah i High pressure in the drywell could indicate a break in the !
- RCPB. A reactor scram is initiated to minimize the ;
3 possibility of fuel damage and to reduce the amount of i
- energy being added to the coolant and the drywell. The 1
- j. Drywell Pressure - High Function is a secondary scram signal r to Reactor Vessel Water' Level - Low, Level 3 for LOCA events I
inside the drywell. However, no credit is taken for a scram initiated from this Function for any of the DBAs analyzed in. j the FSAR. This Function was not specifically credited in d the accident analysis, but it is retained for the overall
, redundancy and diversity of the RPS as required by the NRC approved licensing basis. '
(continued) f 1 HATCH UNIT 1 B 3.3-14 PROPOSED REVISION 7/16/96 l I i
t RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 6. Drywell Pressure - Hiah (continued) SAFETY ANALYSES, LCO, and High drywell pressure signals are initiated from four APPLICABILITY pressure transmitters that sense drywell pressure. The 4 Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment.
- Four channels of Drywell Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function
- on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS, resulting i
in the limiting transients and accidents. l
- 7.a. and 7.b. Scram Discharae Volume Water Level - Hiah The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered.
Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water j from a full core scram. The two types of Scram Discharge
- Volume Water Level - High Functions are an input to the RPS 1
logic. No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the FSAR. However, they are retained 3 to ensure the RPS remains OPERABLE. i SDV water level is measured by two diverse methods. The
- level in each of the two SDVs is measured by two float type level switches and two thermal probes for a total of eight level signals. The outputs of these devices are arranged so i
that there is a signal from a level switch and a thermal , probe to each RPS logic channel. The level measurement instrumentation satisfies the recommendations of
- Reference 8.
The Allowable Value is chosen low enough to ensure that
- there is sufficient volume in the SDV to accommodate the water from a full scram.
4
; (continued)
HATCH UNIT 1 B 3.3-15 PROPOSED REVISION 7/16/96 l
i RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 7.a. and 7.b. Scram Discharae Volume Water Level -- Hiah SAFETY ANALYSES, (continued) LCO, and APPLICABILITY Four channels of each type of Scram Discharge Volume Water Level - High Function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal. These Functions are required in MODES 1 and 2, and in MODE 5 with any control , rod withdrawn from a core cell containing one or more fuel l assemblies, since these are the MODES and other specified l conditions when control rods are withdrawn. At all other times, this Function may be bypassed. 4
- 8. Turbine Stoo Valve - Closure Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on a TSV-Closure signal before the TSVs are completely closed in anticipation of the transients that would result from the closure of these valves. The Turbine Stop Valve - Closure Function is the primary scram signal for the turbine trip event analyzed in Reference 2. For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded.
Turbine Stop Valve - Closure signals are initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram. In addition, certain combinations of two valves closed will result in a half-scram. This Function must be enabled at THERMAL POWER 2: 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. (continued) HATCH UNIT 1 B 3.3-16 PROPOSED REVISION 7/16/96 l
RPS Instrumentation i B 3.3.1.1 BASES APPLICABLE 8. Turbine Ston Valve - Closure (continued) SAFETY ANALYSES, LCO, and The Turbine Stop Valve - Closure Allowable Value is selected APPLICABILITY to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient. E.9h t channels of Turbine Stop Valve - Closure Function, with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if the TSVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is 2 30% RTP. This Function is not required when THERMAL POWER is < 30%.RTP since the Reactor Vessel Steam Dome Pressure - High and the 1 Average Power Range Monitor Neutron Flux - High Functions l are adequate to maintain the necessary safety margins.
- 9. Turbine Control Valve Fast Closure. Trio Oil Pressure - Low- l Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Function is the primary scram signal for the generator load rejection event analyzed in Reference 2. For ,
this event, the reactor scram reduces the amount of energy l required to be absorbed and, along with the actions of the E0C-RPT System, ensures that the MCPR SL is not exceeded. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure transmitter is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic channel. This function must be enabled at THERMAL POWER 2 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. (continued) HATCH UNIT 1 B 3.3-17 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9. Turbine Control Valve Fast Closure. Trio Oil SAFETY ANALYSES, Pressure - Low (continued) LCO, and APPLICABII.ITY The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Allowable Value is selected high enough to
- detect imminent TCV fast closure.
Four channels of Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analysis assumptions, whenever THERMAL POWER is 2: 30% RTP. This Function is not required when THERMAL POWER is < 30% RTP, since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions l are adequate to maintain the necessary safety margins. I
- 10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function i provides signals, via the manual scram logic channels, directly to the scram pilot solenoid power circuits. These 4
manual scram logic channels are redundant to the automatic protective instrumentation channels and provide manual
- reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for 4 the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
The reactor mode switch is a single switch with two - channels, each of which provides input into one of the RPS manual scram logic channels. There is no Allowable Value for this function, since the channels are mechanically actuated based solely on reactor mode switch position. Two channels of Reactor Mode Switch - Shutdown Position 1 Function, with one channel in each manual scram trip system, are available and required to be OPERABLE. The Reactor Mode (continued) HATCH UNIT 1 B 3.3-18 PROPOSED REVISION 7/16/96
RPS Instrumentation ! B 3.3.1.1
- BASES APPLICABLE 10. Reactor Mode Switch - Shutdown Position (continued)
SAFETY ANALYSES, LCO, and Switch - Shutdown Position Function is required to be APPLICABILITY OPERABLE in MODES I and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
- 11. Manual Scram The-Manual Scram push button channels provide signals, via the manual. scram logic channels, directly to the scram pilot solenoid power circuits. These manual scram logic channels are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This l Function was not specifically credited in the accident ;
analysis but it is retained for the overall redundancy and i diversity of the RPS as required by the NRC approved licensing basis. There is one Manual Scram push button channel for each of the two RPS manual scram logic channels. In order to cause a scram it is necessary that each channel in both manual scram trip systems be actuated. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the ! position of the push buttons. i Two channels of Manual Scram with one channel in each manual scram trip system are available and required to be OPERABLE in MODES I and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. ACTIONS A Note has been provided to modify the ACTIONS related to RPS instrumentation channels. Section 1.3, Completion ' Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required (continued) HATCH UNIT 1 B 3.3-19 PROPOSED REVISION 7/16/96 l
RPS Instrumentation B 3.3.1.1 , BASES ACTIONS Actions of the Condition continue to apply for each (continued) additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel. A.1 and A.2 Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours has been shown to
- be acceptable (Refs. 9 and 12) to permit restoration of any l inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the ino)erable channel cannot be restored to OPERABLE status witlin the allowable out of service time, the channel or the I associated trip system must be placed in the tripped condition per Raquired Actions A.1 and A.2. Placing the i inoperable chaniel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability,
, restore capability to accommodate a single failure, and i allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip
- (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken.
, As noted, Required Action A.2 is not applicable for APRM Functions 2.a, 2.b, 2.c, and 2.d. Inoperability of one required APRM channel affects both trip systems; thus, Required Action A.1 must be satisfied. This is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. (continued) HATCH UNIT 1 B 3.3-20 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) Condition B exist: when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accoe_odate a single failure in either trip system. i Required Actions B.1 and 8.2 limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function). The reduced reliability of this logic arrangement was not evaluated in References 9 and 12 for the l 12 hour Completion Time. Within the 6 hour allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system. Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in l References 9 and 12, which justified a 12 hour allowable out l of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with l two inoperable channels could be in a more degraded state ' than a trip system with four inoperable channe.ls if the two inoperable channels are in the same Function while the four inoperable channels are all in different Funt+ inns). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in). If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip. The 6 hour Completion Time is judged acceptable based on the I remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram. ; I i (continued) HATCH UNIT I B 3.3-21 PROPOSED REVISION 7/16/96
4 I RPS Instrumentation B 3.3.1.1 BASES i ACTIONS B.1 and B.2 (continued) Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip system in trip would result in a scram or RPT), Condition D must be entered and its Required Action taken. As noted, Condition B is not applicable for APRM Functions 2.a, 2.b, 2.c, and 2.d. Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM two-out-of-four voter and other non-APRM channels for which condition B applies. For an inoperable APRM channel, Required Action A.1 must be l satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide Required Actions that are appropriate for the inoperability of APRM
. Functions 2.a, 2.b, 2.c, and 2.d, and these Functions are i not associated with specific trip systems as are the APRM i two-out-of-four voter and other non-APRM channels, 4 Condition B does not apply.
f.d j Required Action C.1 is intended to ensure that appropriate
- actions are taken if multiple, inoperable, untripped 4
channels within the same trip system for the same Function
- result in the Function not maintaining RPS trip capability.
A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given i Function on a valid signal. 1 The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. (continued) HATCH UNIT 1 B 3.3-22 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES ACTIONS U (continued) Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action ' of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.1. F.1. and G.1 If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Action E.1 is consistent with the Completion Time provided in LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)." M If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. l (continued) HATCH UNIT 1 B 3.3-23 PROPOSED REVISION 7/16/96 l l
RPS Instrumentation g B 3.3.1.1 l J BASES (continued) i l 1 l i SURVEILLANCE As noted at the beginning of the SRs, the SRs for each RPS ' REQUIREMENTS instrumentation Function are located in the SRs column of . Table 3.3.1.1-1. E The Surveillances are modified by a Note to indicate that , when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to
- 6 hours, provided the associated Function maintains RPS trip
! capability. Upon completion of the Surveillance, or
- expiration of the 6 hour allowance, the channel must be
, returned to OPERABLE status or the applicable Condition j t entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average i time required to perform channel Surveillance. That i analysis demonstrated that the 6 hour testing allowance does i not significantly reduce the probability that the RPS will trip when necessary. j SR 3.3.1.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures f that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter ! indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument ll channels monitoring the same parameter should read i approximately the same value. Significant deviations l between instrument channels could be an indication of i
- excessive ir.strument drift in one of the channels nr !
- something even more serious. A CHANNEL CHECK will detect l gross channel failure; thus, it is key to verifying the !
instrumentation continues to operate properly between each i CHANNEL CALIBRATION.
- Agreement criteria are determined by the plant staff based i on a combination of the channel instrument uncertaintias,
{~ including indication and readability. If a channel is
- outside the criteria, it may be an indication that the instrument has drifted outside its limit.
l-l l (continued) . HATCH UNIT 1 B 3.3-24 PROPOSED REVISION 7/16/96 l 1 1 _. __ ~_, _ . _ _ _ . . _ _ .
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 (continued) REQUIREMENTS The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.1.1.2 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8. A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at 2 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed ; within 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. SR 3.3.1.1.3 i (Not used.) SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. (continued) HATCH UNIT I B 3.3-25 PROPOSED REVISION 7/16/96
;i 1
RPS Instrumentation j B 3.3.1.1 4 1 BASES l 2 1 SURVEILLANCE SR 3.3.1.1.4 (continued)
] REQUIREMENTS As noted, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 i required IRM Functions cannot be performed in MODE 1 without l l' utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve ; hours is based on operating. experience and in consideration
- of providing a reasonable time in which to complete the SR.
{' A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is i based on reliability analysis (Ref. 9). 1 SR 3.3.1.1.5 ! A CHANNEL FUNCTIONAL TEST is performed on each required
)
channel to e m re that the entire channel will perform the
- intended function. A Frequency of 7 days provides an
} acceptable le.'el of system average availability over the i Frequency and is based on the reliability analysis of _ j Reference 9. (The Manual Scram Function's CHANNEL
, FUNCTIONAL TEST Frequency was credited in the analysis to ! extend many automatic scram Functions' Frequencies.)
l SR 3.3.1.1.6 and SR 3.3.1.1.7 4' . These Surveillances are established to ensure that no gaps
- in neutror. flux indication exist from subcritical to power i operation for monitoring core reactivity status.
! The overlap between SRMs and IRMs is required to be
- demonstrated to ensure that reactor power will not be l 1 increased into a neutron flux region without adequate :
i indication. This is required prior to withdrawing SRMs from l
- the fully inserted position since indication is being i j transitioned from the SRMs to the IRMs.
1 4 4 l 4 (continued)
! HATCH UNIT 1 B 3.3-26 PROPOSED REVISION 7/16/96 4
4 y -, -.-,-er -= +-a - - - - - - 4 -
-w+ <---- - --
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.6 and SR 3.3.1.1.7 (continued) REQUIREMENTS The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the
- system design will prevent further increases (by initiating a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APRMs concurrently have onscale readings such that the transition between MODE 1 and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap i between SRMs and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on range 1 before SRMs have reached the j upscale rod block.
As noted, SR 3.3.1.1.7 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap l- requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale orice in MODE 2). If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate 2 channel (s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition < should be declared inoperable. A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs. SR 3.3.1.1.8 t LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) j System. This establishes the relative local flux profile for appropriate representative input to the APRM System. 4 The 1000 effective full power hours Frequency is based on ensuring the nodal power uncertainty is within the licensing
- basis analysis.
i (continued) HATCH UNIT 1 B 3.3-27 PROPOSED REVISION 7/16/96 l 1
4 i RPS Instrumentation 4 B 3.3.1.1 BASES i i SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 4-REQUIREMENTS
- (continued) A CHANNEL FUNCTIONAL TEST is performed on each required
~ channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant 4 specific setpoint methodology. The 92 day Frequency of ! SR 3.3.1.1.9 is based on the reliability analysis of j Reference 9. l The 18 month Frequency is based on the need to perform this
- Surveillance under the conditions that apply during a plant l outage and the potential for an unplanned transient if the l Surveillance were performed with the reactor at power.
1 Operating experience has shown that these components usually l pass the Surveillance when performed at the 18 month l Frequency. SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure tt:at the entire channel will perform the intended function. Dr the APRM Functions, this test supplements the automat c self-test functions that operate i continuously in the APRM and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation fic s processing - applicable to Function 2.b only), the two-out-of-four voter channels, and the interface connections to the RPS trip systems from the voter channels. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 184 day Frequency of SR 3.3.1.1.10 is based on the reliability analysis of Reference 12. (NOTE: The actual voting logic of the two-out-of-four voter channels is tested as part of SR 3.3.1.1.15.) For Function 2.a, a Note that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1 is provided. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated Frequancy is not met per SR 3.0.2. (continued) i j HATCH UNIT 1 B 3.3-28 PROPOSED REVISION 7/16/96
J RPS Instrumentation l B 3.3.1.1 i l l BASES 1 SURVEILLANCE SR 3.3.1.1.11 l REQUIREMENTS
. (continued) This SR ensures that scrams initiated from the Turbine Stop 2
Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently ] bypassed when THERMAL POWER is a 30% RTP. This involves i calibration of the bypass channels. Adequate margins for
- the instrument setpoint methodologies are incorporated into
- the actual setpoint. Because main turbine bypass flow can i affect this setpoint nonconservatively (THERMAL POWER is
] derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THER'ML POWER 2 30% RTP to ensure that the calibration is-valid. If any bypass channel's setpoint is nonconservative-(i.e., the Functions are bypassed at 2 30% RTP, either due ! to open main turbine bypass valve (s) or other reasons), then j the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low i Functions are considered inoperable. Alternatively, the , bypass channel can be placed in the conservative condition j (nonbypass). If placed in the nonbypass condition (Turbine
- Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are enabled), this SR is j met and the channel is considered OPERABLE.
1 i The Frequency of 184 days is based on engineering judgment
- and reliability of the components. ,
i l 1 SR 3.3.1.1.13 l l j A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive ! calibrations, consistent with the plant specific setpoint i methodology. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, this SR also i includes a physical inspection and actuation of the switches. For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated
! recirculation loop flow channel.
(continued) { i HATCH UNIT 1 B 3.3-29 PROPOSED REVISION 7/16/96 l
RPS Instrumentation B 3.3.1.1 1 BASES SURVEILLANCE SR 3.3.1.1.13 (continued) REQUIREMENTS Note 1 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 effective full power , hours LPRM calibration against the TIPS (SR 3.3.1.1.8). A j second Note is provided that requires the IRM SRs to be l performed within 12 hours of entering MODE 2 from MODE 1.
- Te: ting of the MODE 2 IRM Functions cannot be performed in l MODE 1 without utilizing jumpers, lifted leads or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.
Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. l The Frequency of SR 3.3.1.1.13 is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. 4 SR 3.3.1.1.14 , (Not used.) . i SR 3.3.1.1.11 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the i , OPERABILITY of the required trip logic for a specific ' i channel. The functional testing of control rods i ! (LC0 3.1.3), and SDV vent and drain valves (LC0 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant i outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. (continued) HATCH UNIT 1 B 3.3-30 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued) j REQUIREMENTS Operating experience has shawn that these components usually pass the Surveillance when performed at the 18 month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM trip conditions at the two-out-of-four voter channel inputs to check all combinations of two tripped inputs to the two-out-of-four. logic in the voter channels and APRM related redundant RPS relays. SR 3.3.1.1.16
.This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the j accident analysis. This test may.be performed in one '
measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 10.' RPS RESPONSE TIME for APRM Two-out-of-Four Voter Function 2.e includes the output relays of the voter and the associated RPS relays and contactors. (The digital portions of the APRM and two-out-of-four voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration check the time base of the digital electronics.) Confirmation of the time base is adequate to assure required response times are met. Neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. This Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. (continued) HATCH UNIT 1 B 3.3-31 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1'.1 BASES (continued) REFERENCES 1. FSAR, Section 7.2.
- 2. FSAR, Chapter 14.
{ 3. FSAR, Section 6.5.
- 4. FSAR, Appendix M.
- 5. FSAR, Section 14.3.3.
- 6. NED0-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
- 7. FSAR, Sections 14.4.2 and 14.5.5.
4
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.
I 9. NED0-30851-P-A , " Technical Specification Improvement . Analyses for BWR Reactor Protection System," i March 1988. ! 10. Technical Requirements Manual.
- 11. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
i 12. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III. Stability Trip Function," October 1995. HATCH UNIT 1 B 3.3-32 PROPOSED REVISION 7/16/96
4 SRM Instrumentation
- B 3.3.1.2 BASES
- i j APPLICABLE System (RPS) Instrumentation"; IRM Neutron Flux - High and )
1 SAFETY ANALYSES Average Power Range Monitor (APRM) Neutron Flux - High i ! (continued) (Setdown) Functions; and LC0 3.3.2.1, " Control Rod Block I j Instrumentation." . I The SRMs have no safety function and are not assumed to
- function during any FSAR design basis accident or transient 1 i
analysis. However, the SRMs provide the only on scale l monitoring of neutron flux levels during startup and i
- refueling. Therefore, they are being retained in Technical l Specifications.
] 3 ! LCO During startup in MODE 2, three of the four SRM channels are required to be OPERABLE to monitor the reactor flux level prior to and during control rod withdrawal, subcritical multiplication and reactor criticality, and neutron flux level and reactor period until the flux level is sufficient to maintain the IRMs on Range 3 or above. All but one of the channels are required in order to provide a representation of the overall core response during those periods when reactivity changes are occurring throughout the Core. In MODES 3 and 4, with the reactor shut down, two SRM channels provide redundant monitoring of flux levels in the Core. In MODE 5, during a spiral offload or reload, an SRM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region of the core. Thus, CORE ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b), requirement that the bundles being spiral reloaded or spiral offloaded are all in a single fueled region containing at least one OPERABLE SRM is met. Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence). In nonspiral routine operations, two SRMs are required to be OPERABLE to provide redundant monitoring of reactivity (continued) HATCH UNIT 1 B 3.3-34 PROPOSED REVISION 7/16/96
- -- - - ~ . - . . _ - . - . - . - . - - - - . - - - - - .
$ .l j Control R d Block Instrumentatien i , B 3.3.2.1 l ! B 3.3 INSTRUMENTATION l i i i B 3.3.2.1 Control Rod Block Instrumentation j BASES !
- BACKGROUND Control rods provide the primary means for control of !
reactivity changes. Control rod block instrumentation !
- includes channel sensors, logic circuitry, switches, and l
! relays that are designed.to ensure that specified fuel 5 i design limits are not exceeded for postulated transients and e accidents. During high power operation, the rod block j monitor (RBM).provides protection for control rod withdrawal !
- error events. During low power operations, control rod l
] blocks from the rod worth minimizer (RWM) enforce specific l . control rod sequences designed to mitigate the consequences i i of the control rod drop accident (CRDA). During shutdown t j conditions, control rod blocks from the Reactor Mode l t Switch - Shutdown Position Function ensure that all control t rods remain inserted to prevent inadvertent criticalities. \ \ l The purpose of the RBM is to limit control rod withdrawal if ' j localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR i j Safety Limit (SL) violation. The RBM supplies a trip signal l ~;. to the Reactor Manual Control System (RMCS) to appropriately. ; inhibit control rod withdrawal during power operation above ! ! the low power range setpoint. The RBM has two channels, l
- either of which can initiate a control rod block when the '!
channel output exceeds the control rod block setpoint. One- j RBM channel inputs into one RMCS rod block circuit and the j 4 other RBM channel inputs into the second RMCS rod block t circuit. l ! 4 i F The RBM channel signal is generated by averaging a set of ! l local power range monitor (LPRM) signals at various core j heights surrounding the control rod being withdrawn. A i signal from one of the four redundant average power range , monitor (APRM) channels supplies a reference signal for one of the RBM channels, and a signal from another of the APRM channels supplies the reference signal to the second RBM channel. This reference signal is used to determine which RBM range setpoint (low, intermediate, or high) is enabled. , If the APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control ! rod is selected (Ref. 1). A rod block signal is also l I (continued) HATCH UNIT 1 B 3.3-42 PROPOSED REVISION 7/16/96 I
Control R:d Block Instrumentation , B 3.3.2.1 l i BASES I BACKGROUND generated if an RBM Downscale trip or an Inoperable trip (continued) occurs. The Downscale trip will occur if the RBM channel i signal decreases below the Downscale trip setpoint after the i RBM signal has been normalized. The Inoperable trip will !
- occur during the nulling (normalization) sequence, if
- the )
i RBM channel fails to null, too few LPRM inputs are 4 available, a module is not plugged in, or the function j switch is moved to any position other than " Operate." ;
- The purpose of the RWM is to control rod patterns during i i startup and shutdown, such that only specified control rod l sequences and relative positions are allowed over the !
operating range from all control rods inserted to 10% RTP. ) i The sequences effectively limit the potential amount and ! rate of reactivity increase during a CRDA. Prescribed ! i control rod sequences are stored in the RWM, which will ' initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored J sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses , l
- feedwater flow and steam flow signals to determine when the
! reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2). The RWM is a single l
- channel system that provides input into both RMCS rod block l i circuits. '
i With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This i Function prevents inadvertent criticality as the result of a
- control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the
. shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block
- circuit. A rod block in either RMCS circuit will provide a
- control rod block to all control rods.
l i (continued) f
- HATCH UNIT I B 3.3-43 PROPOSED REVISION 7/16/96
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.1 REQUIPEMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control System input. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The Frequency of 184 days is based on reliability analyses (Ref.11). SR 3.3.2.1.2 and SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL' TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with the prescribed sequence and verifying a control rod block occurs. This test is performed as soon as possible after the applicable conditions are entered. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until I hour after any control rod is withdrawn at < 10% RTP in MODE 2, and SR 3.3.2.1.3 is not required to be performed until I hour after THERMAL POWER is < 10% RTP in MODE 1. This allows entry into MODE 2 (and if entered during a shutdown, concurrent power reduction to < 10% RTP) for SR 3.3.2.1.2 and THERMAL POWER reduction to < 10% RTP in MODE 1 for SR 3.3.2.1.3 to perform the required Surveillances if the 92 day Frequency is not met per SR 3.0.2. The I hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The 92 day Frequencies are based on reliability analysis (Ref. 8). SR 3.3.2.1.4 The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's inaut to each RBM channel. Below the minimum power set)oint, tie RBM is automatically bypassed. These power Allowaale Values (continued) HATCH UNIT 1 B 3.3-50 PROPOSED REVISION 7/16/96
Control Rod Block Instrumentation B 3.3.2.1 BASES 2 SURVEILLANCE SR 3.3.2.1.8 (continued) REQUIREMENTS OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible. REFERENCES 1. FSAR, Section 7.5.8.2.3. , 2. FSAR, Section 7.2.2.4.
- 3. NEDC-30474-P, " Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvements (ARTS) Program for Edwin I. Hatch Nuclear Plants,"
. December 1983.
- 4. NEDE-240ll-P-A US, " General Electrical Standard Application for Reload Fuel," Supplement for United States, (revision specified in the COLR).
- 5. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),
" Amendment 17 to General Electric Licensing Topical Report NEDE-240ll-P-A," BWROG-8644, August 15, 1986.
- 6. NED0-21231, " Banked Position Withdrawal Sequence,"
January 1977. 4 ) 7. NRC SER, " Acceptance of Referencing of Licensing i Topical Report NEDE-240ll-P-A," " General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.
- 8. NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"
October 1988. 4
- 9. GENE-770-06-1, " Bases For Changes To Surveillance Test Intervals and Allowed Out-0f-Service Times For Selected Instrumentation Technical Specifications,"
February 1991.
- 10. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 11. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function," October 1995. j i HATCH UNIT 1 B 3.3-53 PROPOSED REVISION 7/16/96
E0C-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Stoo Valve - Closure SAFETY ANALYSES, LCO, and Closure of the TSVs and a main turbine trip result in the APPLICABILITY loss of a heat sink and increases reactor pressure, neutron (continued) flux, and heat flux that must be limited. Therefore, an RPT is initiated on a TSV - Closure signal before the TSVs are completely closed in anticipation of the effects that would result from closure of these valves. E0C-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient. Closure of the TSVs is determined by measuring the position of each valve. While there are two separate position switches associated with each stop valve, only the signal from one switch for each TSV is used, with each of the four channels being assigned to a separate trip channel. The logic for the TSV - Closure Function is such that two or more TSVs must be closed to produce an E0C-RPT. This Function must be enabled at THERMAL POWER 2 30% RTP. This is normally accomplished automatically by pressure transmitter: sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TSV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an E0C-RPT from this Function on a valid signal. The TSV - Closure Allowable Value is selected to detect imminent TSV closure. This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is a 30% RTP. I Below 30% RTP, the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor (APRM) Neutron Flux - l ; High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR Safety Limit. ; i Turbine Control Valve Fast Closure. Trio 011 Pressure - Low j i Fast closure of the TCVs during a generator load rejection results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TCV Fast Closure, Trip 011 Pressure - Low in anticipation of the transients that would result from the closure of these ; l I (continued) ! l HATCH UNIT 1 B 3.3-82 PROPOSED REVISION 7/16/96
E0C-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Control Valve Fast Closure. Trio 011 Pressure - Low SAFETY ANALYSES, (continued) i LCO, and i APPLICABILITY valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient. 1 Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve. There is one pressure transmitter associated with each control valve, and the signal from each transmitter is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip Oil Pressure - Low Function is such that two or more TCVs must be closed (pressure transmitter trips) to produce an E0C-RPT. This Function must be enabled at THERMAL POWER 2: 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TCV Fast l Closure, Trip 011 Pressure - Low, with two channels in each j trip system, are available and required to be OPERABLE to j ensure that no single instrument failure will preclude an i E0C-RPT from this Function on a valid signal. The TCV Fast Closure, Trip Oil Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure. This protection is required consistent with the safety analysis whenever THERMAL POWER is 2 30% RTP. Below 30% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Neutron Flux - High Functions of the RPS are l adequate to maintain the necessary margin to the MCPR Safety Limit. ACTIONS A Note has been provided to modify the ACTIONS related to E0C-RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable E0C-RPT instrumentation channels provide (continued) HATCH UNIT 1 B 3.3-83 PROPOSED REVISION 7/16/96
Recirculation Lecps Operating B 3.4.1 BASES APPLICABLE case (since the intact loop starts at a lower flow rate and
~ SAFETY ANALYSES' the core response is the same as if both loops were (continued) operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement.
The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 14 of the FSAR. A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3). The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop. operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Simulated l Thermal Power - High setpoint is in LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation." Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 5). LCO Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCI caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1,
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"),
(continued) HATCH UNIT 1 B 3.4-3 PROPOSED REVISION 7/16/96
Recirculation Losps Operating B 3.4.1 1 l BASES 1 LCO and APRM Simulated Thermal Power - High setpoint l (continued) (LCO 3.3.1.1) must be applied to allow continued operation consistent with the assumptions of Reference 3. In l addition, core flow as a function of core thermal power must ' be in the " Operation Allowed Region" cf Figure 3.4.1-1 to ensure core thermal-hydrauli: ssciMations do not occur. APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor I Coolant Recirculation System are necessary since there is I considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. I i ACTIONS A.1 and 8.1 Due to thermal-hydraulic stability concerns, operation of the plant with one recirculation loop is controlled by restricting the core flow to 2: 45% of rated core flow when THERMAL POWER is greater than the 76% rod line. This requirement is based on the recommendations contained in GE SIL-380, Revision 1 (Reference 4), which defines the region where the limit cycle oscillations are more likely to occur. If the core flow as a function of core thermal power is in the " Operation Not Allowed Region" of Figure 3.4.1-1, prompt action should be initiated to restore the flow-power cosMnation to within the Operation Allowed Region. The 2 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing core oscillations to be quickly detected. An immediate reactor scram is also required with no recirculation pumps in operation, since all forced circulation has been lost and the probability of thermal-hydraulic oscillations is greater. (continued) HATCH UNIT 1 B 3.4-4 PROPOSED REVISION 7/16/96
SDM Test - Refueling 8 3.10.8 BASES APPLICABLE CRDA analyses assume that the reactor operator follows SAFETY ANALYSES prescribed withdrawal sequences. For SDM tests performed (continued) within these defined sequences, the analyses of References 1 and 2 are applicable. However, for some sequences developed for the 3DM testing, the control rod patterns assumed in the safety analyses of References 1 and 2 may not be met. Thecciore, special CRDA analyses, performed in accordance with an NRC approved methodology, may be required to demonstrate the SDM test sequence will not result in unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain , normal test operations as well as postulated accidents I within the bounds of the appropriate safety analyses (Refs. I and 2). In addition to the added requirements for ; the RWM, Average Power Range Monitors, and control rod coupling, the notch out mode is specified for out of sequence withdrawals. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test. As described in LCO 3.0.7, compliance with Special 0)erations LCOs is optional, and therefore, no criteria of t1e NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases. LC0 As described in LC0 3.0.7, compliance with this Special Operations LC0 is optional. SDM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LC0 or its ACTIONS. For SDM tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is available. To provide additional scram protection beyond the normally required IRMs, the Average Power Range Monitors are also required to be OPERABLE (LC0 3.3.1.1, Functions 2.a, 2.d, and 2.e) as l though the reactor were in MODE 2. Because multiple control rods will be withdrawn and the reactor will potentially become critical, the approved control rod withdrawal sequence must be enforced by the RWM (LCO 3.3.2.1, (continued) HATCH UNIT 1 B 3.10-34 PROPOSED REVISION 7/16/96 I
SDM Test - Refueling B 3.10.8 BASES LC0 Function 2, MODE 2), or must be verified by a second (continued) licensed operator or other qualified member of the technical staff. To provide additional protection against an inadvertent criticality, control rod withdrawals that do not conform to the banked position withdrawal sequence specified in LC0 3.1.6, " Rod Pattern' Control," (i.e., out of sequence control rod withdrawals) must be made in the individual notched withdrawal mode to minimize the potential reactivity insertion associated with each movement. Coupling integrity of withdrawn control rods is required to minimize the probability of a CRDA and ensure proper functioning of the withdrawn control rods, if they are required to scram. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in progress. Furthermore, since the control rod scram function with the RCS at atmospheric pressure relies solely on the CRD accumulator, it is essential that the CRD charging water header remain pressurized. This Special Operations LC0 then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startup/ hot standby position, such that the SDM tests may be performed while in MODE 5. APPLICABILITY These SDM test Special Operations requirements are only applicable if the SDM tests are to be performed while in MODE 5 with the reactor vessel head removed or the head bolts not fully tensioned. Additional requirements during these tests to enforce control rod withdrawal sequences and restrict other CORE ALTERATIONS provide protection against potential reactivity excursions. Operations in all other MODES are unaffected by this LCO. ACTIONS /kl With one or more control rods discovered uncoupled during this Special Operation, a controlled insertion of each uncoupled control rod is required; either to attempt recoupling, or to preclude a control rod drop. This controlled insertion is preferred since, if the control rod fails to follow the drive as it is withdrawn (i.e., is
" stuck" in an inserted position), placing the reactor mode switch in the shutdown position per Required Action B.1 (continued)
HATCH UNIT 1 B 3.10-35 PROPOSED REVISION 7/16/96 l
! SDM Test - Refueling o B 3.10.8 i l BASES (continued) i i 4 SURVEILLANCE SR 3.10.8.1. SR 3.10.8.2. and SR 3.10.8.3 REQUIREMENTS
- LC0 3.3.1.1, Functions 2.a 2.d, and 2.e, made applicable in l this Special Operations LCO, are required to have their
- Surve111ances met to establish that this Special Operations
! LCO is being met. However, the control rod withdrawal j sequences during the SDM tests may be enforced by the RWM L (LC0 3.3.2.1, Function 2, MODE 2 requirements) or by a
- second licensed operator (Reactor Operator or Senior Reactor
! Operator) or other qualified member of the technical staff J (e.g., a qualified shift technical advisor or reactor , engineer). As noted, either the applicable SRs for the RWM i (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3). This latter . ) verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviatioas from the ; specified sequence. These Surveillances provide adeauate i assurance that the specified test sequence is being ; followed. j SR 3.10.8.4 Periodic verification of the administrative controls established by this LC0 will ensure that the reactor is operated within the bounds of the safety analysis. The
- 12. hour Frequency is intended to provide appropriate i assurance that each operating shift is aware of and verifies '
compliance with these Special Operations LCO requirements. SR 3.10.8.5 Coupling verification is performed to e.sure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the full-out notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This , Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved, as well as operating experience related to uncoupling events. (continued) HATCH UNIT 1 B 3.10-37 PROPOSED REVISION 7/16/96
l RPS Instrumentation
- B 3.3.1.1 j BASES APPLICABLE Averaae Power Ranae Monitor (APRM)
SAFETY ANALYSES, , LCO, and The APRM channels provide the primary indication of neutron APPLICABILITY flux within the core and respond almost instantaneously to (continued) neutron flux increases. The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous
- indication of average reactor power from a few percent to a greater than RTP.
- The APRM System is divided into 4 APRM channels and 4 two- I out-of-four voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter .
channels are divided into two groups of two each, with each ! i group of two providing inputs to one PPS trip system. The ! APRM System is designed to allow one LPRM channel, but no voter channels, to be bypassed. A trip from any one i unbypassed APRM will result in a " half-trip" in all four ; voter channels, but no trip inputs to either RPS trip : system. A trip from any two unbypassed APRM channels will ' result in a full-trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip
. logic channel (A1, A2, B1, and 82). Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate I
coverage of the entire core, consistent with the design bases for APRM Functions 2.a 2.b, and 2.c, at least 17 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be OPERABLE for each APRM channel. La. Averaae Power Ranae Monitor Neutron Flux - Hiah (Setdown) For operation at low power (i.e., MODE 2), the Average Power Range Monitor Neutron Flux - High (Setdown) Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide a secondary scram to the Intermediate l Range Monitor Neutron Flux - High function because of the relative setpoints. With the IRMs at Range 9 or 10, it is (continued) HATCH UNIT 2 B 3.3-7 PROPOSED REVISION 7/16/96
l RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Averaae Power Ranae Monitor Neutron Flux -- Hiah SAFETY ANALYSES, (Setdown) (continued) LCO, and l , APPLICABILITY possible that the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide the primary trip l signal for a corewide increase in power. No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux - High (Setdown) l Function. However, this Function indirectly ensures that before the reactor mode switch is-placed in the rea position, reactor power does not exceed 25% RTP (SL 2.1.1.1) > when operating at low reactor pressure and low core flow. Therefore, it indirectly prevents fuel damage during significant reactivity increases with THERMAL POWER
< 25%'RTP.
I The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 25% RTP. The Average Power Range Monitor Neutron Flux - High (Setdown) Function must be OPERABLE during MODE 2 when l control rods may be withdrawn since the potential for-criticality exists. In MODE 1, the Average Power Range Monitor Neutron l : Flux - High Function provides protection against reactivity ! transients and the RWM and rod block monitor protect against i control rod withdrawal error events.
?.b. Averaae Power Ranae Monitor Simulated Thermal l
)ower - Hiah The Average Power Range Monitor Simulated Thermal Power - l High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. The '
APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. Changes to fuel design include an evaluation of the time constant to determine if the electronic filter requires replacement. The trip level is_ varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed (continued) HATCH UNIT 2 B 3.3-8 PROPOSED REVISION 7/16/96
l RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Averaae Power Ranae Monitor Simulated Thermal l' SAFETY ANALYSES, Power - Hiah (continued) LCO, and APPLICABILITY control rod pattern) but is clamped at an upper limit that is always lower than the Average Power Range Monitor Neutron l Flux - High Function Allowable Value. The Average Power Range Monitor Simulated Thermal Power - l High Function provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event) and protects the fuel cladding integrity by ensuring that the MCPR SL is not exceeded. During these events, the THERMAL POWER increase does not significantly lag the neutron flux response and, because of j a lower trip setpoint, will initiate a scram before the high i neutron flux scram. For rapid neutron flux increase events, the THERMAL POWER lags the neutron flux and the Average ; Power Range Monitor Neutron Flux - High Function will l i provide a scram signal before the Average Power Range Monitor Simulated Thermal Power - High Function setpoint and l associated time delay are exceaded. Each APRM channel uses one total drive flow signal l representative of total core flow. The total drive flow i signal is generated by the flow processing logic, which is part of the APRM channel. The flow is calculated by summing two flow transmitter signals, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel OPERABILITY requirements for this Function. The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power - High Function for the mitigation of the loss'of feedwater heating event. The time constant is based on the fuel heat transfer dynamics and provides a signal , proportional to the THERMAL POWER. The Average Power Range Monitor Simulated Thermal Power - l High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL ap)11 cable to high pressure and core flow conditions (MCPR S.). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity. (continued) HATCH UNIT 2 B 3.3-9 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES 4 APPLICABLE 2.c. Averaae Power Ranae Monitor Neutron Flux - Hiah SAFETY ANALYSES, LCO, and The Average Power Range Monitor Neutron Flux - High Function APPLICABILITY is capable of generating a trip signal to prevent fuel (continued) damage or excessive RCS pressure. For the overpressurization protection analysis of Referrace 4, the Average Power Range Monitor Neutron Flux - High Function is l assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety / relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 7) takes credit for the Average Power Range Monitor Neutron Flux - High Function to terminate the CRDA. The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses. The Average Power Range Monitor Neutron Flux - High Function l is required to be OPERABLE in tiODE 1 where the potential consequences of the analyzed ;ransients could result in the SLs (e.g., MCPR and RCS prer ure) being exceeded. Although the Average Power Range Mcaitor Neutron Flux - High Function l is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High (Setdown) Function conservatively bounds the assumed trip l and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor l Neutron Flux - High Function is not required in MODE 2. Lg. Averaae Power Ranae Monitor - Inoo This Function (Inop) provides assurance that the minimum number of APRM channels is OPERABLE. For any APRM channel, any time: 1) its mode switch is in any position other than " Operate," 2) an APRM module is ; unplugged, or 3) the automatic self-test system detects a ! critical fault with the APRM channel, an Inop trip signal is 1 sent to all four voter channels. Inop trips from two or i more unbypassed APRM channels result in a trip output from i all four voter channels to their associated trip system. 1 (continued) j l HATCH UNIT 2 B 3.3-10 PROPOSED REVISION 7/16/96 i
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.d. Averaae Power Ranae Monitor - Inoo (continued) SAFETY ANALYSES, LCO, and This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and APPLICABILITY i diversity of the RPS as required by the NRC approved ! licensing basis. ! l There is no Allowable Value for this Function. , This Function is required to be OPERABLE in the MODES where l the APRM Functions are required. 2.e. Two-out-of-Four Voter The Two-out-of-Four Voter Function provides the interface between the APRM Functions and the final RPS trip system logic. As such, it is required to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions. Therefore, the Two-out-of-Four Voter Function is required to be OPERABLE in MODES 1 and 2. All four voter channels are required to be OPERABLE. Each voter channel also includes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, an Inop trip is issued from that voter channel l to the associated trip system. i There is no Allowable Value for this Function.
- 3. Reactor Vessel Steam Dome Pressure - Hiah An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 4, reactor scram (the analyses conservatively (continued)
HATCH UNIT 2 B 3.3-11 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 3. Reactor Vessel Steam Dome Pressure - Hiah (continued) SAFETY ANALYSES, LCO, and assume scram on the Average Power Range Monitor Neutron l APPLICABILITY Flux - High signal, not the Reactor Vessel Steam Dome Pressure - High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits. High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure - High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event. Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will 3reclude a scram from this Function on a valid signal. Tie Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.
- 4. Reactor Vessel Water Level - Low. Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 3). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel. Four channels of Reactor Vessel Water Level - Low, level 3 Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. (continued) HATCH UNIT 2 B 3.3-12 PROPOSED REVISION 7/16/96 ,
RPS Instrumentation )! B 3.3.1.1 , BASES l APPLICABLE 4. Reactor Vessel Water Level - Low. Level 3 (continued) ' i SAFETY ANALYSES, LCO, and The Reactor Vessel Water Level - Low, Level 3 Allowable 1 APPLICABILITY Value is selected to ensure that (a) during normal operation - the steam dryer skirt is not uncovered (this protects 1 available recirculation pump net positive suction head '
, (NPSH) from significant carryunder) and, (b) for transients i involving loss of all normal feedwater flow, initiation of
- the low )ressure ECCS subsystems at Reactor Vessel Water .ow Low Low, Level I will not be required.
