ML20137L130
| ML20137L130 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 05/31/1995 |
| From: | FLORIDA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20137K821 | List:
|
| References | |
| FOIA-96-485 JPN-PSL-SENP-95, JPN-PSL-SENP-95-049, JPN-PSL-SENP-95-49, NUDOCS 9704070159 | |
| Download: ML20137L130 (28) | |
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1 JPN-PSL-SENP-95-049 REVISICN 0 PAGE 2 OF 27 i
REVIEW AND APPROVAL RECORD - - -
1 PLANT _ _ _
ST. LUCIE UNIT I
, TITLE ALTERNATE NIS EXCORE DETECTOR ARRANGEMENT i
l LEAD DISCIPLINE NUCLEAR i
t I i ENGINEERING ORGANIZATION PRODUCTION ENGINEERING GROUP REVIEW / APPROVAL:
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JPN-PStr8EMP-95-049 REVISION O PAGE 3 0F 27 TABLE OF CONTENTS l
EECTION TITLE PAGE
-- Cover 1 -
- Review and Approval Record 2 i
3
~~ Table of Contents )
- Abstract 4 1.0 Purpose and Description 6 2.0 Licensing Requirements 7 3.0 Analysis of Effects on Safety 9 4.0 Failure Modes and Effects Analysis 20 5.0 Plant Restrictions 20 6.0- Effect on - .ical Specificatiot.s. 21 7.0 Unreviewed Safety Question 22 Determination 8.0 Actions Required 24 9.0 References 26 Kttachmentas 1 Cable Routing Instructions 2 pages Fleures 1 Detector Well Locations 2 Cable Routing: Option 1 i
3 Cable Routing: Option 2 4 Jumper Routing: Option 1 & 2
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i JPN-pSL-SENP-95-049
! REVISIoM 0 PAGE 4 OF 27 i
l ABSTRACT I
! The St. Lucie Unit i reactor protection system (RPS) normally operates with two-out-of-four (2/4) reactor trip logic. Since i
channel D is out of service due to shorting of its cable and/or j detector, the RPS is in one-out-of-three (1/3) trip logic as
- required by Technical Specifications. With the RPS in 1/3 trip
- logic a spurious signal will result in reactor trip and unnecessary <
l exercise of plant safety systems.
4 This safety evaluation provides an assessment of an alternative NIS excore detector arrangement for the RSS. This involves connecting the detector signal from power range control channel 2 to linear
- power range channel D. This restores channel D to operable status and reactor trip logic from 1/3 to 2/4 coincidence. This i
j arrangement is feasible since the linear power range and power range control channels use the same design for detectors and i
! cables.
The use of the alternative arrangement involves changes to certain plant systems, including the safety related nuclear instrumentation i system. The change to the nuclear instrumentation system is the
- detector location. The control channel de
- ector is located at 202 l
degrees and th7 detector for the failed channal is at 255 degrees.
The effects of the change in detector location ce such items as i
j axial. shape.index, - + hat power tilt and limiting safety system i
settings have been secu sined to be acceptabloc
! other plant systems affected are the reactor regulating system, the low power feedwater system, the Power ratio calculator and recorders JR-010 & 012. These systems are classified non safety j
j related. The effects on each of these systems with respect to safe
! plant operation have been evaluated and determined to be l acceptable.
The .-Iternative arrangement may be implemented using either of two i cable routing options. Option 1, which may be installed at power, l
involves installing a jusper in the control room and running the D
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channel cable in the same raceway as the E channel cable (both powered from train B) for a portion of the cable route. This modified routing for channel D has been evaluated with respect to
- safe plant operation and found to be acceptable. This i
configuration will only remain in place until the next planned or j unplanned unit shutdown. During the next shutdown, options include
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restoring the detector to service or implementing the second option
! covered by this evaluation.
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l Option 2 involves installing a jumper in the containment instrument
- tunnel applying the existing channel swparation criteria'for the J
entire channel cable route. - - - -
! This safety evaluation demonstrates that safe plant operation is
! maintained using either of the two options. It neither involves an
- unreviewed safety question nor requires a change to Technical j specifications. Therefore, implementation may be performed without j prior NRC approval.
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it JPN-PSL-SENP-95-049 i REVISION 0 i PAGE 6 0F 27 1.0 DBasosa men n:;merion Intreduet(gg 1 This safety evaluption provides for an instrumentatio.1 system alternative nuclear substitutes power range (NIS) excore detecter - rangement which !
range channel D (LRD) control channel 2 (CC2) for linear power Lucie Unit 1 (see Figure 1).in the reactor protection system (RPS) of St channel D to operable status and reactor trip logic from one-out-Th4:
of-three (1/3) to two-out-of-four (2/4) coincidence.
The status alternative due to shorting arrangement is necessary of its cable since the LRD is in trip and/or detector. !
located and cannot between the containment be repaired penetration with the reactor at power. andThetheshort reactoris cavity l i
Specifications require an inoperable af t.sr 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to either be restoredlinear power range channel Technical placed in trip status. to operable status or be status, the RPS is using 1/3 coincidencerange With a linear power channel in trip for trip logic, and therefore, spurious l
will result in reactor trip and unnecessary exercisesignals or failures l
safety systems. of plant will be scheduled for the next refueling outage. Replacement of the ch The cables control as thechannel detee' utilises the same Class 1E design detectors and power range channels A, B, C & D..d cables used in tne safety related linear requirements and qualifications of the linear power range and powerF range provided control in Sections channel 2 &cables,
- 3. power supplier and sensors are detector to channel D.The alternative arrangement involves connecting a con different cable routing It may be implemented using either of two options as shown in Figures 2 and 3.