The Function is required in MODES 1 and 2 where considerable i energy exists in the RCS resulting in the limiting i transients and accidents. ECCS initiations at Reactor
;' Vessel Water Level - Low Low, Level 2 and Low Low Low,
- Level 1 provide sufficient protection for level transients in all other MODES.
i
- 5. Main Steam Isolation Valve - Closure 4
MSIV closure results in loss of the main turbine and the i condenser as a heat sink for the nuclear steam supply system i and indicates a need to shut dowr. the reactor to reduce heat l generation. Therefore, a reactor scram is initiated on a '
- Main Steam Isolation Valve -- Closure signal before the MSIVs
- are completely closed in anticipation of the complete loss
! of the normal heat sink and subsequent overpressurization
; transient. However, for the overpressurization protection i analysis of Reference 4, the Average Power Range Monitor i Neutron Flux - High Function, along with the S/RVs, limits l the peak RPV pressure to less than the ASME Code limits.
- That is, the direct scram on position switches for MSIV j closure events is not assumed in the overpressurization i analysis. Additionally, MSIV closure is assumed in the j transients analyzed in Reference 2 (e.g., low steam line j pressure, manual closure of MSIVs, high steam line flow).
l' The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel )eak cladding temperature remains below the limits of 10 C;R 50.46.
]
i a (continued) i i HATCH UNIT 2 B 3.3-13 PROPOSED REVISION 7/16/96 f
RPS Instrumentation B 3.3.1.1 BASES l APPLICABLE 5. Main Steam Isolation Valve - Closure (continued) SAFETY ANALYSES LCO, and MSIV closure signals are initiated from position switches APPLICABILITY located on each of the eight MSIVs. Each MSIV has two ! position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve - Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve - Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur. In addition, certain combinations of valves closed in two lines will result in a half-scram. The Main Steam Isolation Valve - Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient. Sixteen channels of the Main Steam Isolation Valve - Closure Function, with eight channels in each trip system, are , required to be OPERABLE to ensure that no single instrument l failure will valid signal. preclude the scram This Function from is only this Function required in MODEon1a since, with the MSIVs open and the heat generation rate high, a pressurization transient can occur if the MSIVs close. In MODE 2, the heat generation rate is low enough so that the other diverse RPS functions provide sufficient protection.
- 6. Drywell Pressure - Hiah High pressure in the drywell could indicate a break in the RCPB. A reactor scram is initiated to minimize the possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell. The Drywell Pressure - High Function is a secondary scram signal '
to Reactor Vessel Water Level - Low, Level 3 for LOCA events inside the drywell. However, no credit is taken for a scram initiated from this Function for any of the DBAs analyzed in the FSAR. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis. (continued) HATCH UNIT 2 B 3.3-14 PROPOSED REVISION 7/16/96 l
. i ' RPS Instrumentation
- 8 3.3.1.1
-BASES i
'. APPLICABLE _6 . Drvwell Pressure - Hiah (continued) l l SAFETY ANALYSES,
- LCO, and High drywell pressure signals are initiated from four
- APPLICABILITY pressure transmitters that sense drywell pressure. The
- Allowable Value was selected to be as low as possible and indicative of a LOCA inside primary containment. !
Four channels of Drywell Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required in MODES 1 and 2 where considerable energy exists in the RCS, resulting in the limiting transients and accidents. 7.a. and 7.b. Scram Discharae Volume Water Level - Hiah The SDV receives the water displaced by the motion of the CRD pistons during a reactor scram. Should this volume fill to a' point where there is insufficient volume to accept the displaced water, control rod insertion would be hindered. Therefore, a reactor scram is initiated while the remaining free volume is still sufficient to accommodate the water from a full core scram. The two types of Scram Discharge Volume Water Level - High Functions are an input to the RPS logic. No credit is taken for a scram initiated from these Functions for any of the design basis accidents or transients analyzed in the FSAR. However, they.are retained to ensure the RPS remains OPERABLE. SDV water level is measured by two diverse methods. The level in each of the two SDVs is measured by two float type level switches and two thermal probes for a total of eight level signals. The outputs of these devices are arranged so that there is a signal from a level switch and a thermal probe to each RPS logic channel. The level measurement instrumentation satisfies the recommendations of Reference 8. The Allowable Value is chosen low enough to ensure that there is sufficient volume in the SDV to accommodate the water from a full scram. (continued) HATCH UNIT 2 B 3.3-15 PROPOSED REVISION 7/16/96 l
l - RPS Instrumentation B 3.3.1.1 i 3
. BASES APPLICABLE 7.a. and 7.b. Scram Discharae Volume Water Level - Hiah SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY Four channels of each type'of Scram Discharge Volume Water Level - High Function, with two channels of each type in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from these Functions on a valid signal. These Functions are required in MODES I and 2, and in M00E 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. At all other times, this Function may be bypassed.
)
- 8. Turbine Ston Valve - Closure !
Closure of the TSVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on a TSV-Closure signal before the TSVs are completely closed in anticipation of the transients that - would result from the closure of these valves. The Turbine Stop Valve - Closure Function is the primary scram signal for the turbine trip event analyzed in Reference 2. For this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded. Turbine Stop Valve - Closure. signals are initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram. In addition, certain combinations of two valves closed will result in a half-scram. This Function must be enabled at THERMAL POWER a: 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. (continued) HATCH UNIT 2 B 3.3-16 PROPOSED REVISION 7/16/96 l l
RPS Instrumentation l B 3.3.1.1 1 BASES APPLICABLE 8. Turbine Stoo Valve - Closure (continued) SAFETY ANALYSES, LC0 and The Turbine Stop Valve - Closure Allowable Value is selected i , APPLICABILITY to be high enough to detect imminent TSV closure, thereby l
- reducing the severity of the subsequent pressure transient.
1 Eight channels of Turbine Stop Valve - Closure Function, with four channels in each trip system, are required to be , OPERABLE to ensure that no single instrument failure will l preclude a scram from this Function if the TSVs should l l close. This Function is required, consistent with analysis I assumptions, whenever THERMAL POWER is 2 30% RTP. This i Function is not required when THERMAL POWER is < 30% RTP 1 since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High. Functions l i are adequate to maintain the necessary safety margins. l
- 9. Turbine Control Valve Fast Closure. Trio Oil I Pressure - Low l 1
Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these l
, valves. The Turbine Control Valve Fast Closure, Trip 011 l Pressure - Low Function is the primary scram signal for the i generator load rejection event analyzed in Reference 2. For '
this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the E0C-RPT System, ensures that the MCPR SL is not exceeded. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low signals are initiated by the electrohydraulic control (EHC) i fluid pressure at each control valve. One pressure j transmitter is associated with each control valve, and the
, signal from each transmitter is assigned to a separate RPS i
logic channel. This Function must be enabled at THERMAL POWER 2 30% RTP. This is normally accomplished l automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. (continued) HATCH UNIT 2 B 3.3-17 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES l 1 APPLICABLE 9. Turbine Control Valve Fast Closure. Trio 011 SAFETY ANALYSES, Pressure - Low (continued) LCO, and APPLICABILITY The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure. Four channels of Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be
- OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This function is required, consistent with the analysis assumptions, whenever THERMAL POWER is 2: 30% RTP. This l Function is not required when THERMAL POWER is < 30% RTP, !
since the Reactor Vessel Steam Dome Pressure - High and the l Average Power Range Monitor Neutron Flux - High Functions l are adequate to maintain the necessary safety margins.
- 10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, to each of the four RPS logic channels, which are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
The reactor mode switch is a single switch with four channels, each of which provides input into one of the RPS logic channels. There is no Allowable Value for this Function, since the channels are mechanically actuated based solely on reactor mode switch position. Four channels of Reactor Mode Switch - Shutdown Position Function, with two channels in each trip system, are available and required to be OPERABLE. The Reactor Mode (continued) HATCH UNIT 2 B 3.3-18 PROPOSED REVISION 7/16/96
i l RPS Instrumentation
- B 3.3.1.1 4
! BASES APPLICABLE 10. Reactor Mode Switch - Shutdown Position (continued) a SAFETY ANALYSES, LCO, and Switch - Shutdown Position Function is required to be APPLICABILITY OPERABLE in MODES I and 2, and MODE 5 with any control rod a withdrawn from a core cell containing one or more fuel l- assemblies, since these are the MODES and other specified ! conditions when control rods are withdrawn. j 11. Manual Scram i The Manual Scram push button channels provide signals, via the manual scram logic channels, to each of the'four RPS logic channels, which are redundant to the automatic protective instrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis. There is one Manual Scram push button channel for each of the four RPS -logic channels. In order to cause a scram it is necessary that at least one channel in each trip system be actuated. There is no Allowable Value for this Function since the channels are mechanically actuated based solely on the position of the push buttons. Four channels of Manual Scram with two channels in each trip system arranged in a one-out-of-two logic are available and
-required-to be OPERABLE in MODES I and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn.
ACTIONS A Note has been provided to modify the ACTIONS related to RPS instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required l l (continued) I HATCH UNIT 2 B 3.3-19 PROPOSED REVISION 7/16/96 l
,m .m_ _ .._ _ _ _ _ _ . . . _ _ _ _ . _ . . . _ . _ _ . _ _ _ . _ _ _ _ _ _ _ _
i RPS Instrumentation i B 3.3.1.1 j BASES i ACTIONS: Actions of the Condition continue to apply for each , (continued) additional failure, with Completion Times based on initial-
- entry into the Condition. However, the Required Actions for J inoperable RPS instrumentation channels provide appropriate com>ensatory measures for separate inoperable channels. As
- suci, a Note has been provided that allows separate i Condition entry for each inoperable RPS instrumentation j channel.
l A.1 and A.2 l j Because of the diversity of sensors available to provide i trip signals and the redundancy of the RPS design, an ) allowable out of service time of 12 hours has been shown.to be acceptable (Refs. 9 and 13) to permit restoration of. any l inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to 1 Required Actions B.1, B.2, and C.1 Bases). If the I ino>erable channel cannot be restored to OPERABLE status witiin the allowable out of service time, the channel. or the associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system)-in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken. As noted, Required Action A.2 is not applicable for APRM Functions 2.a, 2.b, 2.c., and 2.d. Inoperability of one ! required APRM channel affects both trip systems; thus, Required Action A.1 must be satisfied. This is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. l i (continued) l l HATCH UNIT 2 B 3.3-20 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accommodate a single failure in either trip system. Required Actions B.1 and B.2 limit the time the RPS scram logic, for any Function, would not accommodate single I failure in both trip systems (e.g., one-out-of-one and l one-out-of-one arrangement for a typical four channel ! Function). The reduced reliability of this logic l arrangement was not evaluated in References 9 and 13 for the l 12 hour Completion Time. Within the 6 hour allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system. Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in References 9 and 13, which justified a 12 hour allowable out l of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in). If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip. The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverso Functions, and the low probability of an event requiring the initiation of a scram. (continued) HATCH UNIT 2 B 3.3-21 PROPOSED REVISION 7/16/96 i
RPS Instrumentation l B 3.3.1.1 , l BASES l ACTIONS B.1 and B.2 (continued) Alternately, if it is not desired to place the inoperable ! channels (or one trip system) in trip (e.g., as in the case j i where placing the inoperable channel or associated trip, ' system in trip would result in a scram or RPT), Condition D must be entered and its Required Action taken. l l As noted, Condition B is not applicable for APRM Functions -l ! 2.a. 2.b, 2.c, and 2.d. Inoperability of an APRM channel affects both trip systems and is not associated with a l specific trip system, as are the APRM two-out-of-four voter and other non-APRM channels for which Condition B applies, i For an inoperable APRM channel, Required Action A.1 must be i satisfied, and'is the only action (other than restoring l OPERABILITY) that will restore capability to accommodate a 4 l single failure. Inoperability of more than one required l APRM channel results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each ! channel. Because Conditions A and C provide Required l Actions that are appropriate for the inoperability of APRM l Functions 2.a 2.b, 2.c, and 2.d, and these Functions are l not associated with specific trip systems as are the APRM two-out-of-four voter and other non-APRM channels, l Condition B does not apply. fu.1 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. i The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. (continued) I IIATCH UNIT 2 B 3.3-22 PROPOSED REVISION 7/16/96 1 -
I RPS Instrumentation
- B 3.3.1.1 1
, BASES , j ACTIONS M (continued) 1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other-specified condition dependent and may change as the Required Action of a previous Condition is completed. Each _ time an inoperable channel has not met any Required Action j of Condition A, B, or C and the associated Completion Time 4 has expired, Condition D will be entered for that channel
- and provides for transfer to the appropriate subsequent i
Condition. E.1. F.1. and G.1 If the channel (s) is not restored to OPERABLE status or + placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, ! the Completion Time of Required Action E.1 is consistent l with the Completion Time provided in LC0 3.2.2, " MINIMUM ) CRITICAL POWER RATIO (MCPR)." M If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect . the reactivity of the core and are, therefore, not required l to be inserted. Action must continue until all insertable ! control rods in core cells containing one or more fuel i assemblies are fully inserted. (continued) ] HATCH UNIT 2 B 3.3-23 PROPOSED REVISION 7/16/96 l
RPS Instrumentation j B 3.3.1.1 BASES (continued) SURVEILLANCE As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.1.1-1. The Surve111ances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains RPS trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RPS will trip when necessary. i i SR 3.3.1.1.1 l Performance of the CHANNEL. CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter snould read approximately the same value. Significant deviations i between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the instrument has drifted outside its limit. (continued) HATCH UNIT 2 B 3.3-24 PROPOSED REVISION 7/16/96 l
l RPS Instrumentation l B 3.3.1.1 j i l BASES SURVEILLANCE SR 3.3.1.1.1 (continued) REQUIREMENTS The' Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. SR 3.3.1.1.2 1 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8. A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at 2 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat i balance when < 25% RTP. At low power levels, a high degree - of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with l SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met ; per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. SR 3.3.1.1.3 (Not used.) SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. (continued) HATCH UNIT 2 B 3.3-25 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 i BASES SURVEILLANCE SR 3.3.1.1.4 (continued) REQUIREMENTS As noted, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without l utilizing jumpers, lifted leads, or movable links. This , allows entry into MODE 2 if the 7 day Frequency is not met '
- per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from NODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR.
- A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9).
4 SR 3.3.1.1.5 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A Frequency of 7 days provides an , acceptable level of system average availability over the
- Frequency and is based on the reliability analysis of Reference 9. (The Manual Scram Function's CHANNEL
- FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.)
I SR 3.3.1.1.6 and SR 3.3.1.1.7 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status. l The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be
, increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from ;
the fully inserted position since indication is being transitioned from the SRMs to the IRMs. (continued) i HATCH UNIT 2 8 3.3-26 PROPOSED REVISION 7/16/96 1 I J
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.6 and SR 3.3.1.1.7 (continued) REQUIREMENTS The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (by initiating a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APRMs l concurrently have onscale readings such that the transition j " between MODE 1 and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on range 1 before SRMs have reached the upscale rod block. As noted, SR 3.3.1.1.7 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2). If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate channel (s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition should be declared inoperable. A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs. i SR 3.3.1.1.8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 effective full power hours Frequency is based on ' ensuring the nodal power uncertainty is within the licensing basis analysis. l (continued) ! HATCH UNIT 2 B 3.3-27 PROPOSED REVISION 7/16/96 l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of Reference 9. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b l only), the two-out-of-four voter channels, and the interface connections to the RPS trip systems from the voter channels. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 184 day Frequency of SR 3.3.1.1.10 is based on the reliability analysis of Reference 13. (NOTE: The actual voting logic of the two-out-of-four voter channels is tested as part of SR 3.3.1.1.15.) For Function 2.a, a Note that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1 is provided. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. (continued) HATCH UNIT 2 B 3.3-28 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 REQUIREMENTS ; (continued) This SR ensures that scrams initiated from the Turbine Stop : Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is a 30% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER 2 30% RTP to ensure that the calibration is val id , ! If any bypass channel's setpoint is nonconservative , (i.e,', the Functions are bypassed at 2 30% RTP, either due i to open main turbine bypass valve (s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low functions are enabled), this SR is met and the channel is considered OPERABLE. The Frequency of 18 months is based on engineering judgment I and reliability of the components. j l SR 3.3.1.1.13 l A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, this SR also includes a physical inspection and actuation of the switches. For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated recirculation loop flow channel. (continued) HATCH UNIT 2 B 3.3-29 PROPOSED REVISION 7/16/96
=- . - _ . . .-_ _
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.13 (continued) REQUIREMENTS Note I states that neutron detectors are excluded from
, CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric .
calibration (SR 3.3.1.1.2) and the 1000 effective full power ! hours LPRM calibration against the TIPS (SR 3.3.1.1.8). A 1 second Note is provided that requires the IRM SRs to be l performed within 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 IRM Functions cannot be performed in l MODE 1 without utilizing jumpers, lifted leads or movable ! , links. This Note allows entry into MODE 2 from MODE 1 if ! . the associated Frequency is not met per SR 3.0.2.
; Twelve hours is based on operating experience and in i consideration of providing a reasonable time in which to i
complete the SR. ; I The Frequency of SR 3.3.1.1.13 is based upon the assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.1.1.14 (Not used.) 1 SR 3.3.1.1.15 The LOGIC SYSTEM FUN'TIONAL C TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LC0 3.1.3), and SDV vent and drain valves (LC0 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the i Surveillance were performed with the reactor at power. l (continued) l HATCH UNIT 2 B 3.3-30 PROPOSED REVISION 7/16/96 l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued) REQUIREMENTS Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e
~
simulates APRM trip conditions at the two-out-of-four voter channel inputs to check all combinations of two tripped inputs to the two-out-of-four logic in the voter channels and APRM related redundant RPS relays. SR 3.3.1.1.16 This SR ensures that the individual channel response times , are less than or equal to the maximum values assumed in the' l accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 10. RPS RESPONSE TIME for APRM two-out-of-four Voter Function 2.e includes the output relays of the voter and the associated RPS relays and contactors. (The digital portions of the APRM and two-out-of-four voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration check the time base of the digital electronics.) Confirmation of the time base is adequate to assure required response times are met. Neutron detectors are excluded from RPS RESPONSE-TIME testing because the principles of detector operation virtually ensure an instantaneous response time. Note 1 allows neutron detectors to be excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. Note 2 allows channel sensors for Reactor Vessel Steam Dome Pressure - High and Reactor Vessel Water Level - Low, Level 3 (Functions 3 and 4)_to be excluded from RPS RESPONSE TIME testing. This allowance is supported by Reference 12 which concludes that any significant degradation of the channel sensor response time can be detected during the performance of other Technical Specifications SRs. RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS Frequency to be determined based on four channels per trip system, in lieu of the eight channels specified in (continued) HATCH UNIT 2 B 3.3-31 PROPOSED REVISION 7/16/96 l
RPS Instrumentation B 3.3.1.1
- BASES SURVEILLANCE SR 3.3.1.1.16 (continued)
- REQUIREMENTS l Table 3.3.1.1-1 for the Main Steam Line Isolation i Valve - Closure Function. This Frequency is based on the
- logic interrelationships of the various channels required to i produce an RPS scram s!gnal. This Frequency is consistent
! with the typical industry refueling cycle and is based upon ! plant operating experience, which shows that random failures i of instrumentation com)onents causing serious response time degradation, but not c1annel failure, are infrequent i occurrences. REFERENCES 1. FSAR, Section 7.2. I 2. FSAR, Chapter 15.
- 3. FSAR, Section 6.3.3.
- 4. FSAR, Supplement 5A.
1 5. FSAR, Section 15.1.12. I 6. NED0-23842, " Continuous Control Rod Withdrawal in the i Startup Range," April 18, 1978. 4
- 7. FSAR, Section 15.1.38.
! 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram j Discharge System Safety Evaluation," December 1,1980.
- 9. NED0-30851-P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System,"
March 1988. ;
- 10. Technical Requirements Manual.
- 11. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 12. NED0-32291, " System Analyses for Elimination of Selected Response Time Testing Requirements,"
January 1994.
- 13. NEDC-32410P-A, " Nuclear Measurement Analysis and i Control Power Range Neutron Monitor (NUMAC PRNM) ,
Retrofit Plus Option III Stability Trip Function," i October 1995. HATCH UNIT 2 B 3.3-32 PROPOSED REVISION 7/16/96
SRM Instrumentation B 3.3.1.2 i BASES i 1 APPLICABLE System (RPS) Instrumentation"; IRM Neutron Flux - High and j- SAFETY ANALYSES Average Power Range Monitor (APRM) Neutron Flux - High (continued) (Setdown) Functions; and LCO 3.3.2.1, " Control Rod Block Instrumentation." . The SRMs have no safety function and are not assumed to function during any FSAR design basis accident or transient 4 analysis. However, the SRMs provide the only on scale monitoring of neutron flux levels during startup and refueling. Therefore, they are being retained in Technical i Specifications. During startup in MODE 2, three of the four SRM channels are LCO required to be OPERABLE to monitor the reactor flux level ! prior to and during control rod withdrawal, subcritical multiplication and reactor criticality, and neutron flux level and reactor period until the flux level is sufficient i to maintain the IRMs on Range 3 or above. All but one of the channels are required in order to provide a i representation of the overall core response during those i periods when reactivity changes are occurring throughnt the Core. l l i In MODES 3 and 4, with the reactor shut down, two SRM channels provide redundant monitoring of flux levels in the core. In MODE 5, during a spiral offload or reload, an SRM outside l the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the 1 fueled region of the core. Thus, CORE ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an. adjacent 4 quadrant provided the Table 3.3.1.2-1, footnote (b), ! requirement that the bundles being spiral reloaded or spiral offloaded are all in a single fueled region containing at least one OPERABLE SRM is met. Spiral reloading _ and 4 offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence). In nonspiral routine operations, two SRMs are required to be OPERABLE to provide redundant monitoring of reactivity a (continued) HATCH UNIT 2 B 3.3-34 PROPOSED REVISION 7/16/96 l
-- , + w7
l l l Control Rod Block Instrumentation ' B 3.3.2.1 B 3.3 INSTRUMENTATION B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Cor. trol rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and . relays that are designed to ensure that specified fuel l design limits are not exceeded for postulated transients and accider,ts. During high power operation, the rod block l monitor (RBM) provides protection for control rod withdrawal ' error events. During low power operations, control rod , blocks from the rod worth minimizer (RWM) enforce specific ! control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch - Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities. The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One ' RBM channel inputs into one RMCS rod block circuit and the other RBM channel inputs into the second RMCS rod block l circuit. l The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core i heights surrounding the control rod being withdrawn. A signal from one of the four redundant average power range l monitor (APRM) channels supplies a reference signal for one l of the RBM channels, and a signal from another of the APRM channels supplies the reference signal to the second RBM , channel. This reference signal is used to determine which ! RBM range setpoint (low, intermediate, or high) is enabled. If the APRM is indicating less than the low power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 1). A rod block signal is also generated if (continued) HATCH UNIT 2 B 3.3-42 PROPOSED REVISION 7/16/96
l I Centrol Rod Bleck Instrumentation B 3.3.2.1 , 1 BASES BACKGROUND an RBM Downscale trip or an Inoperable trip occurs. The (continued) Downscale trip will occur if the RBM channel signal decreases below the Downscale trip setpoint after the RBM signal has been normalized. The Inoperable trip will occur during the nulling (normalization) sequence, if: the RBM channel fails to null, too few LPRM inputs are available, a module is not plugged in, or the function switch is moved to i any position other than " Operate." The purpose of the RWM is to control rod patterns during i startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. The sequences effectively limit the potential amount' and rate of reactivity increase during a CRDA. Prescribed control rod sequences are stored in the RWM, which will initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored : sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses . feedwater flow and steam flow signals to determine when the ! reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block , circuits. With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents inadvertent criticality as the result of a ' control rod withdrawal during MODE 3 or 4, or during MODE 5 when the reactor mode switch is required to be in the i shutdown position. The reactor mode switch has two ' channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods. ' i l I (continued) HATCH UNIT 2 B 3.3-43 PROPOSED REVISION 7/16/96 . _ _ _ _ _ _ - . _ _ _ _ _ + _ . - , . _.,.._m_i _.7.- i- -._ - 9 - , . m 7 y
k i Control Rod Block Instrumentation i l B 3.3.2.1 BASES 1 SURVEILLANCE SB 3.3.2.1.1 REQUIREMENTS : ) (continued) A CHANNEL FUNCTIONAL TEST is performed for each RBM channel
- to ensure that the entire channel will perform the intended j function. It includes. the Reactor Manual Control System j input.
{ Any setpoint adjustment shall be consistent with the l assumptions of.the current plant specific setpoint - ! methodology. The Frequency of 184 days is based on reliability analyses (Ref.11). l { SR 3.3.2.1.2 and SR 3.3.2.1.3 l A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. i The CHANNEL FUNCTIONAL TEST for the RWM is performed by
- attempting to withdraw a control rod not in compliance with
! the prescribed sequence and verifying a control rod block
- occurs. This test is performed as soon as possible after i the applicable conditions are entered. As noted in the SRs,
! SR 3.3.2.1.2 is not required to be performed L.ntil I hour l after any control rod is withdrawn at < 10% RTP in MODE 2, and SR 3.3.2.1.3 is not required to be performed until I hour after THERMAL POWER is < 10% RTP in MODE 1. This allows entry into MODE 2 (and if entered during a shutdown, concurrent power reduction to < 10% RTP) for SR 3.3.2.1.2 and THERMAL POWER reduction to < 10% RTP in MODE 1 for SR 3.3.2.1.3 to perform the required Surveillances if the 92 day Frequency is not met per SR 3.0.2. The 1 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the SRs. The 92 day Frequencies are based on reliability analysis (Ref. 8). SR 3.3.2.1.4 ; The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to l each RBM channel. Below the minimum power setpoint, the RBM is automatically bypassed. These power Allowable Values i (continued) HATCH UNIT 2 B 3.3-50 PROPOSED REVISION 7/16/96
l
- Control Rod Block Instrumentation B 3.3.2.1 i
BASES SURVEILLANCE SR 3.3.2.1.8 (continued) i REQUIREMENTS i OPERABLE following loading of sequence into RWM, since this i is when rod sequence input errors are possible. i , i l REFERENCES 1. FSAR, Section 7.6.2.2.5.
- 2. FSAR, Section 7.6.8.2.6.
! 3. NEDC-30474-P, " Average Power Range Monitor, Rod Block j Monitor, and Technical Specification Improvements (ARTS) Program for Edwin I. Hatch Nuclear Plants," December 1983.
- 4. NEDE-24011-P-A-US, " General E1cetrical Standard Application for Reload Fuel," supplement for United States, (revision specified in the COLR). l
- 5. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),
" Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," BWROG-8644, August 15, 1986.
- 6. NED0-21231, " Banked Position Withdrawal Sequence," I January 1977. l l
- 7. NRC SER, " Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," " General Electric Standard Application for Reactor Fuel, Revision 8, )
Amendment 17," December 27, 1987. l
- 8. NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"
October 1988.
- 9. GENE-770-06-1, " Bases for Changes To Surveillance Test Intervals And Allowed Out-0f-Service Tines For Selected Instrumentation Technical Specifications," !
February 1991.
- 10. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 11. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function," October 1995. HATCH UNIT 2 B 3.3-53 PROPOSED REVISION 7/16/96
l , E0C-RPT Instrumentation B 3.3.4.1 l BASES i
- l APPLICABLE Turbine Stoo Valve - Closure SAFETY ANALYSES, i
LCO, and Closure of the TSVs and a main turbine trip result in the 1 APPLICABILITY loss of a heat sink and increases reactor pressure, neutron ! (continued) flux, and heat flux that must be limited. Therefore, an RPT is initiated on a TSV - Closure signal before the TSVs are completely closed in anticipation of the effects that would I result from closure of these valves. EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient. Closure of the TSVs is determined by measuring the position of each valve. While there are two separate position switches associated with each stop valve, only the signal from one switch for each TSV is used, with each of the four channels being assigned to a separate trip channel. The logic for the TSV - Closure Function is such that two or
, more TSVs must be closed to produce an E0C-RPT. This Function must be enabled at THERMAL POWER 2 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure;
- therefore, opening of the turbine bypass valves may affect this Function. Four channels of TSV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an E0C-RPT from this Function on a valid signal.
The TSV -- Closure Allowable Value is selected to detect imminent TSV closure. This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWER is 2 30% RTP. Below 30% RTP, the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor (APRM) Neutron Flux - l High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR Safety Limit. l Turbine Control Valve Fast Closure. Trio Oil Pressure - Low Fast closure of the TCVs during a generator load rejection results in the loss of a heat sink that produces reactor ! pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TCV Fast Closure, Trip 011 Pressure - Low in anticipation of the transients that would result from the closure of these (continued) 4 HATCH UNIT 2 8 3.3-82 PROPOSED REVISION 7/16/96
EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Control' Valve Fast Closure. Trin Oil Pressure -- Low SAFETY ANALYSES, (continued) LCO, and APPLICABILITY valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient. Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control valve. There is one pressure transmitter associated with each control valve, and the signal from each transmitter is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip 011 Pressure - Low Function is such that two or more TCVs must be closed (pressure transmitter trips) to produce an E0C-RPT. This Function must be enabled at THERMAL POWER 2: 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TCV Fast Closure, Trip 011 Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an E0C o.PT from this Function on a valid signal. The TCV Fast Closure, Trip Oil Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure. This protection is required consistent with the safety analysis whenever THERMAL POWER is 130% RTP. Below 30% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Neutron Flux - High Functions of the RPS are l adequate to maintain the necessary margin to the MCPR Safety Limit. ACTIONS A Note has been provided to modify the ACTIONS related to E0C-RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable E0C-RPT instrumentation channels provide (continued) HATCH UNIT 2 B 3.3-83 PROPOSED REVISION 7/16/96
. Recirculation Loops Operating B 3.4.1
- BASES APPLICABLE case (since the intact loop starts at a lower flow rate and SAFETY ANALYSES the core response is the same as if both loops were (continued) operating at a lower flow rate), a small mismatch has been determined to be . acceptable based on engineering judgement.
The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the FSAR. , A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core . cooling, provided the APLHGR requirements are modified ! accordingly (Ref. 3). The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation j (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR i requirements are modified. During single recirculation loop i operation, modification to the Reactor Protection System (RPS) average power range raonitor (APRM) instrument setpoints is also required to account for the different
. relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop
, operation are specified in the COLR. The APRM Simulated l Thermal Power - High setpoint is in LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation." a Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 5). LC0 Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LC0 3.2.1, 4 " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"), (continued) HATCH UNIT 2 B 3.4-3 PROPOSED REVISION 7/16/96
Recirculation Loops Operating B 3.4.1 BASES LC0 and APRM Simulated Thermal Power - High setpoint l l (continued) (LC0 3.3.1.1) must be applied to allow continued operation consistent with the assumptions of Reference 3. In addition, core flow as a function of core thermal power must be in the " Operation Allowed Region" of Figure 3.4.1-1 to ensure core thermal-hydraulic oscillations do not occur. I i APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor i Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. l In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. ACTIONS A.1 and B.1 l l Due to thermal-hydraulic stability concerns, operation of the plant with one recirculation loop is controlled by restricting the core flow to 2: 45% of rated ccre flow when THERMAL POWER is greater than the 76% rod line. This requirement is based on the recommendations contained in GE SIL-380, Revision 1 (Reference 4), which defines the region where the limit cycle oscillations are more likely to occur. If the core flow as a function of core thermal power is in the " Operation Not Allowed Region" of Figure 3.4.1-1, prompt action should be initiated to restore the flow-power combination to within the Operation Allowed Region. The 2 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing core oscillations to be quickly detected. An immediate reactor scram is also required with no recirculation pumps in operation, since all forced circulation has been lost and the probability of thermal-hydraulic oscillations is greater. (continued) HATCH UNIT 2 8 3.4-4 PROPOSED REVISION 7/16/96
I SDM Test - Refueling B 3.10.8 BASES I APPLICABLE CRDA analyses assume that the reactor operator follows , SAFETY ANALYSES prescribed withdrawal sequences. For SDM tests performed. ! (continued) within these defined sequences, the analyses of References 1 i and 2 are applicable. However, for some sequences developed ! for the SDM testing, the control rod patterns assumed in the ! safety analyses of References 1 and 2 may not be met. Therefore, special CRDA analyses, performed in accordance
- with an NRC approved methodology, may be required to 1 1 demonstrate the SDM test sequence will not result in 1
- unacceptable consequences should a CRDA occur during the i testing. For the purpose of this test, the protection
, provided by the normally required MODE 5' applicable LCOs, in - addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents 4 i within the bounds of the appropriate safety analyses l (Refs. I and 2). In addition to the added requirements for l the RWM, Average Power Range Monitor, and control rod ! coupling, the notch out mode is specified for out of sequence withdrawals. Requiring the notch out mode limits withdrawal steps to a single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may occur during the test. i l As described in LC0 3.0.7, compliance with Special ; Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A ; discussion of the criteria satisfied for the other LCOs is ; provided in their respective Bases. ' i LCO As described in LC0 3.0.7, compliance with this Special 0)erations LC0 is optional. SDN tests may be performed w111e in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS. For SDM ; tests performed while in MODE 5, additional requirements ' must be met to ensure that adequate protection against potential reactivity excursions is available. To provide additional scram protection beyond the normally required IRMs, the Average Power Range Monitors are also required to be OPERABLE (LC0 3.3.1.1, Functions 2.a. 2.d, and 2.e) as l though the reactor were in MODE 2. Because multiple control rods will be withdrawn and the reactor will potentially become critical, the approved control rod withdrawal sequence must be enforced by the RWM (LC0 3.3.2.1, (continued) HATCH UNIT 2 B 3.10-34 PROPOSED REVISION 7/16/96
1 SDM Test - Refueling l B 3.10.8 ! BASES LC0 Function 2, MODE 2), or must be verified by a second (continued) licensed operator or other qualified member of the technical i staff. To provide additional protection against an ! inadvertent criticality, control rod withdrawals that do not I conform to the banked position withdrawal sequence specified in LC0 3.1.6, " Rod Pattern Control," (i.e., out of sequence i control rod withdrawals) must be made in the individual notched withdrawal mode to minimize the potential reactivity insertion associated with each movement. Coupling integrity l of withdrawn control rods is required to minimize the probability of a CRDA and ensure proper functioning of the withdrawn control rods, if they are required to scram. Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in progress. Furthermore, since the control rod scram function with the RCS at atmospheric pressure relies solely on the CRD accumulator, it is essential that the CRD charging water header remain pressurized. This Special Operations LC0 then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startup/ hot standby position, such that the SDM tests may be performed while in MODE 5. APPLICABILITY These SDM test Special Operations requirements are only applicable if the SDM tests are to be performed while in MODE 5 with the reactor vessel head removed or the head bolts not fully tensioned. Additional requirements during these tests to enforce control rod withdrawal sequences and restrict other CORE ALTERATIONS provide protection against potential reactivity excursions. Operations in all other MODES are unaffected by this LCO. ACTIONS Ad With one or more control rods discovered uncoupled during this Special Operation, a controlled insertion of each uncoupled control rod is required; either to attempt recoupling, or to preclude a control rod drop. This controlled insertion is preferred since, if the control rod fails to follow the drive as it is withdrawn (i.e., is
" stuck" in an inserted position), placing the reactor mode switch in the shutdown position per Required Action 8.1 (continued)
HATCH UNIT 2 B 3.10-35 PROPOSED REVISION 7/16/96 l
l SDM Test - Refueling j B 3.10.8 ; BASES (continued) l SURVEILLANCE SR 3.10.8.1. SR 3.10.8.2. and SR 3.10.8.3 REQUIREMENTS LC0 3.3.1.1, Functions 2.a. 2.d, and 2.e, made applicable in l l this Special Operations LCO, are required to have their l Surveillances met to establish that this Special Operations I LCO is being met. However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LC0 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff (e.g., a qualified shift technical advisor or reactor engineer). As noted, either the applicable SRs for the RWM (LC0 3.3.2.1) must be satisfied according to the applicable i Frequencies (SR 3.10.8.2), or the proper movement of control ' rods must be verified (SR 3.10.8.3). This latter ; verification (i.e., SR 3.10.8.3) must be performed during l control rod movement to prevent deviations from the ; specified sequence. These Surveillances provide adequate l assurance that the specified test sequence is being ) followed. l SR 3.10.8.4 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The 12 hour Frequency is intended to provide appropriate assurance that each operating shift is aware of and verifies compliance with these Special Operations LC0 requirements. SR 3.10.8.5 Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control l rod is withdrawn to the full-out notch position, or prior to declaring the control rod OPERA 3LE after work on the control rod or CRD System that could affect coupling. This i Frequency is acceptable, considering the low probability ' that a control rod will become uncoupled when it is not being moved, as well as operatir.g experience related to uncoupling events. (continued) HATCH UNIT 2 B 3.10-37 PROPOSED REVISION 7/16/96
~- -. - - - - - - . - - - . - - . - - - . . - - - _ _ . -
1 i RPS Instrumentation B 3.3.1.1
- BASES i
i APPLICABLE Averaae Power Ranae Monitor (APRM) SAFETY ANALYSES,
- LCO, and The APRM channels provide the primary indication of neutron i APPLICABILITY flux within the core and respond almost instantaneously to j (continued) neutron flux increases. The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power i
- distribution and local power changes. The APRM channels 4
average these LP 'n to > rov de a continuous I indication of a age eac r r a few percent to greater than R P. 9y The APRM System divided into 4 AP channels and 4 two-out-of-four voter . RM channel provides inputs to each of the four voter channels. The four voter
- channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The
} APRM System is designed to allow one APRM channel, but no i v unbypa e channels, to be bypassed. A trip from any one
,PRM will . result in a " half-trip"- in all four
) rN lN SE RT
- E '
4 voter channels, but no trip inputs to either RPS trip ! system.W A- ip from any two unbypassed APRM channels will r g' v+i t full-trip in each of the four voter channels, t turn results in two trip inputs into each RPS trip
- system e of the four APRM channels- and all four_ of the lN S ERTtp.
- votergan l t is are required to be OPERABLE to ensure that no i sin 6.ure will preclude a scram on a valid signal. In ition, to provide adequate coverage of the entire core, i consistent with the design bases for APRM Functions 2.a,
- 2.b, and 2.c, at least 17 LPRM inputs, with at least t i LPRM inputs from each of the four axial levels the j LPRMs are located, must be OPERABLE for eac RM channel. g l IN SERT *G ">
1
- 2.a. Averaae Power Ranae Monitor Neutron Flux - la ,
j (Setdown)
- x
- For operation at low power (i.e., MODE 2), the Average Power Range Monitor Neutron Flux - High (Setdown) Function is l capable of generating a trip signal that prevents fuel j damage resulting from abnormal operating transients in this ;
- power range. For most operation at low power levels, the i
- Average Power Range Monitor Neutron Flux - High (Setdown)
Function will provide a secondary scram to the Intermediate-K
; Range Monitor Neutron Flux - High Function because of the 1 i !
3 (continued) ; l HATCH UNIT 1 8 3.3-7 PROPOSED REVISION 7/16/96 s 5 _ -_ - - - _ . .-. - _ _ _ _ - a
; Insert 'D' - Bases B 3.3.1.1 Average Power Range Monitor (APRhD ;- Each APRM also includes an Oscillation Power Range Monitor (OPRM) Upscale Function which monitors small groups ofLPRM signals to detect thermal-hydraulic instabilities.
Insert 'E' - Bases B 3.3.1.1 Average Power Range Monitor (APRM) APRM trip Functions 2.a,2.b,2.c, and 2.d are voted independently from OPRM Upscale 2 Function 2.f. Therefore, any Function 2.a,2.b,2.c, or 2.d . Insert 'F' - Bases B 3.3.1.1 Average Power Range Monitor (APRM) Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full-trip from each of the four voter channels. Insert 'O' - Bases B 3.3.1.1 Average Power Range Monitor (APRM) l For OPRM Upscale Function 2.f, LPRMs are assigned to " cells" of three detectors. A ! minimum of three cells, each with a minimum of two LPRMs, must be OPERABLE for l OPRM Upscale Function 2.f to be OPERABLE. i a i 3 k 1 4 ) 4
,m.
l l 4 RPS Instrumentatien B 3.3.1.1 ' BASES t APPLICABLE 2.b. Averaae Power Ranae Monitor Simulated Thermal i SAFETY ANALYSES,
%[
Power - Hiah (continued) LCO, and APPLICABILITY control rod pattern) but is clamped at an uppe m at s is always lower than the Average Power Ran Flux - High Function Allowable Value. g onitor Neutr gg [ l The Average Power Range Monitor Simu ated T a ! High Function provides protection ag inst transients where THERMAL POWER increases slowly (suc as the loss of feedwater heatina event) and prote J the fuel cladding integrity by ensuring that theQ4CP QiC is not exceeded. During these events, the THERMAL P WER increase does not MOlLN significantly lag the neutron flux response and, because of I C(2.1TTLAL a lower trip setpoint, will initiate a scram before the high neutron flux scram. For rapid neutron flux increase events, On the THERMAL POWER lags the neutron flux and the Average l brK) Power Range Monitor Neutron Flux - High Function will provide a scram signal before the Average Power Range Y ! Monitor Simulated Thermal Power - High Function setpoint and g ' associated time delay are exceeded. Each APRM channel uses one total drive flow signal representative of total core flow. The total drive flow signal is generated by the flow processing logic, which is part of the APRM channel. The flow is calculated by summing two flow transmitter signals, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel 0PERABILITY N ( ,
- requirements for this Function. /
The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power - High Function for the mitigation of the loss of 4 feedwater heating event. The time constant is based on the fuel heat transfer dynamics and provides a signal , proportional to the THERMAL POWER. i The Average Power Range Monitor Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when
- there is the possibility of generating excessive THERMAL 4 POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.