Option implemented 1 provides in Modefor 1. the connection in the control room and may be the containment building and may be implemented in Node 3 or 4 Further detail on the cable routing options .
Installation instructions are contained in Attachment 1.as provided below.
In the alternative arrangement, detectors would not provide inputone therefore, of control to its the control channel channel, and affected. the systems supported by that control channel are safety related functions, the effects are considered acceptab discussed in Section 3'and 6.
be supplying input to the RPS.Using the alternative arrangement, the LRD be providing the input signal. In its place, the CC2 detector will The location of the excore detectors in detector degrees (#10) and LRD at 255 degrees (#8) wells around the asreactor shownvessel in Figure is1.CC2 at 202 The effects of this change in detector location on such items as axial shape index, azimuthal power tilt and limiting safety system settings have been determined to be acceptable.
The effect on each
JPN-PSL-SENP-95-049 REVISION O PAGE 7 OF 27 of these items and others are explained in Sections 3 and 6.
cable Routina Ontions option 1 (Figure 2) option 1 involves connecting the output from the CC2 detector to the LRD drawer in the control room. Three coaxial cables (i.e. ,
upper signal, lower signal and high voltage) are disconnected from RTGB 104, jumpered to the RPS and connected to the LRD drawer i inside the RPS MD cabinet. Cables are uupported by structural '
steel and attached with approved tie-wraps. Option 1 may be implemented in Mode 1. Installation instructions are contained in Attachment 1. ,
Option 2 (Figure 3)
Option 2 involves connecting the output from the CC2 detector to the LRD output signal cable in the containment building. The CC2 ,
l detector cables are disconnected from the CC2 detector in its !
junction box and reconnected to the LRD, cables in the LRD junction j box within the electrical tunnel of the containment building.
] option 2 may be implemented in Mode 3 or 4. Installation j instructions are contained in Attachment 1. l Prior to implementation of option 2 the location of the short must i: be confirmed to be '
the detector. This option cannot be implemented if the su , is between the detector junction box and the containment penetration.
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3.O LICEN8ING REOUIREMaarts j Technical Snecifications i
The Unit 1 Technical Specifications include the following limiting conditions for operation (LCOs) and surveillances which pertain to the use of the alternative NIS excore detector arrangement:
- 1) T/s 3.3.1.1 (Ref 9.2, page 3/4 3-1)
This Technical Specification addresses the operability requirements for the linear power range channels and is included to explain the required actions related to an inoperable channel.
- LCOs "With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
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- a. The inoperable channel is placed inFor either the purposes of the Ly g d or tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
testing and maintenance, the inoperable channel any be i
l bypassed for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> from the time of initial loss of OPERABILITY; however, the inoperable channel aball i
then be either restored to OPERABLE status or placed in 1
the tripped condition."
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- 3) T/8 4.2.4.3 (Ref 9.2, page 3/4 2-12)
This Technical Fpecification provides for surveillance of azimuthal 1
power tilt using the incore detector system when an excore channel e is inoperable. -
i - semesrLLames:
2 "The ASINUTHAL POWER TILT shall be determined to be within the
' limit by:
- c. Using the incore detectors to determine the AZIMUTHAL POWER TILT at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when one excore channel is inoperable and THERMAL POWER is greater than 7st of RATED THERNAL POWER."
t l 3) T/S 4 3.1 3 (aaf 9 3, page 3/4 2-1) i This Technical Sys.. ,stion provides for the surveillance of i linear heat rate when using the excore detector system option to y
l monitor axial shape index.
- scavarummes:
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" .. Verifying that the AXIAL SHAPE INDEX is maintained within
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the allowable limits..."
- 4) T/8 4.3.3.4 (Ref 9.3, page 3/4 2-10) i j
This Technical specification provides for surveillance of total
- integrated radial peaking factor using azimuthi.1 power tilt.
- somenrummes i; F', is made "T, shall be determined each time a calculation of using a non-full core power distribution analysis code. The used to determine F', in this case shall be the j value nessuredof vaT,lue of T,."
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- 5) T/S 4.3.5.1 (Ref 9.3, page 3/4 2-13) i j This Technical Specificatic provides for surveillance of AXIAL SHAPE INDEX using the excore detector system.
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gyyg-pst,-SENP-95-049 i I
REVISION 0 PAGE 9 0F E7
- scRVEZI&ANoss "Each of the parameters ... [ axial shape index) ... shall be .
verified to be within their limits by instrument readout at ~
least oncw per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />." ,
bion 6.
The effects on Technical specifications are discussed i.1 Se ISAE The reactor protection system (RPS) and the nuclear instrumentation system (NIS) are described in section 7.3 of the FSAR (Ref 9.1) the.
and The RPS complies with the requirements of 1EEE 279-1971 General Design criteria as discussed in Sostion 7.2.2 of the FSAR.
These requirements include independence of the protective channels, which is metThe in part by physical separation of wiring and i
components.
system and component separation is described in .
Section 3 addresses the specifics of the FSAR chapters 7 and 8. regard to electrical proposed alternative arrangement with separation. .
OM saFETT 3.0 EntYSTE OF arraum i
Li===r ;eJer mance mma ;;;;g names centrei ekammel f.ematiens f
The NIs consists includes the linearof p -the four power range, sets ofrange azoore detector control, widechannels range which stor system. Thess detector channel sets and the excore neutrc..
provide the means for control of the fission process by monitoring core power levels and generating the appropriate trips and alarms for various phases of reactor operating and shutd/Am conditions.
' The four liticar power range channels are dedicated safety channels in that they provide a linear These power inputssignal are provided to various byRPS dual trip and section calculator functions. The ion chamber signal is converted, uncompensated ion chambers. ,
amplified and used byThe indication uses forand theprotectionoutput of linear equipment powerlocated range in the control room. 1 channels are outlined below:
Input is used by the core protection calculator to determine 1.
axial shape index (ASI). ASI is an input to the thermal marr;in ' low pressure and the local power density trips, and is indkomced on the RPS cabinets.
Input is used for the determination of Q, (the highest of nuclear power or thermal delta power, AT). Q , is input to 2.
l thermal margin / low pressure trip, high local power density is also used in the power trip, and high power trip. Q,,,
i ratio calculator for power level determination.
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- REVISION 0 l PAGE 10 0F 27 )
- 3. Input is used to establish the permissive signals for high start-up rate trip disable circuntry, reactor trip enable on turbine trip, and the high local power density trip.
- . 4. Input is used by the CIA drop detection circuit to. provide rod drop light indication on the NI drawer front paneT,~~and is i used in the dropped CIA alara (K-
- s7) and the auto withdrawal prohibit alarm (K-14) .
S. Input is used in the determination of the sub-channel !
deviation (Mi & Ni-M1) alar.a (L-20) . -
l The two power range control channels located in the reactor i
! regulating systes (Ras) cabinets son' tor reactor power and provide '
i a linear power signal that is utilized by the RRS. The linear l
power signal is provided by dual section uncompensated ion chambers
- of the same design as the linear power range channel detectors.
j The uses for the power range control channels are outlined below:
I 1. Input is used by the RAS for automatic rod control between 15%
and 1004 power.
! 2. Input is used by the low power feedwater control system for 4
8/G level control between 24 and 15% reactor power.
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- 3. Input is used by the power ratio calculator to determine the measured power rette. Note that the asasured power ratio is another name fe- *ial shape index determined using the input from the conh , channels. The measured power ratio is compared with ASI limits in the power ratio calculator to j verify compliance with the DNS LCo. A reactor power deviation i
alarm (L-34) is annunciated if the asasu:.ed power ratio is
- outside of ASI limits. The ASI limits and measured power
! ratio are recorded on JR-012.
l 4. Input is recorded as percent power on JR-010.
1 l Outlined below are those items related to the alternative NIS
- excore detector arrangement that have been evaluated for their
- potential effect on safety
i j h qar power manse mad power mamme control r*hannel cualification I
! The CC2 and LRD detectors currently in use have the same model #,
M&S # and are part of the same safety related purchase order (Raf j 9.4 & 9.7). The CC2 and IJtD cables (Ref 9.19, 9.21) and connectors
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provided by vendor (CE) are functionally equivalent and qualified for their intended function. Therefore, the linear power range detectors used by CC2 are valid substitutes for the LRD.
I j The CC2 detector support tube is seismically designed (Ref 9.11),
and is therefore acceptable f or use in the alternative arrangement.
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- 1 JPN-PSL-SENP-95-049 REVISION O PAGE 11 OF 27 i
i j The cables and connectors to be installed are qualified for classDu i 13 service. the cables will not be adversely affected by any connection points,iated with a seismic event. Installation of the
- displacement assoccables will not adversely affect the structural integrity of any attachment peint due to the negligible weight of the--sables (approximately 0.2 pounds per foot for 3 coaxiaA cables, Ref 9.12) .
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t.t === e r .;_r annae chammel Desien ameruirements timetrieal Train Senaration Section 8.3.1.2.3 of the FSAR states that physical separation is l
provided between the electrical A and B (as well as the swing AB) l load groups for protection against Physical loss of the A and separation of B trains load electrical fron a single f ailure or event.
i groupe is provided by theThe useRPS of spatial separation measurement and/orsatisfy channels erectionthis
! of physical barriers. j I requirement by physically separating the MA and MC channels from the ME and MD channels which are powered from the A and B trains,
! The cable routing options for the alternative respectively. arrangement satisfy the A'and B train separation criteria.
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N annal Ynda==ndance and Senaration I
The reactor protective system is designed to the requirements of !
l IEEE 279-1971 and the General Design FSAR criteria which 7.2.2.2.6 Sections include 4
independence of the pr * +1ve channels.
. reactor protective system measurement and 7.2.1.1 channels arestate that independent and isolatedincluding and that the the cables are separately from each other elentrical routed penetration area. Physical .eparation as8.3.1.2.3, a p. art of" crennel Physical l
independence is defined in FSAR Section i
separation". It defines the minimum spatial separation criteria 4' i for uncovared redundant raceways to be 18" horizontally and cannot be Where spatial separation i criteria i vertically.
maintained, solid tray covers are provided for low level signal i
trays 'Ref 9.19) . Using covered cable trays, theThe minimum cable routing spatial i
separation criteria is reduced to 6" (Ref 9.19) .for option 2 satisf i The jumper cable for option 2 is an extension to the existing LRD Therefore, detector. The jumper cable is not routed in conduit.
- the jumper cable required between the LRD otherjunction RPS box i
i Is" horizontal separation from redundant A train and 1
channel cabling.