(continued) HATCH UNIT 1 B 3.3-9 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 j BASES APPLICABLE 2.c. Averaae Power Ranae Monitor Neutron Flux - Hiah SAFETY ANALYSES, LCO, and The Average Power Range Monitor Neutron Flux - High Function APPLICABILITY is capable of generating a trip signal to prevent fuel (continued) damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux - High Function is % i assumed to terminate the main steam isolation valve (MSIV)
- closure event and, along with the safety / relief valves 4
(S/RVs), limits the peak reactor pressure vessel (RPV) i pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 7) takes credit for the Average Power Range Monitor Neutron Flux - High Function to terminate the CRDA. The Allowable Value is based on the Analytical Limit assumed K in the CRDA analyses. The Average Power Range Monitor Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential A consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although i the Average Power Range Monitor Neutron Flux - High Function is assumed in the CRDA analysis, which is applicable in l MODE 2, the Average Power Range Monitor Neutron Flux - High (Setdown) Function conservatively bounds the assumed trip g' and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor g~ Neutron Flux - High Function is not required in MODE 2. 2.d. Averaae Power Ranae Monitor - Inoo 4 ' I his Function (Inop) provides assurance that the minimum s > number of APRM channels is OPERABLE. For any APRM channel, any time: 1) its mode switch is in any position other than " Operate," 2) an APRM module is unplugged, or 3) the automatic self-test system detects a critical fault with the APRM channel, an Inop trip signal is sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from j all four voter channels to their associated trip system.
}E CE k NSS ChCLMd5 Gnd GM kut th i
( was 8'b ae yea + 6e ometre eache (continued) HATCH UNIT 1 B 3.3-10 PROPOSED REVISION 7/16/96 i
RPS Instrumentation B 3.3.1.1 BASES j APPLICABLE 2.d. Averaae Power Ranae Monitor-Inoo (continued) SAFETY ANALYSES,
, LCO, and This Function was not specifically credited in the accident APPLICABILITY analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
There is no Allowable Value for this Function. This Function is required to be OPERABLE in the MODES where the APRM Functions are required. 2.e. Two-out-of-Four oter > in[ % oPRAA The Two-out-of-Four Vo r unc Dn~pran des the interface between the APRM Funct ons an the final RPS trip system logic. As such, it is r red to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those / Functions. Therefore, the Two-out-of-Four Voter Function is required to be OPERABLE in MODES 1 and 2. y/ All four voter channels are required to be OPERABLE. Each voter channel also incudes self-diagnostic functions. If any voter channel detects a critical fault in its own N processing, an Inop trip is issued from that voter channel t gu to the associated trip system.
- There is no Allowable Value for this Function.
(.-3%EVT 3. Reactor Vessel Steam Dome Pressure - Hiah An ir. crease in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to . increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor , Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, . counteracting the pressure increase by rapidly reducing core power. For the overpressurization' protection analysis of Reference 4, reactor scram (the analyses conservatively 4 (continued) HATCH UNIT 1 B 3.3-11 PROPOSED REVISION 7/16/96
1 i j Insert 'H' - Bases B 3.3.1.1 - 2.e.' Two-out-of-Four Voter . , The Two-out-of-Four Voter Function votes APRM Functions 2.a,2.b,2.c, and 2.d
- . independently of Function 2.f. The voter also includes separate outputs to the RPS for th
- !
i two independently voted sets of Functions, each of which is redundant (four total inputs), j l Voter Function 2.e must be declared inoperable if any ofits functionality is inoperable. ! j However, due to the independent voting of APRM trips and the redundancy of outputs, - j there may be conditions where Voter Function 2.e is inoperable, but trip capability for one j or more of the other APRM Functions through that voter is still maintained. This may be
- considered when determining the condition of other APRM Functions resulting from partial inoperability of Voter Function 2.e. ;
2 n t : ! Insert 'I' - Bases B 3.3.1.1 2.f . Oscillation Power Ranne Monitor (OPRM) Unscale ! j 2.f Oscillation Power Range Monitor (OPRM) Unscale
]
1 i j The OPRM Upscale Function provides comphance with GDC 10 and GDC 12, thereby ; providing protection from exceeding the fuel MCPR SL due to anticipated thermal-j hydraulic power oscillations. a i References 13,14, and 15 describe three algorithms for detecting thermal-hydraulic ! i instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in i the OPRM Upscale Function, but the safety analysis takes credit only for the period based
- detection algorithm. The remaining algorithms provide defense in depth and additional j protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY
- for Technical Specifications purposes is based only on the period based detection
- algorithm.
j The OPRM Upscale Function receives input signals from the LPRMs within the reactor j core, which are combined into " cells" for evaluation by the OPRM algorithms. The OPRM Upscale Function is required to be OPERABLE when the plant is in MODE 1. Within the region of power-flow operation where anticipated events could lead . j_ to thermal-hydraulic instability and related neutron flux oscillations, the automatic trip is j . enabled when THERMAL POWER, as indicated by APRM Simulated Thermal Power, is i i > 25% RTP and reactor core flow, as indicated by recirculation drive flow, is < 60% of rated flow. 1 i f 1 l n -.. -
i I Insert 'I' - Bases B 3.3.1.1 2.f. Oscillation Power Range Monitor (OPRM) Upscale 2.f Oscillation Power Range Monitor (OPRM) Upscale (Continued) i An OPRM Upscale trip is issued from an APRM channel when the period based detection
- algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the i trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the
- channel ifeither the growth rate or amplitude based algorithms detects growing oscillatory j changes in the neutron flux for one or more cells in that channel.
Three of the four channels are required to be OPERABLE. Each channel is capable of-detecting thermal-hydraulic instabilities by detecting the related neutron flux oscillations and issuing a trip signal before the MCPR SL is exceeded. There is no Allowable Value for this Function. 1 i
RPS Instrumentation B 3.3.1.1 BASES-ACTIONS expressed in the Condition, discovered to be inoperable or { (continued) not within limits, will not result in separate entry into l the Condition. Section 1.3 also specifies that Required i Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for l inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate l Condition entry for each inoperable RPS instrumentation l channel. D A.1 and A.2 gM lb Because of the diver o sensors - e to provide l trip signals and redund cy the RPS design, an i allowable out of service ti of 2 hours has been shown to be acceptable (Re s. inoperable channel 9g1 PERAB t permit restoration of any tatus. However, this out of I service time is only accep able provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the
- inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the '
l associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in l trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and i allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip l (e.g., as in the case where placing the inoperable channel l in trip would result in ondition D must be entered and its Req Action taken. and 2.4 l As noted, Acti l 2.a, 2.b, 2.c MiP2.d[. Inoper biA.2 isrequired of one not, applicable APRM or A l channel affect both t ip ems; thus, Required Action A.1 l must be satisf is the only action (other than
. 1 restoring OPERABILITY) that will restore capability to Q' accommodate a single failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and entry into Condition C, as l well as entry into Condition A for each channel.
(continued) HATCH UNIT 1 B 3.3-20 PROPOSED REVISION 7/16/96
RPS Instrumentatien B 3.3.1.1 BASES ; ACTIONS B.1 and B.2 , (continued) Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel ! per trip system is OPERABLE, the RPS still maintains trip ! capability for that Function, but cannot accommodate a i single failure in either trip system. Required Actions B.1 and B.2 limit the time the RPS scram ! logic, for any Function, would not accommodate single , failure in both trip systems (e.g., one-out-of-one and l one-out-of-one arrangement for a typical four chann l Function). The reduced reliability of this 4 pd b l 4 arrangement was not evaluated in Reference 9pWg12ffor 12 hour Completion Time. Within the 6 hou allowan he
)l[ !
associated Function wil e all required c e s OPERABLE , or in trip (or any inatio ) in one trip system. l Grd llo i Completing on these qu red Actions restores RPS to a , evel e valen to that evaluated in Xf reliabili l Referenc s % gpW 1 g whic justified a 12 hour allowable out l of servi time as pre ted in Condition A. The trip , system in egraded state should be placed in trip 1 or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with . two inoperable channels could be in a more degraded state l than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state , should be based on prudent judgment and take into account ! current plant conditions (i.e., what MODE the plant is in). j If this action would result in a scram or RPT, it is ' permissible to place the other trip system or its inoperable channels in trip. i The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability i of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event , requiring the initiation of a scram. ! (continued) ) l HATCH UNIT 1 B 3.3-21 PROPOSED REVISION 7/16/96
l l RPS Instrumentation ) B 3.3.1.1 i BASES l ACTIONS B.1 and B.2 (continued) Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip system in trip would r s PT), Condition D , must be entered an s Required Action ta n. l y and. 2.f \ As noted, con nBi not applicable r APRM Functions 2.a.2.b,2.,end2.d[.Inoperab of an APRM channel affects both trip syste a not associated with a specific trip s st are the APRM two-out-of-four voter and other non-AP channels for which Condition B applies. For an inoperable APRM channel, R tb satisfied, and is the only ac (oth r tha rest ring y OPERABILITY) that will rest cappbili i single failure. Inoperabilityv8f d 'JtThan ;,t,o one accommodate recuired pgg a 4 APRM channel results in los trip capabilityranc entry into Condition C, as well as tr int Cond tion for each channel. Because Condi Actions that are appr iate for thejnpgra ility of APRM Functions 2.a, 2.b, 2 c, end- 2.d(4T t% se unctions are not associated with sp ;ific tr p sys ems s are the APRM two-out-of-four voter a o P channels, Condition B does not apply. L.1 Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilitie:, The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. l l (Continued) HATCH UNIT 1 B 3.3-22 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 BASES ACTIONS M (continued) Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Ccndition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.1. F.1. D V I If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. The allowed Completion Times are reasonable, based on operating experience e th specified condition from full power con ions an rder manner and without allenging plant sy ems. In addition, and J.1 am the Completion Ti f Required Action .1 % consistent with the Completio ime provided in LC 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)." M If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel as s ully inserted. T IN SE'R*7^ 'J " _A ^
^
(continued) HATCH UNIT 1 B 3.3-23 PROPOSED REVISION 7/16/96 l
3 Insert 'J' - Bases B 3.3.1.1 Actions I.1 and I.2 1 L1 ! t If OPRM Upscale trip capability is not maintained, Condition I exists. Reference 12 justifies use of an alternate method to detect and suppress oscillations for a limited period of time. The alternate method is procedurally established consistent with the guidelines identified in Reference 17 requiring manual operator action to scram the plant if cenain predefined events occur. The 12 hour Completion Time is based on engineeringjudgment to allow orderly transition to the alternate method while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in place. Based on the small probability of an instability event occurring, the 12 hour Completion Time isjudged to be reasonable. Il The alternate method to detect and suppress oscillations implemented in accordance with Required Action I.1 was evaluated based on use up to 120 days (Ref.12). The evaluation, based on engineeringjudgment, concluded that the likelihood of an instability event that could not be adequately handled by the alternate methods during this 120 day period is negligibly small. The 120 day period is intended to be an outside limit to allow for the case where design changes or extensive analysis may be required to understand or correct some unanticipated characteristic of the instability detection algorithms or equipment. This action is not intended to be, and was not evaluated as, a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the Completion Times allowed for Required Actions for Conditions A and B.
RPS Instrumentation B 3.3.1.1 l BASES i SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the asshmptions of the current plant , specific setpoint methodology. The 92 day Frequency of l SR 3.3.1.1.9 is based on the reliability analysis of ' Reference 9. i The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant j outage and the potential for an unplanned transient if the ' Surveillance were performed with the reactor at power. Operating experience has shown that these components usually i pass the Surveillance when performed at the 18 month Frequency. SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required i channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test l supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b , only), the two-out-of-four voter channels, and the interface i connections to the RPS trip systems from the voter channels. l Any setpoint adjustment shall be consistent with the assumptions of the current plant specific se A methodology. The 184 day Frequency of SR .3.1.1.10M based on the reliability analysis of Refe n l2 NOTE: The actual voting logic of the two-out-of fo@ur voter channels is tested as part of SR 3.3.1.1. lQ, For Function 2.a, a Note that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1 is provided. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if , the associated Frequency is not met per SR 3.0.2. I (continued) HATCH UNIT 1 B 3.3-28 PROPOSED REVISION 7/16/96 i l l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued) REQUIREMENTS Operating experience has shown t these components usually pass the Survei w per o a the 18 month Frequency.
'The LOGIC FUNCTIONAL TE T f AP Function 2.e simulates STE[tr RM c he wo-out-of-four voter channel i ut to c eck all combinations of two tripped inputs to -out-of-four logic in the voter channels h/
and APRM related redundant RPS relays. SR 3.3.1.1.16 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement cr in overlapping segments, with verification that all componeds are tested. The RPS RESPONSE TIME acceptance criteria ce included in Reference 10. RPS RESPONSE TIME for APRM Two-out-of-Four Voter Function 2.e includes the output relays of the voter and the associated RPS relays and contactors. (The digital portions of the APRM and two-out-of-four voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration check the time base of the digital electronics.)' Confirmation of the time base is adequate to assure required hf response times are met. Neutron detectors are excluded from
~RPS RESPONSE TIME testing because the principles of detector operation virtually-ensure an instantaneous response time.
RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. This Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of
--tr en on c onents causing serious response time degrada on, no annel failure, are infrequent occurrences.
S uV2KT
- N
- REFERENCE 2.
- 2. FSAR, Chapter 14.
(continued) HATCH UNIT 1 B 3.3-31 PROPOSED REVISION 7/16/96
j f Insert 'K' - Bases B 3.3.1.1 SR 3.3.1.1.17 i SR 3.3.1.1.17 l l This SR ensures that scrams initiated from OPRM Upscale Fenction 2.f will not be inadvertently bypassed when THERMAL POWER, as indicated by APRM Simulated Thermal Power, is 2 25% RTP and core flow, as indicated by recirculation drive flow, is l
< 60% rated core flow. This normally involves confirming the bypass setpoints. l Adequate margins for the instrument setpoint methodologies are incorporated into the j
j actual setpoint. The actual surveillance ensures that the OPRM Upscale Function is - l enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow. Other smveillances ensure that the APRM Simulated Thermal i Power and recirculation flow properly correlate with THERMAL POWER and core flow, i respectively. If any bypass setpoint is nonconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power is 2 25% and recirculation drive flow is_ < 60% rated), the affected channel is considered inoperable for the OPRM Upscale Function. Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative condition (unbypass). If placed in the unbypass condition, this SR is met and the channel is considered OPERABLE. The 18 month Frequency is based on engineeringjudgment and component reliability.
)
i
I RPS Instrumentation B 3.3.1.1 BASES REFERENCES 3. FSAR, Section 6.5. (continued)
- 4. FSAR, Appendix M.
- 5. FSAR, Section 14.3.3.
- 6. NEDO-23842, " Continuous Control Rod Withdrawal in the l Startup Range," April 18, 1978. l
- 7. FSAR, Sections 14.4.2 and 14.5.5.
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1,1980.
- 9. NED0-30851-P-A , " Technical Specification Improvement Analyses for BWR Reactor Protection System,"
March 1988.
- 10. Technical Requirements Manual.
- 11. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 12. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) N Retrofit Plus Option III Stability Trip Function," ' [ ;
, ,, ctober 1995.
HATCH UNIT 1 B 3.3-32 PROPOSED REVISION 7/16/96
! i ( 1
- Insert 'L' - Bases 3.3.1.1 References )
i
- 13. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing l
. Methodology," November 1995.
I
- 14. NEDO-31960-A, Supplement 1,"BWR Owners' Group Long-Term Stability l Solutions Licensing Methodology," November 1995. j
- 15. NEDO-32465-A, "BWR Owners' Group Long-Term Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications,"
March 1996.
- 16. NEDO-32410P, Supplement 1," Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," May 1996. l 1
- 17. Letter, L. A. England (BWROG) to M. J. Virgilio, "BWR Owners' Group Guidelines for Stability Interim Corrective Action," June 6,1994.
I I l l l l l l r
I Recirculation Loops Operating B 3.4.1 BASES APPLICABLE case (since the intact loop starts at a lower flow rate and SAFETY ANALYSES the core response is the same as if both loops were (continued) operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgem m t. The recirculation system is also assumed to have sufficut flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 14 of the FSAR. l A plant specific LOCA analysis has been performed assuming l only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3). The transient analyses of Chapter 15 of the FSAR have also l been performed for single recirculation loop operation l (Ref. 3) and demonstrate sufficient flow coastdown l characteristics to maintain fuel thermal margins during the L abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop l operation, modification to the Reactor Protection System l (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Simulated Thermal Power - High setpoint is in LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation." K Recirculation lo s operating sat sfies Criterion 2 of the NRC Policy Stat nt(Ref.K). 4 LC0 V Two recirculation loops are normally required to be in l i operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis'are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LC0 3.2.1,
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"),
l 4 (continued) HATCH UNIT 1 B 3.4-3 PROPOSED REVISION 7/16/96 s
Recirculation Lo:ps Operating B 3.4.1 l l BASES l LC0 (continued) and APRM Simulated Thermal Power - High setpoint (LC0 3.3.1.1) must be-1giplied to ati ntinu t-on [ Itent'TvTibtdassumptions of Referenc . M-maan < e. n ow me / rnon+4mn nr enrgthor 2Vnnw.dne
~ 4 4E $ am.. IE d11nbE M E4aa n
- z : : / m n m x x 4vavi f4nnr d au5z:/.d .~[ 7 i #
w .a u s s swi e 6nus mu s nyu c.eu s s w s s ativxrw uv ovy vmvu
)
l P APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor ! Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. f- - A.? A!W 1 / ACTIONS .n u m .1 l l7'aOc^ /c thes , al-nydrau l t,
/C
>63uil j/cncel..E,
^
0"*"Sti"a ef-pl :". ith one ^^sciiCglati n lG5p a CGntrolled by-
]tge+*4e+4 a
/ THERMA, /CM,.;Tt h eg/
fi nw tn >y
.~ -
.ARY nf
../..-/
~
2t.A co-- #'----'--
-.'I.~"/_'.'Y" """"/
ruwm spreateron f rea"i ya==at i:t ::ed en t .N,an one y reo iine y inis me.n l
-- v u te-
, _ _an_ . o. .n. .a . .u. . s u n
7 .. uu g nca oncp iun _ N. _.._ti ,O__. .... inmy
. nmo 3".". M,7,masfi'N. *M"""
l ^A .mummR.f,7 s.u re 6 gion" cf T'y vu- ENl.u*F F is in l actian 4ctild Sc nitiat-d c restere $egre3.4.7,1,p[ryt-the"Onor__tionN[et11c;;cd[R~yvi fic.: 7 de:cr -
-combify[tien
?hnm.ennle+fn7j.c'_b: to ' thin thujvperetienp;)luwed eden ,, le. =vvavil ipe - Rpion ,
d acp/entocpdringdg[eg,9 this t peried[cncrc[i'cfa- cn b - he to ceAplete the f gired fc .cn, cad on freqv t ccy f
=11 mi f x. - nuiliations to De p=enito[ri-jbyope[recr:
au4tkl eteded. An iii...;dipactop'screiii ij' el sug' react ed .;u, ,,ufi.prcpety p= pun operatjon, sj..ce all-40 irs.usatiun f nas pee ny ui a- M G 5C U * ' y:e mlost =r ater. any Lne pro piilt or ~ i:f
.4+p.,....i l
(continued) ! HATCH UNIT 1 B 3.4-4 PROPOSED REVISION 7/16/96 i 1 1
Recirculation Loops Operating B 3.4.1 BASES ACTIONS b re the LC0 not mel Mp :, 'M,'
,k MMidi$ d;,M."X e recirculata lo s ust be restored to operation wi matched flows wit in 2 . A ir tio considered not in operation when the pump in tha loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses.
Therefore, only a limited time is allowed to restore the inoperable loop to operating status. Alternatively, if the single loop requirer.ents of the LCO are applied to operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence. The 24 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable tin to ccmplete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump. 6 II.1 A ny(RequiredAcionandassociatedCompletionTimeof Condi iorfW0 n met, the plant must be brought to a MODE i which the C0 does not apply. To achieve this status, n must be brought to MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of Design (continued) I HATCH UNIT 1 B 3.4-5 REVISION 0
Recirculation Losps Operating I B 3.4.1 i BASES
- 6 -
ACTIONS g_d continued) asis Accidents and minimal dependence on the recirculation l loop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and-without challenging plant systems. SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins t:.s the fuel cladding integrity Safety Limit such that the pote.ntial adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows , from all of the jet pumps associated with a single ' recirculation loop. The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless' during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner. (plot tred. R 3.4.1.2 Y Y m /co......, u. . . <, / . , 4.-- + 4 - < s --- A--.4 z_ za:i..u : :2_~_xz :.r;;i;7,n 17 CZ Z '2'!AG"7 3 L T Z G"M7' Z ., G _ ZLV'4 "'C_-.~ZY."iX, 7 ;'I's Z T; Z ~ X U 'i7Z..- si
/ 7l, Z-"ZE 'i"I';C E ';. Z'JE'; 77"ZZZ2 ' ' ' -
i id[2h555554 [t5 W Z!i TyESE @ EAI'EIIty Fij r: 3.[11i:d:::ddtheMid-:d:'lidd4 l l / (continued) HATCH UNIT 1 8 3.4-6 REVISION O i
l Recirculation Loops Operating i B 3.4.1 l l BASES en n , / s . . . As , SURVEILLANCE 6 e ., - . , . . . e. . . . . z. ,- REQUIREMENTS I /
,...Z. > .u .. f.. . a . 2 :. .
..Z ... .. . .; Z ...
... /.. .4.... .. , L- . . u.,,,u...
. : ., . .. ..,. a.
. ./' ,, .. .,. .. ... ..
4.. u. ,.....m.. ... ,, . . . 41,. ut. . .
- u. .......J......_...,. f ., .1
/.Z ., .. z. . . . . f. 2.
, . _ . . _ . _ . ,tvK . .
. .Z- . . . .
. Z,...
- 4. ,. ,. . ./
u.........
.u..... .
- 4. ... ....u.
,.2..
/.. .u. ~_
.,. .I.
....4
,.a 1. . .ea.o 4. Z. ..,.v.
__ m u.. 4
....__...;u,,.. ....I " ,__ /_.
.A,/u ... ...... ..
u. Z... u..... 2...
.: Z. .. . . . .
- E.. J ..,....,..
. . . ./. . d. .e /.
- a. . Z. . ,.
. _ ..../ ~. ._ u.. ,u. . L. . . .. .. . .. ,.. . .: . . a.L . . ,
f ,. 7;' .. . . . , , a. . .4 ._.
,e,... ..... .
- s. ..u..i.....,...
, ,n,
. , . . , .... . . . .,.Z 4,
. ,. .. ..e.. ..,
4/ 4. , ... .u.. _ . . . - . .X. u, ... 4.,.
. . . /4 . g /,e4. .u 2. F. 6. .. .u ,. ,.1.
. /.4 e p ,u.. gk1. m. 1...A2 ,
.,.. . . . . , . . . 1. a uq,_.,.,_., .s. ._ _ ag'e ' :- -..diti:::. [ . .
+e(,E '. .': . . ,:ft-
. 7.::t: i ng 7. n., !
A_ - 4 REFERENCES 1. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," December 1986. l l 2. FSAR, Section 4.3.5. ' 1
- 3. NED0-24205, "E.I. Hatch Nuclear Plant nits 1 d2 oo ion " gust F. .
./.2. A.,
I a,, / e de.,. l. a. . L. .... ,.. .7,./.
. . .,. / . .,.
. . . .. . ,n -K,, ,.n.
Ar.f . . J. .,. . n i f i-s. M.- x J .v... c.1. c . . L 11 6 s.
, ,u.;,y';" 7 ' ""' "y'"/' ' ' ' '3/'" '7 '
l . 1
,K " ' 'Z ','D,
,..7, .. .
&I. NRC No. 9 - , al Policy Statement on Technical Specification Improvements," July 23, 1993.
i l l l l 1 1 f i i i HATCH UNIT 1 B 3.4-7 REVISION O
^
l I RPS Instrumentation ! B 3.3.1.1 1 i i l BASES APPLICABLE bFo'wehlance Monitor (APRM) I l SAFETY ANALYSES, /nSer#
- C
- LCO, and 2.a. Mverad~e) Power Ranne Manitor Neutron Flux - H l APPLHJ.BILITY (continued) Th:
V V N ch::::1s-receive input :t,n:1: fra. ... 1 ::1 ;: : l r: ;: :: iter: (LP"":) within th: re::t:r ::r: te presid: :n-f adication of th= p-- r distributie 2d 10 :1 p=:- 05: ; :. The ^"P.". ch::::Is :v:r:;: th::: L""" :ign:1: t:j previde : ::: tine::: indi :t10: Of :ver:;: r---t:r ;;x::, fr : : fee per:::t t: ;re:ter th: ",TP oFor o M on1t 1 low power (i.e., 0 i?utLon_S i Pf; J )th'ii onAwiageis capable Power Range Monitor of generating a trip signal that prevents fuel d resulting from abnormal operating transients in is pow ange. For most operation at low power levg verage Power Range Monitor Neutron Flux - idfuunc io will provide a sect.ndary scram to the te Range Monito Neutron Flux -. High Function because of the relative sitpoin . l With the IRMs at Range 9 or 10, it is possib that the-Average Power Range Monitor Neutron Flux - H h* Function (Sd Q will provide the primary trip signal for a c itle 4 acf H sa t in power. No specific safety analyses take direct cred for tha - Average Power Range Monitor Neutron Flux f- HLh Fuh' ion (gh)
! However, this Function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 25% RTP (SL 2.1.1.1) when operating at low reactor pressure and low core flow. Therefore, it indirectly prevents fuel damage during significant l reactivity increases,with THE. LROWER < 25%
Of - The ^.P"" S i- divid ir* te: ree 6"-t h: =r d in J. t-4::ch +4p :p:4=f .ch:r41: 4 :y-teeith l y-thr:4:EP[P 4: -
...,1 ' l.e::h E f 4 t J. ,....i- A..:d n ./ .
.te
. .file"",C.
, o,.u. . ?.: Z. 4:::
.,L. ..g. . L,.y,. :1: .,.. 2. .....
. .i7.fp -"...t e: f. ..b:
g .y ,y.,
- =
.. .e te}.ip. ..
'N,...f.:::y*..d ty,p ?Z"'. :yp,?,.$5..E.,5.
.crZ..cy!E$.ng$.,E ../7'
..?.?.i ?'T.i.. E,N..."5.5.
W . In c e s war,. +kn+ . An c4n a #s 41 elma ue 1 Y monel d a n/e m.- E=== hkke7t's.E ((a a u 14A nel 7 f= 55a [ m J -
- .Z.7.j. 'Z.' ' C.
~
i I G 7 '
;;iL.'ss.Z. _;_Z. . Z,_'.F. jv. K. 7 :3. 7.f. .n 4 7a".Z.'.
_m_ nea+ s v.a ma e via ame M e ====1 + k 7+ 1 aj.++ +w 7 7i 41.57.::.: . le. '. .'.' ' :'I.. . c. ' i.i. .I. 'e. .".". om . ".". ' ;.C.
. C. ."' '. ". H.
i L . . .": cre ..,. 7.c ated. / G-L -
/ N (continued)
HATCH UNIT 2 B 3.3-7 REVISION 0 \
i l l Insert 'C' Bases B 3.3.1.1 Averane Power Ranne Monitor (APRM) ! 1 The APRM channels provide the primary indication of neutron flux within the core and i respond almost instantaneously to neutron flux increases. The APRM channels receive i input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM ] channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP. The APRM System is divided into four APRM channels and 4 two-out-of-four voter l channels. Each APRM channel provides inputs to each of the four voter channels. The i four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The APRM system is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one unbypassed APRM will result in a " half-trip" in all four voter channels, but no trip inputs to either RPS trip system. A trip from any two unbypassed APRM channels will result in a full-trip in each of the four voter channels, which in turn results in two trip inputs into each ! RPS trip logic channel (Al, A2, B1, or B2). Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the entire core, consistent with the design bases for APRM Functions 2.a,2.b, and 2.c, at , least 17 LPRM inputs, with at least three LPRM inputs from each of the four axial levels i at which the LPRMs are located, must be OPERABLE for each APRM channel. 1 I I l I i
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Averaae Power Ranae Monitor Neutron Flux - Hin (Se/doun) SAFETY ANALYSES, (continued) LCO, and APPLICABILITY The Allowable Value is based on preventing signi can increases in power when THERMAL POWER is < 25% P. (Sefdown) The Average Power Range Monitor Neutron Flux - 1 (Function must be OPERABLE during MODE 2 when control rods mayW withdrawn since the potential for critica yeertsts. In MODE 1, the Average Power Range Montt r M utron I Flux - High Function provides protection a activity transients and the Rim and rod block monitor protect against control rod withdrawal error events. Lb. Averaae Power Ranae Monito imulated Thermal Power - Hiah ' j . The Average Power Range Monit r . f mulated Thermal Power - High Function ttert.m n flux to approximate the THCMAL POWER being transferred to the reactor coolant. Ths APRM neutron f1 filtered with a time constant r s31 c . icaMy sentative of the fue1 N heat transfer dynamics to gene ate a si==1 pr-eti- 1 to , the THERMAL POWER in the react r.t The trip level is varied _Tn;c4 " as a function of recirculation flow (i.e. at lower core flows, the setpoint is reduced p al reduction in power experienced as core flow is reduced with a xad,ceatt rol rod pattern) but is clamped at an upper mit that WMways lower than t AvJeaePowerRange Monitor $Nejutron Flux - H owable Value. lP 'T e Averaga_ Power Range Monit r,pcjty inulated Ihe Power - High Function . b,.. a
. p._ ection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event) and protects the fuel !
cladding integrity by ensuring that the MCPR SL is not l exceeded. During these events, the THERMAL POWER increase : does not significantly lag the neutron flux response and, ! because of a lower trip setpoint, will initiate a scram ; before the high neutron flux scram. For rapid neutron flux ' increase events, the TH the neutron flux and the Average Power R Monito N utron Flux - High Function will prov de a,s g [r,# g nal before the Average Power Range Monito f' 7 .y .,_ S i lated Thermal i I (continued) il l HATCH UNIT 2 B 3.3-8 REVISION 0
t i e Insert 'D' - Bases B 3.3.1.12.b. Average Power Range Monitor Simulated Thermal Power - High Changes to fuel design include an evaluation of the time constant to determine if the electronic filter requires replacement. ! \ l l l i I l l l i l l l 1 f i 4 5
.- = = - .__ . - . . . _ - .-
RPS Instrumentation B 3.3.1.1 BASES l l l APPLICABLE 2sb. SAFETY ANALYSES, Averaae Power Ranae Monit r FM'-i5 mulated Thermal Power - Hiah LCO, and (continued APPLICABILITY t Power - High Function setpoint and associated time delay are exceaded.7 iL. Annu e..6- J. J2..J_J 2_A- s.m_ N ' N
- nw .ans au a wy,L.-- a b urn1 ia usy ausu snbv 6wv yivuya ya s u_fu r nLs . w s.kO2 .a w sLenr-
_+L
.4.
w.is ww J..J An.nu ru us warusse sw a
- J 4.-ne. 4- ...L us yuwa sv w wwss es sy ay a b srus ,
ema- .o.a._ TL. -o.4 a r rw ay a bwurs { 4- .1 1.m. ___ ,L.__-1 2-ra ---L amJ- .o,4m_ 6- L-kn.... ww a Ju ysisu A-obw -- assww Annu vns wesussurs s en w wwss bs ay ay a t wass bv us wJyua**we ony vow os nus wounuwe -L.--_1 4- = 6m2. ,o,+-- -- .o.. bL_ en u we sy ey a w wsn won www-w ,
._---1.6 2 s1 ,._s. a- A_1_
basb usaVwsubsu bs sy r.._ L-___1. f \ m.. aI
- 66s 6v 6 6 =' '583 v
.-- n...__ n.___ u _ _ 2 A _'s mvssp s wws e neurys ssVssa b us_ c 1 .' 8. Fan 2__'vu'e s sWW WsuJuW
_J 2 _' .'_"A 1 _ J TL_' 1 I
' n. --wa m. . . . y U.s4s s , L..m 4... -L.m 2- -- L
.J s SITU s a b su s ess s issu s i 4n s
- w. J, 6 L bsu w.
1- w . , u s , ._ _1 - nssa su s ews s Ab ., sy a,o.e.- J . . a b w.us u .. s...,wu- J i J nn.a_nn+__n#._4.u.
- . . w-._ v -w wysw us s ss
.-- ---..J- 6.
bw us L_ #inrnantr L sw1p u s by n_.._ w..... w +w. u. u. +es. --
.4..,,_
s sw sssm . s.. _ - - . q usus s s w uf . ,k. . _ _ .-__,..). .,,,VI . wwa un. f _
+w u.s e ,ae . . - - u -- -_ . ...w s s b, 12 ,-_ , ara ys sb w
,- su _ ms,__ uus u si v .n s vssw w a wrs vs a u ...vusau
-. J--..... _-.._____ , u - -_.,__ a-___ s yasu a . ,
-.4J- A ss uvu s b s vis , LU l
[ y wv suw .. ,__-. l 11 uwwquwww wwyst uys vs bsis sus h s s E uvs G g ub ssmak
.ne, 1
w D. Bu. 4. n , c.+. +ma
- i. w.
.ru w i n u.w s w .,.4-J u s s su sv f-. -..L w uw s s ma A nn,u s is
,b is . .
mh yu, n4.+ .n Wuw I w , nu. a 4--...., f__ -L f u, f. r3.u., ,-..1 .1 ; b s w_ s ru s_ w-w s . r isyu w. a s ...s b uw.s .s n . v , m.u.was s-w eunusi nau. --- ,_. APRM chan 11v.www oE 6. m. gw m b+a essw mss uw .waysu.
. ....m total drive flow si alKrepresent3iveof o;a ..
l
'The total drive f1 signa 1Kaeeg v rated yVf N *.' d E' i n n. 4. +- c _., +- wn. n f. uk 4. e.k. en
-- 'Ib e4an=lelen +ka + = =O , + m . A Annu. - - r r ' r-- -- a - - - - =' w ** ** y-=v w j
.mL 41. 6La -6L- eu. ...--lo en
. 4, - - = 1 ,, +w.k. w +w4n
/ ma ne i s , wvs e i w biss v wiss s www a i auyymy a ynea tw w. iy l c gu .e. + . -n.e.
E k 4--.1 . D A DDM..e . wse.n #.1.. au n .ns. .m 4 + syn-s e 4 ,, y ,.-w', 4s-w- J . A ko un y
- - - - -e n- 4 n.o, y +.. k.n. r. i. -nu .
,4n y... ,1.,. . c.wn-a ,
+L-wnw +m. -~4--mi-+4--
v swvs ==s-wswn 1.nane 7-- T.n ak+s4.n w --.- + k. . a
. w m..a
- w e. .
, + .n-,--o.64o- --f--_--a 4-- wW.IewI VEs w g vb a s a w s wsaww .R egi3u5ag
+ha OA6 I El m. . 4. 3m f.__- AL- &._ fl -. . ..20_
wesu w w w in a s s ww. asy I___-^I-S J u t. +wks , , +m4. .u,+.- sueaJ-.- s i vues 2L J bass bww a5ww uss 5 ha guaawweabsu u
.L-.s --_ __..A-J A- . 1 -. .
es sy ay a w wsu me uwaws susu uwvusf us s avubsu bv u s wys
..,+4
, - .1 -. . 2 6 ...--2.6 J .2AL --_L annu r_ L annus_
.w w s wie wassuab e eae+
unmalam4, a www e a bsu w a bas sawss ni nsse L uw e s nsIVs a
--. 4am i wams 4
- m. 4 6 wauwus6 asawwwe
&L-b u rb 1.--.- f &L-swwse vs bsss b yrv s 5 vvy uss a b
&..- fi - . ...2&
e4 Enm H,a ee +ka em .- O 4- ._mf_---=_
- . agn n a l. .e iwi www uw wisw w w s um ha sy a s e ss snus s vs f-= 4k4 basu b nset4en1m.
r-- - - - - - - ADDM. ===" > i Esek n n4 mJ Au .nm n... D. -m u--46-- w3=***= o *s usw s wwwi ri m , D4.. J eJ-.1.6 J TL-__1 n-..__ n=s>Uw a sw"
- 6 v s u aw nas _L _L.- 1 -_1m usudsus w s uuru s a b s u s s sw a ariu s swwws - as s yn wasusases a vus s y I
- c. a yn n t. e.w n .e .. n4 m. u. y.w m n + t O w- n w- nnenADtE fi m._. u w e s.s w iw s s. a vww unsw 46, .4---
a i sswi 4La 6ses _4 m A 4 4 J . . 1 Annu .L.---1 . 411 -f-_ 4L- JmJ [ i v enu : ruuma os su a wnunnus wasa yws susm besu J ass 6%nuwu f..--+4m-iunw*iwu anlu non _ u 4. +. k. w. sg view nDEDADIE ..4+ 4-...e U...-...m J. J-- vs s s w iws w f1 a uvmm.s. est s e suyuwe s awww w w s , s es vs uws en == 4m+=4- m i-nla f = 41. . . . 4+m 4. .. J-.. 4L-J .k=... vv enuinsusas aany w f--
+ka ri...+4m. .+ i. n ,, + m..
a u a a us w we a6ws eu ud usaws swww uwwww a vs
.ss. . wi s s . .tung uw wu w wnw
.-.4J An--.-. n-u== D.---
i u__a+-m ri ... nt._ J e4-.1.s s w qu i . wu newsmyw avv.ws nussyw J 1 n-.... s avis s 6 v e e aww wsnasu w s inu e e b w u
+
TL.fEea a ess awwws - t_fest a. L .L..- yss wasunssw i 1 4, - . w.w
. e .k. 4.yn cu og ,e ,.w +n, m..u .c +wwkn w- essskl
.. . ---.a a f. m..,s 4. m. +. s 4. n. 4. n.-..an
- . . n,
_nnrD RDt E fi m.m ..44 vs u s w iw i s. i sw. unsw 4 ,. - = 1, 4. m. 4= ma
.wsyn= nu.n=+ af , -
sw w w.s. w. _f s 41. :s wa a f. -man g
.s.-w. u e + 4. s
.n a n e w 4. .w ee n. 4. +. , a- m. - f. l. -au- u n. 4 4. , 4, n +..k.w a .s .c ,ewwn e 4. ..m= +-'n. d tw4n \
'r e
,jo e + .w.
n es,t gw.n., ia p 4. [. g [l an s . m. 4. +. 4e
-. 4i nman-mahla,
-y w a ows. am wn .a. n[ +ba vs vnw
+un
.-w
- r. a ny s e 4. r. -n. d A.u.vn e.sna g=
D an.a ea
. --wa O. . .s3% m. a isw ua=4+m n fl ou DJ..mJ P2-.1.& J
. sswvi a aww uieasu w sinu s a bsu d
(continued) HATCH UNIT 2 B 3.3-9 REVISION 0 I
- . -- _ - - . - . - - . - = .-- -. - ..-. .
l l l Insert 'E' - Bases B 3.3.1.12 b. Average Power Range Monitor Simulated Thermal Power-High (continued) Y
.... .. the flow processing logic, which is pan of the APRM channel. The flow is calculated by summing two flow transmitter signals, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is pan of the APRM channel OPERABILITY requirements for this Function.
l i t {
'l I
L l l l , l l l 1 I I l , t , I l 1 i I l d i l 1 1
l RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Averaae Power Ranae Monitor Flow Biased Simulated SAFETY ANALYSES, 3hermal Power - Hiah i Y LCO, and N(continued) ON APPLICABILITY pn yThcgcp gg i
=y'... ]' .we 7,c y 4
tpiyg],,v I 7ex The clamped Allowable Value is based on a ke pe credit for the Average Power Range Monit J b ,F i Af Simulated Thermal Power - High Function ' '
'on of the loss of feedwater heating event. The time constant is based on the fuel heat transfer dynamics and provides a l signal proportional to the THERMAL POWER.
TheAveragePowerRangeMonitr(d$['Vc,- f 'mulated Thermal Power - High Function e e OPERABLE in ! MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity. 2.c. Averaae Power Rance Monitor o Flux / Hioh r Nnu M, am 2 4-. . ..[ zz2: : z _cx :n t A... _a. . ...t..x, ..x. . :x...x.:z. , m.y" l f ,n^. 7,ro-,gn. . ux.h4.x. o
- .. 7,,c rc,vuoverage-P.ower
.,m, 7' m o t, .
nge-M6nito V b"edfeltron-FliixM.. The Function is capable of generating
-tetp signal to prevent fuel damage or exces eJCS pressure. For the overpressurization pro cti asysis of Reference 4, the Average Power Range Moni or peutron Flux-HighFunctionisassumedtoterminaethe3ainsteam e isolation valve (MSIV) closure event and, along with the safety / relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code The control rod drop accident (CRDA) analysis
( 7) takes credit for the Average Power Range Monitor J Ne tron Flux - H unction to terminate the CRDA.
-The-A , di"'dd into Ta-gro D h E::_^fR ne g ac psjyd;"
=,hg __ up l
I rz : : *7p" r=r"r
- i'::z,"!.:;gua:m;":c.:p' L"z:_.;.:
"y ::z:.w:::" r,:z r"'".:"r,u"."no.3mm
- n. c. o3 m m 2ovn: X,su' C. . 'm:'uMIt:
meum v, .c'., u
"" ' / " " " L' 'As o" '
o . 37 -
.s s' f (continued) l HATCH UNIT 2 B 3.3-10 REVISION 0
RPS Instrumentation B 3.3.1.1 BASES m APPLICABLE 2.c. Averaae Power Ranae Monito NN tron Flux - Hiah T NALYSES, entinu P w Z APPLICABILITY b tu ch:r.p 1: irj::htry:y:
- r: f quir:d in g::n: ::t ..
g
'y.,f e 1 g) j .: b: g.,:rr:n ."^SLE[:n.. 7 th:t
.ing1p in:trp :nt
^ ' f:'7.
ign:1 r: ;) 7.1 In :Z.ti:n, preeppr: v': p:r:: fr p:f..:thi:7'
..:n_:n Tun'Z,. .,. .I ._. .veli" _..... .. ,, J, , -nr. f/
2_
,.w .. . . ,/. r. ... L. .._ ,/.. _. .
2..h_. ._ p . ._ , .,X. _1 _ _ _ . 4, ,,_1. t/nnu
, . . . a. ., . . , , . 2.,./._ .J f. ._. ,I . L.... -.,,.. .L.__I1.,
......,,. o 2. &..L /. 6
- 1. 46 am.
- r- - --
___ 1 _ /. A . ./r. 1 _ . . _1. r. - . .L. Iw % .A. s Abe .L. .r 2 L. L. r-
.w . . A .L. ri.u. . _. m2.
,E .
. .. 2. _1.E.
_ " . ._. _ l, _ _ J. . .w. .L.
. r n
. . . . . . . 2. .
The Allowable Value is based on the Analytical Limit essumed
- in the CRDA analyses, i
' The Average Power Range Monit d,"// eu N ron Flux - High Function is required to be OPE .E5 -N6DE 1 where the potential consequences of the ana' ed transients c result in the SLs (e.g., MCPR and RCS pressure) be .g exceeded. Although the Averag r Range Monit Neutron Flux - High Function assu n the CRD analysis, which is applicabl 14Jg)D12,theAveragePower Range Monitor Neutron F'icx HfgfFFuneionconservatively bounds the assumed t and, with the assumed IRM l trips, provides a u tection. Therefore, the Average Power Range Monit N utron Flux - High Function is not required in .
,2/ .... / ___ _2_.__/___,.__ _ . .,_
, i: p. :1 : r:: t th:r: :d:q.
"::try itch t-
" nt'7, ring p:.te:
tecti:n)ier 9 .. ran 7 t f..:p1 .d g :itier.j the in y[ter J..h
" "."": thp f.
- dp . = :ing ter :p 9.,.:
witp., in run,j.a """"y p.1:.[" ::
- ig j'. ::incidj..t with ..:::::i:) dint:j..:di:t:$
I b=ag,%...:nito/r
. , .y 4.__.., n. . . e.
- e. . :tr:n F ux -- "J[.,:.r
....m.
i- ____ 4 . , , m In Z.'.3:r:te- pA. ' . gn:1 l _. . , ...y ,24.. ,... Ik-Z.t =1yZ. but it_t. ' .,. ,:. ,.4.T.u..
- 7. .
_Z :::i'" th , 1 2. r: t :i n:l.._ f: r .t t.,.4._.E.u.:.,.v=_/u _2 24..
- 7,... . . g.
_ppr:v . ..::::: 7.:.4..m,
... . .. . o.n.e
. .. . L.
itj """," Sy:'7,. i: dip: dint 1: .th p..;; input int :::7. trip y.'7 :. T :ytte- : d :d tagr[::.Ofc[h.n:byp:p[.e i
- . 211
- ,..y .: ch = :p in = h t
! -("=v r, the p;t:n.ici exi:t.7 ; :yn.: :: :.j:nd . to byp:: : b "r"" :ing (continued) I HATCH UNIT 2 B 3.3-11 REVISION O I l
1 RPS Instrumentation B 3.3.1.1 l BASES
, ~ rm -
\
APPLICABLE 1..J C ^I r: f: ";.Ir ":::.l '"; nit-l - 0=n: w::10 \ l SAFETY ANALYSES /
/ /
/ / / /
(: atint: 1 , LC0, and ,/.. .. . ni/ L J
/a / r_ . 1, . . n_..__ ,(____
^ " I- . '-
- i. s_. m _. __. _ _._ ,.
In... "' .u. '" , . I '7'. . . _ _ L. . _. ._.
. . #I' .. . . . .
. . . I , ,,
"..-.'.,l..,,..____
APPLICABILITY .. ^.
- rr
.,y . . I
, n,,neL.
,y j .. ..
n;:7 _ ,i n :J.:_ L. a. __
... n:- ,....-......
;u'
_,_ .... A......p.-'y.
-cf-tw.-/z.__1;gi:.,',..-.p....
1, r qu'rp., d to r 7 ,..
;:li'7....__ .. . . .
r...f... .f t
/ ;L...
r . . . .,.=. .1s.,T
, _ n.er.
. n. ,. _. '7.i _
,., _ s_, ."ur'L.
...s., .
.........,..~.,
i ___ ca :]j:
%. sign:1.
,J. , u. _
a.
. . .,1.
_ f.Th: ., ._. ._ _,/ r..__6A___ d__ , _I _f r p.. ..i, ___&. v. A.. L. y.s AnfTn
.. An p..i.
.f fa. .T T. p_w . f. &.nyb L Jm A . . _ _1_ _ i
_.._L n__ n_1____,_ . . . j. , . I
,/n....y....., . .c L.u.... . . . . J _ . ____; rf.__A2__
. 2. A L _ I TE 'FAL__ JE / I
...L. _ .r. n.o7.L . . . _ . _ _ .f, _ . ._,__
_ . , ....C.,,._,......./..... T_ . . . . I. . . .u .. . . .,./ . . . . .//. . . . /_ _. n _Z .
. u__
. . . . /..__ 4. .. /_ _ . ./ , . /
. E.. . .. +n.L ,. ..,, . .. . . .L _. . .
n r..... )
. . . . . a. , . . 'in. . . . , ._. .. . ,_ , _I.
_ . , . . . . ....., .a _,.__ .f....u, 'L. , . . . . . . _
+ u. +J,
. ., . . . n. . . . ], + / .. ..J...,,./
,IL ._ _.. A., L... T nM.. J.. _. ._, _y ._.,L. ._ L,m 4...s,.
.. / _ /_ 1, . ./ - > > 4
,i. . . .L. . ,_1.,.L.
.. ,F. n.M. .4...,/+ ..-., ..
J d..___,__
/ ./ s ___
A_..__ y . 6. fy s . L. ._ u . 1 2. u A. ._ ns ..s.y. I /
. . . mf._. _, y_ _
/ "c. .= i cr 0; a:ui: .h=nci .: ::n .dcr:d in p: .ble '
,,,_... _, 2_ _;
- y v.L . m. . . .nn l
- u. . . . .7 ...L
- n. ..i. f_as_tr: o...f, _ _ L,. . 7,_ ._..:d: L7 n. ,,j..
- 7. - r.. 3 i th- f. . g: _ _.,fers . :re _
t LL. .. ._ n .... _ _ . Xno.l _ i =_ .n .. _:r 3::. . . . . . 1 , no, .1 :
- v. u. . ._, _a
.r
- a. . . . . .y. i. . _.. m VI n.,5nr,.n. ,p _
.7. . . q . .A .U 3
+.. k. 4. ., 4,,./ oL. . . .
- 5. .
%f E.r . pp 5 . ... J 6.. L. - . annul. . . . .. . . _ _. . .. v 2.. I ._ ._ 2/;2. _&___
1r,
;=. A _ ,. .n.
.p _ _ _ _ A. ._y_ j i
- 2. . a M 'oe n runnon civy :0 .ne
'#provides assu that e minimum number of m
fu py APRM ebl, This APRMfvare g bPERABLE.+h gnytim = ^""" mode switch is = :d t: V any position other t an " Operate,"[n APRM module is InSegvp ,, un .- -' -+ al ugg ed g +. '.L..o.. .,, ._,,,.A.n. L,. . . -. -. .. .-'. './.,' r.
- ., . ,___ . - ". .,1.
. . , '. .. -f .- . L... /-
I. -
.n ,- __
_. u.., m.
.Y. .r ., _ 1_.L . . , ,
J._ ._ 2. . . J / L,,. .
....,A._. anhu 2
_w_ f, #. _. _.. .I.
. ._2 2 1.h., _ _ / _ _ , .L ._ n nnJ.
A. . L. r/n n a_ n,e. ,, , . 31. ._ ._. ._
. ..L..a JuL. . . mL. . . . J. J "J..7. __ ' * *_ ']' _ _ ' ' ' ' 'f_ _ "_ ' ' Jjv _w "_ 'f' *f*'} _ . . _ _ *J *_ . . . . fJ E /g gg g l
- 7. ."."y.' '. .,;"."L, . . ."."- "f "'. .',"!. _. .,7 T.
Ton Qas-netspecif4cally crM p ... , A,,n, ?, '. . w ,..
- n t W accident i
analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis, i r1 - _d n_. <_ D k._ z,.
, "z...y"7. ";y: 'r..";.. ":r, ,.-=., .r... '. _'_' z. ..,.li-"In
_x n nwr :";
+' ..y . ;
/.lud: j . 11 p y. .. 7:c.. e su Z:" .
y . .. c,..Z. s. ,. Z.._. '.. ne . ,. a..
,. L.ngl e '.yl ,
ur:A. , r=, . r= l
/L -
% d (continued)
HATCH UNIT 2 B 3.3-12 REVISION O l
i I i I Insert 'F' - Bases B 3.3.1.1 2.d. Averane Power Ranne Monitor - Inon i
?
..or 3) the automatic self-test system detects a critical fault with the APRM channel, an Inop trip signal is sent to all four voter channels. Inop trips from two or more non-
, bypassed APRM channels result in a trip output from all four voter channels to their , l associated trip system. ! l t l ' I l l l i i l l l l l 1 i f i
RPS Instrumentation B 3.3.1.1 1 BASES e APPLICABLE d2 SAFETY ANALYSES, 2.i./JAverace Power Ranoe Monitor - Inoo (continued) ! LCO, and APPLICABILITY re is no Allowable Value for this Function. Thi Func ion is required to be OPERABLE in the MODES where e APR Fun tions are required. !
/n ser+
- G " \
A earAor Vessel Steam Dome Pressure - Hioh An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive ! l reactivity insertion. This causes the neutron flux and ' THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection sis Reference 4, reactor scram (the analyses cons rvja iye,1y 7 l Neutron Flux - High signal, not the Reactorassumescramonthe l tea I Dome Pressure-High signal), along with the S/RVs, 3 ts the peak RPV pressure to less than the ASME Section III Code ' limits. High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure - High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event. Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Function is required to be OPERABLE in MODES I and 2 when the RCS is pressurized and the potential for pressure increase exists. 1 i (continued) HATCH UNIT 2 B 3.3-13 REVISION O
Insert 'O' - Bases 3.3.1.12.e 2-out-of-4 Voter i i 2.e Two-out-of-Four Voter
- The Two-out-of-Four Voter Function provides the interface between the APRM i l Functions and final RPS trip system logic. As such, it is required to be OPERABLE in the ,
MODES where the APRM Functions are required and is necessary to support the safety I analysis applicable to each of those Functions. Therefore, the Two-out-of-Four Voter l Function needs to be OPERABLE in MODES I and 2. All four voter channels are required to be OPERABLE. Each voter channel also includes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, I-an Inop trip is issued from that voter channel to the associated trip system. There is no Allowable Value for this Function. I 1 I i 4
RPS Instrumentation : B 3.3.I.I i BASES j APPLICABLE 5. Main Steam Isolation Valve - Closure SAFETY ANALYSES, i LCO, and MSIV closure results in loss of the main turbine and the APPLICABILITY condenser as a heat sink for the nuclear steam supply system ; (continued) and indicates a need to shut down the reactor to reduce heat generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs ) are completely closed in anticipation of the complete loss i of the normal heat sink and subsequent overpressurization it. However, for the overpressurization protection < papsp of Reference 4, the Average Power Range Monitor i f r,. N tutron Flux - High Function, along with the S/RVs, 1 the peak RPV pressure to less than the ASME Code , imits. That is, the direct scram on position switches for j MSIV closure events is not assumed in the overpressurization i analysis. Additionally, MSIV closure is assumed in the ' transients analyzed in Reference 2 (e.g., low steam line pressure, manual closure of MSIVs, high steam line flow). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS, ensures that the fuel peak. cladding temperature remains below the limits of 10 CFR 50.46. J MSIV closure signals are initiated from position switchei located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve - Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve - Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur. In addition, certain combinations of valves closed in two lines will result in a half-scram. The Main Steam Isolation Valve - Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient. Sixteen channels of the Main Steam Isolation Valve - Closure Function, with eight channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude the scram from this Function on a (continued) HATCH UNIT 2 B 3.3-15 REVISION 0
6 RPS Instrumentatien l B 3.3.1.1 BASES APPLICABLE 8. Turbine Sten Valve - Closure (continued) SAFETY ANALYSES, , LCO, and for the turbine trip event analyzed in Reference 2. For APPLICABILITY this event, the reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded. Turbine Stop Vaive - Closure signals are initiated from position switches located on each of the four TSVs. Two , independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram. In addition, certain combinations of two valves closed will result in a half-scram. This Function must be enabled at THERMAL POWER = 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. The Turbine Stop Valve - Closure Allowable Value is selected ) to be high enough to detect imminent TSV closure, thereby , reducing the severity of the subsequent pressure transient. Eight channels of Turbine Stop Valve - Closure Function, l with four channels in each trip system, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function if the TSVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is a 30% RTP. This Function is not required when T L POWER is < 30% RTP since the Reactor Vessel S Average Power Range Monit f p r .,ilp ressure
. tron Flux - High
- Highand the Functionsareadequatetomain[~ain e necessary safety margins.
- 9. Turbine Control Valve Fast Closure. Trio Oil Pressure - Low Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux (continued)
HATCH UNIT 2 B 3.3-18 REVISION 0
i RPS Instrumentation i B 3.3.I.I ! I { BASES l l l 9. APPLICABLE Turbine Control Valve Fast Closure. Trio 011
- SAFETY ANALYSES, Pressure - Low (continued) i LCO, and 3
1 APPLICABILITY transients that must be limited. Therefore, a reactor scram i
- is initiated on TCV fast closure in anticipation of the ;
- transients that would result from the closure of these 3 i valves. The Turbine Control Valve Fast Closure, Trip 011 '
i Pressure - Low Function is the primary scram signal for the i generator load rejection event analyzed in Reference 2. For j this event, the reactor scram reduces the amount of energy
- required to be absorbed and, along with the actions of the j EOC-RPT System, ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast Closure, Trip 011 Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure transmitter is associated with each control valve, and the signal from each transmitter is assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POWER a 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure. Four channels of Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Function with two channels in each trip system arranged in a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analysis assumptions, whenever THERMAL POWER is a 30% RTP. This , Function is not required when THERMAL POWER is < 30% RTP, since the Reactor Vessel Pressure - High and the Average Power Range Monit r tron Flux - High l Functions are adequate to aintain he necessary safety ] margins. '
- 10. Reactor Mode Switch - Shutdown Position 1 The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, to (continued)
HATCH UNIT 2 B 3.3-Ig REVISION 0
RPS Instrumentation B 3.3.1.1
-BASES APPLICABLE 11. Manual Scram (continued)
SAFETY ANALYSES, LCO, and There is no Allowable Value for this Function since the APPLICABILITY channels are mechanically actuated based solely on the position of the push buttons. Four channels of Manual Scram with two channels in each trip system arranged in a one-out-of-two logic are available and required to be OPERABLE in MODES I and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. ACTIONS A Note has been provided to modify the ACTIONS related to RPS instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel. A.1 and A.2 gg Because of the diver y of sensors available to provide trip signals and th cy of the RPS design, an allowable out of vlce time f 12 hours has been shown to be acceptable (Re to it restoration of any inoperable channel 0? BLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped (continued) HATCH UNIT 2 B 3.3-21 REVISION 0
RPS Instrumentation 4 B 3.3.1.1 . BASES i
- s ACTIONS A.] and A.2 (continued) a condition per Required Actions A.1 and A.2. Placing the 2 inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and 4-allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel '
i j in tri would result in a full scram), Condition D must be nt red a~ Required Action taken. ' l/)Ser} N'
.1 nd1i.
i Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip l system. In this condition, provided at least one channel i per trip system is OPERABLE, the RPS still maintains trip I capability for that Function, but cannot accommodate a t 2 single failure in either trip system. I Required Actions B.1 and B.2 limit the time the RPS scram l logic, for any Function, would not accommodate single I j failure in both trip systems (e.g., one-out-of-one and l one-out-of-one arrangement for a typi r channels andl3 Function). The reduced reliability f this i I arrangement was not evaluated in Ref rene or the 2 hour l Completion Time. Within the 6 hour associated Function will have all required channels OPERABLE i i or in trip (or any combination) in one trip system. an4 6 i Completin ese Required Actions restores RPS to a reliabi level e ivalent to that evaluated in Referen '9, which ustified a 12 hour allowable out of service esented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in). (continued) HATCH UNIT 2 B 3.3-22 REVISION 0 l
i. I i Insert 'H' - Bases 3.3.1.1 ACTIONS A.1 and A.2 , As noted, Action A.2 is not applicable for APRM Functions 2.a,2.b,2.c, and 2.d. Inoperability of one required APRM channel affects both trip systems; thus, Required Action A.1 must be satisfied. This is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. ; Inoperability of more than one required APRM channel of the same trip function results in l loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. , L i l 1 l
i i RPS Instrumentation B 3.3.1.1 l 'l { BASES ' i I ACTIONS B.1 and B.2 (continued) i i- If this action would result in a scram or RPT, it is ; j permissible to place the other trip system or its inoperable channels in trip. I d t a The'6 hour Completion Time is judged acceptable based on the : i i remairiing capability to trip, the diversity of the sensors availa' ole to provide the trip signals, the low probability , l' of extansive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event ' requiring the initiation of a scram. i Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip i system in trip would result in a scram or RPT), Condition 0 j i pe en and its Required Action taken. 1 \ In terf 'T * ] Required Action C.1 is intended to ensure that appropriate 3 ac'. ions are taken if multiple, inoperable, untripped d.amals within the same trip system for the same Function , l result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both i trip systems will generate a trip signal from the given Function on a valid signal. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. D.i.1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each (continued) HATCH UNIT 2 B 3.3-23 REVISION 0
_ . . _ . . _ . - _ . .______.-__.._.__.-__m___.. . _ _ _ . __ _._ .. l Insert 'I' - Bases 3.3.1.1 Actions B I and B.2 i As noted, Condition B is not applicable for APRM Functions 2.a,2.b, 2.c, and 2.d. l Inoperability of an APRM channel affects both trip systems and is not associated with a i specific trip system as are the APRM two-out-of-four voter and other non-APRM I channels for which Condition B applies. For an inoperable APRM channel, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one ; required APRM channel results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide 4 Required Actions that are appropriate for the inoperability of APRM Functions 2.a, 2.b, 2.c, and 2.d, and these Functions are not associated with specific trip systems as are the APRM two-out-of-four voter and other non-APRM channels, Condition B does not apply. l l i
,~ . . . - . - _ - __ - .- _ _ - - -. .- - - ... - - - ) i RPS Instrumentation I B 3.3.1.1 i l BASES ' SURVEILLANCE SR 3.3.1.1.2 REQUIREMENTS i (continued) To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor i power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, ! which could affect the APRM reading between performances of I SR 3.3.1.1.8. t A restriction to satisfying this SR when < 25% RTP is ' provided that requires the SR to be met only at 2: 25% RTP because it is difficult to accurately maintain APRM , indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2: 25% RTP, : the Surveillance is required to have been satisfactorily ' performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. e I s The
- v. A :r:;;
- t. . ._,_Z. .... P:;L...,.n:n--
,.,w. , ._ . . _ _ neait--
.c.,T. . _ . , . ,Z..._ . .,
rl ;Z..:_:._:.:_e
>=,1:t
,. Z.. ...,.
..n ...
7i___ Jsom E 1. . . . 64 o . .m /a TL V. en -_/..--_ ,ALsA X M I15"$ [ 3af 'ggri..'.' _
&_2J M _IIi __A 2Za' ""/r" '
- i n' 7 "n I.S 'E I'_ '". _ IZ '
- ':'l.
. .'". ,'."l
- y. [""...y "%. .'.:. M. . _
}y ::tp. , t::7 10; : .:1ip,""l,,,"
te prify7. .".M.y%..L fi: j.;::1 J. ip'. ,"'"g. 'T . '
..j ' , . . .
i
- tp- ther:f: th: ^
refE,.t:d sul e t: tF7 ' E rRM Fun 'E .ica tier::Itr:tel'
.Zhr:quir .a...Ta::.tp # ., ., Kasu::.:Ju.r.: ..._Z.-a. :f fic/ If
/
. L as /
.Z. ..' .g '. :'" E.. .. 6.lul' ', ,' .' Z"L. .'. ; ' _'21. .. .".r. .' .'. -" . ",N. .'I.
. "". ". "Z. "".' I. .'. ' 72. l.
- 21. .... . . J. 6 m.a
. m .A..1. L.a 2 n.. - - _ J. L .1. w, .
Th .r s;;.; ;f T e. 1 s .:
! .1.
. a... ..L.,..s...
. 7,' .2... a..u 1.. , , =u L.;..in: Eu ri g: '.7_:=t,
'.00tr : rtiOC'.
.....m ,. . . c.f o . j (continued)
HATCH UNIT 2 B 3.3-26 REVISION 0
RPS Instrumentation , B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.4 i REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the , intended function. Any setpoint adjustment shall be ! consistent with the assumptions of the current plant specific setpoint methodology. As noted, SR 3. .1.4 is not required to be performed when entering MOD fr 1, since testing of the MODE 2 required IRM ;(jA{lL,ctions F cannot be performed in MODE 1 witho (lizi g jumpers, lifted leads, or movable links. This all ry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. A frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9). SR 3.3.1.1.5 l A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the l intended function. A Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on the reliability analysis of Reference 9. (The Manual Scram Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.) 1 SR 3.3.1.1.6 and SR 3.3.1.1.7 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status. The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from (continued) HATCH UNIT 2 B 3.3-27 REVISION 0
I RPS Instrumentation j B 3.3.1.1 f BASES l 5 SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 REQUIREMENTS 1 (continued) A CHANNEL FUNCTIONAL TEST is performed on each required . channel to ensure that the entire channel will perform the
- intended function. Any setpoint adjustment shall be
- consistent with the assumptions of the current plant
- specific setpoint methodology. The 92 day Frequency of 4 SR 3.3.1.1.9 is based on the reliability analysis of
- Reference 9.
j The 18 month Frequency is based on the need to perform this
- Surveillance ander the conditions that apply during a plant
- outage and the potential for an unplanned transient .if the
! Surveillance were performed with the reactor at power.
- Operating experience has shown that these components usually pass,Ahe4urveillance when performed at the 18 month i equency lA50 b jvlove thh subbn +o follow sR 3.7././.ll
= 3.3.1.1.10 er.d. E (.3.1.1.13 A BRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint i methodology. For MSIV-Closure, SDV Water Level-High i (Float Sw M,d TSV-Closure Functions, this SR also include a physica17 spection and actuation of the switch .jo,gg oga Note 1 sta es ha neutron detectors are excluded M CHANNEL CALIBRATION because they are passive devid s, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 effec ull power hours .1.1) . A secondLPRM Note iscalibration against provided that re i the TIPS e J.-]f Ut SRs to be performed within 12 ho so" enter from MODE 1. Testing of the MODE be performed in MODE 1 withou ,,.T1I ut
$I jumpers, lifted Functions cannot leads or movable links. This No e allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.
(continued) HATCH UNIT 2 B 3.3-29 REVISION 0
Insert 'J' - Bases 3.3.1.1 Surveillance Requirements SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure 1 that the entire channel will perform the intended function. For the APRM Functions, this j test supplements the automatic self-test functions that operate continuously in the APRM l and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM i channels (including recirculation flow processing -- applicable to Function 2.b only), the j two-out-of-four voter channels, and the interface connections into the RPS trip systems l from the voter channels. Any setpoint adjustment shall be consistent with the assumptions ' of the current plant specific setpoint methodology. The 184 day Frequency of SR 3.3.1.1.10 is based on the reliability analysis of Reference 13. (Note: The actual voting logic of the two-out-of-four voter channels is tested as part of SR 3.3.1.1.15.) For Function 2.a, a Note is provided that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizingjumpers or lifted leads. This Note allows entiy into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Insert 'K' - Bases 3.3.1.1 Surveillance Requirements SR 3.3.1 1.13 For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated recirculation loop flow channel. ) i
- RPS Instrumentation B 3.3.1.1
- BASES SURVEILLANCE REQUIREMENTS
'F2dE1'6ddbR 3.3.1.1.13 (continued)
Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to 4 4 complete the SR. v vf e.- ~. . ., _ c e o 2 e i e in u ,-u .+v.,,,4+u 3 GX2?'T."7."; 2" ~1"2:::"A T~X 'W~":.'2;..C 7 i W ' : ZZ, . .i. . 'Z. . ' . ,'." 2.
, . . . . . . , Z,. ,...~' . !.~"Z, , I, ". .',J, , , J. . . .'.'", Z. . . ,'.I '/. .".7 The Frequency of SR 3.3.1.1.13 is based upon the assumption i
of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. A SR 3.3.1.1.11 This SR ensures that scrams initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is 2 30% RTP. This invokes i calibration of the bypass channels. Adequate matgins for the instrument setpoint methodologies are incorporated into i the actual setpoint. Because main turbine bypass flow can i affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine l bypass valves must remain closed during the calibration at , THERMAL POWER 2 30% RTP to ensure that the calibration is ! valid. I If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at a 30% RTP, either due to open main turbine bypass valve (s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure-Low Functions are enabled), this SR is met and the channel is considered OPERABLE. The Frequency of 18 months is based on engineering judgment and reliability of the components. 7 hvsEt7 SR. 3.3./.I./.f HTRE ., (continued) HATCH UNIT 2 B 3.3-3: REVISION O l
i i RPS instrumentation i B 3.3.1.1 i BASES 5 14 CK-SURVEILLANCE SP
"(dof
^"^ v5ed.
REQUIREMENTS (continued) --- a-)"- 'a---- "---- - *---""-'" 4 4Z.Zi' Z '.~ Z < "'7E.' Tl2 '.' ,"'. '.Z 'l Z ' T' '_'c X"..
- EI+ 0 hhdr: E55ihEp'r' rile.:
l[nlE,""2~ I c< +
"[~% <.
-LP0"7<
--- Z +f.
+ Z..[.
.c +the uf e...,
- 5h.A[N5
"."".".,.rien[rer, u
+.
flu [
/ -
. ZZ'i,.' A';X '. 427;Z"ZC',' AU. E.'Z"y/ . . . .
'CTX Gl' .'".:';7en '"' %% ";7.' + T' ~ L . i"'4"l' " 7
. . < /s.a T + u. cE, e . /-. . . u . L . . a e a 1 + . . . Z. .. + J. +
iG%i;'Z..f. J. "'4. 2.. ','+ "Z. , ".M. .'J.4. C..,' I L Z.. 'I '
.. I.'r','. Z'.I.. ..
- ne cf I d';;'th*kb"**d ~
1 ,,heydcring,jh:
. :::, . re, .":bily .f th: ;I';=:=:ik..t:<'=
- : r1 iT'
=I "/'
N /' SR 3.3.1.1.15 4 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods
!.LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8),
overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month ncy.
])9 In sert *L "
....s This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification 1 components are tested. The RPS RESPONSE TIME ac rD TT lnser+cepta q e criteria are included in Reference 10.
v Not allows neutron detectors to be excluded from RPS l SPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. (continued) HATCH UNIT 2 B 3.3-31 REVISION 4
. - - . _ . - - - - - . . - . . _ _ - - _ . .= - _. - _ _ . - - - -
I f I i
- Insert 'L' - Bases 3.3.1.1 Surveillance Requirements SR 3.3.1.1.15 l The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM
trip conditions at the two-out-of-four voter channel inputs to check all combinations of two tripped inputs to the two-out-of-four logic in the voter channels and APRM related redundant RPS relays.
, j Insert 'M' - Bases 3.3.1.1 Surveillance Requirements SR 3.3.1.1.16 i
j RPS RESPONSE TIME for APRM Two-out-of-Four Voter Function 2.e includes the ) output relays of the voter and the associated RPS relays and contactors. (The digital l portions of the APRM and two-out-of-four voter channels are excluded from RPS I RESPONSE TIME testing because self-testing and calibration check the time base of the i
, digital electronics.) Confirmation of the time base is adequate to assure required response j times are met. Neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time.
iM 9 1 i b i 4 d l l
RPS Instrumentation 8 3.3.1.1 , BASES SURVEILLANCE SR 3.3.1.1.16 (continued) REQUIREMENTS , RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS l Frequency to be determined based on four channels per trip system, in lieu of the eight channels specified in Table 3.3.1.1-1 for the Main Steam Line Isolation Valve-Closure - Function. This Frequency is based on the logic , interrelationships of the various channels required to produce an RPS scram signal. This Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time 7 degradation, but not channel failure, are infrequent : occurrences. REFERENCES 1. FSAR, Section 7.2.
- 2. FSAR, Chapter 15. ,
- 3. FSAR, Section 6.3.3. ,
- 4. FSAR, Supplement 5A.
i
- 5. FSAR, Section 15.1.12. l
- 6. NEDO-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
- 7. FSAR, Section 15.1.38.
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1,1980.
- 9. NEDO-30851-P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System,"
March 1988.
- 10. Technical Requirements Manual.
- 11. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993. .
l
- 12. NEDO-32291, " System Analyses for Elimination of lee Response Time Testing Requirements,"
an 94. In se r +uary'N W HATCH UNIT 2 B 3.3-32 REVISION 4
. . - . . - . - . . . ~ - . - . . - . . . - . - . . . _ . . - - . - - ....-.._. . .. . . -. .- . ._ - . _ - -
l l Insert 'N' - Bases 3.3.1.1 References
- 13. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range
- Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995.
i I l i.
l' j SRM Instrumentation B 3.3.1.2 i { BASES i APPLICABLE System (R ) Instrumentation"; IRM Neutron Flux - High and I SAFETY ANALYSET Average P er Range Monitor (APRM) Neutron Flux - Hig , ! l (continued) l (Setdown)Fonctions; and LCO 3.3.2.1, " Control Rod Bloc l Lation." 1 The SRMs have no safety function and are not assumed to
- function during any FSAR design basis accident or transient analysis. However, the SRMs provide the only on scale monitoring of neutron flux levels during startup and refueling. Therefore, they are being retained in Technical Specifications.
LCO During startup in MODE 2, three of the four SRM channels are required to be OPERABLE to monitor the reactor flux level prior to and during control rod withdrawal, subcritical multiplication and reactor criticality, and neutron flux level and reactor period until the flux level is sufficient to maintain the IRMs on Range 3 or above. All but one of the channels are required in order to provide a representation of the overall core response during those periods when reactivity changes are occurring throughout the Core. In MODES 3 and 4, with the reactor shut down, two SRM channels provide redundant monitoring of flux levels in the Core. In MODE 5, during a spiral offload or reload, an SRM outside the fueled region will no longer be required to be OPERABLE, since it is not capable of monitoring neutron flux in the fueled region of the core. Thus, CORE ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b), requirement that the bundles being spiral reloaded or spiral offloaded are all in a single fueled region containing at i least one OPERABLE SRM is met. Spiral reloading and ! offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence). In nonspiral routine operations, two SRMs are required to be OPERABLE to provide redundant monitoring of reactivity (continued) HATCH UNIT 2 B 3.3-34 REVISION 0
i 1 Centrol Rod Block Instrumentation 1 i B 3.3.2.1 : i 3 3.3 INSTRUMENTATION l B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod block monitor (RBM) provides protection for control rod withdrawal < error events. During low power operations, control rod blocks from the rod worth minimizer (RWM) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch - Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities. The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint t during control rod manipulations. It is assumed to function : to block further control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the q other channel inputs into the second RMCS rod block circuit. The RBM channel signal is generate avera ing a r range no nals a vario s core heights sur i n e control rod being withdrawn. of the k,- redvodant signalfromonefaveragepowerrang3 monitor (APRM)channe1V8
.+m n -s m. --s q cgs ,c rnony c g+
sup t.gies Mn 4t;;:r.aThis reference reference signal signalTrkhe is used RBM channeldO to determine e set oint low, i termedi te, o igh enabled. If t i the ow power range setpoint, the RBM is automatically bypassed. The RBM is also automatically bypassed if a peripheral control rod is selected (Ref.1). A rod block signal is also generated if an RBM Downscale trip or an Inoperable trip occurs. The Downscale trip will occur if the RBM channel signal I (continued) HATCH UNIT 2 B 3.3-42 REVISION 0
Insert 'O' - Bases 3.3.2.1 Background . . . . ,and a signal from another of the APRM channels supplies the reference signal to the second RBM channel. l l
Control Rod Block Instrumentation l B 3.3.2.1 i i i ! BASES . 1 ! i' i BACKGROUND decreases below the Downscale trip setpoint after the R8M l ! (continued) signal has been normalized. The Inoperable trip will occur j during the nulling (normalization) sequence, if: the RBM
- channel fails to null, too few LPRM u ' able, a 1 ei ) i rt ction swi is v l any pos tion ther than "Op ate." 7 3"
- ::l ;h Ti;;d -D:1 y
-t 't 't: n-t:11-9 :4;a .:1 45.:.p /n.:J:
- L:.:Au:.t
.i ,. L. 2- 1.s. 4d
., .J +Lt1s.t._L'Is .
l LAl"sT ".' 'Z '"'al'L'XT(. I 'C"'LL"' C"'i ':l': Z'"di: i ="- - 1 ! 7"- r-C, - . M"'.y"R.. . "'l.,l. .,", ",Z
-- . . '. ". /' ""' " ' ' ' ' "
The purpose of eR s to control rod patterns during i startup and shutdown, such that only specified control rod i sequences and relative positions are allowed over the i operating range from all control rods inserted to 10% RTP. l The sequences effectively limit the potential amount and
! rate of reactivity increase during a CRDA. Prescribed 4 control rod sequences are stored in the RWM, which will i initiate control rod withdrawal and insert blocks when the j actual sequence deviates beyond allowances from the stored
! sequence. The RWM determines the actual sequence based position indication for each control rod. The RWM also uses feedwater flow and steam flow signals to determine when the i reactor power is above the preset power level at which the l RWM is automatically bypassed (Ref. 2). The RWM is a single i channel system that provides input into both RMCS rod block i circuits. With the reactor mode switch in the shutdown position, a i control rod withdrawal block is applied to all control rods j to ensure that the shutdown condition.is maintained. This ! Function prevents inadvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 I when the reactor mode switch is required to be in the ! shutdown position. The reactor mode switch has two l i channels, each inputting into a separate RMCS rod block l ! circuit. A rod block in either RMCS circuit will provide a l control rod block to all control rods. i 1 i 1 l i (continued) HATCH UNIT 2 B 3.3-43 REVISION 0 i I _ ___ ..
Control Rod biock Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.1 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control System input. Any setpoint adjustment shal stent with the assumptions of the curren l lant spyJ ic setpoint ' methodology. The Freque cy of 4Fdrfs s based on reliability analyses (Re . E). ! N ' SR 3.3.2.1.2 and SR 3.3.2.1.3 - A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. , The CHANNEL FUNCTIONAL TEST for the RWM is performed by ' attempting to withdraw a control rod not in compliance with t the prescribed sequence and verifying a control rod block occurs. This test is performed as soon as possible after the applicable conditions are entered. As noted in the SRs, SR 3.3.2.1.2 is not required to be performed until I hour after any control rod is withdrawn at < 10% RTP in MODE 2, ! and SR 3.3.2.1.3 is not required to be performed until , I hour after THERMAL POWER is < 10% RTP in MODE 1. This allows entry into MODE 2 (and if entered during a shutdown, : concurrent power reduction to < 10% RTP) for SR 3.3.2.1.2 and THERMAL POWER reduction to < 10% RTP in MODE 1 for SR 3.3.2.1.3 to perform the required Surveillances if the 92 day Frequency is not met per SR 3.0.2. The 1 hour ' allowance is based on operating experience and in , consideration of providing a reasonable time in which to ; complete the SRs. The 92 day Frequencies are based on reliability analysis (Ref. 8). SR 3.3.2.1.4 I The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values automatically change are based on the APRM signal's input to each RBM channel. Below the minimum power setpoint, the RBM is automatically bypassed. These power Allowable Values (continued) i HATCH UNIT 2 8 3.3-50 REVISION O
l' Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.8 (continued) REQUIREMENTS OPERABLE following loading of sequence into RWM, since this i is when rod sequence input errors are possible. J l a l
; REFERENCES 1. FSAR, Section 7.6.2.2.5.
- 2. FSAR, Section 7.6.8.2.6.
- 3. NEDC-30474-P, " Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvements (ARTS) Program for Edwin I. Hatch Nuclear Plants,"
December 1983.
- 4. NEDE-24011-P-A-US, " General Electrical Standard Application for Reload Fuel," Supplement for United States, (revision specified in the COLR).
- 5. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),
" Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," BWROG-8644, August 15, 1986. l
- 6. NEDO-21231, " Banked Position Withdrawal Sequence,"
January 1977.
- 7. NRC SER, " Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," " General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.
- 8. NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation,"
October 1988.
- 9. GENE-770-06-1, " Bases for Changes To Surveillance Test Intervals And Allowed Out-0f-Service Times For Selected Instrumentation Technical Specifications,"
February 1991.
- 10. NRC No. 93-102, " Final Policy Statement on Technical if4c1 o provements," July 23, 1993.
Mserf 'P kd HATCH UNIT 2 B 3.3-53 REVISION 0
r Insert 'P' - Bases 3.3.2.1 References
- 11. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range ,
Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995. l h I t i 4 l l 1 l l j i
,l
x i EOC-RPT Instrumentation ! B 3.3.4.1 1 BASES f
- i l APPLICABLE Turbine Ston Valve - Closure
- SAFETY ANALYSES, l LCO, and Closure of the TSVs and a main turbine trip result in the APPLICABILITY loss of a heat sink and increases reactor pressure, neutron i (continued) flux, and heat flux that must be limited. Therefore, an RPT l l is initiated on a TSV - Closure signal before the TSVs are l completely closed in anticipation of the effects that would l l result from closure of these valves. EOC-RPT decreases l
- reactor power and aids the reactor scram in ensuring that '
j the MCPR SL is not exceeded during the worst case transient. l Closure of the TSVs is determined by measuring the position j of each valve. While there are two separate position ,
- switches associated with each stop valve, only the signal j j from one switch for each TSV is used, with each of the four l l channels being assigned to a separate trip channel. The !
- logic for the TSV - Closure Function is such that two or '
i more TSVs must be closed to produce an E0C-RPT. This
- Function must be enabled at THERMAL POWER k 30% RTP. This j is normally accomplished automatically by pressure i j transmitters sensing turbine first stage pressure; l
- therefore, opening of the turbine bypass valves may affect !
! this Function. Four channels of TSV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TSV - Closure Allowable Value is selected to detect imminent TSV closure. This protection is required, consistent with safety analysis assumptions, whenever THERMAL is a RTP. Below 30% RTP, the Reactor Vessel Ste Dome Pressur High and "$14eut on Flux the Average
- High Power Functions Range of the Monitor Reactor F otec(A RM) N Syst (RPS) are adequate to maintain the necessary margin to the PR Safety Limit. ;
Turbine Control Valve Fast Closure. Trio Oil Pressure - Low Fast closure of the TCVs during a generator load rejection , results in the loss of a heat sink that produces reactor I pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TCV Fast Closure, Trip 011 Pressure - Low in anticipation of the transients that would result from the closure of these (continued) HATCH UNIT 2 B 3.3-82 REVISION 1
1
- \
EOC-RPT Instrumentation , B 3.3.4.1 l 4 l BASES j APPLICABLE Turbine Control Valve Fast Closure. Trio Oil Pressure - Low j SAFETY ANALYSES, (continued) i LCO, and 3 APPLICABILITY valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded
- during the worst case transient.
- Fast closure of the TCVs is determined by measuring the electrohydraulic control fluid pressure at each control i valve. There is one pressure transmitter associated with
! each control valve, and the signal from each transmitter is
! assigned to a separate trip channel. The logic for the TCV i Fast Closure, Trip 011 Pressure - Low Function is such that two or more TCVs must be closed (pressure transmitter trips) to produce an EOC-RPT. This Function must be enabled at THERMAL POWER 2: 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TCV Fast Closure, Trip 011 Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure will preclude an E0C-RPT from this Function on a valid signal. The TCV Fast Closure, Trip 011 Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.
This protect s required consistent with the safety analysi ever ERMAL POWER is 1 30% RTP. Below 30% , th Vessel Steam Dome Pressure - High and PRM,y,ejeactodeutonFlux-HighFunctionsoftheRPSare the I adeq te to maint n the necessary margin to the MCPR Safety I Limit ! ACTIONS A Note has been provided to modify the ACTIONS related to EOC-RPT instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable EOC-RPT instrumentation channels provide (continued) HATCH UNIT 2 B 3.3-83 REVISION 1
i , Recirculation Loops OpGrating 2 B 3.4.1 ' i i BASES ~ l APPLICABLE case (since the intact loop starts at a lower flow rate and l SAFETY ANALYSES the core response is the same as if both loops were ; (continued) operating at a lower flow rate), a small mismatch has been j determined to be acceptable based on engineering judgement. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal-i margins during abnormal operational transients (Ref. 2),
. which are analyzed in Chapter 15 of .the FSAR. .