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JPN-PSL-SENP-95-049 REVISION 0 PAGE 12 0F 27 l
1 j option 1 of the alternativt m'trangement does not meet the FSAR
- channel separation criteria sfrt a portion of its cable routing.
' Channel separation is satisfied for the portion of cable routing l
that is contained within conduit embedded in concrete Separationfrom the is also j reactor cavity to the detector junction box.
l taintained in the electrical tunnel where cable is routed between j the detector junction box and the pull box which is mounted on the-l outside of the electrical tunnel wall. From the outside electrical i
tunnel wall through the secondary shield wall, the MB and MD cables
! are routed in the same conduit embedded in concrete. Similarly, routing of the MB and MD cables are in the same raceways from the secondary shield wall to the control root RTG14 104. Within the j control room, the MD jumper cables will be separated by a minimum of 4' vertical and 18" :.orizontal from any A train cables between the RPS cabinet and RTGB 104.
l Since the physical separation ~ criteria is not satisfied for a portion of the MB and MD cables, their potential susceptibility to single failures as defined in the FSAR are evaluated below.
Internal Missiles The postulated containment internal missiles are discussed in FSAR Section 3.5.3 and listed in Table 3.5-1. There are no postulated missiles which may damage safety related equipment outside the secondary shield well since all of the postulated missiles would be contained by berriers or would not have adequate energy to cause such damage. Missiles are not postulated to occur in tk- -1ectrical tunnel and the MB and MD cables are routed in cu.. .. : which is embedded in concrete from the electrical tunnel to the outside of the secondary shield wall. Therefore, the cables will not be jupected by any postulated containment intern.1 missiles. In tn.: reactor auxiliary building (RAB), internal missiles are postulated at the -10 ft, elevation only. Since the MB and MD cables are not routed through this area, RAB internal missiles are not a concern.
Appendix R Fires The linear power range and the power range control channels and associated sensors, power supplies and cables are not used to satisfy the requirements of 10 CFR 50 Appendix R (Ref 9.22), and therefore, the common routing of the MB and MD cables does not affect the plant's ability to respond to Appendix R fires including those postulated to occur in the control room.
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! 9 JPN-PSL-5EMP-95-049 l REVISICW 0 PAGE 13 0F 27 {,
f Righ EmerTy Line Breaks / Enviressental goalificaties ]
l j FSAR Section 15.4 accident scenarios do not credit the use of 1 the linear power range detectors to initiate a reactor trip. ,
! - Trip signals are initisted from low pressuriser pressure and :
l low steam generator level. In addition, linear power range l
detectors are not required to mitigate post-aooident conditions as defined in the Environsaatal Therefore, high energy line Qualification breaks and program (met 9 23).
i harsh environment design basis ocenarios are not impacted by j
the alternative NIs excore detector arrangement described in _
j Options 1 and 2.
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4 Electrieal Transients
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Routing MD with Ma cabler is accepuble based on the low-j power / low-current operation of these cables. The CC2 and IJtD j
detector power supplies. operate at a maximum of 5 mil 11 amperes and 1000 volts for a total of approximately 5 watts output j l
(Ref 9.14). This signal output lacks sufficient power to i
result in a short circuit current which could cause damage to j
the MD cable or the adjacent MB cables. The other cables l
contained in the MB raceways consist of signal cables (Ref i
9.19); electrical contM 1 and power cables are not permitted i to be routed in the same raceways as signal cables. Any
! postulated electrical failure in one of the other signal oables would not have sufficient energy to result in damage to l any other cable contained in the MB raoevays.
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protaction svah= r* ara . . dependence: GDC 22
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As documented in FSAR Sections 7.2.2.1 and 7.2.2.2.6, the reactor protection system meets the requhements of ceneral essign criteria 22 and IEEE 279-1971. It consists of four independent linear power j a range channels complete with sensors, sensor Channelpower supplies and independence is bistable trip units for each channel.
l maintained in both options using the alternative arrangement since all four NI sensors remain independent ard connected separately to tour WI trip channels. The routing of MB cables with MD cables in l
i Dption i satisfies channel independence based on the discussion The provided in the Channel Independence and separation s>ection.
l eennection of the CC2 detector cables to the IJtD at the junction i box in Option 2 satisfies channel independence, due to the use of i separate routes from each detector to its associated junction box.
t nc.taccian avstan Re11 ability and,Testabilltvr CDC 21 As documented in FSAR 8ection 7.2.2.1, 7.2.2.2.2 and 7.2.2.3, the i
reactor protection system meets the requirements of General Design f l
Criteria 21 and IEEE 279-19713 i
~l 1. No single failure results in loss of the protection function and,
- 2. Removal from service of any component or channel does not J result in loss of the required minimum redundancy.
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.7PN-PSL-SENP-95-049 REVISION O PAGE 14 0F 27 A single failure will not result in loss of protection functic.n l
since (i) the substitute to the channel D sensor sensor andand and cable; associated connectors cable is identical are equivalent, I
(ii) the four linear power range channels are still independent and separately connected to four trip channels for both options, (iii) ,
routing of MB cables with MD cables in Option 1 does not. affect l system reliability and testability based on the discussion provided j l
in the channel Independence and separation section, and (iv) l l
connection of the CC2 detector cables to the LRD at the junction I l
box in option 2 satisfies reliability and testability, due to the use of separate routes from each detector to its associated - l junction box.