} A plant specific LOCA analysis has been performed assuming ! only one operating recirculation loop. This analysis has l demonstrated that, in the event of a LOCA caused by a pipe i i break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core l cooling, provided the APLHGR requirements are modified j accordingly (Ref. 3), $ The transient analyses of Chapter 15 of the FSAR have also j been performed for single recirculation loop operation j (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR l requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System-
- (RPS) average power range monitor (APRM) instrument 4
setpoints is also required to account for the different
- relationships between recirculation drive flow or j
i core flow. The APLHGR and MCPR setpoints fo spy operation are specified in the COLR. The AP ,,, _ i Simulated Thermal Power-High setpoint is in L 3. 1 .1, ! " Reactor Protection System (RPS) Instrumentation. 1 Recirculation loops operating satisfies Criterion 2 of the i NRC Policy Statement (Ref. 5). i l I LCO Two recirculation loops are normally required to be in l operation with their flows matched within the limits j specified in SR 3.4.1.1 to ensure that during a LOCA caused
- by a break of the piping of one recirculation loop the i assumptions of the LOCA analysis are satisfied. ,
- Alternately, with only one recirculation loop in operation, I
, modifications to the required APLHGR limits (LCO 3.2.1, i j " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR l l limits (LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"), j I
) (continued) 1 l
2 l HATCH UNIT 2 B 3.4-3 REVISION 0 4 , k l 1 _ - ~ ._
i i j Recirculation Loops Operating , B 3.4.1 l ] BASES 1 LCO and AP WWf imulated Thermal Power - High setpoint applied to allow continued operation i (continued) (LCO 3.3. . 4 consistent with the assumptions of Reference 3. In ! addition, core flow as a function of core thermal power must i be in the " Operation Allowed Region" of Figure 3.4.1-1 to j ensure core thermal-hydraulic oscillations do not occur. , i
- APPLICABILITY In MODES I and 2, requirements for operation of the Reactor
- Coolant Recirculation System are necessary since there is 4
considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. ACTIONS A.1 and B.1 l Due to thermal-hydraulic stability concerns, operation of the plant with one recirculation loop is controlled by restricting the core flow to 2: 45% of rated core flow when THERMAL POWER is greater than the 76% rod line. This l requirement is based on the recommendations contained in GE SIL-380, Revision 1 (Reference 4), which defines the region 1 where the limit cycle oscillations are more likely to occur. If the core flow as a function of core thermal power is in j the " Operation Not Allowed Region" of Figure 3.4.1-1, prompt l action should be initiated to restore the flow-power combination to within the Operation Allowed Region. The 2 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing core oscillations to be quickly detected. An immediate reactor scram is also required with no recirculation pumps in operation, since all forced circulation has been lost and the probability of thermal-hydraulic oscillations is greater. (continued) HATCH UNIT 2 B 3.4-4 REVISION 5
.)
. _ . _ _ . _ . . _ __ _. _ _ _ . ~ _ _ _ _ . _ . __ __ _ _ _ _._. . _ _ .. _ _ _
J j SDM Test - Rer'ueling B 3.10.8 i i BASES
! . APPLICABLE CRDA analyses assume that the reactor operator follows
! SAFETY ANALYSES prescribed withdrawal sequences. For SDM tests performed !' (continued) within these defined sequences, the analyses of References 1 and 2 are applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the
- safety analyses of References 1 and 2 may not be met.
4 Therefore, special CRDA analyses, performed in accordance f with an NRC approved methodology, may be required to 1 demonstrate the SDM test sequence will not result in ! unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection ! provided by the normally required MODE 5 applicable LCOs, in
. addition to the requirements of this LCO, will maintain i normal test operations as well as postulated accidents
! within the bounds of the appropriate safety analyses
] (Refs. 1 and 2). In addition to the added requirements for j the RWM, Average Power Range Monitor, and control rod l coupling, the notch out mode is specified for out of sequence withdrawals. Requiring the notch out mode limits ! withdrawal steps to a single notch, which limits inserted
! reactivity, and allows adequate monitoring of changes in j neutron flux, which may occur during the test. l As described in LCO 3.0.7, compliance with Special
! Operations LCOs is optional, and therefore, no criteria of ) the NRC Policy Statement apply. Special Operations LCOs ) provide flexibility to perform certain operations by j appropriately modifying requirements of other LCOs. A j discussion of the criteria satisfied for the other LCOs is j provided in their respective Bases.
l 4 ! LC0 As described in LCO 3.0.7, compliance with this Special { Operations LCO is optional. SDM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without
! meeting this Special Operations LCO or its ACTIONS. For SDM
- tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against
- potential reactivity excursions is available. To provide i additional scram protection beyond the UpHy repaired m are also reauired to 2.d IRMs, the Average Power Range Monito 3
, be OPERABLE (LC0 3.3.1.1, Functions . a *and 2.e) as though j the reactor were in MODE 2. Because htrn1A.,
will be withdrawn and the reactor will potentially become
! critical, the approved control rod withdrawal sequence must 4
- (continued) j HATCH UNIT 2 B 3.10-34 REVISION 0 i
---m. y -
y --w ----w r e- m- ""e
l i i SDM Test - Refueling i B 3.10.8 l l BASES (continued) l I SURVEILLANCE SR 3.10.8.1. SR 3.10.8 G SR 3.10.8.3 REQUIREMENTS 2.4, I l LCO 3.3.1.1, Functions .a/and .e, made applicable in this I j Special Operations LCO, uired to have their
- Surveillances met to establish that this Special Operations j LCO is being met. However, the control rod withdrawal
! sequences during the SDM tests may be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff (e.g., a qualified shift technical advisor or reactor ], engineer). As noted, either the applicable SRs for the RWM j (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3). This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations fron the specified sequence. These Surveillances provide adequate assurance that the specified test sequence is being followed. SR 3.10.8.4 Periodic verification of the administrative controls established by tuts LCO will ensure that the reactor is operated within the bounds of the safety analysis.. The 12 hour Frequency is intended to provide appropriate assurance that each operating shift is aware of and verifies compliance with these Special Operations LC0 requirements. SR 3.10.8.5 Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control . rod is withdrawn to the full-out notch position, or prior to I declaring the control rod OPERABLE after work on the control l rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability j that a control rod will become uncoupled when it is not ' being moved, as well as operating experience related to uncoupling events. (continued) HATCH UNIT 2 B 3.10-37 REVISION 0 i j
Enciosure 4B Edwin 1. Hatch Nuclear Plant Request to Revise Technical Specifications: I i i OPRhi Bases Changes and Associated bfarkups 1 l l l l 1 I l I l l l l I
. HL-5054 E4B-1
RPS Instrumentation B 3.3.1.1 ) ? i BASES - APPLICABLE Averaae Power Ranae Monitor (APRM) SAFETY ANALYSES, LCO, and The APRM channels provide the primary indication of neutron
- APPLICABILITY flux within the core and respond almost instantaneously to (continued) neutron flux increases. The APRM channels receive input j signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP. Each APRM also includes an Oscillation i Power Range Monitor (0PRM) Upscale Function which monitors
. small groups of LPRM signals to detect thermal-hydraulic j instabilities.
3 The APRM System is divided into 4 APRM channels and 4 two-out-of-four voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The APRM System is designed to allow one APRM channel, but no voter channels, to be bypassed. A trip from any one unbypassed APRM will result in a " half-trip" in all four voter channels, but no trip inputs to either RPS trip system. APRM trip Functions 2.a, 2.b, 2.c, and 2.d are voted independently of OPRM Upscale Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d trip from any two unbypassed APRM channels will result in a full-trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip logic channel (A1, A2, B1, and B2). Similarly, a function 2.f trip from any two unbypassed APRM channels will result in a full-trip from each of the four voter channels. Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the entire core, consistent with the design bases for APRM Functions 2.a, 2.b, and 2.c, at least 17 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be OPERABLE for each APRM channel. For OPRM Upscale Function 2.f, LPRMs are assigned to " cells" of three detectors. A minimum of three cells, each with a minimum of . two LPRMs, must be OPERABLE for OPRM Upscale Function 2.f to ! be OPERABLE. ! 4 (continued) HATCH UNIT 1 B 3.3-7 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Averaae Power Ranae Monitor Neutron Flux -- Hiah SAFETY ANALYSES, (Setdown) LCO, and APPLICABILITY For operation at low power (i.e., MODE 2), the Average Power (continued) Range Monitor Neutron Flux - High (Setdown) Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide a secondary scraa to the Intermediate Range Monitor Neutron Flux - High Function because of the relative setpoints. With the IRMs at Range 9 or 10, it is possible that the Average Power Range Monitor Neutron Flux - High (Setdown) Function will prov',de the primary trip signal for a corewide increase in power. No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux - High (Setdown) Function. However, this function indirectly ensures that before the reactor mode switch is placed in the run position, reactor power does not exceed 25% RTP (SL 2.1.1.1) when operating at low reactor pressure and low core flow. Therefore, it indirectly prevents fuel damage during significant reactivity increases with THERMAL POWER
< 25% RTP.
The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 25% RTP. The Average Power Range Monitor Neutron Flux - High 4 I (Setdown) Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for criticality exists. In MODE 1, the Average Power Range Monitor Neutron Flux - High Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events. 2.b. Averaae Power Ranae Monitor SimulateChermal Power - Hiah The Average Power Range Monitor Simulated Thermal Power - High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. (continued) HATCH UNIT 1 B 3.3-8 PROPOSED OPRM 7/31/96 l
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Averaae Power Ranae Monitor Simulated Thermal SAFETY ANALYSES, Power - Hiah (continued) LCO, and APPLICABILITY The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level is varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed control rod pattern) but is clamped at an upper limit that is always lower than the Average Power Range Monitor Neutron Flux - High Function Allowable Value. l The Average Power Range Monitor Simulated Thermal Power - High Function provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event) and protects the fuel cladding integrity by ensuring that the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit (SL) is not exceeded. During these events, the THERMAL POWER increase does not significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram before the high neutron flux scram. For rapid neutron flux increase events, the THERMAL l POWER lags the neutron flux and the Average Power Range l Monitor Neutron Flux - High Function will provide a scram signal before the Average Power Range Monitor Simulated i Thermal Power - High Function setpoint and associated time delay are exceeded. Each APRM channel uses one total drive flow signal representative of total core flow. The total drive flow signal is generated by the flow processing logic, which is part of the APRM channel. The flow is calculated by summing two flow transmitter signals, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel OPERABILITY requirements for this function. The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heating event. The time constant is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER. (continued) HATCH UNIT 1 B 3.3-9 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Averaae Power Ranae Monitor Simulated Thermal SAFETY ANALYSES, Power - Hiah (continued) LCO, and APPLICABILITY The Average Power Range Monitor Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity. 2.c. Averaae Power Ranae Monitor Neutron Flux - Hiah The Average Power Range Monitor Neutron Flux - High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux - High Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety / relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 7) takes credit for the Average Power Range Monitor Neutron Flux - High Function to terminate the CRDA. The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses. The Average Power Range Monitor Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux - High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High , (Setdown) Function conservatively bounds the assumed trip ! and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Neutron Flux - High Function is not required in MODE 2. l (continued) HATCH UNIT 1 B 3.3-10 PROPOSED OPRM 7/31/96 l
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.d. Averaae Power Ranae Monitor - Inoo SAFETY ANALYSES, LCO, and Three of the four APRM channels and all four voter channels APPLICABILITY are required to be OPERABLE for each of the APRM Functions. (continued) This function (Inop) provides assurance that the minimum number of APRM channels is OPERABLE. For any APRM channel, any time: 1) its mode switch is in any position other than " Operate," 2) an APRM module is unplugged, or 3) the automatic self-test system detects a critical fault with the APRM channel, an Inop trip signal is sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from all four voter channels to their associated trip system. This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis. There is no Allowable Value for this function. This Function is required to be OPERABLE in the MODES where the APRM Functions are required. 2.e. Two-out-of-Four Voter The Two-out-of-Four Voter Function provides the interface between the APRM Functions, including the OPRM Upscale Function, and the final RPS trip system logic. As such, it is required to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions. Therefore, the Two-out-of-Four Voter Function is required to be OPERABLE in MODES 1 and 2. , All four voter channels are required to be OPERABLE. Each voter channel also incudes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, an Inop trip is issued from that voter channel to the associated trip system. The Two-out-of-Four Voter Function votes APRM Functions 2.a, 2.b, 2.c, and 2.d independently of Function 2.f. The voter also includes separate outputs to the RPS for the two independently voted sets of Functions, each of which is (continued) HATCH UNIT 1 B 3.3-11 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.e. Two-out-of-Four Voter (continued) SAFETY ANALYSES, LCO, and redundant (four total inputs). Voter Function 2.e must be APPLICABILITY declared inoperable if any of its functionality is inoperable. However, due to the independent voting of APRM trips and the redundancy of outputs, there may be conditions where Voter Function 2.e is inoperable, but trip capability for one or more of the other APRM Functions through that voter is still maintained. This may be considered when determining the condition of other APRM Functions resulting from partial inoperability of Voter Function 2.e. There is no Allowable Value for this Function. 2.f Oscillation Power Ranae Monitor (0PRM) Uoscale a . The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the i fuel MCPR SL due to anticipated thermal-hydraulic power
- oscillations.
J
- References 13,14, and 15 describe three algorithms for I detecting thermal-hydraulic instability related neutron flux
, oscillations: the period based detection algorithm, the
! amplitude based algorithm, and the growth rate algorithm.
All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specifications purposes is based only on the perico based detection algorithm. The OPRM Upscale Function receives input signals from the LPRMs within the reactor core, which are combined into
" cells" for evaluation by the OPRM algorithms.
The OPRM Upscale Function is required to be OPERABLE when the plant is in MODE 1. Within the region of power-flow operation where anticipated events could lead to thermal-hydraulic instability and related neutron flux oscillations, the automatic trip is enabled when THERMAL POWER, as indicated by APRM Simulated Thermal Power, is 2 25% RTP and reactor core flow, as indicated by recirculation drive flow, is < 60% of rated flow. (continued) HATCH UNIT 1 B 3.3-12 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.f Oscillation Power Ranae Monitor (0PRM) Voscale SAFETY ANALYSES, (continued) LCO, and APPLICABILITY An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithm detects growing oscillatory changes in the neutron flux for one or more cells in that channel. Three of the four channels are required to be OPERABLE. Each channel is capable of detecting thermal-hydraulic instabilities by detecting the related neutron flux oscillations and issuing a trip signal before the MCPR SL is exceeded. There is no Allowable Value for this Function.
- 3. Reactor Vessel Steam Dome Pressure -- Hiah :
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of (continued) HATCH UNIT 1 B 3.3-12a PROPOSED OPRM 7/31/96
. - __ _ _ _ -~ ._ _ _ _ _ _ _ . _ _ _ _ _ _ _ _
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 3. Reactor Vessel Steam Dome Pressure - Hiah (continued) ! SAFETY ANALYSES, 4 LCO, and Reference 4, reactor scram (the analyses conservatively 4 APPLICABILITY assume scram on the Average Power Range Monitor Neutron Flux - High signal, not the Reactor Vessel Steam Dome ;
- Pressure - High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code j limits.
High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure - High Allowable Value is chosen to provide a sufficient margin to the ASME Section III Code limits during the event. Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ' ensure that no single instrument failure will preclude a l scram from this Function on a valid signal. The Function is ! required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists.
- 4. Rgactor Vessel Water level - Low. Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPV water level decrease too far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat generated in the fuel from fission. The Reactor Vessel Water Level - Low, Level 3 Function is assumed in the ,
analysis of the recirculation line break (Ref. 3). The i reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46. Reactor Vessel Water Level - Low, Level 3 signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variabie leg) in the vessel. (continued) HATCH UNIT 1 B 3.3-12b PROPOSED OPRM 7/31/96 l
RPS Instrumentation B 3.3.1.1 BASES ACTIONS Actions of the Condition continue to apply for each (continued) additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel. A.1 and A.2 Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours has been shown to be acceptable (Refs. 9, 12, and 16) to permit restoration of l any inoperable channel to OPERABLE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped 1 condition per Required Actions A.1 and A.2. Placing the l inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken. As noted, Required Action A.2 is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f. Inoperability of one l required APRM channel affects both trip systems; thus, Required Action A.1 must be satisfied. This is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability ; of more than one required APRM channel of the same trip ; function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. (continued) HATCH UNIT 1 B 3.3-20 PROPOSED OPRM 7/31/96
. . - - - . - - . - - - . - - - - - . _ ~ . - . - - . - - - -
E
.RPS Instrumentation B 3.3.1.1 BASES .
ACTIONS 8.1 and B.2 , (continued) Condition B exists when, for any one or more Functions, at : least one required channel is inoperable in each trip system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accommodate a . single failure in either trip system. ! Required Actions B.1 and 8.2 limit the time the RPS scram logic, for any Function, would not accommodate single failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel. Function). The reduced reliability of this logic arrangement was not evaluated in References 9, 12, and 16 l for the 12 hour Completion Time. Within the 6 hour ' allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system. Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in References 9,12, and 16 which justified a 12 hour allowable .l out of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip or, alternatively, all the inoperable channels in that- trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channe's are all in different Functions). The-decision of which trip system is in the more degraded state should be based on prudent judgment and take.into account current plant conditions (i.e., what MODE the plant is in). If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip. The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors , available to provide the trip signals, the low probability ; of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram. (continued) HATCH UNIT 1 B 3.3-21 PROPOSED OPRM 7/31/96
)
I RPS Instrumentation i B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip system in trip would result in a scram or RPT), Condition D must be entered and its Required Action taken. As noted, Condition B is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f. Inoperability of an APRM l channel affects both trip systems and is not associated with a specific trip system, as are the APRM two-out-of-four voter and other non-APRM channels for which Condition B applies. For an inoperable APRM channel, Raquired Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restora capability to accommodate a single failure. Inoperability of a Function in more than one required APRM channel resulty in loss of trip capability for that Function and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide Required Actions that are appropriate for the inopertbility of APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f, and these Functions l are not associated with specific trip systems as are the APRM two-out-of-four voter and other non-APRM channels, l Condition B does not apply. 1 l M l Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both . trip systems will generate a trip signal from the given l Function on a valid signal. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The I hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. (continued) HATCH UNIT 1 B 3.3-22 PROPOSED OPRM 7/31/96
I RPS Instrumentation B 3.3.1.1 BASES ACTIONS M (continued) Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.1. F.1. G.I. and J.1 l If the channel (s) is not restored to OPERABLE status or ; placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. The allowed Coinpletion Times are reasonable, based on operating experience, to reach the > specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Times of Required Actions E.1 and J.1 are l consistent with the Completion Time provided in LC0 3.2.2,
" MINIMUM CRITICAL POWER RATIO (MCPR)."
M If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. This is done by immediately initiating I action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. (continued) 4 HATCH UNIT 1 B 3.3-23 PROPOSED OPRM 7/31/96
RPS Instrumentation 4 B 3.3.1.1 BASES i i ACTIONS L1 l (continued) If OPRM Upscale trip capability is not maintained, , Condition I exists. Reference 12 justifies use of an !
- alternate method to detect and sup';ress oscillations for a limited period of time. The alternate method is '
procedurally established consistr.nt with the guidelines ! identified in Reference 17 requiring manual operator action to scram the plant if certain predefined events occur. The ! 12 hour Completion Time is based on engineering judgment to ' allow orderly transition to the alternate method while limiting the period of time during which no automatic or alternate detect and suppress trip capability is formally in l place. Based on the small probability of an instability event occurring, the 12 hour Completion Time is judged to be reasonable. Ll The alternate method to detect and suppress oscillations implemented in accordance with Required Action I.1 was evaluated based on use up to 120 days (Ref. 12). The evaluation, based on engineering judgment, concluded that the likelihood of an instability event that could not be adequately handled by the alternate method during this 120 day period is negligibly small. The 120 day period is intended to be an outside limit to allow for the case where design changes or extensive analysis may be required to , understand or correct some unanticipated characteristic of l the instability detection algorithm or equipment. This ! action is not intended to be, and was not evaluated as, a routine alternative to returning failed or inoperable j equipment to OPERABLE status. Correction of routine ; equipment failure or inoperability is expected to normally ' be accomplished within the Completion Times allowed for Required Actions for Conditions A and B. 4h* (continued) HATCH UNIT 1 B 3.3-24 PROPOSED OPRM 7/31/96 l i
- . - _ . . - - - - - - - - - ~. .- . - - .
RPS Instrumentation B 3.3.1.1 BASES (continued) 1 SURVEILLANCE As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.1.1-1. The Surveillances are modified by a Note to indicate that ; when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains RPS trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RPS will trip when necessary. SR 3.3.1.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, ; including indication and readability. If a channel is ! outside the criteria, it may be an indication that the instrument has drifted outside its limit. (continued) HATCH UNIT 1 B 3.3-25 PROPOSED OPRM 7/31/96 l
RPS Instrumentation l B 3.3.1.1 , I 4 BASES l 1 SURVEILLANCE SR 3.3.1.1.1 (continued) REQUIREMENTS i The Frequency is based upon operating experience that ' demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of - channels during normal operational use of the displays l associated with the channels required by the LCO. SR 3.3.1.1.2 4 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor
- power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of j SR 3.3.1.1.8.
I A restriction to satisfying this SR when < 25% RTP is i provided that requires the SR to be met only at a 25% RTP because it is difficult to accurately maintain APRM i indication of core THERMAL POWER consistent with a heat
- balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2 25% RTP, i the Surveillance is required to have been satisfactorily i performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met
. per SR 3.0.2. In this event, the SR must be performed . j within 12 hours after reaching or exceeding 25% RTP. Twelve ! ) hours is based on operating experience and in consideration of providing a reasonabie time in which to complete the SR. SR 3.3.1.1.3 (Not used.) SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. (continued) HATCH UNIT 1 B 3.3-26 PROPOSED OPRM 7/31/96 l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.4 (continued) REQUIREMENTS As noted, SR 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. A Frequency of 7 days provides an acceptable level of system average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9). SR 3.3.1.1.5 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on the reliability analysis of j Reference 9. (The Manual Scram Function's CHANNEL 1 FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram functions' Frequencies.) SR 3.3.1.1.6 and SR 3.3.1.1.7 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status. The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be I increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from the fully inserted position since indication is being transitioned from the SRMs to the IRMs. l (continued) j HATCH UNIT 1 B 3.3-27 PROPOSED OPRM 7/31/96 l
_ ~ . . _ . - - - . _ - _ _ . _ - - . . - . . . - . .. - .- -. . . RPS Instrumentation i B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.6 and SR 3.3.1.1.7 (continued) REQUIREMENTS . The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (by initiating i, a rod block) if adequate overlap is not maintained. Overlap ! between IRMs and APRMs exists when sufficient IRMs and APRMs d concurrently have onscale readings such that the transition between MODE I and MODE 2 can be made without either APRM 1 downscale rod block, or IRM upscale rod block. Overlap 4 between SRMs and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on range 1 before SRMs have reached the upscale rod block. 3 As noted, SR 3.3.1.1.7 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to . the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2). If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the '
- Surveillance should be determined and the appropriate channel (s) declared inoperable. Only those appropriate channels that are required in the current MODE or condition
- should be declared inoperable.
A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs. i
- SR 3.3.1.1.8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP)
System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 effective full power hours Frequency is based on
- ensuring the nodal power uncertainty is within the licensing basis analysis.
4 4 (continued) HATCH UNIT 1 B 3.3-28 PROPOSED OPRM 7/31/96 l
i
- RPS Instrumentation B 3.3.1.1
, BASES SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 REQUIREMENTS
- ! (continued) A CHANNEL FUNCTIONAL TEST is performed on each required I
channel to ensure that the entire channel will perform the , intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant . specific setpoint methodology. The 92 day Frequency of j
- SR 3.3.1.1.9 is based on the reliability analysis of ,
a Reference 9. l The 18 month Frequency is based on the need to perform this , 3 Surveillance under the conditions that apply during a plant i outage and the potential for an unplanned transient if the , Surveillance were performed with the reactor at power. 1 Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b only), the two-out-of-four voter channels, and the interface connections to the RPS trip systems from the voter channels. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint , methodology. The 184 day frequency of SR 3.3.1.1.10 is ' based on the reliability analysis of References 12 and 16. l (NOTE: The actual voting logic of the two-out-of-four voter channels is tested as part of SR 3.3.1.1.15.) For Function 2.a, a Note that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1 is provided. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. (continued) HATCH UNIT 1 B 3.3-29 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 1 REQUIREMENTS (continued) This SR ensures that scrams initiated from the Turbine Stop . Valve - Closure and Turbine Control Valve Fast Closure, Trip l Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is 2 30% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can i affect this setpoint nonconservatively (THERMAL POWER is I derived from turbine first stage pressure), the main turbine i bypass valves must remain closed during the calibration at i THERMAL POWER 2 30% RTP to ensure that the calibration is valid. , If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at 2 30% RTP, either due to open main turbine bypass valve (s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low functions are considered inoperable. Alternatively, the 1 bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low functions are enabled), this SR is met and the channel is considered OPERABLE. The Frequency of 184 days is based on engineering judgment and reliability of the components. l SR 3.3.1.1.13 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, this SR also includes a physical inspection and actuation of the switches. For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated recirculation loop flow channel. (continued) HATCH UNIT 1 B 3.3-30 PROPOSED OPRM 7/31/96 l
RPS Instrumentation B 3.3.1.1 BASES { i SURVEILLANCE SR 3.3.1.1.11 (continued) )' REQUIREMENTS Note I states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 effective full power hours LPRM calibration against the TIPS (SR 3.3.1.1.8). A second Note is provided that requires the IRM SRs to be t performed within 12 hours of entering MODE 2 from MODE 1. Testing of the MODE 2 IRM Functions cannot be performed in , MODE 1 without utilizing jumpers, lifted leads or movable l links. This Note allows entry into MODE 2 from MODE 1 if ' the associated Frequency is not met per SR 3.0.2. i Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. The Frequency of SR 3.3.1.1.13.is based upon the assumption ! of an 18 month calibration interval in the determination of 3 the magnitude of equipment drift in the setpoint analysis. ' SR 3.3.1.1.14 > (Not used.) SR 3.3.1.1.15 ; The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the , OPERABILITY of the required trip logic for a specific i channel. The functional testing of control rods (LC0 3.1.3), and SDV vent and drain valves (LC0 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. (continued) HATCH UNIT 1 B 3.3-31 PROPOSED OPRM 7/31/96 l i
RPS Instrumentation B 3.3.1.1 1 BASES l SURVEILLANCE SR 3.3.1.1.15 (continued) REQUIREMENTS , The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e I simulates APRM and OPRM trip conditions at the two-out-of- l l four voter channel inputs to check all combinations of two tripped inputs to the two-out-of-four logic in the voter channels and APRM related redundant RPS relays. SR 3.3.1.1.16 , l This SR ensures that the individual channel response times l are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME I acceptance criteria are included in Reference 10. 1 RPS RESPONSE TIME for APRM Two-out-of-Four Voter Function 2.e includes the output relays of the voter and the i associated RPS relays and contactors. (The digital portions ~ I of the APRM and two-out-of-four voter channels are excluded 1 from RPS RESPONSE TIME testing because self-testing and calibration check the time base of the digital electronics.) 1 Confirmation of the time base is adequate to assure required response times are met. Neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. This Frequency is consistent with the typical industry refueling cycle and is based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time ; degradation, but not channel failure, are infrequent l occurrences. ; l SR 3.3.1.1.17 This SR ensures that scrams initiated from 0PRM Upscale . Function 2.f will not be inadvertently bypassed when THERMAL POWER, as indicated by APRM Simulated Thermal Power, is 125% RTP and core flow, as indicated by recirculation drive flow, is < 60% rated core flow. This normally involves (continued) HATCH UNIT 1 B 3.3-32 PROPOSED OPRM 7/31/96
1 i RPS Instrumentation , B 3.3.1.1 j 1 BASES SURVEILLANCE SR 3.3.1.1.17 (continued) l REQUIREMENTS confirming the bypass setpoints. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual Surveillance ensures that the 1 OPRM Upscale Function is enabled (not bypassed) for the ) correct values of APRM Simulated Thermal Power and 1 recirculation drive flow. Other Surveillances ensure that the APRM Simulated Thermal Power and recirculation flow properly correlate with THERMAL POWER and core flow, respectively. If any bypass setpoint is nonconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power is 125% and recirculation drive flow is < 60% rated), the affected channel is considered inoperable for the OPRM Upscale Function. Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative condition (unbypass). If placed in the unbypass condition, this SR is met and the channel is considered OPERABLE. The 18 month Frequency is based on engineering judgment and i component reliability. , 1 REFERENCES 1. FSAR, Section 7.2.
- 2. FSAR, Chapter 14.
- 3. FSAR, Section 6.5.
- 4. FSAR, Appendix M.
- 5. FSAR, Section 14.3.3.
- 6. NED0-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
- 7. FSAR, Sections 14.4.2 and 14.5.5.
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1,1980.
(continued) HATCH UNIT 1 B 3.3-32a PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES REFERENCES 9. NED0-30851-P-A , " Technical Specification Improvement (continued) Analyses for BWR Reactor Protection System," March 1988.
- 10. Technical Requirements Manual.
- 11. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 12. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function," October 1995.
- 13. NED0-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
- 14. NED0-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"
November 1995.
- 15. NED0-32465-A, "BWR Owners' Group Long-Term Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," March 1996.
- 16. NED0-32410P, Supplement 1, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," May 1996.
- 17. Letter L.A. England (BWROG) to M.J. Virgilio, "BWR Owners, Group Guidelines for Stability Interim Corrective Action," June 6, 1994.
HATCH UNIT 1 B 3.3-32b PROPOSED OPRM 7/31/96
Recirculation Loops Operating B 3.4.1 BASES APPLICABLE case (since the intact loop starts at a lower flow rate and SAFETY ANALYSES the core response is the same as if both loops were (continued) operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal
- margins during abnormal operational transients (Ref. 2),
which are analyzed in Chapter 14 of the FSAR. A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3). The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Simulated , l Thermal Power - High setpoint is in LCO 3.3.1.1, " Reactor Protection System (RPS) Instrumentation." , l Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 4). l LCO Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1,
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"),
(continued) HATCH UNIT 1 B 3.4-3 PROPOSED OPRM 7/31/96
. - - - - - . - -- -- -- - --. . _- . _ . . = _ _ . ._
i Recirculation Loops Operating 8 3.4.1 ! BASES l LC0 and APRM Simulated Thermal Power - High setpoint I (continued) (LC0 3.3.1.1) must be applied to allow continued operation l consistent with the assumptions of Reference 3. l l l APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. ACTIONS L1 With the requirements of the LC0 not met, the recirculation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow l coastdown and resultant core response may not be bounded by I the LOCA analyses. Therefore, only a limited time is allowed to restore the inoperable loop to operating status. Alternatively, if the single loop requirements of the LC0 are applied to operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LC0 and the initial conditions of the accident sequence. The 24 hour Completion Time is based on the low probability of an accident occurring during this time period, on a i reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. l { l (continued) i HATCH UNIT 1 B 3.4-4 PROPOSED OPRM 7/31/96
i> circulation Loops Operating B 3.4.1 BASES ACTIONS M (continued) l This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump. M I , 1 With any Required Action and associated Completion Time of I Condition A not met, the plant must be brought to a MODE in l l which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of Design Basis Accidents and rrinimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.1.1 ; i REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements ; provide larger margins to the fuel cladding integrity Safety i Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop. (continued) HATCH UNIT 1 B 3.4-5 PROPOSED OPRM 7/31/96
R: circulation Lorps Operating , B 3.4.1 , I BASES SURVEILLANCE SR 3.4.1.1 (continued) REQUIREMENTS The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered not in operation. The SR is not required when both loops are not in operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner. SR 3.4.1.2 (Not used.) l 4 MW (continued) HATCH UNIT 1 B 3.4-6 PROPOSED OPRM 7/31/96
R: circulation Loops Operating B 3.4.1 BASES (continued) REFERENCES 1. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," , December 1986.
- 2. FSAR, Section 4.3.5.
- 3. NED0-24205, "E.I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation," August 1979.
- 4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
I e HATCH UNIT 1 B 3.4-7 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE Averaae Power Ranae Monitor (APRM) SAFETY ANALYSES, LCO, and The APRM channels provide the primary indication of neutron APPLICABILITY flux within the core and respond almost instantaneously to (continued) neutron flux increases. The APRM channels receive input signals from the local power range monitors (LPRMs) within the reactor core to provide an indication of the power distribution and local power changes. The APRM channels average these LPRM signals to provide a continuous indication of average reactor power from a few percent to greater than RTP. Each APRM also includes an Oscillation Power Range Monitor (OPRM) Upscale Function which monitors small groups of LPRM signals to detect thermal-hydraulic instabilities. The APRM System is divided into 4 APRM channels and 4 two-out-of-four voter channels. Each APRM channel provides inputs to each of the four voter channels. The four voter channels are divided into two groups of two each, with each group of two providing inputs to one RPS trip system. The APRM System is designed to allow one APRM channel, but no votor channels, to be bypassed. A trip from any one unbypassed APRM will result in a " half-trip" in all four voter channels, but no trip inputs to either RPS trip system. APRM trip Functions 2.a 2.b, 2.c, and 2.d are voted independently of OPRM Upscale Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d trip from any two unbypassed APRM channels will result in a full-trip in each of the four voter channels, which in turn results in two trip inputs into each RPS trip logic channel (A1, A2, B1, and B2). Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full-trip from each of the four voter channels. Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the entire core, consistent with the design bases for APRM Functions 2.a, 2.b, and 2.c, at least 17 LPRM inputs, with at least three LPRM inputs from each of the i four axial levels at which the LPRMs are located, must be i OPERABLE for each APRM channel. For OPRM Upscale ! Function 2.f, LPRMs are assigned to " cells" of three i detectors. A minimum of three cells, each with a minimum of l two LPRMs, must be OPERABLE for OPRM Upscale Function 2.f to 1 be OPERABLE. (continued) HATCH UNIT 2 B 3.3-7 PROPOSED OPRM 7/31/96 1
l RPS Instrumentation l B 3.3.1.1 BASES ; APPLICABLE 2.a. Averaae Power Ranae Monitor Neutron Flux - Hiah l SAFETY ANALYSES, (Setdown) LCO, and APPLICABILITY For operation at low power (i.e., MODE 2), the Average Power (continued) Range Monitor Neutron Flux - High (Setdown) Function is capable of generating a trip signal that prevents fuel damage resulting from abnormal operating transients in this power range. For most operation at low power levels, the : Average Power Range Monitor Neutron Flux - High (Setdown) : Function will provide a secondary scram to the Intermediate l Range Monitor Neutron Flux - High Function because of the relative setpoints. With the IRMs at Range 9 or 10, it is j possible that the Average Power Range Monitor Neutron ! Flux - High (Setdown) Function will provide the primary trip ; signal for a corewide increase in power. i l No specific safety analyses take direct credit for the ! Average Power Range Monitor Neutron Flux - High (Setdown) Function. However, this Function indirectly ensures that l before the reactor mode switch is placed in the run ' position, reactor power does not exceed 25% RTP (SL 2.1.1.1) when operating at low reactor pressure and low core flow. ; Therefore, it indirectly prevents fuel damage during ! significant reactivity increases with THERMAL POWER l
< 25% RTP. )
l The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 25% RTP. The Average Power Range Monitor Neutron Flux - High (Setdown) Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for criticality exists. In MODE 1, the Average Power Range Monitor Neutron Flux - High Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events. 2.b. Averaae Power Ranae Monitor Simulated Thermal Power - Hiah The Average Power Range Monitor Simulated Thermal Power - High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. (continued) HATCH UNIT 2 B 3.3-8 PROPOSED OPRM 7/31/96 l i
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Averaae Power Ranae Monitor Simulated Thermal SAFETY ANALYSES, Power - Hioh (continued) LCO, and APPLICABILITY The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level is varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed control rod 2 pattern) but is clamped at an upper limit that is always lower than the Average Power Range Monitor Neutron Flux - High Function Allowable Value. The Average Power Range Monitor Simulated Thermal Power - High Function provides protection against transients where THERMAL POWER increases slowly (such as the loss of feedwater heating event) and protects the fuel cladding integrity by ensuring that the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit (SL) is not exceeded. During these events, the THERMAL POWER increase does not significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram before the high neutron flux scram. For rapid neutron flux increase events, the THERMAL POWER lags the neutron flux and the Average Power Range Monitor Neutron Flux - High Function will provide a scram signal before the Average Power Range Monitor Simulated Thermal Power - High Function setpoint and associated time delay are exceeded. Each APRM channel uses one total drive flow signal representative of total core flow. The total drive flow signal is generated by the flow processing logic, which is part of the APRM channel. The flow is calculated by summing two flow transmitter signals, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel 0PERABILITY , requirements for this function. The clamped Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power - High Function for the mitigation of the loss of feedwater heating event. The time constant is based on the fuel heat transfer dynamics and provides a signal proportional to the THERMAL POWER. (continued) HATCH UNIT 2 B 3.3-9 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES I APPLICABLE 2.b. Averaae Power Ranae Monitor Simulated Thermal SAFETY ANALYSES, Power - Hioh (continued) l LCO, and APPLICABILITY The Average Power Range Monitor Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity. 2.c. Averaae Power Ranae Monitor Neutron Flux - Hiah The Average Power Range Monitor Neutron Flux - High Function is capable of generating a trip signal to prevent fuel damage or excessive RCS pressure. For the ; overpressurization protection analysis of Reference 4, the Average Power Range Monitor Neutron Flux - High Function is I assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety / relief valves l (S/RVs), limits the peak reactor pressure vessel (RPV) l pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 7) takes credit for the Average Power Range Monitor Neutron Flux - High Function to terminate the CRDA. The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses. The Average Power Range Monitor Neutron Flux - High Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux - High Function is assumed in the CRDA analysis, which is applicable in MODE 2, the Average Power Range Monitor Neutron Flux - High (Setdown) Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Range Monitor Neutron Flux - High Function is not required in MODE 2. (continued) HATCH UNIT 2 B 3.3-10 PROPOSED OPRM 7/31/96 l
RPS Instrumentation B 3.3.1.1 1' BASES APPLICABLE 2.d. Averaae Power Ranae Monitor -- Inoo SAFETY ANALYSES, l LCO, and Three of the four APRM channels and all four voter channels APPLICABILITY are required to be OPERABLE for each of the APRM Functions. 1 (continued) This Function (Inop) provides assurance that the minimum ) number of APRM channels is OPERABLE. For any APRM channel, any time: 1) its mode switch is.in any position other than " Operate," 2) an APRM module is
! unplugged, or 3) the automatic self-test system detects a critical fault with the APRM channel, an Inop trip signal is I
sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from ! all four voter channels to their associated trip system. ? This Function was not specifically credited in the accident analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis. 1 There is no Allowable Value for this Function. t This Function is required to be OPERABLE in the MODES where the APRM Functions are required. 7
- 2.e. Two-out-of-Four Voter j The Two-out-of-Four Voter Function provides the interface -
between the APRM Futetions, including the OPRM Upscale Function, and the fina. ilPS trip system logic. As such, it is required to be (MRAF.LE in the MODES where the APRM Functions are required and is necessary to support the ' i safety analysis applicable to each of those Functions.
- Therefore, the Two-out-of-Four Voter Function is required to be OPERABLE in MODES 1 and 2.
All four voter channels are required to be OPERABLE. Each voter channel also includes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, an Inop trip is issued from that voter channel to the associated trip system. The Two ou.-of-Four Voter Function votes APRM Functions 2.a. 2.b, 2.c, and 2.d independently of Function 2.f. The voter also includes separate outputs to the RPS for the two independently voted sets of Functions, each of which is (continued) HATCH UNIT 2 B 3.3-11 PROPOSED OPRM 7/31/96
RPS Instrumentation 1 B 3.3.1.1 BASES APPLICABLE 2.e. Two-out-of-Four Voter (continued) SAFETY ANALYSES, LCO, and redundant (four total inputs). Voter Function 2.e must be
- APPLICABILITY declared inoperable if any of its functionality is inoperable. However, due to the independent voting of APRM trips and the redundancy of outputs, there may be conditions where Voter Function 2.e is inoperable, but trip capability for one or more of the other APRM Functions through that voter is still maintained. This may be considered when j determining the condition of other APRM Functions resulting i from partial inoperability of Voter Function 2.e.
There is no Allowable Value for this Function. 2.f Oscillation Power Ranae Monitor (0PRM) Uoscale ! The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR SL due to anticipated thermal-hydraulic power oscillations. References 14, 15, and 16 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specifications purposes is based only on the period based detection algorithm. The OPRM Upscale Function receives input signals from the LPRMs within the reactor core, which are combined into
" cells" for evaluation by the OPRM algorithms.
The OPRM Upscale Function is. required to be OPERABLE when the plant is in MODE 1. Within the region of power-flow operation where anticipated events could. lead to thermal-hydraulic instability and related neutron flux oscillations, the automatic trip is enabled when THERMAL POWER, as indicated by APRM Simulated Thermal Power, is 125% RTP and reactor core flow, as indicated by recirculation drive flow, is < 60% of rated flow. (continued) HATCH UNIT 2 B 3.3-12 PROPOSED OPRM 7/31/96
I RPS Instrumentation B 3.3.1.1 BASES j APPLICABLE 2.f Oscillation Power Ranae Monitor (0PRM) Voscale SAFETY ANALYSES, (continued)
- LCO, and APPLICABILITY An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects ,
oscillatory changes in the neutron flux, indicated by the ! combined signals of the LPRM detectors in a cell, with ! period confirmations and relative cell amplitude exceeding specified setpoints. One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithm detects
, growing oscillatory changes in the neutron flux for one or j more cells in that channel.