Removal from service of any component or channel will not result in loss of the required minimum redundancy since the four linear power range channels are stil1 independent and permit testing without loss of the protective function.
j Erotection System Failure Modes: CDC 23 As documented in FSAR Sections 7.2.2.1 and 7.2.2.2.2, the reactor protection system meets the requirements of General Design Criteria 23 and IEEE 279-1971 in that protective system trip channels are designed to fail into a fail safe state or into a state established as acceptable in the event of loss of power supply. The failure modes of the linear power range channels are addressed in FSAR Table 7.2-6. Using any one of the alternative arrangements, new failure modes are not created .al since to the (1) the subetitute channel D sensorsensor and and cable; associated cable is id and connectors are equivaient, (ii) routing of MB cables with MD I cables in option 1 is acceptable based on the discussion provided in the Channel Independence and Separation section, and (iii) connection of the CC2 detector cables to the LRD at the junction box in option 2 does not affect system failure modes due to the use of separate routes from each detector to its associated junction box.
I Protection System Functions: GDC 20 As documented in FSAR Sections 7.2.2.1 and 7.2.'2.2.1, the reactor protection system meets the requirements of Genera) Design criteria 20 and IEEE 279-1971 in that (i) the reactor protection system monitors reactor operating conditions and automatically initiates a reactor trip when the monitored variable exceeds a setpoint, and (ii), the reactor trip setpoints are selected to ensure that the anticipated operational occurrences Using do not either cause acceptable of the alternative fuel design limits to be violated.
arrangements, the linear power range channels continue to provide the protective function of Since reactor trip and the reactor trip CC2 is at a different location setpoints are not affected.than LRD, the effect of the change in detector locatio j
have been determined to be acceptable T,.
as discussed in Sec i
Effects of Detector Location on .iSI and I
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REVISION 0 l PAGE 15 0F 27 i
Senaration of Protection and control Systemst GDC 24
$ As documented in FSAR Sections 7.2.2.1 and 7.2.2.2.7, the reactor j protection system meets the requirements of General Design Criteria l '
24 and IEEE 279-1971 in that the protection systems are separated i from the control instrumentation systems such that failure of any j control system component does not inhibit tne function of the 1 protection system. Using either one of the alternative l arrangements, a control channel detector will provide the input 1 signal to channel D, however, the control channel detector will be 1 disconnected from its normally alignr1 control channel circuit such j that separation of protection and control systems are maintained.
1 Channel Intaarity
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l As discussed in FSAR Section 7.2.2.2.5, the reactor protection system meets the requirements of IEEE 279-1971 with respect to l! channel integrity since the linear power range channels maintain j the functional capability required under applicable extremes of
- conditions relating to environment, energy supply, malfunctions and j accidents.
i j The alternative arrangement does not affect the functional j capability of the linear power range channels since (i) the 4 substitute sensor and associated cable is identical to the channel 1 D sensor and cable; and connectors are equ Lvalent, (ii) the linear i power range channels remain independent arm separately connected to i the trip channels for W options, (iii), routing cf MD with MB
! cables in Option 1 :. . .ies channel integrity baced on the
! discussion provided in the Channel Independence and Coparation j section, and (iv) connection of the CC2 detector cables to the LRD at the junction box in Option 2 catisfies channel i.cegrity due to the use of separate routes from each detector to its associated i
1 j junction box.
j Effects of the Loss of a control channel i
Reactor Raoulatina System l
i j The reactor regulating system (RRS) consists of control cabinets i RRS 1 and RRS 2. Either RRS 1 or RRS 2 may be selveted to provide
- signals for the automatic insertion of control element assemblies i (CEAs) between 15% to 100% power based on inputs from nuclear
- power, turbine power, Tm and Tao. The CEAs may be operated in a
! manual or an automatic sequential mode. Automatic control includes 4 preventing automatic CEA movement until reactor power has increased i above 15% power and blockir.g CEA movement below 11% power. CC1 j" provides the nuclear pover input signal to the RRS 1 cabinet and CC2 provides the nuclear power signal to the RRS 2 cabinet. In 4 addition, the power range control channels provide input to the low j power Auto CEA Motion Prohibit alarm (K-12).
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i .77WaySI,=SENP-95-049 i REVISION 0
- -- PAGE 16 0F 27 If CC2 is not available and ans 3 is selected, manual operation of i
j the CEAs is required and annunalator window R-13 will be energized.
If CC2 is not available and ans 1 is selected, automatio operation of the CEAs is unaffected. Since the control element drive system i
2 1s normally off, operation with either RR8 1 or aus 2 is l acceptable.
Note During monthly calibration of the enaffected control l
channel, there will not be a control channel input to the RRS.
l Low Power Feedvater contral Svetam -
i The low power feedwater control system (LPFCS) provides steam generator level control between 2 and 154 power power based and on inputs feedwater from steam generator level, nuclear, j
temperature. CC1 provides nuclear power Anput for steam generator i 1A level control and CC2 provides nuclear power input for staan l The LPFCs for each steam generator has y generator 18 level control. On selection of auto control, the LPFCS an auto-manual controller.
provides automatic control of steam generator level. On selection 4
i j of manual, the operator provides control by manual operation of the feedvater bypass valve. Operation of LPFCs in manual is an j
acceptable method.
l With a control channel not available, manual control of steam
- generator level is required. Power operation below 15%, if required, is performed manually for the affacted steam generator.
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If automatic control of LMCS is desired, nuclear power input from the unaffected control ..nel can be jumpe.ed to the affected control channel in the RTGB.
See section 8 for the required action.
I Note: During monthly calibration of the unaffected control j
channel, there will not be a control channel input to the LPFCS.
Power Ratio Calculator & Recorder JR-012 Power range control channels 1 and 2 provide input to theControl power l
ratio calculator which is located in the RPS cabinet D.
channel inputs to the power ratio calculator can be switched out by i
the toggle switch at the power ratio calculator. An average of j
both control channels is used to determine the measured power index i ratio. (Note the measured power ratio is the axialMeasured shape power determined using input from the control channels).
l ratio is defined as L-U/L+U where L is the input from lower
{ detector and U is the input from upper detector of each control
] channel. The measured power ratio is then compared with the power limits in the power ratio j
dependent positive and negative AsI j
calculator.
The reactor power deviation alatz (L-34) is provided i
if the measured ASI is outside the positive or negative limits.
J The average measured power ratio, positive ASI and negative ASI limit are provided on JR-012 (3 pen recorder), RTGB 104.
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.7PN-PSL-SENP-95-049 REVISION O PAGE 17 0F 27 With CC2 not available, the measured power ratio recorded on Recorder JR-012 is from CC1 only. The reactor power deviation alarm (L-34) will represent power deviation as detected by the
- unaffected control channel and not by the average of CC1 and CC2.
The core loading pattern design is syssetric for eacE" quadrant (quarter core rotationally symmetric), and therefore, the ASI as determined by the power ratio calculator from either CC1 or CC2 (180* apart) will be approximately the same. Use of only one control channel is thus an acceptable means of ASI surveillance for ~
DNB verification. l Note During monthly calibration of the unaffected control channel, there will not be any ASI input to the Recorder. See Section 6 for impact on technical specification surveillance.
Recorder JR-010 !
I Power range control channels 1 and 2 provide input to RecorderWith JR-010 on RTGB-104 which record reactor power as a percentage.
a control channel unavailable, the output from the unaffected control channel is recorded. The loss of a control channel input i to the recorder is not an operational concern. l If a jumper is provided for automatic con.:rol of LPFCS between two control channel inputs at JY-009-1 and JY-010-1 as discussed in the Ient Power Feedwater Control System section and section 8, then both pens of this recor' *til read input frcm the unaffected control channel and display .ae same value. If it is desired that no display be produced by the pen associated with the affected channel a jumper can be provided for the affected pen. See Section 8 for the required action.
Effects of Detector'iocation on ASI and TS Avial Shane Inder (ASI)
The use of a contL,1 channel detector which is at a different location than the LRD detector affects the assembly weighting factors used in the setpoint calculations. In Reference 9.6 it was determined that, for the Cycle 12 power distrimutions, the maximum differences in excore measured ASI resulting from replacing the LRC safety channel signal with either the CC1 or CC2 control channels signal was insignificant. Since the peripheral power distributions are very similar for Cycle 12 and Cycle 13, there will be an ,
insignificant difference in ASI attributed to thethe Therefore use impact of the on CC2 the l
detector instead of the LRD detector.
asasured ASI expected from the proposed change is bounded by the measurement uncertainties for this parameter, as concluded in Reference 9.6.
l i JPN-PSL-SEMP-95-049
} REVISICOI 0
$ PAGE 18 OF 27 l Animuthal Power Tilt (T,)
J
{i The linear power range channel detectors located circumferential1y around the reactor vessel at 75,165, 255, and 345 degrees from due
! - north, respectively, provide the input for the calculation of
- azimuthal power thit. These detectors respond to -the power t distribution near the periphery of the core adjacent to the j detector. If the CC2 detector is used as a substitute for LRD in i the calculation of T, , the southwest quadrant of the core is
! considered acceptably monitored, and therefore, the sub channel .
' deviation alars is operable provAded a correction is applied to the
- alara setpoint (Ret 9.18) . See also item 16, Actions Required (Section 8) of this document.
Effect of Detector Location om FS&R Ameidaat Analysis This section discusses the effect of the alternative arrangement on FSAR Chapter 15 analyses and provides verification that the current l boundaries of the LSSS and LCO remain applicable. Since core power i indication and axial shape index asasurement are affected by the
! only those transients which involve j
change in significant asymmetribal var detector location,iations in these parameters require
- detailed evaluation. An evaluation of impact on the different FSAR
! Chapter 15 event categories is summarised as follows:
i Tncrease in Heat S--wel by the Baeepdarv Rystem t
i These transients re in depressurisatien of both the primary and secondary systems e ..J. are terminated by reactor trips on low
{ primary or secondary pressure indications which are independent of
- measured core power and ASI and therefore, are not affected.
j Decrease in Heat Damaval by the Secondary System 3,
These transients result in excessive pressurization of both the primary and secondary systems and are terminated by either high RCS pressure or low steam generator level trip signals which do not depend on measured core power or ASI and therefore, are not
. affected.
4 Decrease in Reactor Coolant Flow Rate These events are characterized by loss of power to one or more reactor coolant pumps (RCFs) or an RCP shaft seizure. A rapid decrease in the RCS flow rate follows, resulting in a reactor trip i on low flow within a few seconds. The acceptance criteria for the I RCP shaft seizure is satisfied by ensuring that the reactor trip on low flow occurs almost immediately following the event initiation to limit the number of fuel failures. Events involving loss of power to one or more RCPs require both reador trip on low flow and compliance with the DNB LCO to avoi$ excweding the Specified Acceptable Fuel Design Limi on DNB. Because these events provide the basis for the DNB LCO determination, the effect of the proposed change on these transients is addressed by demonstrating the continued applicability of the current LCO boundaries (see Setpoint Analyses section).