Three of the four channels are required to be OPERABLE. Each channel is capable of detecting thermal-hydraulic instabilities by detecting the related neutron flux oscillations and issuing a trip signal before the MCPR 3L is exceeded. There is no Allowable Value for this Function.
- 3. Reactor Vessel Steam Dome Pressure - Hiah An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor ccolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 4, reactor scram (the analyses conservatively assume scram on the Average Power Range Monitor Neutron Flux - High signal, not the Reactor Vessel Steam Dome Pressure - High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits.
High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The (continued) HATCH UNIT 2 B 3.3-12a PROPOSED OPRM 7/31/96
RPS Instrumentation ! B 3.3.1.1 l l l
- BASES l l
APPLICABLE 3. Reactor Vessel Steam Dome Pressure - Hiah (continued) j SAFETY ANALYSES, LCO, and Reactor Vessel Steam Dome Pressure - High Allowable Value is i APPLICABILITY chosen to provide a sufficient margin to the ASME l l Section III Code limits during the event. : i Four channels of Reactor Vessel Steam Dome Pressure - High l Function, with two channels in each trip system arranged ina
, one-out-of-two logic, are required to be OPERABLE to ensure i that no single instrument failure will preclude a scram from l
< this Function on a valid signal. The Function is required ! !_ to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists. 3 4. Reactor Vessel Water level - Low. Level 3 Low RPV water level indicates the capability to cool the
- fuel may be threatened. Should RPV water level decrease too a far, fuel damage could result. Therefore, a reactor scram is initiated at Level 3 to substantially reduce the heat j generated in the fuel from fission. The Reactor Vessel
- Water Level - Low, Level 3 Function is assumed in the analysis of the recirculation line break (Ref. 3). The 1
reactor scram reduces the amount of energy required to be
- absorbed and, along with the actions of the Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46.
- Reactor Vessel Water Level - Low, Level 3 signals are i initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
j t d 4 (continued) HATCH UNIT 2 B 3.3-12b PROPOSED OPRM 7/31/96 l i 1
1 RPS Instrumentation I B 3.3.1.1 1 BASES ACTIONS Actions of the Condition continue to apply for each (continued) additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel. A.1 and A.2 Because of the diversity of sensors available to provide trip signals and the redundancy of the RPS design, an allowable out of service time of 12 hours has been shown to I be acceptable (Refs. 9,13, and 17) to permit restoration of l l any inoperable channel to OPERABLE status. However, this ) out of service time is only acceptable provided the associated Function's inoperable channel is in one trip , system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, i restore capability to accommodate a single failure, and ! allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel , in trip would result in a full scram), Condition D must be l entered and its Required Action taken. l As noted, Required Action A.2 is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f. Inoperability of one l required APRM channel affects both trip systems; thus, Required Action A.1 must be satisfied. This is the only action (other than restoring 0PERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. (continued) HATCH UNIT 2 B 3.3-20 PROPOSED OPRM 7/31/96 l
)
RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 , (continued) 1 Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip system. In this condition, provided at least one channel I per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accommodate a 1
, single failure in either trip system.
Required Actions B.1 and B.2 limit the time the RPS scram ) logic, for any Function, would not accommodate single ' failure in both trip systems (e.g., one-out-of-one and one-out-of-one arrangement for a typical four channel Function). The reducad reliability of this logic arrangement was not (valuated in References 9, 13, and 17 l for the 12 hour Comple+. ion Time. Within the 6 hour allowance, the associated Function will have all required channels OPERABLE or in trip (or any combination) in one trip system. Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in References 9,13, and 17 which justified a 12 hour allowable l out of service time as presented in Condition A. The trip system in the more degraded state should be placed in trip l or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function wiiile tne four inoperable channels are all in different Functions). The decision of whir.n trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in). If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable channels in trip. The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram. (continued) HATCH UNIT 2 B 3.3-21 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip l system in trip would result in a scram or RPT), Condition D - must be entered and its Required Action taken. As noted, Condition B is not applicable for APRM Functions 2.a, 2.b, 2.c, 2.d, and 2.f. Inoperability of an APRM l channel affects both trip systems and is not associated with a specific trip system, as are the APRM two-out-of-four voter and other non-APRM channels for which Condition B applies. For an inoperable APRM channel, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of a Function in more than one required APRM channel results in loss of trip capability for that Function and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide Required Actions that are appropriate for the inoperability of APRM Functions 2.a. 2.b, 2.c, 2.d, and 2.f, and these Functions l are not associated with specific trip systems as are the j APRM two-out-of-four voter and other non-APRM channels, i Condition B does not apply. ) I 1 bl i Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function l result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip . capability when sufficient channels are OPERABLE or in trip ! (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. The Completion Time is intended to allow the operator time s evaluate and repair any discovered inoperabilities. The
. hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channel s .
(continued) HATCH UNIT 2 B 3.3-22 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES ACTIONS M (continued) Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each time an inoperable channel has not met any Required Action of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.1. F.1. G.I. and J.1 l If the channel (s) is not restored to OPERABLE status or j placed in trip (or the associated trip system placed in ! trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. The allowed Completion Times are reasonable, based on operating experience, to reach the specified condition from full power conditions in an orderly manner and without challenging plant systems. In addition, the Completion Time of Required Actions E.1 and J.1 are l consistent with the Completion Time provided in LC0 3.2.2,
" MINIMUM CRITICAL POWER RATIO (MCPR)."
M If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be placed in a MODE or other specified condition in which the LC0 does not apply. This is done by immediately initiating action to fully insert all insertable control rods in core ; cells containing one or more fuel assemblies. Control rods i in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. (continued) , HATCH UNIT 2 B 3.3-23 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 l BASES I ACTIONS M (continued) If OPRM Upscale trip capability is not maintained, Condition I exists. Reference 13 justifies use of an ; alternate method to detect and suppress oscillations for a limited period of time. The alternate method is i procedurally established consistent with the guidelines ; identified in Reference 18 requiring manual operator action ' to scram the plant if certain predefined events occur. The 12 hour Completion Time is based on engineering judgment to : allow orderly transition to the alternate method while limiting the period of time during which no automatic or l alternate detect and suppress trip capability is formally in pl ace. Based on the small probability of an instability event occurring, the 12 hour Completion Time is judged to be reasonable. M The alternate method to detect and suppress oscillations implemented in accordance with Required Action I.1 was evaluated based on use up to 120 days (Ref. 13). The evaluation, based on engineering judgment, concluded that the likelihood of an instability event that could not be i adequately handled by the alternate method during this i 120 day period is negligibly small. The 120 day period is l intended to be an outside limit to allow for the case where l design changes or extensive analysis may be required to , understand or correct some unanticipated characteristic of ! the instability detection algorithm or equipment. This action is not intended to be, and was not evaluated as, a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment failure or inoperability is expected to normally be accomplished within the Completion Times allowed for Required Actions for Conditions A and B. (continued) l HATCH UNIT 2 B 3.3-24 PROPOSED OPRM 7/31/96
1 RPS Instrumentation B 3.3.1.1 BASES (continued) SURVEILLANCE As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS instrumentation Function are located in the SRs column of Table 3.3.1.1-1. The Surveillances are modified by a Note to indicate that when a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours, provided the associated Function maintains RPS trip capability. Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RPS will trip when necessary. SR 3.3.1.1.1 Performance of the CHANNEL CHECK once every 12 hours ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read i approximately the same value. Significant deviations between instrument channels could be an indication of excessive instrument drift in one of the channels or ! something even more serious. A CHANNEL CHECK will detect ; gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may b1 an indication that the , instrument has drifted outside its limit. (continued) HATCH UNIT 2 B 3.3-25 PROPOSED OPRM 7/31/96 l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.1 (continued) REQUIREMENTS The Frequency is based upon operating experience that l demonstrates channel failure is rare. The CHANNEL CHECK l supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. i SR 3.3.1.1.2 To ensure that the APRMs are accurately indicating the true core average power, the APRl's are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of l SR 3.3.1.1.8. l A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at 2 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2 25% RTP, I the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. SR 3.3.1.1.3 (Not used.) SR 3.3.1.1.4 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. (continued) HATCH UNIT 2 B 3.3-26 PROPOSED OPRM 7/31/96 l
l 4 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.4 (continued) REQUIREMENTS As noted, 4 3.3.1.1.4 is not required to be performed when entering MODE 2 from MODE 1, since testing of the MODE 2 required IRM Functions cannot be performed in MODE 1 without , utilizing jumpers, lifted leads, or movable links. This allows entry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. A Frequency of 7 days provides an acceptable level of system j average unavailability over the Frequency interval and is ' based on reliability analysis (Ref. 9). I J SR 3.3.1.1.5 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. A Frequency of 7 days provides an acceptable level of system average availability over the , Frequency and is based on the reliability analysis of i Reference 9. (The Manual Scram Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.) SR 3.3.1.1.6 and _SR 3.3.1.1.7 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status. The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from the fully inserted position since indication is being transitioned from the SRMs to the IRMs. (continued) HATCH UNIT 2 B 3.3-27 PROPOSED OPRM 7/31/96 l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.6 and SR 3.3.1.1.7 (continued) REQUIREMENTS The overlap between IRMs and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (by initiating a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APRMs concurrently have onscale readings such that the transition between MODE 1 and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap between SRMs and IRMs similarly exists when, prior to withdrawing the SRMs from the fully inserted position, IRMs are above mid-scale on range 1 before SRMs have reached the upscale rod block. As noted, SR 3.3.1.1.7 is only required to be met during entry into MODE 2 from MODE 1. That is, after the overlap requirement has been met and indication has transitioned to the IRMs, maintaining overlap is not required (APRMs may be reading downscale once in MODE 2). If overlap for a group of channels is not demonstrated (e.g., IRM/APRM overlap), the reason for the failure of the Surveillance should be determined and the appropriate . channel (s) declared inoperable. Only those appropriate l channels that are required in the current MODE or condition should be declared inoperable. A Frequency of 7 days is reasonable based on engineering i judgment and the reliability of the IRMs and APRMs. SR 3.3.1.1.8 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 effective full power hours Frequency is based on ensuring the nodal power uncertainty is within the licensing basis analysis. ! l (continued) HATCH UNIT 2 B 3.3-28 PROPOSED OPRM 7/31/96 l
1 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of Reference 9. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing - applicable to Function 2.b only), the two-out-of-four voter channels, and the interface connections to the RPS trip systems from the voter channels. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint , methodology. The 184 day Frequency of SR 3.1.1.1.10 is ! based on the reliability analysis of References 13 and 17. l (NOTE: The actual voting logic of the two-out-of-four voter channels is tested as part of SR 3.3.1.1.15.) For Function 2.a, a Note that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1 is provided. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from M0DE 1 if the associated frequency is not met per SR 3.0.2. (continued) HATCH UNIT 2 B 3.3-29 PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 REQUIREMENTS (continued) This SR ensures that scrams initiated from the Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is 2 30% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER 2 30% RTP to ensure that the calibration is valid. If any bypass channel's setpoint is nonconservative (i.e., the functions are bypassed at 2 30% RTP, either due to open main turbine bypass valve (s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nnnbypass). If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are enabled), this SR is met and the channel is considered OPERABLE. The Frequency of 18 months is based on engineering judgment and reliability of the components. SR 3.3.1.1.13 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive . calibrations, consistent with the plant specific setpoint j For MSIV - Closure, SDV Water Level - High methodology. (Float Switch), and TSV - Closure Functions, this SR also includes a physical inspection and actuation of the j switches. For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated recirculation loop flow channel. l (continued) HATCH UNIT 2 B 3.3-30 PROPOSED OPRM 7/31/96 l
RPS Instrumentation l B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.13 (continued) l REQUIREMENTS Note 1 states that neutron detectors are excluded from ! CHANNEL CALIBRATION because they are passive devices, with
- minimal drift, and because of the difficulty of simulating a i meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 effective full power hours LPRM calibration against the TIPS (SR 3.3.1.1.8). A second Note is provided that requires the IRM SRs to be performed within 12 hours of entering MODE 2 from MODE 1.
Testing of the MODE 2 IRM Functions cannot be performed in MODE 1 without utilizing jumpers, lifted leads or movable i links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. The Frequency of SR 3.3.1.1.13 is based upon the' assumption of an 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.1.1.14 (Not used.) SR 3.3.1.1.15 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the l OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LC0 3.1.3), and SDV vent and drain valves (LC0 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. (continued) HATCH UNIT 2 B 3.3-31 PROPOSED OPRM 7/31/96 l l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued)- REQUIREMENTS The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM and OPRM trip conditions at the two-out-of- l four voter channel inputs to check all combinations of two tripped inputs to the two-out-of-four logic in the voter channels and APRM related redundant RPS relays. SR 3.3.1.1.16 This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 10. RPS RESPONSE TIME for APRM two-out-of-four Voter Function 2.e includes the output relays of the voter and the associated RPS relays and contactors. (The digital portions of the APRM and two-out-of-four voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration check the time base of the digital electronics.) Confirmation of the time base is adequate to assure required . response times are met. Neutron detectors are excluded from i RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. Note 1 allows neutron detectors to be excluded from RPS i RESPONSE TIME testing because the principles of detector l operation virtually ensure an instantaneoas response time. i l Note 2 allows channel sensors for Reactor ressel Steam Dome Pressure - High and Reactor Vessel Water Lrvel - Low, Level 3 (Functions 3 and 4) to be excluded from RPS RESPONSE TIME testing. This allowance is supported by Reference 12 which concludes that any significant degradation of the channel sensor response time can be detected during the performance of other Technical Specifications SRs. RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS Frequency to be determined based on four channels per trip system, in lieu of the eight channels specified in Table 3.3.1.1-1 for the Main Steam Line Isolation I (continued) HATCH UNIT 2 B 3.3-32 PROPOSED OPRM 7/31/96
)
RPS Instrumentation l B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 (continued) l REQUIREMENTS J Valve - Closure Function. This Frequency is based on the logic interrelationships of the various channels required to produce an RPS scram signal. This Frequency is consistent , with the typical industry refueling cycle and is based upon i plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent ) occurrences. l 1 SR 3.3.1.1.17 I This SR ensures that scrams initiated from OPRM Upscale Function 2.f will not be inadvertently bypassed when THERMAL POWER, as indicated by APRM Simulated Thermal Power, is 2 25% RTP and core flow, as indicated by recirculation drive flow, is < 60% rated core flow. This normally involves confirming the bypass setpoints. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual Surveillance ensures that the OPRM Upscale Function is enabled (not bypassed) for the i correct values of APRM Simulated Thermal Power and I recirculation drive flow. Other Surveillances ensure that the APRM Simulated Thermal Power and recirculation flow properly correlate with THERMAL POWER and core flow, respectively. ' If any bypass setpoint is ncnconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power is 2 25% and recirculation drive flow is < 60% rated), then the affected channel is considered inoperable for the OPRM Upscale Function. Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative condition (unbypass). If placed in the unbypass condition, this SR is met and the channel is considered OPERABLE. The 18 month Frequency is based on engineering judgment and component reliability. REFERENCES 1. FSAR, Section 7.2.
- 2. FSAR, Chapter 15.
- 3. FSAR, Section 6.3.3.
(continued) HATCH UNIT 2 8 3.3-32a PROPOSED OPRM 7/31/96
RPS Instrumentation B 3.3.1.1 BASES REFERENCES 4. FSAR, Supplement 5A. (continued)
- 5. FSAR, Section 15.1.12.
- 6. NED0-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
- 7. FSAR, Section 15.1.38.
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1,1980.
- 9. NED0-30851-P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System,"
March 1988.
- 10. Technical Requirements Manual.
- 11. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
- 12. NED0-32291, " System Analyses for Elimination of Selected Response Time Testing Requirements,"
January 1994.
- 13. NEDC-32410P-A, " Nuclear tieasurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Trip Function," October 1995.
- 14. NED0-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
- 15. NED0-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology,"
November 1995.
- 16. NED0-32465-A, "BWR Owners' Group Long-Term Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," March 1996.
- 17. NED0-32410P, Supplement 1, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," May 1996.
- 18. Letter L.A. England (BWROG) to M.J. Virgilio, "BWR Owners, Group Guidelines for Stability Interim Corrective Action," June 6, 1994.
HATCH UNIT 2 B 3.3-32b PROPOSED OPRM 7/31/96 I
4 ,.
-L ,_ , - _ . _l a s La_$ ._-s . ia 4....,_ _ u_. .i- En m m Recirculation loops Operating B 3.4.1 BASES APPLICABLE case (since the intact loop starts at a lower flow rate and SAFETY ANALYSES the core response is the same as if both loops were (continued) operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement.
The recirculation system is also assumed to have sufficient 4 flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2),
- which are analyzed in Chapter 15 of the FSAR.
A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency
- Core Cooling System response will provide adequate core
- cooling, provided the APLHGR require.nents are modified accordingly (Ref. 3).
The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Simulated Thermal Power - High setpoint is in LC0 3.3.1.1, " Reactor Protection System (RPS) Instrumentation." ! Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement (Ref. 4). l LC0 Two recirculation loops are normally required to be in operation with their flows matched within the limits , specified in SR 3.4.1.1 to ensure that during a LOCA caused j by a break of the piping of one recirculation loop the ! assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1,
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR l limits (LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"), I (continued) i HATCH UNIT 2 B 3.4-3 PROPOSED OPRM 7/31/96
- 1 Recirculation Loops Operating B 3.4.1 BASES
- LC0 and APRM Simulated Thermal Power - High setpoint
- (continued) (LC0 3.3.1.1) must be applied to allow continued operation
; consistent with the assumptions of Reference 3.
APPLICABILITY In MODES 1 and 2, requirements for operatica of the Reactor 4 Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. 1 In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the j recirculation loops are not important. l ! ACTIONS Ad With the requirements of the LC0 not met, the recirculation loops must be restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the , mismatch between total jet pump flows of the two loops is i greater than required limits. The loop with the lower flow 1
- must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow ,
coastdown and resultant core response may not be bounded by ) the LOCA analyses. Therefore, only a limited time is allowed , to restore the inoperable loop to operating status. j Alternatively, if the single loop requirements of the
- LC0 are applied to operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LC0 and the initial conditions of the accident sequence.
The 24 hour Completion Time is based on the low probability of an accident occurring during this time period, on a y reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. (continued) HATCH UNIT 2 B 3.4-4 PROPOSED OPRM 7/31/96 l
i Recirculation Loops Operating B 3.4.1 i BASES ACTIONS M (continued) l
- This Required Action does not require tripping the i
recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is groter than the required limits. However, in cases where large . flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump. j M I i With any Required Action and associated Completion Time of
- Condition A not met, the plant must be brought to a MODE in l which the LCO does not apply. To achieve this status, the ll plant must be brought to MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be i operating because of the reduced severity of Design Basis j Accidents and minimal dependence on the recirculation loop a coastdown characteristics. The allowed Completion Time of j 12 hours is reasonable, based on operating experience, to 1
reach MODE 3 from full power conditions in an orderly manner
- and without challenging plant systems.
4 l
- SURVEILLANCE S_R 3.4.1.1 REQUIREMENTS t This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety i Limit such that the potential adverse effect of early i boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of
- rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop.
(continued) HATCH UNIT 2 B 3.4-5 PROPOSED OPRM 7/31/96
Recirculation Loops Operating
- B 3.4.1 1
4 BASES SURVEILLANCE SR 3.4.1.1 (continued)
- The mismatch is measured in terms of percent of rated core l fl ow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is considered not in operation. l
, The SR is not required when both loops are not in operation ^ since the mismatch limits are meaningless during single loop
- or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation.
2 The 24 hour Frequency is consistent with the Surveillance i Frequency for jet pump OPERABILITY verification and has been , shown by operating experience to be adequate to detect off I normal jet pump loop flows in a timely manner. SR 3.4.1.2 l (Not used.) 1 i l i i i l l 1 i i (continued) HATCH UNIT 2 B 3.4-6 PROPOSED OPRH 7/31/96
l , RPS Instrumentation ! B 3.3.1.1
]
l BASES (continued) l ! REFERENCES 1. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 l i SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," i December 1986.
- 2. FSAR, Section 5.5.1.4.
j 3. NED0-24205, "E.I. Hatch Nuclear Plant Units 1 and 2 l Single-Loop Operation," August 1979. I i 4. NRC No. 93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993. ] I i l 1 i HATCH UNIT 2 B 3.4-7 PROPOSED OPRM 7/31/96 l l
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE race h Ranae Monitor (APRM) SAFETY ANALYSES, /nSeN 90 % i LCO, and . VeranvPower Ranoe Monitor Neutron Flux - Hi (gerdou>n) APPLICABILITY (continued) Th: "".". 0h::::1: r:::iv: . ;;t :i; :1: ..;. th: .x:1 ;xx r;;;; :::it r: (L""":) within th: r:::tr er: t: pr;;id: = indi::ti = -f th: ; = = di:tributic d 10:21 ; r r ch::;::. The ^""" ch::::1: evers; thee: LPa" zi; 21 : t: pr vide : :=tte := indi::ti= cf .v:r:/ :=tr ;xx
'r ;7:t r th = "J K FEf'operatT '
M= : '-"(i . eg ' ' y, e Verage Power Range Monitor Neutron Flux Hi hv ufrc is capable of generating a trip signal that prev damage resulting from abnormal operating transients in range. For most operation at low power v g he Average Power Range Monitor Neutron Flux - H vFunct n will provide a secondary scram to the In ate Range Monitor Neutron Flux - High Function because of the relative With the IR"e at Range 9 or 10, it is possi t A"1 rage Pow-- Range Monitor Neutron Flux - y$q"th r ti n will provide the primary trip signal for a co increase in power. No specific safety analyses take direct credi or the - Average Power Range Monitor Neutron Flux - Hi However, this Function indirectly ensures that fFunction.(Setdo (e' reactor mode switch is placed in the run position, reactor power does not exceed 25% RTP (SL 2.1.1.1) when operating at low reactor pressure and low core flow. Therefore, it indirectly prevents fuel damage during significant reactivity increases with HERMAL POWER < 25% RTP. Tu noo , . . i..u,.a 4.+. +C. .... . D M , D
.cza..n=2; iZ= .:tc:x=.= r=2:=-
17 Z,7Zla +7,'ii,J:Z 2."'"4 Z" ' Z . .. ".'!It ' '= /" C,. :Z. R ". L. . .,*. ': 7 2.n. 'J Z
- a. J, .... J.s.2 L. .. . . . .
T..'7.1"."%.1"
'I i.a;.L
. . ..'.".."i.1
- s. .
!C.'s,2 l.C r; Z'
' W"'"KI+4
' "L. . .'L.f. ..".
e . . ./y .u.--. 6 -
,.E'.'Ji/ iz,;E 'r ;C n_Z Z'Z..:4': 2 + T' WoiF +/
- Z.:'h2; Z ;<ZZ 7,IiZ '17,;'Z'Hi'a 7, .'",7c:2 7 a..'.i. '. 'y' Z. . . '. I. , . entir
" ,Z. . ".' M. . i. i. ' Z.gi i.y.=, . . ., i. ,:7d. . .
ide'";_ t:<. Z::.
, Z" 3. age-of- .,,u Nou=0,
,. . . < . a . u.. Z', lent . . a + f .,L+s.
+ L """ y
' ~ '
/ -w-8R5'in[s'e= EA[h tee. of-th7four s OTd'Mvd:t e5h j thi I !
x - . (continued) HATCH UNIT 1 B 3.3-7 REVISION O
. _ . .- 1
4 l Insert 'C' - Bases B 3.3.1.1 Average Power Range Monitor (APRM) ! _ The APRM channels provide the primary indication of neutron flux within the core and respond almost instantaneously to neutron flux increases. The APRM channels receive , inptit signals from the local power range monitors (LPRMs) within the reactor core to
- provide an indication of the power distribution and local power changes. The APRM
- channels average these LPRM signals to provide a continuous indication of average
{ reactor power from a few percent to greater than RTP. The APRM System is divided into 4 APRM channels and 4 two-out-of-four voter channels Each APRM channel provides inputs to each of the four voter channels. The i four voter channels are divided into two groups of two each, with each group of two j providing inputs tc one RPS trip system. The APRM System is designed to allow one - , APRM channel, bu' io voter channels, to be bypassed A trip from any one unbypassed j APRM will result t. . half-trip" in all four voter channels, but no trip inputs to either i RPS trip system. A trip from any two unbypassed APRM channels will result in a full-l trip in each of the four voter channels, which in turn results in two trip inputs into each l RPS trip logic channel (A1, A2, B1, and B2). Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure that no single failure will preclude a scram on a valid signal. In addition, to provide adequate coverage of the i entire core, consistent with the design bases for APRM Functions 2.a,2.b, and 2.c, at , least 17 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, must be OPERABLE for each APRM char.nel. I
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Averaae Power Ranae Monitor Neutron Flux - Hi SAFETY ANALYSES, e own LCO, and (continued) l APPLICABILITY The Allowable Value is based on preventing significant increases in power when THERMAL POWER is < 25% . The Average Power Range Monitor Neutron Flux h
- nc H (S,eidon) must be OPERABLE during MODE 2 when control rods y be withdrawn since the potential for criticali ts. l In MODE 1, the Average Power Range Monit Flux - High Function provides protection gai st M utron I activity !
transients and the RWM and rod block moni protect against control rod withdrawal error events. l 1 l 2.b. Averaae Power Ranae Monito ""^^ imulated-Thermal Power - Hiah 4 The Average Power Range Monit OG@d imulated Thermal Power - High Function ! nito n flux to ! approximate the THERMAL POWER b ' , transferred to the reactor coolant. The APRM neutron flux is electronically ! filtered with a time constant representative of the fuel heat transfer dynamics to generate _ a sianal oronortiaa=1 tn i
\
f ggd the THERMAL POWER in the reactor.t The trip level is varied as a function of recirculation drive flow (i.e., at lower ! core flows, the setpoint is reduced proportional to the reduction in power experienced as core flow is reduced with a fixed n r i rod pattern) but is clamped at an upper it hatisllwayslowerthantheAveragePowerRange gMA M it fi'#'7eutron Flux Monit er Range - i g y owable Value. 5.- .. . S mulated ! Mhermal Power - High Function ovide .prot tion against transients where THERMAL POWER in ases s owly (such as the loss of feedwater heating event) and protects the fuel cladding integrity by ensuring that the MCPR SL is not exceeded. During these events, the THERMAL POWER increase ' does not significantly lag the neutron flux response and, because of a lower trip setpoint, will initiate a scram For rapid neutron flux l before the high increase events, neutron the THERMAL flux sc am.Y1 gh e neutron flux an the Average Power Ran 140n1 WEJ Function will prov' e a scrim'qr f ,44 Neu re on Flux - High the Average PowerRangeMonitr/}f,'Sfa;6y 7 imulated Thermal I i j (continued) HATCH UNIT 1 B 3.3-8 REVISION 0
Insert 'D' - Bases B 3.3.1.12.b. Average Power Range Monitor Simulated Thermal Power - High Changes to fuel design include an evaluation of the time constant to determine if the electronic filter requires replacement. i 1 s
+- - _ -
- - - . - . - -. - - - _ - - - - - .~ . . _- ,
l RPS instrumentation B 3.3.1.1 BASES 1 a
,m APPLICABLE 2.b.
SAFETY ANALYSES, Averaae Power Ranae Monito I M M 8 S'mulated ' ' LCO, and Thermal Power - Hiah (continued F APPLICABILITY
- Power - High Function setpoint and associated time delay are
, exceeded. m, rke no u e,m.,+www. w m 4, .n :,_A :_._ .... _ . . . . . , _ c 'w . . w. . . . ) , ._.: i., w.v i wm ..sww en. ,i wy. v w.L. -
.u...--
wi ww A n. sm n u, us _L.u.___, :. _.s_yw . m,. _ ._ ___L
.,_2_
i-.w. L. vv .
, sy , . , . _ ._.
ssisw. ou sum, Je Ja.,4-- J +. .1 w w.,s T. s i w ., . . w. w... ._ _L
- i. ww eyesww ww w .1. awn . . ___ v.sw wasutslis t
___1 _ __L A_2. 2 _% suunt be sy dJ d w wr88 6. 5
..,_A_ L.
ww wE k,m r...w. ,,,-J. A ,. . v ._s w-,As .n..nu.
.L.__
,,_ .-J
.i
+_2-m . u , ,, , s , , ,_ .
,s, . m.
u _
.s . ,
y . . . ,i
.6%... _ .._
ww., wu. w
+..L.w- .. v 4. .- .. sv , . . . . _ _ .. ._2_ r_.._
n...
.~ u .w ., ay. a wwa _L..._.._.,s,.
..su _
w c. A....,-
.w. . w a _.
n.___ n2 __J eJ_
. mas 3% isw u ._. ;. 6. w_. _sr s1 wy _ . ,. usuosm 1 n_. -
_L w s . ..u 1. u.www + J T L. ii w_ a s.i s v .m . U. . :. 3 , .n. 2 i 4ssL. 6.... _L.___1_ ;_ m._L _ 4n =% a m. ,=. asi+ _ at + .- 1._ . --. wm m.u.ss,ss. s as s ums t 6, . J.r ,u.em_ w w s.. .-u .s u.,_
.=__J wi ww . we w .w w ---.2__J A. L- AMPn A nt P A-swysw us w wyv s a su kw uw vu 6 3 V 1M b b bV n-,..-m +k+
w.i..i w wieww v wm , 4. ... , 1,w.
- m. _.,6,..___s,
,as. f.J1. - ..:
vu ,1,1 --
, asm s suaavs%
- c. r. e.s=.. Fw. nm..i 4-i w . , r s _- .1. ._. J - u-
. . w +. L. 4. e E. n.i .w. - +w . . . J +n
---..;J- v s u vw 1, J, w . 4s , , , =,1 i f
.. m. .AA4+4.a =
r.wu suw -J_....+- _v.i wei , 6 _ _ . . . . .__ _f L yuu w 2 _ m,m_
-ww A, s . s. w s . A, .., w, uw .4 n u3s w, ssw 1wwe4 .=
- s. i. i nou u. . r- ....
. . . . . . .--.4..__J . ,. _.w. -_f .u Anou. ,.u._ ..,.. m -4. . +. 6,+..
_ i. wu m o , +. eum w...i . .. wnw t n.nu m us Ji _ssru . 6,. fs _ w_ m_s
, w, AesL, ._ f, w_us . . m un. .uJ 1 1.u 1, 4
m.u n _ L . L. a n
- h. . . w . st+1-drfve
.... us
- u. w c . . us Each APRLc annelv.yM.w.o_.9
.nu www r
w .1 6
=nw
-s4 TXre se tative of o al l
-totahdrhe ' ' al generated by h m n at. _gg j '
..-4+,
wi i n v w.,
+m-wnw w=
c . .u _ u . . _ _ , . . .:___,. ._ .L. . :. ,m,+-_ A Annu, +L, siewas uyyay a a y s s u a ., ww winw wi iy ./.=w" '
- ru su ew y .res .L i21 .6L s.._ _.,__1.. _;__
e aw w e sw w wasu s s uv v
- 1. 6, 66. 4 co,+a- D Annu. awyysJ J u yesu a a ww wiew w c.,L .
-y..w"
,..J_-
- .. a.=wn c1....
i w as was 2,.w ., 4.yi - _v u.1, 4, i . ye w .u.4A.A --- ku g 1
. . - 4L. f1 . . .2__.1 f_- j 4
. unan i s s y uy waaw i awv a e yesu s a e a wm wasw AL. 6...
wnw e wwav. 4 , n w====<v41 , + 4 a c "" l
- l. a. n.r.n e . T. a. nk+=4- _-,+ ....--.. 64um --tama-aa i
.. e +..k.w a w.=
+u +-+., e,.. .:...,.
_is.w. vuwsv. . w ws wiew w e. d.,...
. a -
- 1. ., . i
' wirw wwmus s awn f--_ au_ a... f,... .._:s- i.....a.s-J ; 4 syisusa a i um b a ns b uvu 6 s v us r.s s i s b s gua.wwauwww s ,+ } _ u 4. +. h - + .. w 4.r nj .co..a- =n u. J-ww - J,L_J L _.. \ .- -- 6 J ea - 1
.%..+ 4n- a4mm.4+
wi www uwwwsj we w i vuwww ww = sw-'a. n w w w w i wie
... .J.6.J ..?EL -..L Annu E.,6 Annute
%9.m+ d n .sk wa= wuew uwwww: wwww is i w a s s uben ni av i s wwwsu ru vu' 1
d dwmA$+ ,-Iha4. 3 _... _[ +Lm u u w w e wie wsa wuuw .ws%wwa 6w .ksw- v vu s s wa
+..sA n1.5 A.- +
wasw w vv w i 1 wn _P4n-9 P II P A +bn
=i. **
.vges= = - .MM we =.= 9.P mi w wws, m un m . m, &md.- ss m.famamma [nm +b94 y aw ws ..sww ivs wii==
- yu.n = =w + 4.wuius e a s 1 * =to Annu E..L -__..J-_J A.. . . n um- n.--. u.-Je sua. .
- E1 - . . n2-..J P2-.1.a
- w. u b u s a s qu s s su muss uys a w rv u s swussyw s 8v i e w e a s w vs wausww wamuaubww J rL_ _ _1 n_.._._ u 2 ,,, L ,. L . - - o 1 m1u iss a suu u s v uv s e aaiyes wasuressw wan.4wn, == wi' iJ i w ig - e i ww -i.
4.r..+
' == i wns t. .- wiiw
.. tinPn A nt r fi vs uswiwa a m.
s s w ww
..-Je, uss a w .e4-en s isw w ek-we'*
ekg---1 -_mf jam _4 n.-- . A 4 t---r 4. A n As DDu. e l. w i s -i u w .n 21,1 i i yw. s wi m
- eLa me w 4
ses e.nwnwww fan-e+4^= i-ow=iwm ue+L a-lu -- n iw. wie ej w..aw finEn wi usw A.D1 ww a svn E fi m. .. J A u, is% 2..e. L s ,twnwww
- m. _ . u n ,e , 4- n=Anm
&n s eiy u -- ===
s4-+=4.. l- e. ww "=s sw. 4 .y i w f. w.isu,
. 4 1. w wi. . . m i, wws ,. .
, ww..w. 4kn.A Am.
iw = sknun ----- (n= E. -i
+..k.wa i
==sw= ,+4.n., wi.
.+ 1 ..+ -_.
ww iwuew w ais __J A.. _ _ _ . .s wgu s a su nvwa wys a wwwa '5**'3w
... n-..-- n.---
u.n . >4=ws+ m = El wi i i wn now=wwwu Di..aJ e4-ol.+-J wsmusuwww TL -- 1 s sss e anu a nm. a w vs u
. Li t a k k.---1 4- ,L . i i gis w e'u s sa ru a
. &it- .o,e. -o,+ .-.L1- t sie wwwm we ay vjemwm nu,6 L, wi wwyuu w we - .4.us+w%4r ivn4i n..n enu s e.
A nt P fl .. .._26 .:__.1 g -. MVI [en.nubb .s a a v v1 was s b a s yssu s sus bass sub
- 2. L_ _..
"M
. (awe41ame *we w af v'
un'
...64- .:~.:+ - . r1 ..
u uw w s was wei wus=, wa u a s w rvus.._aw, 6 :. iis A N $w,n,4.+,A ba wwaw=ww
+=4.
wr sy i c.,u .e.. + n.r m. /n.a., y 4, f .a (1. . a _.n . . 4, 6w s 2.
. u. .s J,wpew A*e r -s=sw a af wi
+ka v.sw +w "n w a
i ___.4__J Am ..-- n_..-_ n.-__ u__. s wyuei ww gw a w vu s - A c _..i.. J ri,wi a s us ry w a #was s'h D e.mU I _wwu w a n'u a as 5 w u (continued) HATCH UNIT 1 B 3.3-9 REVISION 0 J
Insert 'E' - Bases B 3.3.1.12.b. Average Power Range Monitor Simulated Thermal Power-High (continued)
.. . the flow processing logic, which is part of the APRM channel. The flow is calculated by summing two flow transmitter signals, one from each of the two recirculation loop flows. The flow processing logic OPERABILITY is part of the APRM channel OPERABILITY requirements for this Function.
)
l l l l
RPS Instrumentation B 3.3.1.1 BASES mm APPLICABLE 2.b. Averaae Power Ranae Monit I./ # d imulated SA ETY NALYSES, Thermal Pnw - Hinh u APPLICABILITY T'--d- "'A ^ " - d '# #- l #'l" d'l 2 "--
.Z.
.. "Z Z. . ".'X. .2.. , L'
. . X ..',Z.. ".lY. .,'./, . " ' "* ' ~ r 'r ' ' ' " "P" ~
A The c amped Allowable Va ue is based on a T! e credit for the Average Power Range Monito P-.y./,l-d Simulated Thermal Power - High Function f r the miti on of the loss of feedwater heating event, me constant is based on the fuel heat transfer dynamics and provides a signal proportional to the THER . R. The Average Power Range Monit 1 Thermal Power - High Function ['[./0.'[Smulated s re
~
e OPERABLE 1:1 MODE 1 when there is the possibi y of generating excessive THERMAL POWER and potentially exceeding the SL applicable to high pressure and core flow conditions (MCPR SL). During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.' 2.c. Averaae Power Ranae Manit r // M Ne tron Flux - Hiah 3, r,u::yuma. u .... , ./. ...a a. w. . aW.a ...a.. .(m. .s a .a.. .
'ciZ '.'IA I7EZ.EZ.J- " Z S '.T Z ..I Z Z 2 2 .T C
-(: W d U d :' 2~
t on
- 50 [ ' fa M M U Function is capable of generating signal to prevent fuel damage or ex 'v CS pressure.
For the overpressurization prot ct of ana sis of Reference 4, the Average Power Range Moni or,}.7./;f Ne tron Flux - High Function is assumed to termin he steam isolation valve (MSIV) closure event and, along with the safety / relief valves (S/RVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limi The control rod drop accident (CRDA) analysis JFf.7)takescreditfortheAveragePowerRangeMonitor ff,7jf/Ne tron F un-d- iigh FunQion to terminate the CRDA.
,. ~
ar x.x:rz .... ..., zuz L a ..- -.
- xr xzx.:"rmtr,-
7".'X'"IZZxL'ZC';ZL:'L-".'X: TD. '2' + L ) MI.Z7"'Z'J '7Eou"X'" 1'"Z . '" !Z" " 7.7 Z~ n.Z ;tz;ZZ.7a ZI . ~X'"'""'G 'Ai!.E II'""d'"/ _z. . .. x rxr
- x. .. x. _ _.z. . ...... . .. 2. . .. :m. . r.a. .. n. . z. .u. . .7. r " r. x.. . n
~ n A (continued)
HATCH UNIT 1 B 3.3-10 REVISION O
1 RPS Instrumentation 4 s B 3.3.1.1 4 i ! BASES ! ~m APPLICABLE 2.c. Averaae Power Ranae Monit I'.ftd utron Flux - Hiah i SAFETY ANALYSES, LCO, and g"(continued) N t.I , ch =:'L:.in =h t-ip :yL:o;r. (Q sA 1 l APPLICABILITY
.Z4 - . l.2.__21..
X: M'; ')..'_1'C l'. !L '" ~7 "' :"rzJ .,'Z.' ;"7' L.
-t "r ."-
o.:L. e:n;T..in. .L.=. L J .t :/. - N
'.'U.
! 4 l. "'
- i . .. l. !. l ,C...' ..
" :"Tl' .
p ..',,Z. ,"i,.:.2 'L 7 ' ; _ "E' ' - Z _ ' :Z' -~ lZ- 'T ~
/ 2d y'; :t ij 7r =r:7.;: "7 t.5:juu rl:'"":' 7"n" f
' ' r='t 1' " 7 "'T .J 1
7 ith77 t1 W E ;"t " u
'E cir:7
. nnol_ ,_22 $=. _27,1._._....
for#.:h^g""d.nn:b,"w-(hb:t4:n th:7u,r:di:1 s th :If 7 ,5 .: The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses. The Average Power Range Moni r #' U2'Ne tron Flux - High Function is required to be OPE BE ODE 1 where the potential consequences of the a d transients co result in the SLs (e.g., MCPR and RCS pressure) (ei g exceeded. Although the Average Range Monitor Neutron Flux - High Function ssum - in the Cl analysis, which is applicabl i dgpE A the Average e Range Monitor Neutron Flux - hvFtfn"c't'i n conservatively bounds the assumedJr4g A d, tog ith the assumed IRM trips, provides adeq at r ection. Therefore, the Average Power Range Monit6r tron Flux - High Function is not required in D . 2.e. = t= e = ,= m a = a = = s I
- z. .:i:1 -x a ..7:: z.:t t. t ,,i #.;.;;.;.'r. _::.tr ne. .n T
e
/r . . .:::
/ 5."..
r 7I i..'.'. y.Z.. .:'.!:
. . . ,.". .,,r, y . n. . . ,'n. . R. . ..,n,. ..s;,
. . . , . r. ._ I. :. T. ". ; ~
l e e. . ,I Lf 4 4.L. 4 . .. J_ _ . . . j d^y.n
. . a . L.
/J. ...
- L. . ....A n n.
7 :. . L..i;n:'j...... /_ ,
" .i ter "d;: tron [:' y ...,,,, :in d h}7..- ht:j
..at wi .. =[]=
t Pfen=[1 = >
'A gh .
Ir. p i n:l,=[;tu: j j, j
/ e;:Z.
+. u.. g=,./4hu
. . -. u.. ... .,J.. ...., . , . . .. ,1.
r.. u u... ... ,
=,/ =s.,
=__Z xp =_a
. , . . . . . . . . au,f .,L. ..y= w =_. ,
..j 1 ._. u,e/ = :
. J.. ..J. nwJ ..o 4.
.. .J 2. m _/ . a. a by .s 2.m .L.
kus Wr e. .a _ ._L..2-s q. e s D Lm
.y A. L .
- sw
.sw u n r*
; bu n .
u!.0"Od'.:=in , 1= .=; m th = p. su:u uu e w
.= e 2 :=u pu =, p .i, p t= j ,u .y
.re=g,=e=uv p: m e g L
u:,d!=r,'[ =/:. up,MP"2139
=.=
K p= 4:.= tent 1==ht: ,tri,[,m t h yp='
- =f/ d ..
(continued) 7 l HATCH UNIT 1 B 3.3-11 REVISION 0
4 5 i ! RPS Instrumentation i i J B 3.3.1.1 1 i BASES
- V v sf-
! APPLICABLE 2.d.,y;;,#r:- .j m r "h::: ".:r.j.ter SAFETY ANALYSES C= :cale -{c:rth"..P i LCO and = f'n"/---- -" + / ' c-"/ "-- /- -' ^"/--- " ./-a---- l APPLICABILITY U CI'57' UE'T'I'EA7- ~"Z"75II 'ZE T T"+"2" C" ' 2".a 4 Z~;"" T...T !" IC I"'iT Z. '".ZI".I l'A'. " T - 1 In' u,. rz.. E .4 ;
- w. .w G ,I.". 4. Z.,Y.,T. ,. '."4. . . , T. . ..;T'. ,,.. H.2J. . '_al ._
Sn. dun .'. + kZ 4 e7.J 2..J.4. 7.. e ed. .ma /Caa.e+ r .. . uAl4a e4 d.1 Tk./ J "Z 2 , .. TZ..i..Q 7 ...
/- ._ i... . + :"V.f..u.. G. .
. , ...l.iL.. '"u'..Z'C...'e. i.. :K ' L. . ./ .,.u.1.
T. t
- .r.... . . .
. . ... J. .. . ..7.I n.
. w. .
J ..
-.-. I.. n,,arnane. ,....
.. . L. i .r yu. . d.m .L ..- 6...
. ..T..
. .. . J. w. .. ,7 ..J y n .,.. 7.w . , uww.
g . .w n .. . 4s. www . . .yu . s w. s u..y a van . ay . E d. . + 4 .. . ,7 .....e I,..J . .l. u w.ye. . wa f A /. . ...J t. 4.k. . . d .Y .n.M. 7 , s, .
.. 1 41,., /4.L.
n... . . f. . ..... . . . . . . ]. . . .n-. . . . . .l . . . a l. . 4 u..T
.. . c.s...
-.+4, ..lii ou .u 4./.s.. + u +,J.. ..,...a444... ,..
- . 4. _. 4...,..
. l +...s. . . . ./- . . ,- ...
1 X .M..u.'+uG."7...f..:...'.,2. : c.. T..E.:1+ . . J. s .1R. l'i::f....T:ZL".'.12,. ". n..;).'h. n. .'2 '.:' L s. r.., ...
..a n...
1 n- a.. . L. .i.. i. G. T A1 ' T.. '. I. '2. 4..'.!..4. ".7., .'.'..' 4. Z. Z.1 4..
.. .. .. . . . '.'.'.Z, _'.._k._ .-. .
T.. u.. i. ,. __ u i u. , ...... . u...a . . . . . .. . u. ,a - ,
. w.... 4. . .u..
.,. ... ....J- l. . 21 . ..
+ f. . . . l _ , u..... . ... T. . ..e.
4.... u_y-. f4...L ......u.. nn&
. . . . . . . . . ..a_ ....
_ . . I.nu .l. c... . _ a.+. e T. .k. 4.
.4._ _,
+4.
./ .u...
.. 4. . ,4.. .. 4..
.4..
. . 7.i . . ' ., l . .+ . .
+ k. 4 ,
._.. +kZ_ ..y n.on.s...
. Z.... w.u.
+
. 7u.E..ennern
.4.-..
_.y 4.241..._J m 3
- 2. . Averaae Power Ranae Monitor - Inon
--' fut1ction (Inop O .r- +he This:';M vides assuranc that e'm C.
^^-if (nimum number of for W ggy APRMe4 BLE.Qlhytime" any po;> tion other than " Operate,"kn.d= APRM module switch is is ;.;nd t; in Io ser+ af,,') - un> lugged ' ' - + - - - ' - - - - - + ' - - " ' + - - - * - ' - '--'"
.. D:X=DC.Z M "Y2I G:U: 4f"'
ri, pq. 3
.s.. 1. 7a. ,
7' Y.wf u.Z.a. u
.1 ../..a. u..p.f.
su nc g '.L A A ..L. p g l.'..,5 7
+u l.'n...L4Jy .. .. .
. ,' -.'f . .a '/' .P,4 4. .' ".,./.l..u,..'./v ,A..n""B.A,F.
T ! .' m .'. A.. hl'''
+
" 'I'
' " d'.' 'm. ed',
. .w
.4 AM
. ./ n.
- 4. . .
,4m
...I.&34J.e,we.m + ., . u.
\
. u -mo L g w uJ w s's, .w us vu e u pr #
Qi Tunction-was-not'specif,.A.
,y . _. u, ,4_ 4. +.ul. '.
.7
+
. . _ . . ....w...,
. + 4 2. 4. ... .L.u i'cahy-cddited'irr-the-a,cEr.
n.a. / +aw'Z
. .4 .f7 /4.
usungL.r
. 4. . .'
. .y
.-damt T s analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.
- e. . .i. ....
.. .k. ,A. .... . _,.. ,... J . z.,i.n .
.1...
. . . , . . . ,Jh, . .. =. . . .,. . . . .,. . ,.u. . ,
,R B....,A...
E , a
. f.
_-....,7.. . 4. / ... .. la.i7 4.m y7 44. .u.fe. r,...-.C....J/4..fL.ar.
. 53 f wy f w. e,, m. r a. hw
..,d. . . A 6 ../ ar,7j.,
K w 4. . n.sj is . =ya l. l. i.
/I . .f -
. . . 4. L. -. .1.j.lJ ./
. .?4..12[...l..l., .,. . l*
. ,. 7 ?.e n .,. K. .U
.N J.u. 4.. . . fc. .. u.I. w wu u7 w ...J."
/
A . (continued) HATCH UNIT 1 B 3.3-12 REVISION 0
Insert 'F' - Bases B 3.3.1.1 2.d. Average Power Range Monitor - Inop
. ...or 3) the automatic self-test system detects a critical fault with the APRM channel, an Inop trip signal is sent to all four voter channels. Inop trips from two or more unbypassed APRM channels result in a trip output from all four voter channels to their associated trip system.
l l l l l l l t i f i l I f f
RPS Instrumentation B 3.3.1.1 BASES O. APPLICABLE 2. SAFETY ANALYSE eraoe Power Ranoe Monitor - InoD (Continued) LCO, and APPLICABILITY hers)is no Allowable Value for this Function. This Function is required to be OPERABLE in the MODES where ftOPS tions are required. lnserf 'G"
- 3. (Rea todssel Steam Dome Pressure - Hioh An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and .
THERMAL POWER transferred to the reactor coolant to ' increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection anal f Reference 4, reactor scram (the analyses conse el assume scram on the Average Power Range Monit yf' Neutron Flux - High signal, not the Reactor V Dome Pressure-High signal), along with the S/RVs, limits the peak RPV pressure to less than the ASME Section III Code limits. High reactor pressure signals are initiated from four pressure transmitters that sense reactor pressure. The Reactor Vessel Steam Dome Pressure - High Allowable Value is i chosen to provide a sufficient margin to the ASME Section III Code limits during the event. Four channels of Reactor Vessel Steam Dome Pressure - High Function, with two channels in each trip system arranged in a one-out-of-two logic, are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. The Functan is required to be OPERABLE in MODES 1 and 2 when the RCS is { pressurized and the potential for pressure increase exists. i I (continued) HATCH UNIT 1 B 3.3-13 REVISION O I i
- - . _~ _ ... .- .. . . - . . -
r l Insert 'G' - Bases 3.3.1.12 e 2-out-of-4 Voter
,1.e Two-out-of-Four Voter The Two-out-of-Four Voter Function provides the interface between the APRM l
^ Functions and the final RPS trip system logic. As such, it is required to be OPERABLE in the MODES where the APRM Functions are required and is necessary to support the safety analysis applicable to each of those Functions. Therefore, the Two-out-of-Four Voter Function is required to be OPERABLE in MODES I and 2. 1 All four voter channels are required to be OPERABLE. Each voter channel also includes self-diagnostic functions. If any voter channel detects a critical fault in its own processing, an Inop trip is issued from that voter channel to the associated trip system. There is no Allowable Value for this Function. l 1 1 1 l l i I l
i l l RPS Instrumentation l B 3.3.1.1 ! BASES APPLICABLE 5. Main Steam Isolation Valve - Closure SAFETY ANALYSES, j LCO, and C4 SIP clo:ure results in loss of the main turbine and the
- APPLICABILITY condenser as a heat sink for the nuclear steam supply system 4
(continued) and indicates a need to shut down the reactor to reduce heat ' generation. Therefore, a reactor scram is initiated on a Main Steam Isolation Valve - Closure signal before the MSIVs are completely closed in anticipation of the complete loss
- f normal heat sink and subsequent overpressurization i
transi . However, for the overpressurization protection anal is of Reference 4, the Average Power Range Monitor j Ne tron Flux - High Function, along with the S/RVs,
- limits t e peak RPV pressure to less than the ASME Code j limi . That is, the direct scram on position switches for i
closure events is not assumed in the overpressurization i analysis. Additionally, MSIV closure is assumed in the
- transients analyzed in Reference 2 (e.g., low steam line j pressure, manual closure of MSIVs, high steam line flow).
- The reactor scram reduces the amount of energy required to
- be absorbed and, along with the actions of the ECCS, ensures
- that the fuel peak cladding temperature remains below the j limits of 10 CFR 50.46.
! MSIV closure signals are initiated from position switches located on each of the eight MSIVs. Each MSIV has two position switches; one inputs to RPS trip system A while the other inputs to RPS trip system B. Thus, each RPS trip system receives an input from eight Main Steam Isolation Valve - Closure channels, each consisting of one position switch. The logic for the Main Steam Isolation Valve - Closure Function is arranged such that either the inboard or outboard valve on three or more of the main steam lines must close in order for a scram to occur. In , addition, certain combinations of valves closed in two lines ! will result in a half-scram. The Main Steam Isolation Valve -Closure Allowable Value is specified to ensure that a scram occurs prior to a significant reduction in steam flow, thereby reducing the severity of the subsequent pressure transient. Sixteen channels of the Main Steam Isolation Valve - Closure Function, with eight channels in each trip system, are , required to be OPERABLE to ensure that no single instrument ' failure will preclude the scram from this Function on a I (continued) HATCH UNIT 1 B 3.3-15 REVISION 0
-- - . . . . . - . - _ _ _ . _ _ . . = _ _ - _ . ...
l 4 RPS Instrumentation B 3.3.1.1 BASES ) APPLICABLE 8. Turbine Stoo Valve - Closure (continued) i
- SAFETY ANALYSES,
- LCO, and for the turbine trip event analyzed in Reference 2. For
- APPLICABILITY l this event, the reactor scram reduces the amount of energy
{ required to be absorbed and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded. Turbine Stop Valve - Closure signals are initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be closed to produce a scram. In addition, certain combinations of two valves closed will result in a half-scram. This Function must be enabled at THERMAL POWER m 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. The Turbine Stop Valve - Closure Allowable Value is selected to be high enough to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient. Eight channels of Turbine Stop Valve - Closure Function, l with four channels in each trip syn % are required to be ! OPERABLE to ensure that no single ategant failure will preclude a scram from this Function if in TSVs should I close. This function is required, consistent with analysis l assumptions, whenever THERMAL POWER is a 30% RTP. This 1 Function is not required whe E L POWER is < 30% RTP i since the Reactor Vessel a 'ressure - High and the Average Power Range Mon tor Neutron Flux - High Functions are adequate o maintain ie necessary safety margins.
- 9. Turbine Control Valve Fast Closure. Trio Oil Pressure - Low Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flur (continued)
HATCH UNIT 1 B 3.3-18 REVISION 0
i l, - j RPS Instrummtation { B '4.3.1.1 BASES l 1 APPLICABLE 9. Turbine Control Valve Fast Closure. Trio Oil
- l. SAFETY ANALYSES, Pressure - Low (continued) i LCO, and
! APPLICABILITY transients that must be limited. Therefore, a reactor scram l 1s initiated on TCV fast closure in anticipation of the l transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Function is the primary scram signal for the ! generator load rejection event analyzed in Reference 2. For l this event, the reactor scram reduces the amount of energy
- required to be absorbed and, along with the actions of the
. EOC-RPT System, ensures that the MCPR SL is not exceeded. l Turbine Control Valve Fast Closure, Trip 011 Pressure - Low signals are initiated by the electrohydraulic control (EHC) ! fluid pressure at each control valve. One pressure i transmitter is associated with each control valve, and the j- signal from each transmitter is assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POWER 2 30% RTP. This is normally accomplished l automatically by pressure transmitters sensing turbine first , stage pressure; therefore, opening of the turbine bypass
- valves may affect this Function.
The Turbine Control Valve Fast Closure, Trip 011 j Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure. I i Four channels of Turbine Control Valve Fast Closure, Trip ! Oil Pressure - Low Function with two channels in each trip
- system arranged in a one-out-of-two logic are required to be 1 OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function is required, consistent with the analycis i assumptions, whenever THERMAL POWER is a 30% RTP. This l 7
Function is not required when L POWER is < 30% RTP, l since the Reactor Vessel eam ressure - High and the Average Power Range Moni or Nei tron Flux - High , Functions are adequate t aintain tle necessary safety l margins.
- 10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, (continued)
HATCH UNIT 1 B 3.3-19 REVISION 0
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 11. SAFETY ANALYSES, Manual Scram (continued) LCO, and APPLICABILITY There is no Allowable Value for this function since the channels are mechanically actuated based solely on the position of the push buttons. Two channels of Manual Scram with one channel in each manual scram trip system are available and required to be OPERABLE in MODES 1 and 2, and in MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies, since these are the MODES and other specified conditions when control rods are withdrawn. ACTIONS A Note has been provided to modify the ACTIONS related to RPS instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or l not within limits, will not result in separate entry into l the Condition. Section 1.3 also specifies that Required ' Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel. A.1 and A.2 Because of the diversity of sensors available to provide ' trip signals an. ,
.. ncy of the RPS design, an allowable out f serv s be acceptabl: Ref. 9,)ice 'tb ,1me p itofrestoration 12 hours has of been any shown to inoperable cha -1 . ** * :LE status. However, this out of service time is Snly acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip sys'cem must be placed in the tripped (continued)
HATCH UNIT 1 B 3.3-21 REVISION 0 mra
1 i RPS Instrumentation B 3.3.1.1 i BASES ACTIONS A.1 and A.2 (continued) j condition per Required Actions A.1 and A.2. Placing the i inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, i restore capability to accommodate a single failure, and allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel ! in t, rip would result in a full scram), Condition D must be i 3 nt4redNa s Required Action taken. In ser + *H ' B. ndAEL 2 Condition B exists when, for any one or more Functions, at i least one required channel is inoperable in each trip system. In this condition, provided at least one channel j per trip system is OPERABLE, the RPS still maintains trip capability for that function, but cannot accommodate a single failure in either trip system. Required Actions B.1 and B.2 limit the time the RPS scram logic, for any Function, would not accommodate single
- failure in both trip systems (e.g., one-out-of-one one-out-of-one arrangement for a typical fou c an el Function). The reduced reliability of thi ^^j 4 arrangement was not evaluated in Reference pl ic for the Completion Time. Within the 6 hour allowan , he i associated Function will have all required channels OPERABLE j !
or in trip (or any combination) in one trip system. !
! Complet ne o these Required Actions restores RPS to a a
reliab' ity le Refere ceS9$yek uivalent to that evaluated in
~
hit'h ustified a 12 hour allowable out of servic ti as ented in Condition A. The trip system i in the e aded state should be placed in trip or, alternatively, all the inoperable channels in that trip system should be placed in trip (e.g., a trip system with two inoperable channels could be in a more degraded state than a trip system with four inoperable channels if the two inoperable channels are in the same Function while the four inoperable channels are all in different Functions). The decision of which trip system is in the more degraded state should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in). (continued) HATCH UNIT 1 B 3.3-22 REVISION 0
( J . l Insert 'H' - Bases 3.3.1.1 Actions A.1 and A.2 As noted, Required Action A.2 is not applicable for APRM Functions 2.a,2.b,2.c, and j 2.d. Inoperability of one required APRM channel affects both trip systems; thus, Required i Action A.1 must be satisfied. This is the only action (other th:.n restoring ; OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel of the same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for l each channel. b i 4 8 i i i
RPS Instrumentation B 3.3.1.1 W BASES l ACTIONS B,1 and 8.2 (continued) If this action would result in a scram or RPT, it is 1 permissible channels to place the other trip system or its inoperable in trip. l The 6 hour Completion Time is judged acceptable based on the 1 remaining capability to trip, the diversity of the sensors available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all diverse Functions, and the low probability of an event requiring the initiation of a scram. Alternately, if it is not desired to place the inoperable channels (or one trip system) in trip (e.g., as in the case where placing the inoperable channel or associated trip systentin trip would result in a scram or RPT), Condition D
*t be ente and its Required Action taken.
/Asert "I C.
Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function result in the Function not maintaining RPS trip capability. A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both trip systems will generate a trip signal from the given Function on a valid signal. ; The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. D.1 Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is function and MODE or other specified condition dependent and may change as the Required Action of a previous Condition is completed. Each (continued) HATCH UNIT 1 B 3.3-23 REVISION 0
Insert 'I' - Bases 3.3.1.1 Actions B.1 and B 2 As noted, Condition B is not applicable for APRM Functions 2.a,2.b,2.c, and 2.d. Inoperability of an APRM channel affects both trip systems and is not associated with a specific trip system as are the APRM two-out-of-four voter and other non-APRM channels for which Condition B applies. For an inoperable APRM channel, Required Action A.1 must be satisfied, and is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than one required APRM channel results in loss of trip capability for that Function and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A
)
and C provide Required Actions that are appropriate for the inoperability of APRM Functions 2.a, 2.b, 2.c, and 2.d, and these Functions are not associated with specific trip systems as are the APRM two-out-of-four voter and other non-APRM channels, 1 Condition B does not apply. I l 1 1 1
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.2 REQUIREMENTS (continued) To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days is based on minor changes in LPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8. A restriction to satisfying this SR when < 25% RTP is provided that requires the SR to be met only at 2 25% RTP because it is difficult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 25% RTP. At low power levels, a high degree of accuracy is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2 25% RTP, the Surveillance is required to have been satisfactorily performed within the last 7 days, in accordance with SR 3.0.2. A Note is provided which allows an increase in THERMAL POWER above 25% if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. SP3T'13( Ihlof used n- m. _- o_..__ n ___ u. _ u _ _ m _. . o1.._2 c4 ..i...
'" E ~ ", i.,
"' X 2'.L~"'M"'"W
. . T,. Z..... " j. , ' ,T.l'l .'3'. "._l.,.f.""I,
+LL + s.X 1, 12 1, . .! rn .. I . y
'::7:'.'..Z1""u"I:Tll"')E"
. . ,,'/Z. " s_ ",f,"' , '."y.7
- r. . .
17h
' E. . . ,Z. ' M., , , ,. . . . _. . , . . . .
,- L1, r 1- . L ./ E 1,. . L . - l . . f. . . J .h "7; '
. ;2 ,1"I 'l Z. ' .l. 7 ."u.
2..r . . " i..:1._' 7:'7. v'.r...j..7. r r . ?0:'l". '. ' . '.Z, "._ .."7,f1Z
%je-
=t;d.ier int -f +j:t fl 7 i AM"fjth- unct 7 I . ;=1 r:" =,t: 6.d,
+ J. z 1. .. :u r-Eir: D_. 4,thA idefd: .,7, .uuL tp1. l..t._ .t M fd=i 'r f
2 =.L I .fi"E s u s .1 J j
.' ..7'. 'G.
'": Z. Z :
.'.f 'I.=l:
.. 3."7.
. " . . ; ' . I , ". .. y:A 7.:". . r. !.7. /". " C. . .T 7.
punit ."t bc fi 7. . p/.fr /=y / 7 f. idsi=1 b . r:bl e =1=.;i= 4=
,.p=:t,i
- u. + . . 1. r ".+.=_,+ u.. .
p:,. i = =q=d t h,. rel i .,b i l i ty,=f t O .. (continued) HATCH UNIT 1 B 3.3-26 REVISION 0
RPS instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.4 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. l As noted, SR 3 .. 4 is not required to be performed when entering M 2 from 1, since testing of the MODE 2 required I MODE 1 with ;/j'),'l uti i z,TJFu ctionslifted jumpers, cannot leads, beorperformed movable in l links. This al ntry into MODE 2 if the 7 day Frequency is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours after entering MODE 2 from MODE 1. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. i 4 Frequency of 7 days provides an acceptable level of system I average unavailability over the Frequency interval and is based on reliability analysis (Ref. 9). SR 3.3.1.1.5 A CHANNEL FUNCTIONAL TEST is perforrred on each required channel to ensure that the entire channel will perform the intended function. A Frequency of 7 days provides an acceptable level of system average availability over the Frequency and is based on the reliability analysis of Reference 9. (The Manual Scram Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the analysis to extend many automatic scram Functions' Frequencies.) SR 3.3.1.1.6 and SR 3.3.1.1.7 These Surveillances are established to ensure that no gaps in neutron flux indication exist from subcritical to power operation for monitoring core reactivity status. The overlap between SRMs and IRMs is required to be demonstrated to ensure that reactor power will not be increased into a neutron flux region without adequate indication. This is required prior to withdrawing SRMs from (continued) HATCH UNIT I B 3.3-27 REVISION O l l
l RPS Instrumentation : 2 B 3.3.1.1 ! 1
! BASES SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required i
i channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of SR 3.3.1.1.9 is based on the reliability analysis of , Reference 9. ! The 18 month Frequency is based on the need to perfom this i Surveillance under the conditions that apply during a plant i outage and the potential for an unplanned transient if the Surveillance were perfomed with the reactor at power. l Operating experience has shown that these components usually ! pass the Surveillance when performed at the 18 month ! Frequency. . i 1 b82 b $6cIIUw fo
. SR 3.3.1.1.10 andISR 3.3.1.1.13 S R33. l. I. /t i lmd O , y J
' A CHANNEL CALIBRATION is a complete check of the instrument i loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary i range and accuracy. CHANNEL CALIBRATION leaves the channel ! adjusted to account for instrument drifts between successive j. " calibrations, consistent with the plant specific setpoint methodology. For MSIV-Closure, SDV Water Level-High i (Float Swi includes
~)W Closure Functions, this SR also physical ins ection and actuation of the switches /neef */f #
Note 1 s neutron detectors are excluded fron ' CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Changes ir, neutron detector sensitivity are compensated for by performing the 7 day calorimetric i calibration (SR 3.3.1.1.2) and the 1000 effecti e full power i hours LPRM calibration against the TIPS R y . second Note is provided that 8). A t , . ., . , gg SRs to be performed within 12 ho rs o' e from MODE 1. Testing of the MOD 2 ,v .,;," RM Functions cannot ' be performed in MODE 1 with t ti jumpers, lifted leads or movable links. This o e allows entry into MODE 2 l from MODE 1 if the associated Frequency is not met per 3.0.2. (continued) l HATCH UNIT 1 B 3.3-29 REVISION 0 k _ - -
Insert 'J' - Bases 3.3.1.1 Surveillw-;e Requirements SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. For the APRM Functions, this test supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The APRM CHANNEL FUNCTIONAL TEST covers the APRM channels (including recirculation flow processing -- applicable to Function 2.b only), the two-out-of-four voter channels, and the interface connections to the RPS trip systems from the voter channels. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. The 184 day Frequency of SR 3.3.1.1.10 is based on the reliability analysis ofReference 12. (NOTE: The actual voting logic of the two-out-of-four voter channels is tested as part of SR 3.3.1.1.15.) For Function 2.a, a Note that requires this SR to be performed within 12 hours of entering MODE 2 from MODE 1 is provided. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted leads. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2. Insert 'K' - Bases 3.3.1.1 Surveillance Requirements SR 3.3.1.1.13 For the APRM Simulated Thermal Power - High Function, this SR also includes calibrating the associated recirculation loop flow channel. 1 i i i
RPS Instrumentation ; 8 3.3.1.1 l BASES I SURVEILLANCE W 23' 4.lWa4 SR 3.3. ) 0OM E REQUIREMENTS A (centin= @ * % O I U hwelve hours is based on operating experience and in consideration of providing a reasonable time in which to l complete the SR. k rt/ d ..___.. _t en ,n ,s. ,n 2_ t___2 .___,.u_,..._..m_ T MWT.T_:' KLt'12M:E T:"f 5:7 '"1 :7~."7c"7
\
'!. lI. . _...' :. M. . . ._E
. .y. _ ' "f.
.. I_' Z ,'.:..'_'.A....".PE. . . " ' .4. . ," . T. . . ".I. ,". . Z. ;. ""Z i
The Frequency of SR 3.3.1.1.13 is based upon the assumption 1 of an 18 month calibration interval in the determination of {the magnitude of equipment drift in the setpoint analysis. 1 SR 3.3.1.1.11 l This SR ensures that scrams initiated from the Turbine Stop I Valve - Closure and Turbine Control Valve Fast Closure, Trip ; 011 Pressure - Low functions will not be inadvertently ' bypassed when THERMAL POWER is a 30% RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER h 30% RTP to ensure that the calibration is valid. If any bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at a 30% RTP, either due to open main turbine bypass valve (s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure, Trip 011 Pressure-Low Functions are enabled), this SR is met and the channel is considered OPERABLE. The Frequency of 184 days is based on engineering judgment and reliability of the components. fInser+- so s.3. i . I. r5 hee) (continued) HATCH UNIT 1 B 3.3-30 REVISION O
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS S W yred.) 3.1.1l b , (continued) ' -~~o~4 n-s-- u Zi+-- x 'N. rA-' n4'--' ci- -+ d N d 'C"T 7' ' "7l,, L' T V _'Z';::".'.J. :"". "l':X.:.';I" 'a TX . , ' Z.Z; LZl, T Z" C 4.IT'T' "Z:1?".ZM.!L':'/:1~.T JX"';' ' 7r ' SZ"L 'R"'/ . ': Z"' 2: / 7,L ./ n 1 R y :" : 1.,. ,,1
'X. .'."
7. 4,Z".
+ .L
- .' a. ' ' 7',, ."u :. .LL7, . s~ a .".'.'.
.u.' . 7u
'.1. ". :' l."' :
+ . :i.
.l. .7.'..:
+ 1. . .../:".7 L.1. . c.
. . 1 X T,';/E"Z'.:Enu"7 XLX 'T;":7.7' "C".Z' 71"T"/
__ .T JL "!_'/ "L',4'X"'.'"'/ ' ""'"Z_ .' "'1J;"~1 ::"2 ' :'/. '!d . / f.C. ': X. .:.".. ' 7:
. . . s:":_Z". "'Z,". '."."'. ms
,'L"'s. .:.'] l.L..l'E . ::I. :L.:. 1. '. 2 TL c-.____., f i _AL- '
L_.
- rz. :r.y::w ::S.,r. .::: u r: :v.=..vc" a:-
._ ..J. .J 4 .J r= r rn m
SR 3.3.1.1.15 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific channel. The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant l outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Fr P Inse,+ L ) vSk3.3 El .16/ This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME cpeftance7tiperi are included in Reference 10. 7)P In ser+ *M " UW (continued) HATCH UNIT 1 B 3.3-31 REVISION O
)
i Insert 'L' - Bases 3.3.1.1 Surveillance Requirements SR 3.3.1.1 15 The LOGIC SYSTEM FUNCTIONAL TEST for APRM Function 2.e simulates APRM trip conditions at the two-out-of-four voter channel inputs to check all combinations of l two tripped inputs to the two-out-of-four logic in the voter channels and APRM related l redundant RPS relays. Insert 'M' - Bases 3.3.1.1 Surveillance Requirements SR 3.3.1.1.16 i RPS RESPONSE TIME for APRM Two-out-of-Four Voter Function 2.e includes the output relays of the voter and the associated RPS relays and contactors. (The digital ; portions of the APRM and two-out-of-four voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration check the time base of the digital electronics.) Confirmation of the time base is adequate to assure required response times are met. Neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. 1 I 1 i l ! l l i { 1 l l l l l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 (continued) REQUIREMENTS RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. This Frequency is consistent with the typical industry refueling cycle and is. based upon plant operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. REFERENCES 1. FSAR, Section 7.2.
- 2. FSAR, Chapter 14.
- 3. FSAR, Section 6.5.
- 4. FSAR, Appendix M.
- 5. FSAR, Section 14.3.3.
- 6. NE00-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
- 7. FSAR, Sections 14.4.2 and 14.5.5.
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram Discharge System Safety Evaluation," December 1, 1980.
- 9. NED0-30851-P-A , " Technical Specification Improvement Analyses for BWR Reactor Protection System,"
March 1988.
- 10. Technical Requirements Manual.
- 11. NRC No. 93-102, " Final Policy Statement on Technical SpetTfTeaitTori rovements," July 23, 1993.
In se r + *N* b HATCH UNIT 1 B 3.3-32 REVISION 0
.- . . . _ - _ .m__ . . . _ . . _ . _ . _ .- . _ _ . _ _ _ _ _ _ . . _ . _ . _ . _ _ . _ . . _ . . . _ _ . . _ _ . _ .
l I Insert 'N' - Bases 3.3.1.1 References l 12. NEDC-32410P-A," Nuclear Measurement Analysis and Control Power Range , Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip l Function," October 1995. i , f i ( k i I ) i i j I i, b l 1 l
SRM Instrumentation B 3.3.1.2 BASES APPLICABLE System RPS) Instrumentation"; IRM Neutron Flux - High and SAFETY ANALYSES Average mer Range Monitor (APRM) Neutron Flux - Hig , (continued) (Setdown)Fanctions; and LCO 3.3.2.I, " Control Rod Block Instrume tation." ) The SRMs have no safety function and are not assumed to function during any FSAR design basis accident or transient analysis. However, the SRMs provide the only on scale monitoring of neutron flux levels during startup and refueling. Therefore, they are being retained in Technical Specifications. l l LCO During startup in MODE 2, three of the four SRM channels are i required to be OPERABLE to monitor the reactor flux level i prior to and during control rod withdrawal, suberitical ! multiplication and reactor criticality, and neutron flux level and reactor period until the flux level is sufficient to maintain the IRMs on Range 3 or above. All but one of l the channels are required in order to provide a l representation of the overall core response during those l periods when reactivity changes are occurring throughout the l Core. j In MODES 3 and 4, with the reactor shut down, two SRM l channels provide redundant monitoring of flux levels in the i core. 1 In MODE 5, during a spiral offload or reload, an SRM outside the fueled region will no longer be required to be OPERABLE, l since it is not capable of monitoring neutron flux in the fueled region of the core. Thus, CORE ALTERATIONS are allowed in a quadrant with no OPERABLE SRM in an adjacent quadrant provided the Table 3.3.1.2-1, footnote (b), requirement that the bundles being spiral reloaded or spiral ) offloaded are all in a single fueled region containing at least one OPERABLE SRM is met. Spiral reloading and offloading encompass reloading or offloading a cell on the edge of a continuous fueled region (the cell can be reloaded or offloaded in any sequence). In nonspiral routine operations, two SRMs are required to be OPERABLE to provide redundant monitoring of reactivity (continued) HATCH UNIT I B 3.3-34 REVISION O
~ . . _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _ . _ _ . _ . _ _ . _ _ _ . _ _ - . _ _ _ _ . . _
! Control R1d Block Instrumentation 4 B 3.3.2.1 j B 3.3 INSTRtalENTATION ; 4 B 3.3.2.1 Control Rod Block Instrumentation BASES i BACKGROUND Control rods provide the primary means for control of i i reactivity changes. Control rod block instrumentation i includes channel sensors, logic circuitry, switches, and ! relays that are designed.to ensure that specified fuel ! design limits are not exceeded for postulated transients and j accidents. During high power operation, the rod block i i monitor (RBM) provides protection for control rod withdrawal
- l error events. During low power operations, control rod i i blocks from the rod worth minimizer (RWM) enforce specific <
! control rod sequences designed to mitigate the consequences i of the control rod drop accident (CRDA). During shutdown l l conditions, control rod blocks from the Reactor Mode : Switch - Shutdown Position Function ensure that all control rods remain inserted to prevent inadvertent criticalities. The purpose of the RBM is to limit control rod withdrawal if localized neutron flux exceeds a predetermined setpoint ! during control rod manipulations. It is assumed to function l to block further control rod withdrawal to preclude a MCPR l Safety Limit (SL) violation. The RBM supplies a trip signal { to the Reactor Manual Control System (RMCS) to appropriately ) inhibit control rod withdrawal during power operation above the low power range setpoint. The RBM has two channels, either of which can initiate a control rod block when the channel output exceeds the control rod block setpoint. One RBM channel inputs into one RMCS rod block circuit and the ( other RBM channel inputs into the second RMCS rod block
. h hannel al is nerated by averaging a set of loca power r monit PRM t various core heights surrounding the control rod being wi .
sgnayromong}ayragepowerrangemo,jtjrgRMganne{S supplies a reference signal foMeWehainiid45WF **#0"
' " " - - - This reference signal is used to determine ich range setpoint (low, intermediate, or high) i e . If th APRM is ndicati less than t low range setpo ,t is tic b ed. he RBM is also automatically bypassed if a peripheral control rod is selected (Ref. 1). A rod block signal is also generated if an RBM Downscale trip or an Inoperable trip occurs. The Downscale trip Jill occur if the RBM channel signal (continued)
HATCH UNIT 1 B 3.3-42 REVISION 0 l
)
L l l l i ,
- 4
- Insert 'O' - Bases 3.3.2.1 Background -
l
. . ., and a signal from another of the APRM channels supplies the reference signal to the second RBM channel.
l. I t i i i l. t i 1 !~ l-l t t (' I l I l e e f I i, , , .- ,_ . __
! Control Rod B1cck Instrumentation B 3.3.2.1 i BASES
- BACKGROUND decreases below the Downscale trip setpoint after the RBM
, (continued) signal has been normalized. The Inoperable trip will occur during the nulling (normalization) sequence, if: t RBM
. channel fails to null, too few LPRM i 11a , 1 i t ged or he ion switch is moved to y pos tion other
...u...
an rat " A.s s ue ..J.u ..n . < .s <Th-(.typ'7L
.' ...s
#TF0 sZ +0' -f1/ I l '": .'Z2': XC "Z ::7Z *3Z ';C'2'L'II"'s 'ZZ/s u./ l l 4I17+C ".'Z..'"'E'"RZ"" X~'1E' Z LC'IZ'Z, ~Z
, . . . "/ '"' '/ ' ~ 7" '"7 '""
Z. . ,.J.."'.I. l.f..."'Z. . L, ,- ".y'X. 3 The purpose of the RWM is to control rod patterns during startup and shutdown, such that only specified control rod sequences and relative positions are allowed over the operating range from all control rods inserted to 10% RTP. i The sequences effectively limit the potential amount and rate of reactivity increase during a CRDA. Prescribed i control rod sequences are stored in the RWM, which will ! initiate control rod withdrawal and insert blocks when the actual sequence deviates beyond allowances from the stored sequence. The RWM determines the actual sequence based i position indication for each control rod. The RWM also uses ' feedwater flow and steam flow signals to determine when the reactor power is above the preset power level at which the RWM is automatically bypassed (Ref. 2). The RWM is a single channel system that provides input into both RMCS rod block circuits. With the reactor mode switch in the shutdown position, a control rod withdrawal block is applied to all control rods to ensure that the shutdown condition is maintained. This Function prevents iv.dvertent criticality as the result of a control rod withdrawal during MODE 3 or 4, or during MODE 5 wher, the reactor mode switch is required to be in the shutdown position. The reactor mode switch has two channels, each inputting into a separate RMCS rod block circuit. A rod block in either RMCS circuit will provide a control rod block to all control rods. (continued) HATCH UNIT I B 3.3-43 REVISION O
_ _ . . _ _ . _ _ _ . _ _ _ _ _ _ . ~ . . _ _ . _ _ _ _ _ . . _ . _ . _ __.._ ___ f Control Rod Block Instrumentation j B 3.3.2.1
; BASES I
i i SURVEILLANCE SR 3.3.2.1.1 ! REQUIREMENTS i ^ (continued) A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended function. It includes the Reactor Manual Control System [ input. l i ; j Any setpoint adjustment shall be istent with the { ! assumptions of the current speg c setpoint ' methodology. The Freque of W difs based on { reliability analyses (Re . E). ll { SR 3.3.2.1.2 and SR 3.3.2.1.3 i , A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure i that the entire system will perform the intended function. i The CHANNEL FUNCTIONAL TEST for the RWM is performed by
- j. attempting to withdraw a control rod not-in compliance with
- the prescribed sequence and verifying a control rod block
- occurs. This test is performed as soon as possible after' 1 the applicable conditions are entered. As noted in the SRs, i SR 3.3.2.1.2 is not required to be performed until I hour l after any control rod is withdrawn at < 10% RTP in MODE 2, 4
and SR 3.3.2.1.3 is not required to be performed until 3 1 hour after THERMAL POWER is < 10% RTP in MODE 1. This j allows entry into MODE 2 (and if entered during a shutdown, ! 1~ concurrent power reduction to < 10% RTP) for SR 3.3.2.1.2 i and THERMAL POWER reduction to < 10% RTP in MODE 1 for
)
l j SR 3.3.2.1.3 to perform the required Surveillances if the j 92 day Frequency is not met per SR 3.0.2. The 1 hour ); allowance is based on operating experience and in i 4 . consideration of providing a reasonable time in which to j complete the SRs. The 92 day Frequencies are based on i reliability analysis (Ref. 8). SR 3.3.2.1.4 i j The RBM setpoints are automatically varied as a function of power. Three Allowable Values are specified in i Table 3.3.2.1-1, each within a specific power range. The power at which the control rod block Allowable Values i automatically change are based on the APRM signal's input to ; each RBM channel. Below the minimum power setpoint, the RBM l is automatically bypassed. These power Allowable Values ! (continued) HATCH UNIT 1 B 3.3-50 REVISION 0
l Control Rod Block Instrumentation B 3.3.2.1 l l BASES I SURVEILLANCE SR 3.3.2.1.8 (continued) , REQUIREMENTS ! OPERABLE following loading of sequence into RWM, since this l is when rod sequence input errors are possible. ! REFERENCES 1. FSAR, Section 7.5.8.2.3.
- 2. FSAR, Section 7.2.2.4. I l
- 3. NEDC-30474-P, " Average Power Range Monitor, Rod Block Monitor, and Technical Specification Improvements (ARTS) Program for Edwin I. Hatch Nuclear Plants,"
December 1983. 1
- 4. NEDE-24011-P-A-US, " General Electrical Standard j Application for Reload Fuel," Supplement for United i States, (revision specified in the COLR). !
- 5. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),
" Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A," BWROG-8644, August 15, 1986.
- 6. NED0-21231, " Banked Position Withdrawal Sequence,"
lanuary 1977.
- 7. NRC SER, " Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," " General Electric Standard Application for Reactor Fuel, Revision 8, Amendment 17," December 27, 1987.
- 8. NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Control Red Block Instrumentation,"
October 1988.
- 9. GENE-770-06-1, " Bases For Changes To Surveillance Test l Intervals and Allowed Out-0f-Service Times For ;
Selected Instrumentation Technical Specifications," l February 1991. I
- 10. NRC No. 93-102, " Final Policy Statement on Technical pitTeat y ovements," July 23, 1993.
In ser+ *P " i HATCH UNIT 1 B 3.3-53 REVISION 0 s
l 1- i
- l l l i
Insert 'P' - Bases 3.3.2.1 References 1 i i i ' l1. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range i Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995. .; i l 1 i e !
- -I j 1 J
} i l 1 4 4 i i i i T U
EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Ston Valve - Closure SAFETY ANALYSES, LCO, and Closure of the TSVs and a main turbine trip result in the APPLICABILITY loss of a heat sink and increases reactor pressure, neutron (continued) flux, and heat flux that must be limited. Therefore, an RPT is initiated on a TSV- Closure signal before the TSVs are completely closed in anticipation of the effects that would result from closure of these valves. EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient. Closure of the TSV; .s determined by measuring the position of each valve. While there are two separate position switches associated with each stop valve, only the signal from one switch for each TSV is used, with each of the four channels being assigned to a separate trip channel. The logic for the TSV-Closure Function is such that two or more TSVs must be closed to produce an EOC-RPT. This Function must be enabled at THERMAL POWER a: 30% RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this function. Four channels of TSV - Closure, with two channels in each trip system, re available and required to be OPERABLE to ensure that no : 'gle instrument failure will preclude an EOC-RPT from this Ft..ction on a valid signal. The TSV - Closure Allowable Value is selected to detect imminent TSV closure. This protection is required, consistent with the safety analysis assumptions, whenever THERMAL POWE g 30% RTP. Below 30% RTP, the Reactor Vessel Steam omejejsre-High ; and the Average Power Range Monitor (AP ) ,.,7 N tron I Flux - High Functions of the Reactor P otection tem (RPS) are adequate to maintain the necessary o the MCPR Safety Limit. Turbine Control Valve Fast Closure. Trio Oil Pressure - Low Fast closure of the TCVs during a generator load rejection results in the loss of a heat sink that produces reactor 4 pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is initiated on TCV Fast Closure, Trip 011 Pressure - Low in anticipation of the transients that would result from the closure of these (continued) HATCH UNIT I B 3.3-82 REVISION 1
i EOC-RPT Instrumentation B 3.3.4.1 j BASES
-APPLICABLE Turbine Control Valve Fast Closure. Trio 011 Pressure - Low
- SAFETY ANALYSES, (continued)
> LCO, and APPLICABILITY valves. The EOC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL is not exceeded during the worst case transient. . Fast closure of the TCVs is determined by measuring the
- electrohydraulic control fluid pressure at each control valve. There is one pressure transmitter associated with .