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JPN-PSL-SENP-95-049 I
REVISION 0 1
PAGE 19 OF 27 Decrease in Reactor coolant System Inventerv e
l Asymmetric occurrences such as Steam Generator Tube Rupture and
- Small/Large B7 eak LOCA, which result in TM/LP reactor trip, do not j involve significant changes in measured core power or ASI. The low RCS pressure trip setpoint which determines the time to trip for
]
e these transients corresponds to the minimum value allowed by the
~ TM/LP which is not dependent on ASI or measured core power, and therefore, current analyses results for these events remain valid.
Reactivity and Power Distribution Anomaligg f These events include (i) symmetric core reactivity changes such as l
the CEA withdrawal, (ii) asymmetric core rscctivity changes which j
affect radial power distribution of the core such as the rod drop 1 event, and (iii) more severe localized core reactivity changes such as the ejected CEA event which causes significant variations in l
both the axial and radial power distribution of the core.
l The detection of symmetric reactivity and power distribution
- anomalies is independent of excore detector location. similarly, i
relatively minor asymmetric reactivity variations, such as occur l during the rod drop event, do not cause sufficient core power changes to trip the reactor, nor do they significantly change the measured ASI since both the top and bottom halves of the core are almost simultaneously affected. Events in this category rely on the DNB LCO for protection since the only parameter which changes j significantly is the rad .ower distribution. Therefore, the i
consequences of these transients are not affected provided the j
current LCO limits remain applicable (see Setpoint Analyses below) .
l The acceptance criteria for the CEA ejection event ars based on the j maximum calculated fuel enthalpy and peak RCS pressure. The
- analysis of record (Ref 9.8) utilized a generic control rod l l
ejection methodology. Ths results of that analysis showed that the '
j maximum calculated fuel enthalpy is limited by Doppler feedback and is insensitive both to reactor trip CEA insertion and small changes i
in the trip delay time that may occur for this event using the alternative arrangement. In addition, due to the magnitude of the
! power excursion necessary to achieve significant anargy deposition in the fuel, the azimuthal location of the safety channels have a ,
1 very small effect on the time of reactor trip initiation. l j Therefore, it is concluded that the current analysis results for i maximum calculated fuel ithalpy during a CEA ejection event i
support cycle 13 operaticLa with the alternative arrangement. The calculated peak RCS pressure is largely unaffected because it is l dependent on the total deposited energy prior to trip 'which, because of the essentially unchanged time to trip, does not change l
- significantly. Therefore current analysis conclusions remain j applicable with respect to peak RCS pressurization during a CEA
! ejection event (Ref. 9.6).
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i JPN-PSL-SENP-95-049 REVISION O PAGE 20 OF 27 setnelnt inaivnae Due to the syssetric loading patterns used in the core, the fcur linear range safety channel detectors see nearly identical AsI's.
However, since the proposed replacement detector is not.Jocated in a symmetric location, it will detect a slightly different ASI than the safety channel detectors. The geometry weighting factors for
~
the CC2 detector are nearly identical to the In addition, lowcorresponding values power flux reduction for the failed LRD channel. aziel power distributions _ provide assemblies with similar l approxir.ately 70% of the signal to both the LRD and CC2' detectors. !
Additionally, a previous analysis for Cycle 12 (Reference 9.6) I involving replacement of the LRC channel with either the CC1 or CC2 channels concluded that the resulting ditference in excore ASI represents only a small fraction of the measurement uncertainty credited for this parameter in the Setpoint Analyses (Ref 9.9).
Since there are no major design differences between cycles 12 and 13, the corresponding peripheral power distributions are similar and the conclusions of Reference 9.6 remain applicable, safety classification This evaluation is classified as Safety Related since the excore detector arrangement described in this evaluation is used to assure the capability to shut down the reactor (Ref. 9.3).
nTimEMoDEstw JTS AlpkLYSIS 4.0 The alternative NIS excore detector arrangement is identical to the original configuration of four linear power ranne channels with the exception of the physical location The of the seiscted detectors, control channel instrumentation wells, for LRD and cable routing.
cabling and circuitry pro the same design and meet the same specifications and qualifications, Since 07 tion and 1 therefore, groups MD new with MBfailure modes cables, the are not created.
failure modes associated wath physical separation as discussed in Sec, tion 3 were evaluated and determined notIntoaddition, be credible.
the Therefore, new failure modes are not created.is bound by Technical Specif failure of CC2 as LRD FSAR. Therefore, new failure modes are not introduced.
5.0 _PnwT RESTRICTIONS The use of the alternative NIS excore detector If optionarrangement 1 (control roomis limited until the next refueling outage.is implemented, option 2 aust be i modification) next unplanned outage (Mode 3 or 4) unless complete restoration is If option 2 is implemented after option 1, performed at that time.the connection in the control room must be returned configuration.
JPN-PSL-SEMP-95-049 REVISIoM 0 PAGE 21 0F 27 If either option 1 or 2 is implemented see section a for required actions concerning the sub-channel deviation alarm.
The FSAR, plant drawings and other affected engineering documents will not he-revised due to the limited duration of the alternative arrangement.