- each control valve, and the signal from each transmitter is l l assigned to a separate trip channel. The logic for the TCV l l Fast Closure, Trip Oil Pressure - Low Function is such that '
two or more TCVs must be closed (pressure transmitter trips) 4 to produce an EOC-RPT. This Function must be enabled at THERMAL POWER a: 30% RTP. This is normally accomplished , automatically by pressure transmitters sensing turbine first 1 stage pressure; therefore, opening of the turbine bypass i l valves may affect this Function. Four channels of TCV Fast ! Closure, Trip Oil Pressure - Low, with two channels in each trip system, are available and required to be OPERABLE to ' i ensure that no single instrument failure will preclude an
- EOC-RPT from this Function on a valid signal. The TCV Fast
- Closure, Trip 011 Pressure - Low Allowable Value is selected high enough to detect imminent TCV fast closure.
J l This protection is required consistent with the safety l analysis THERMAL POWER is 1 30% RTP. Below , ! 30% RT t e Reac >r Vessel Steam Dome Pressure - High and j the A $Neu;ron Flux - High Functions of the RPS are i adequ to main in the necessary margin to the MCPR Safety Limit. 1 j ACTIONS A Note has been provided to modify the ACTIONS related to
- EOC-RPT instrumentation channels. Section 1.3, Completion
! Times, specifies that once a Condition has been entered, i subsequent divisions, subsystems, components, or variables ! expressed in the Condition, discovered to be inoperable or ! not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required
- Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial
- entry into the Condition. However, the Required Actions for j inoperable EOC-RPT instrumentation channels provide i
j (continued) HATCH UNIT I B 3.3-83 REVISION 1 l
Recirculation Loops Operating l'
- i. B 3.4.1 i
BASES i i APPLICABLE case (since the intact loop starts at a lower flow rate and i SAFETY ANALYSES the core response is the same as if both loops were ' (continued) operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement. i The recirculation system is also assumed to have sufficient i flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 14 of the FSAR.
- A plant specific LOCA analysis has been performed assuming i only one operating recirculation loop. This analysis has l demonstrated that, in the event of a LOCA caused by a pipe i break in the operating recirculation loop, the Emergency
- Core Cooling System response will provide adequate core '
! cooling, provided the APLHGR requirements are modified l l accordingly (Ref. 3). d i j The transient analyses of Chapter 15 of the FSAR have also 2 been performed for single recirculation loop operation l (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the ! abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop , operation, modification to the Reactor Protection System ! (RPS) average power range monitor (APRM) instrument j setpoints is also required to account for the d nt relationships between recirculation drive f1 and re ctor . core flow. The APLHGR and MCPR setpoints f s operation are specified in the COLR. Simulated Thermal Power-High setpointThe A RM,,ge log /)'
.,. i .y is ir LCO 3.33._1,
" Reactor Protection System (RPS) Instrumenta "
Recirculation loops operating satisfies Criterion 2 of the $ NRC Policy Statement (Rr.f. 5). l LC0 Two recirculation loops are normally required to be in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, ; modifications to the required APLHGR limits (LCO 3.2.1, '
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"),
(continued) HATCH UNIT 1 B 3.4-3 REVISION 0 L
4 Recirculation Loops Operating B 3.4.1 l l BASES LC0 and A 6 f imulated Thermal Power - High setpoint : (continued) (LC0 h3Maug,[ e# applied to allow continued operation consistent with the assumptions of Reference 3. In addition, core flow as a function of core thermal power must be in the " Operation Allowed Region" of Figure 3.4.1-1 to ensure core thermal-hydraulic oscillations do not occur. APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. l ACTIONS A.1 and B.1 Due to thermal-hydraulic stability concerns, operation of , the plant with one recirculation loop is controlled by restricting the core flow to 2: 45% of rated core flow when THERMAL POWER is greater than the 76% rod line. This l requirement is based on the recommendations contained in GE SIL-380, Revision 1 (Reference 4), which defines the region where the limit cycle oscillations are more likely to occur. If the core flow as a function of core thermal power is in the " Operation Not Allowed Region" of Figure 3.4.1-1, prompt action should be initiated to restore the flow-power combination to within the Operation Allowed Region. The 2 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing core oscillations to be quickly detected. An immediate reactor scram is also required with no recirculation pumps in operation, since all forced circulation has been lost and the probability of thermal-hydraulic oscillations is greater. (continued) HATCH UNIT 1 B 3.4-4 REVISION 5
4 i j SDM Test - Refueling B 3.10.8 i BASES APPLICABLE CRDA analyses assume that the reactor operator follows SAFETY ANALYSES prescribed withdrawal sequences. For SDM tests performed (continued) within these defined sequences, the analyses of References 1 and 2 are applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety analyses of References 1 and 2 may not be met. Therefore, special CRDA analyses, performed in accordance with an NRC approved methodology, may be required to descastrate the SDM test sequence will not result in unacceptable consequences should a CRDA occur during the testing. For the purpose of this test, the protection provided by the normally required MODE 5 applicable LCOs, in addition to the requirements of this LCO, will maintain normal test operations as well as postulated accidents , within the bounds of the appropriate safety analyses ! (Refs. I and 2). In addition to the added requirements for the RWM, Average Power Range Monitors, and control rod coupling, the notch out mode is specified for out of , sequence withdrawals. Requiring the notch out mode limits I withdrawal steps to a single notch, which limits inserted i reactivity, and allows adequate monitoring of changes in . neutron flux, which may occur during the test. ) As described in LCO 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of the NRC Policy Statement apply. Special Operations LCOs i provide flexibility to perform certain operations by ! appropriately modifying requirements of other LCOs. A < discussion of the criteria satisfied for the other LCOs is ] provided in their respective Bases. < l I LCO As described in LCO 3.0.7, compliance with this Special Operations LCO is optional. SDM tests may be performed l while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS. For SDM ) tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is ava e. o provide additional scram protection beyond normally quired , IRMs, the Average Power Range Moni ors ar also r uired to 1 be OPERABLE (LCO 3.3.1.1, Functio 2.av in '2.e) s though ! the reactor were in MODE 2. Beca e a itiple co trol rods J will be withdrawn and the reactor ly become
)
critical, the approved control rod withdrawal sequence must i I l (continued) HATCH UNIT 1 B 3.10-34 REVISION O
SDM Test - Refueling B 3.10.8 BASES (continued) SURVEILLANCE SR 3.10.8.1. SR 3.1 2 and SR 3.10.8.3 REQUIREMENTS 2.d LCO 3.3.1.1, Function 2.a"a'nd 2.3e, made applicable in this Special Operations LC , are require to have their Surveillances met to esta i this Special Operations LC0 is being met. However, the control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LCO 3.3.2.1, Function 2, MODE 2 requirements) or by a second licensed operator (Reactor Operator or Senior Reactor Operator) or other qualified member of the technical staff (e.g., a qualified shift technical advisor or reactor engineer). As noted, either the applicable SRs for the RWM (LCO 3.3.2.1) must be satisfied according to the applicable Frequencies (SR 3.10.8.2), or the proper movement of control rods must be verified (SR 3.10.8.3). This latter verification (i.e., SR 3.10.8.3) must be performed during control rod movement to prevent deviations from the specified sequence. These Surveillances provide adequate assurance that the specified test sequence is being followed. SR 3.10.8.4 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is operated within the bounds of the safety analysis. The 12 hour Frequency is intended to provide appropriate assurance that each operating shift is aware of and verifies compliance with these Special Operations LCO requirements. SR 3.10.8.5 Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The verification is required to be performed any time a control rod is withdrawn to the full-out notch position, or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved, as well as operating experience related to uncoupling events. (continued) HATCH UNIT 1 B 3.10-37 REVISION 0
'l
- RPS Instrumentation B 3.3.1.1 i
- j. BASES
} ! l APPLICABLE Averaae Power Ranae Monitor (APRM) ! 1 SAFETY ANALYSES, j LCO, and The APRM channels provide the primary indication of neutron ! ! APPLICABILITY flux within the core and respond almost instantaneously to j (continued) neutron flux increases. The APRM channels receive input signals from the local power range monitors . the reactor core to provide an indication of(LPRMs) the powerwithin j distribution and 1 ! The APRM channels average these sig al to provide continuous { indication of verage reactor power from few percent to greater than TP. k /NSER T *D 1 The APRM System s vid i o4 channels and 4 two-i out-of-four voter a 1s. Ea APRM channel provides ) inputs to each of the four voter channels. The four voter l channels are divided into two groups of two each, with each
- group of two providing inputs to one RPS trip system. The APRM System is-designed to allow one APRM channel, but no te nels, to be bypassed. A trip from any one i un passe RM will result in a " half-trip" in all four INSERT ' E" v ter chann s, but no trip inputs to either RPS trip
- system.9.A.t ip from any two unbypassed APRM channels will
)- ull-trip in each of the four voter channels, wh n urn results in two trip inputs into each RPS tri) IN S E R.T ' F " :::t e.4 ree of the four APRM channels and all four of tie
- voter chan els sre required to be OPERABLE to ensure that no i ure will preclude a scram on a valid signal. In l addition, to provide adequate coverage of the entire core, consistent with the design bases for APRM Functions 2.a, l 2.b, and 2.c, at least 17 LPRM inputs, with at least thr LPRM inputs from each of the four axial levels at c he ;
LPRMs are located, must be OPERABLE for each A channel. I l ! /NSERT 'G l 2.a. Averaae Power Ranae Monitor Neutron Flux ich
~
I i (Setdown) l 1
- For operation at low power (i.e., MODE 2), the Average Powe;-
- Range Monitor Neutron Flux - High (Setdown) Function is
- capable of generating a trip signal that prevents fuel
, damage resulting from abnormal operating transients in this l power range. For most operation at low power levels, the i Average Power Range Monitor Neutron Flux - High (Setdown) ! Function will provide a secondary scram to the Intermediate [ i Range Monitor Neutron Flux - High Function because of the i i l (continued) i
- HATCH UNIT 2 B 3.3-7 PROPOSED REVISION 7/16/96
Insert 'D' - Bases B 3.3.1.1 Average Power Range Monitor (APRM) Each APRM also includes an Oscillation Power Range Monitor (OPRM) Upscale Function which monitors small groups of LPRM signals to detect thermal-hydrauhc instabilities. l 1
.j Insert 'E' - Bases B 3.3.1.1 Average Power Range Monitor (APRM) i APRM trip Functions 2.a,2.b,2.c, and 2.d are voted independently from OPRM Upscale l Function 2.f. Therefore, any Function 2.a, 2.b, 2.c, or 2.d . .... .
)
Insert 'F' - Bases B 3.3.1.1 Averane Power Range Monitor (APRM) Similarly, a Function 2.f trip from any two unbypassed APRM channels will result in a full-trip from each of the four voter channels. l Insert 'G' - Bases B 3.3.1.1 Average Power Range Monitor (APRM) l For OPRM Upscale Function 2.f, LPRMs are assigned to " cells" of three detectors. A minimum of three cells, each with a minimum of two LPRMs, must be OPERABLE for
- OPRM Upscale Function 2.f to be OPERABLE.
I i l e i I __. .-
RPS Ynstrumentation B 3.3.1.1 BASES APPLICABLE 2.d. Averaae Power Ranae Monitor - Inoo (continued) SAFETY ANALYSES, LCO, and This Function was not specifically credited in the accident APPLICABILITY analysis, but it is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis. There is no Allowable Value for this Function. This Function is required to be OPERABLE in the MODES where the APRM Functions are required. 2.e. Two-out-of-Four Voter p mcludi % OPRM g-TheTwo-out-of-FourVoter(unctionp vides the interface between the APRM Functionwand the final RPS trip system , logic. As such, it is required to be OPERABLE in the MODES l where the APRM Functions are required and is necessary to i support the safety analysis applicable to each of those : Functions. Therefore, the Two-out-of-Four Voter Function is s required to be OPERABLE in MODES 1 and 2.
/ ,
All four voter channels are required to be OPERABLE. Each voter channel also includes self-diagnostic functions. If any voter channel detects a critical fault in its own
~ processing, an Inop trip is issued from that voter channel n to the associated trip system.
\ ~DMacT 14, .k There is no Allowable Value for this Function. T5W f3. Reactor Vessel Steam Dome Pressure - Hiah An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This causes the neutron flux and THERMAL POWER transferred to the reactor coolant to increase, which could challenge the integrity of the fuel cladding and the RCPB. No specific safety analysis takes direct credit for this Function. However, the Reactor Vessel Steam Dome Pressure - High Function initiates a scram for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power. For the overpressurization protection analysis of Reference 4, reactor scram (the analyses conservatively (continued) HATCH UNIT 2 B 3.3-11 PROPOSED REVISION 7/16/96
Insert 'H' - Bases B 3.3.1.1 2.e. Two-out-of-Four Voter The Two-out-of-Four Voter Function votes APRM Functions 2.a,2.b,2.c, and 2.d independently of Function 2.f. The voter also includes separate outputs to the RPS for the , two independently voted sets of Functions, each of which is redundant (four total inputs). ! Voter Function 2.e must be declared inoperable if any ofits functionality is inoperable. ! However, due to the independent voting of APRM trips and the redundancy of outputs, ;
- there may be conditions where Voter Function 2.e is inoperable, but trip capability for one I or more of the other APRM Functions through that voter is still maintained. This may be j i
considered when determining the condition of other APRM Functions resulting from 1 partial inoperability of Voter Function 2.e. Insert 'I' - Bases B 3.3.1.1 2.f. Oscillation Power Range Monitor (OPRM) Upscale 2 f Oscillation Power Range Monitor (OPRM) Upscale The OPRM Upscale Function provides compliance with GDC 10 and GDC 12, thereby providing protection from exceeding the fuel MCPR SL due to anticipated thermal-hydraulic power oscillations. References 14,15, and 16 describe three algorithms for detecting thermal-hydraulic instability related neutron flux oscillations: the period based detection algorithm, the amplitude based algorithm, and the growth rate algorithm. All three are implemented in the OPRM Upscale Function, but the safety analysis takes credit only for the period based detection algorithm. The remaining algorithms provide defense in depth and additional protection against unanticipated oscillations. OPRM Upscale Function OPERABILITY for Technical Specifications purposes is based only on the period based detection i algorithm. The OPRM Upscale Function receives input signals from the LPRMs within the reactor i core, which are combined into " cells" for evaluation by the OPRM algorithms. l l The OPRM Upscale Function is required to be OPERABLE when the plant is in MODE 1. Within the region of power-flow operation where anticipated events could lead l to thermal-hydraulic instability and related neutron flux oscillations, the automatic trip is enabled when THERMAL POWER, as indicated by APRM Simulated Thermal Power, is
> 25% RTP and reactor core flow, as indicated by recirculation drive flow, is < 60% of rated flow.
1
Insert 'I' - Bases B 3.3.1.1 2.f. Oscillation Power Range Monitor (OPRM) Upscale 2.f Oscillation Power Range Monitor (OPRM) Upscale (Continued) An OPRM Upscale trip is issued from an APRM channel when the period based detection algorithm in that channel detects oscillatory changes in the neutron flux, indicated by the combined signals of the LPRM detectors in a cell, with period confirmations and relative cell amplitude exceeding specified setpoints One or more cells in a channel exceeding the trip conditions will result in a channel trip. An OPRM Upscale trip is also issued from the channel if either the growth rate or amplitude based algorithm detects growing oscillatory changes in the neutron flux for one or more cells in that channel. Three of the four channels are required to be OPERABLE. Each channel is capable of detecting thermal-hydraulic instabilities by detecting the related neutron flux oscillations and issuing a trip signal before the MCPR SL is exceeded. There is no Allowable Value for this Function. i I I l I i
l l RPS instrumentation B 3.3.1.1 BASES ACTIONS expressed in the Condition, discovered to be inoperable or l (continued) not within limits, will not result in separate entry into ' the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable RPS instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable RPS instrumentation channel. A.1 and A.2 g Because of the diversity of se r i a le to provide trip signsis and the redundancy of the RPS design, an allowable out of service time 4f 12 hours has been shown to beacceptable(Refs.9agd13[topermitrestorationofany inoperable channel to $PERABCE status. However, this out of service time is only acceptable provided the associated Function's inoperable channel is in one trip system and the Function still maintains RPS trip capability (refer to Required Actions B.1, B.2, and C.1 Bases). If the inoperable channel cannot be restored to OPERABLE status within the allowable out of service time, the channel or the associated trip system must be placed in the tripped condition per Required Actions A.1 and A.2. Placing the inoperable channel in trip (or the associated trip system in trip) would conservatively compensate for the inoperability, restore capability to accommodate a single failure, and allow operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a fu cr Condition D must be entered and its Requir ton taken. As r,ctad, Action A. is t a e1[or APRM Functions 2.a 2.b, 2.c., ) (I d(Jnoperabilit of one required APRM channel affects both t i s , t us, Required Action A.1 must be satisfied. This is the only action (other than restoring OPERABILITY) that will restore capability to accommodate a single failure. Inoperability of more than " one required APRM channel of the same trip function results in loss of trip capability and entry into Condition C, as well as entry into Condition A for each channel. - (continued) HATCH UNIT 2 B 3.3-20 PROPOSED REVISION 7/16/96
RPS Instrumentation B 3.3.1.1 i
- BASES i I j
ACTIONS B.1 and B.2 (continued) Condition B exists when, for any one or more Functions, at ; least one required channel is inoperable in each trip system. In this condition, provided at least one channel ! per trip system is OPERABLE, the RPS still maintains trip capability for that Function, but cannot accommodate a i single failure in either trip system, l Required Actions B.1 and B.2 limit the time the RPS scram ! logic, for any Function, would not accommodate single , { failure in both trip systems (e.g., one-out-of-one d ! one-out-of-one arrangement for a typical fou nne cpd_l7 j Function). The reduced reliability of thi o ic l i arrangement was not evaluated in Reference % 1 ! ' l-12 hour Completion Time. Within the 6 hou an owan ,te i associated Functio ve all required e ERABLE ! s or in trip (or a combinat ) in one trip system. I dAMA 17
- Completing o of ese Re red Actions restores RPS to a i reliabilit evel e ival t to that evaluated in 1 i Reference 9g 13, w ch justified a 12 hour allowable out l i of servic t1me as sented in Condition A. The trip system in t degraded state should be placed in trip or, alternatively, all the inoperable channels in that trip i system should be placed in trip (e.g., a trip system with i two inoperable channels could be in a more degraded state j than a trip system with four inoperable channels if the two
- inoperable channels are in the same Function while the four 4 inoperable channels are all in different Functions). The j decision of which trip system is in the more degraded state i should be based on prudent judgment and take into account current plant conditions (i.e., what MODE the plant is in).
l If this action would result in a scram or RPT, it is permissible to place the other trip system or its inoperable chanrels in trip. 4 The 6 hour Completion Time is judged acceptable based on the remaining capability to trip, the diversity of the sensors
- available to provide the trip signals, the low probability of extensive numbers of inoperabilities affecting all 5
diverse Functions, and the low probability of an event requiring the initiation of a scram. b (continued) i HATCH UNIT 2 B 3.3-21 PROPOSED REVISION 7/16/96 1
_ _ _ _._ ___ _ _ .._. _ _ _. _ _ _ _ . _ ._.__ _ __ _ . _ .. _ _ _ _ _ -- _ .m 3 i > i ! RPS Instrumentati m 4 B 3.3.1.1 ! ! l BASES [ l 4 j ACTIONS B.1 and B.2 (continued) I i Alternately, if it is not desired to place the inoperable
- channels (or one trip system) in trip (e.g., as in the case
- where placing the inoperable channel or associated trip l system in trip would resu cram or RPT), Condition D 4 must be entered and i equirodAc}iontaken.
As noted, con M.T. l ion B is t appl for APRM Functions. \/ 2.a, 2.b, 2. , pd 2.d I 111ty of an APRM channel affects both rip sys j j specific trip , and is not associated with a as are the APRM two-out-of-four voter &( j i and other non-APRM channels for which Condition B applies. j l For an inoperable APRM channel, Required Action A.1 must be \ { , satisfied, and is the only action (other than restoring 4 . PERABILITY) that will r=< tare c=ambility to accommodate a l ,, p single failure. Inoperability ofFoore than one required 4 APRM channel results in loss of trip capabilityynd entry into Condition C, as well as entry into Conditioh A for each
/
i
- channel. Because Conditi and C provide Required
( j Actions that are appro ate f inoperability of APRM _) ! Functions 2.a. 2.b, i .c, lyd 2.d, a these Func1 ions are i j not associated with ! ecific tri stems as are he APRM p I two-out-of-four vote and other -APRM i i Condition B does apply. , [ s i f.d dMd oTM i i 2 j Required Action C.1 s inten e o ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same trip system for the same Function ) i result in the Function not maintaining RPS trip capability. ' A Function is considered to be maintaining RPS trip , l- capability when sufficient channels are OPERABLE or in trip l l (or the associated trip syste'n is in trip), such that both ! trip systems will generate a trip signal from the given { Function on a valid signal. The Completion Time is intended to allow the operator time to evaluate and repair any discovered inoperabilities. The ! I hour Completion Time is acceptable because it minimizes { risk while allowing time for restoration or tripping of j channels. j (continued) d h 1 HATCH UNIT 2 B 3.3-22 PROPOSED REVISION 7/16/96 c
RPS Instrumentation B 3.3.1.1 i BASES ACTIONS QJ (continued) Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.1-1. The applicable Condition specified in the Table is Function and MODE or ; other specified condition dependent and may change as the l Required Action of a previous Condition is completed. Each ! time an inoperable channel has not met any Required Action l of Condition A, B, or C and the associated Completion Time has expired, Condition D will be entered for that channel and provides for transfer to the appropriate subsequent Condition. E.1. F. 4 M G.lt Iand J.1 If the c s no re tored to OPERABLE status or placed in trip (or the associated trip system placed in j trip) within the allowed Completion Time, the plant must be l placed in a MODE or other specified condition in which the LC0 does not apply. The allowed Completion Times ar reasonable, based on operating experience, ! a th specified condition from full power ons in an orderly manner and without challenging plan syj;tems. In addition, the Completion Time of Required Ac 'onvE.1(H- consistent gg ! , with the Completion Time provided i LC0 3.2.2, " MIN MUM CRITICAL POWER RATIO (MCPR)." i L1 If the channel (s) is not restored to OPERABLE status or placed in trip (or the associated trip system placed in trip) within the allowed Completion Time, the plant must be l placed in a MODE or other specified condition in which the LCO does not apply. This is done by immediately initiating action to fully insert all tasertable control rods in core cells containing one or more fuel assemblies. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and are, therefore, not required to be inserted. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted. W inse+ *7" mw (continued) HATCH UNIT 2 B 3.3-23 PROPOSED REVISION 7/16/96 l
j Insert 'J' - Bases B 3.3.1.1 Actions I.1 and 1.2 i L1 p: If OPRM Upscale trip capability is not maintained, Condition I exists. Reference 13 I justifies use of an alternate method to detect and suppress oscillations for a limited period l of time. The alternate method is procedurally established consistent with the guidelines identified in Reference 18 requiring manual operator action to scram the plant if certain ! predefmed events occur. The 12 hour Completion Time is based on engineeringjudgment ; l - to allow orderly transition to the alternate method while limiting the period of time during i
- which no automatic or alternate detect and suppress trip capability is formally in place.
- Based on the small probability of an instability event occurring, the 12 hour Completion Time isjudged to be reasonable.
L2 The alternate method to detect and suppress oscillations implemented in accordance with I.1 was evaluated based on use up to 120 days (Ref.13). The evaluation, based on engineering judgment, concluded that the likelihood of an instability event that could not be adequately handled by the alternate methods during this 120 day period is negligibly small. The 120 day period is intended to be an outside limit to allow for the case where design changes or extensive analysis may be required to understand or correct some j unanticipated characteristic of the instability detection algorithms or equipment. This i action is not intended to be, and was not evaluated as, a routine alternative to returning failed or inoperable equipment to OPERABLE status. Correction of routine equipment ! failure or inoperability is expected to normally be accomplished within the Completion Times allowed for Required Actions for Conditions A and B. I l l I i
1 RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.9 and SR 3.3.1.1.12 REQUIREMENTS i (continued) A CHANNEL FUNCTIONAL TEST is performed on each required i channel to ensure that the entire channel will perform the ! intended function. Any setpoint adjustment shall be i consistent with the assumptions of the current plant specific setpoint methodology. The 92 day Frequency of
- SR 3.3.1.1.9 is based on the reliability analysis of-i Reference 9.
j The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant } outage and the potential for an unplanned transient if the i Surveillance were performed with the reactor at power. , Operating experience has shown that these components usually i pass the Surveillance when performed at the 18 month l Frecuency. SR 3.3.1.1.10 A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the [)
- intended function. For the APRM Functions, this test
, supplements the automatic self-test functions that operate continuously in the APRM and voter channels. The APRM i
- l. CHANNEL FUNCTIONAL TEST covers the APRM channels (including I i recirculation flow processing - applicable to Function 2.b only), the two-out-of-four voter channels, and the interface (
! connections to the RPS trip systems from the voter channels. l ? Any setpoint adjustment shall be consistent with t ) I assumptions of the current plant specific set methodology. The 184 day Frequency of SR 3 . 1.10 is f based on the reliability analysis-of Refer nc 3,vtNOTE: The actual voting logic of the two-out-of- dier channels is tested as part of SR 3.3.1.1.15 , g}} l For Function 2.a. a Note that requires this SR to , performed within 12 hours of entering MODE 2 from MODE 1 is t provided. Testing of the MODE 2 APRM Function cannot be performed in MODE 1 without utilizing jumpers or lifted ( i l leads. This Note allows entry into MODE 2 from MODE 1 if f the associated Frequency is not met per SR 3.0.2. (continued) HATCH UNIT 2 B 3.3-28 PROPOSED REVISION 7/16/96
.. . . - _- = .- - . - -
\
l RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued) REQUIREMFNTS Operating experie has shown that these components usually , pass the Surve ance ormed at the 18 month ! Frequency.
@ OfM The LOGIC STEM F NCTI T for APRM Function 2.e simulates RM'tr p c rditions at the two-out-of-four voter channel inpu o check all combinations of two tripped \/
inputs to the two-out-of-four logic in the voter channels N
! and APRM related redundant RPS relays. /
SR 3.3.1.1.16 i This SR ensures that the individual channel response times are less than or equal to the maximum values assumed in the accident analysis. This test may be performed in one , measurement or in overlapping segments, with verification '
- that all components are tested. The RPS RESPONSE TIME acceptance criteria are included in Reference 10.
RPS RESPONSE TIME for APRM two-out-of-four Voter
- Function 2.e includes the output rehys of the voter and the associated RPS relays and contactors. (The digital portions of the APRM and two-out-of-four voter channels are excluded from RPS RESPONSE TIME testing because self-testing and calibration check the time base of the digital electronics.)
Confirmation of the time base is adequate to assure required response times are met. Neutron detectors are excluded from (/
/
RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. Note 1 allows neutron detectors to be excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. Note 2 allows channel sensors for Reactor Vessel Steam Dome Pressure - High and Reactor Vessel Water Level - Low, Level 3 (Functions 3 and 4) to be excluded from RPS RESPONSE TIME testing. Thi.; allowance is supported by Reference 12 which concludes that any significant degradation of the channel sensor response time can be detected during the performance of other Technical Specifications SRs. RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS Frequency to be determined based on four channels per trip system, in lieu of the eight channels specified in (continued) HATCH UNIT 2 B 3.3-31 PROPOSED REVISION 7/16/96
RPS instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.16 (continued) REQUIREMENTS Table 3.3.1.1-1 for the Main Steam Line Isolation . Valve - Closure Function. This Frequency is based on the , logic interrelationships of the various channels required to produce an RPS scram signal. This Frequency is consistent with the typical industry refueling cycle and is based upon 4 plant operating experience, which shows that random failures , of instrumentation components causing serious response time de , not channel failure, are infrequent urrences. . tcose d K" ! REFERENCES 7.2.
- 2. FSAR, Chapter 15.
l 3. FSAR, Section 6.3.3. I
- 4. FSAR, Supplement 5A.
- 5. FSAR, Section 15.1.12.
- 6. NEDO-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
- 7. FSAR, Section 15.1.38.
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram
- Discharge System Safety Evaluation," December 1,1980.
$ 9. NED0-30851-P-A, " Technical Specification Improvement . Analyses for BWR Reactor Protection System," March 1988.
- 10. Technical Requirements Manual.
- 11. NRC No. 93-102, "Finai Policy Statement on Technical Specification Improvements," July 23, 1993.
- 12. NED0-32291, " System Analyses for Elimination of Selected Response Time Testing Requirements,"
January 1994.
- 13. NEDC-32410P-A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
( ( tto Option III Stability Trip Function," p Octo5er 1995. (' NET C HATCH UNIT 2 M B 3.3-32 PROPOSED REVISION 7/16/96
Insert 'K' - Bases B 3.3.1.1 SR 3.3.1.1.17 SR 3.3.1.1.17 This SR ensures that scrams initiated from OPRM Upscale Function 2.fwill not be inadvertently bypassed when THERMAL POWER, as indicated by APRM Simulated Thermal Power, is 2 25% RTP and core flow, as indicated by recirculation drive flow, is
< 60% rated core flow. This normally involves confirming the bypass setpoints.-
Adequate margins for the instrument setpoint methodologies are incorporated into the actual setpoint. The actual surveillance ensures that the OPRM Upscale Function is enabled (not bypassed) for the correct values of APRM Simulated Thermal Power and recirculation drive flow Other surveillances ensure that the APRM Simulated Thermal , Power and recirculation flow properly correlate with THERMAL POWER and core flow, ; respectively. If any bypass setpoint is nonconservative (i.e., the OPRM Upscale Function is bypassed when APRM Simulated Thermal Power is 2 25% and recirculation drive flow is < 60% rated), the affected channel is considered inoperable for the OPRM Upscale Function. , Alternatively, the bypass setpoint may be adjusted to place the channel in a conservative l condition (unbypass). If placed in the unbypass condition, this SR is met and the channel is considered OPERABLE. The 18 month Frequency is based on engineeringjudgment and component reliability. l i i i ) l l l
Insert 'L' - Bases 3.3.1.1 References
- 14. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
- 15. NEDO-31960-A, Supplement 1,"BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," November 1995.
- 16. NEDO-32465-A, "BWR Owners' Group Long-Term Stability Detect and ,
Suppress Solutions Licensing Basis Methodology and Reload Applications," March 1996.
- 17. NEDO-32410P, Supplement 1," Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," May 1996.
I 8. Letter, L. A. England (BWROG) to M. J. Virgilio, "BWR Owners' Group Guidalines for Stability Interim Corrective Action," June 6,1994. i l l l
Recirculation Loops Operating i B 3.4.1 l 1 BASES APPLICABLE case (since the intact loop starts at a lower flow rate and SAFETY ANALYSES the core response is the same as if both loops were (continued) operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgement. The recirculation system is also assumed to have sufficient
. flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2),
which are analyzed in Chapter 15 of the FSAR. A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3). The transient analyses of Chapter 15 of the FSAR have also been performed for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain. fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power range monitor (APRM) instrument setpoints is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR setpoints for single loop operation are specified in the COLR. The APRM Simulated l Thermal Power - High setpoint is in LC0 3.3.1.1, " Reactor Protection System (RP tr ntation." l Recirculation loop operating s isfies Criterion 2 of the l NRC Policy Stateme (Ref.f). ! Jv LC0 Two recirculation loops are normally required to be in ; operation with their flows matched within the limits l specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. Alternately, with only one recirculation loop in operation, modifications to the required APLHGR limits (LC0 3.2.1,
" AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits (LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)"),
(continued) HATCH UNIT 2 B 3.4-3 PROPOSED REVISION 7/16/96 J
i 1 l l Recirculation Loops Operating '
- B 3.4.1 BASES i _ A V
- LCO and APRM Simul ed The Po -
h (continued) (LCO 3.3.1.1) must be applied to allo continued op ai 1 consistent with the assumptions of Reference 3. 1 4 d.. e. A, ,. .. s. i m. - n.. Zs4.. A. /.,-.s.. .Z.,*n'...
/.. Z.,e. .,A.4 .
a .u.. _ .mZ .. 4 . 44u , . . . .Z.,. . M. .,.. Z.. l '.
- 2. 7 ./"...'U. .". '. '2. .-
. ... "u.], . ,. ' Zi. I,."Z. . ;I. 'i.',,.4. 4. , C. .'Z. .'
'. ",7.
%. ,. '/. .'."
i APPLICABILITY In MODES I and 2, requirements for operation of the Reactor
- Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting 4
design basis transients and accidents are assumed to occur. In MODES 3, 4, and 5, the consequences of an accident are ! reduced and the coastdown characteristics of the i e recirculation loops are not important. a A _ ACTIONS ,,./, 2...,/. d Y n s - --- lu.,. .,a. . .d.
. . . . . . . m..,.j.4. s.,.. ... j..,... J.. . c. . . . .
- 4. . > .s.,. . . . .L. Z. /.. s/. ~ d,,Z,,,E,.. 4 su .'.. c, J.
... 4..
f .Z..+.4 y 2T. d.... Z'.u.4.>.. .
.c . 4.A.. 'a...s... , e, L..L m., ..x
. . a.su. u, i
E ,! 3 ~ 2~., A... . .R..
- f. T lJ,,Z.7.
A7; ~ ' ' " s. .". ./,. . A. . .'",./. Z. ' .~4. /42. ." '. ' 'ri. .I. ". "" , .. ~
. 4. .. u . 7.. sum ...--_
2
- s. 4 . . ...+,4...
. . . . . . 4. ... c e i
a
,o,.C.,an.a..L......a. 4.. ,.. A.s.
i n.. c.,. Z.......,d.__..,
.2..
., 4.
a..
. c. J.
- 4. . . .
,.. . . + Z...,.....,.4..
! / ou... s.,k.. 4v. 411.sI... ..,,m< y.f.1. . ,. y 6. I f 'y . c:re1 4. .7. ... . u,
... y ./.... 1,4.t,.v.u,the- :' y:::r i' y-
! of 4
+
. . . n.. : . ,M..,. u.. ,.:s
, . . . . . . . ,.% . ,.lo.
.,,m.unt'
. . .,. 4. ,7,,
,,,nc,. :e. 4..,.... 4. .A. 1._1. y Z.
.. ,.n...+
! /s f. t 4 . wG...ala k/ 4.444 M a +. ./cs... sk. m.. . m... wr
.E[ .. .S, [ ,4, u .u...) n,E,E[ E,E, ,E,1,1, ,b. . . , ,n., .E.,5, E 7, T. k. .
4 -- .E.EN...,7. ,. E,57. 7, '7 7 ! 2 ',tr C;- ilit o
/:d.,ider.+J[Ec,eticy urr g durir.i= A.::d i: tim :n ty .ow pro ^
- re i
.4. . .j. s f,. .e. .. s.......
u
, +7. L.. n Z4 his Z,, cried, f. c
.s. ,n,,. s 4 ,, . . ,
7 :r
. ,.4
- g1:ti
. . . .L. . .=Z.
.7 ..: t
.' s. . , .
g: .
- -ar . r /
- tedp=#$.::110ir
;;d ek'y [ir y7.y ..i = di e reset r.
f j r:';ui . with r fa cire'. On recircul:',en h:: ,4est pumpj[.n ip. Op 1'
. .ty hy o = m = 1..:11 = .g=., .j ,. the reb:7,li 4
1,..- . I k I i i i
)
i 4 (continued) i $ HATCH UNIT 2 B 3.4-4 PROPOSED REVISION 7/16/96 e-g, -- - + - - + sw
i Recirculation Loops Operating B 3.4.1 l ACTIONS
;he LCO not me 6 n..,,,, the recirculati loo s must i
s estored to opera,ti matched flows wi n s. A j r ulatio oo s considered not in operation when the i pump in a loop is idle or when the mismatch between total ! jet pump flows of the two loops is greater than required . i limits. The loop with the' lower flow'aust be considered not , in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant j core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore the. l inoperable loop to operating status. ! Alternatively, if the single loop requirements of the
- LCO are applied to operating limits and RPS setpoints,- -
- operation with only one recirculation loop would satisfy the i requirements of the LCO and the initial conditions of the; ' -
i accident sequence.
- The 24 hour Completion Time is based on the low probability 1
of an accident occurring during this time, period, on az reasonable time to complete the Required Action, and on-l frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. i s This Required Action does not require tripping the ] recirculation pump in the lowest flow loop when the mismatch
- between total jet pump flows of the two loops is greater i than the required limits. However, in cases where large i flow mismatches occur, low flow or reverse flow can occur in j the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition i should be alleviated by changing pump speeds to re-establish j forward flow or by tripping the pump.
i A a y Rep ed,Ac ion and associated Completion Time of Condit n w ,no met, the plant must be brought to a MODE in which the L 0 does not apply. To achieve this status, t M p1pnt ust be brought to MODE 3 within 12 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of Design (continued) HATCH UNIT 2 B 3.4-5 REVISION 0
k 1 Recirculation Lceps Operating 4 l B 3.4.1 i BASES i ACTIONS M continued) j B s Accidents and minimal dependence on the recirculation oop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. l SURVEILLANCE SR 3.4.1.1 REQUIREMENTS l i This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < 70% of rated core flow), the MCPR requirements ! provide larger margins to the fuel cladding integrity Safety i Limit such that the potential adverse effect of early - i,. boiling transition during a LOCA is reduced. - A larger flow ' j mismatch can therefore be allowed when core flow is < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance, is the summation of the flows ! - ' ^ from all of the jet pumps associated with a single recirculation loop. i The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits,
, , . the loop with the lower flow is considered not in operation.
i The SR is-not required when both loops are not in operation i since:the mismatch limits are' meaningless during single loop i or natural circulation operation. The Surveillance must be j performed within 24> hours after both loops are in< operation. i
- The 24 hour Frequency is consistent with the Surveillance I j Frequency for jet pump OPERABILITY verification and has been j shown by operating experience to be adequate to detect off j -
normal jet pump loop flows in a timely manner. _SR 3.4.1.2 W
- v w .4 . ... ..a. su. . 2- a, . . . a..--s ud --- ac-, s
'2"" 7".Zv.';;"'Z:'.'. 7 '" a '. '.'"KZ"C : 7'""'/'"'
t'r.:2".;"'.Mi' ;s C" iCI'.'2UlT.','IX:..- / r.r:rstrr'":s zi'Erz.ji,Zvar mi l
76 T.E ' '? '"';'#"" "C' "" ' 2G": . ' Z .- 2 /
7" z'." V J iit ZM"'EZZ"lT;;"IE ii'/ 7 fi d [ U M IE E5h E "$ 5'[E E EEE 3 5 $ $5 in (continued) HATCH UNIT 2 B 3.4-6 REVISION 0
l l i l Ricirculatien Lo:ps Operating j B 3.4.1 , i + BASES ( ! MAN n n SURVEILLANCE c o., , y A,...e f v
- a. . s... a.. . . .. .. .n, -
v g y V W 1 3 REQUIREMEhTS . I / f..:f ...a s=C s 4, whic"' w. u
.: :::d te d ;;r.d [ ;;r: tic.r in th::
u... e /.--. du...a .- Z.ss.
- Z.4.....Z.E.. '. 4 . ...' E....' Z.Z./. 'L/f,"'Z'. " LZZ.
T 'I'.
'. ~
. . .:u. _ ..F. , . '.!.Z.' "M.. .
< Z,..
...s.., , a. a. . . .,. .
4
. a s. .., . a. . . . e. a. . . s. . d. .'. . .
>,A. . . . . . .
.-..,.y.. n. .. .e n .J. ma..... . . c. a . A ,... . u s. .. s x...s.
- u. . -. . Za ,. , .- .
4
.1-- Z,,..,.:.;:..:..:.tter.
. s. . _ . . . .f. s...u.s. ea.y A. ..
epu...e...
< te<. '..y.4...,
r withe",.
.< J...,>u.
,re;etri y ....,..
] , .e.n, ... _ ..f..-.--
..d. a.1
.1
. ..- ..a
- m. . 1. s.d.
p.. 4.,.-
.a_--. . Z.
, ..,. . .. .Z.- A" i, . < .....,,
.. . . s,a . . , . . . . . . ,a..u. s.7L.
..u.
y.t." ..
. . v...,- - - a. .
a ep .,.ing). e.. l pdrible ofte[.:nteri - ::r, ;t[hegle-le;d.p:rf:enditi th::!d y ...,.E ... ,d :: [n - ]::- i
'h-REFERENCES 1. NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," ,
December 1986.
- 2. FSAR, Section 5.5.1.4.
.3. NED0-24205, "E.I. Hatch Nuclear Plant Units 1 and 2.
e- op n " Au ust '
-.22.. d - , a- u. W n-.W. , l
- . .r n> . -"c -
- /. m. wi .... .Zer. .n"r=..f. r... . .r. 7'.m. ./. . x/. =. . . ,
, l
, . . , . , , a,,A_ a. .. . ,.
4[. C No. 93-102, Specification Improvements," July 23, 1993. inal Policy Statement on Technical I i l
)
HATCH UNIT 2 B 3.4-7 REVISION 0}}