6.0 W ON TECIMICAL BPECIFICATIONE Reference is made to the Technical Specification requirements outlined in Section 2. The alternative NS excore detector
- arrangement is evaluated against each applicable requirement:
With respect to T/S 3.3.1.1, satisfying the LCO for linear power range channels using a control channel detector to perform the function of the LRD detector is acceptable. The Surveillance associated with this LCO is not affected by the alternative ,
arrangement.
With respect to T/8 4.2.4.2, satisfying the operability of sub-Channel Deviation Alarm using the CC2 detector to perform the function of the LRD detector was evaluated (Ref 9.18) and determined to be acceptable if a penalty is applied to the sub-channel deviation alarm setpoint. Item 16 of Section 8 of this evaluation should be followed to calculate the azimuthal power tilt.
With respect to T/S 4.2.1.;, satisfying the Surveillance for linear heat rate using recorder JR-012 with input from one control channel to monitor axial shape index is acceptable. Reading axial shape index from the RPS cabinets is another acceptaole method of monitoring axial shape index when using the excore detector system to satisfy the surveillance.
With respect to T/5 4.2.3g4, the Surveillance for total integrated radial peaking factor, F', , can be satisfied for Options 1 & 2 utilisire full core power distribution analyses codes. This surveillance is not impacted since non-full
- core power distribution analysis codes are not used to determine F ,.
With respect to T/S 4.2.5.1, satisfying the Surveillance for axial shape index using recordstr JR-012 with input from one control channal te acceptable. Reading axial shape index from the RPS cabinets is another acceptable method of monitoring axial shape index.
4
JPN-PSL-SBNP-95-049 REVISION O PAGE 22 OF 27 with respect to T/S 2.2.1, the Limiting safety System settings (LSSS) instrumentation setpoints are not affected using the cc2 detector to perform the function of the LRD detector.
sased on the above discussion, implementation of the alternative NIS excore detector arrangement does not require changes ~to~ the Technical Specifications.
7.o nunnernmen marzTY ounsTram onenmurnarren _
In accordance with 10 CFR 50.59, the respenses to below listed questions serve to determine whether the alternative NIS excore detector arrangement constitutes an unnviewed safety question:
7.1 Does the proposed activity increase the probability of occurrence of an accident previously evaluated in the SAR?
The effects of the alternative arrangement are limited to the l NIS. This instrumentation is not postulated in the SAR to j initiate an accident. Since the NIS is not postulated to initiate an accident, the alternative arrangement does not j increase the probability of occurrence of an accident j
previously evaluated in the SAR.
3 l
7.2 Does the proposed activity increase the consequences of an 4 accident previously evaluated in the SAR?
3 i The consequences x an accident previously evaluated in the l
FSAR are not increased beca.use the alternative arrangement l
does not affect the ability of the NIS to perform its design 1
basis function. The NIS trip setpoints are Snaffected by the alternative arrangement. As discussed in Sections 3 and 4, the alternative arrangement neither degrades nor prevents action used to mitigate SAR accidents. Therefore, the i alternative arrangement does not increase the consequences of I an accident previously evaluated in the SAR.
Does the proposed activity increase the probability of an i
i 7.3 i occurrence of a malfunction of equipment iscortant to safety previously evaluated in the SAR?
l
! The LRD and CC2 detectors have the same design and are
! qualified to the same requirements. In addition, the i alternative arrangement does not create new failure modes as
! discussed in Sections 3 and 4. Therefore, the probability of j occurrence of equipment malfunction important to safety j
previously evaluated in the SAR is not increased.
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JPN-PSL-SENP-95-049 REVISION O PAGE 23 OF 27 7.4 Does the proposed activity increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR7 -
The LRD and CC2 detectors have the same design and - are qualified to the same requirements. In addition, the alternative arrangement does not create new failure modes as discussed in Sections 3 and 4. The alternative arrangement neither degrades nor prevents action used to mitigate SAR accidents. Therefore, the alternative arrangement does~not increase the consequences of a malfunction of equipment important to safety previously evaluated in the SAR.
7.5 Does the proposed activity creat.e the possibility of an accident of a different type than any previously evaluated in the SAR?
As stated in Sections 3 and 4, new failure modes are not created using the alternative arrangement. Since the detectors are the same and the system functions are not affected, new types of accidents are not created that arm different from any already evaluated in the SAR.
7.6 Does the proposed activity create the possibility of a malfunction of equipment important to safety of a different type than any previously evaluated in the SAR?
As stated in Sc 4 3 and 4, new f ailure modes are not created by the alternative arrangement. Since the alternative arrangement uses detectors with the same design and qualification requirements and restores the t. ip logic to 2/4 coincidence, malfunctions C equipment important to safety of a different type are precluded. Therefore, the alternative arrangement does not increase the possibility of a malfunction ,
of equipment important to safety of a different type than l I
previously evaluated in the SAR.
7.7 De ss the proposed activity reduce the margin of safety as defined in the basis for any Technical Specification? l The setpoints are unaffected by the alternative arrangement, and therefore, the proposed activity does not reduce the margin of safety as defined in the basis for any technical specification.
The alternative NIS excore detector arrangement does not adversely affect safe operation of the plant (per Sections 3 and 4), does not constitute an unreviewed safety question (per Section 7) and does not require change to the Technical Specifications (per Section 6) .
Therefore, implementation of the alternative arrangement does not require prior NRC spproval.
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.- ~ _ . . . . - . - . . . - . . _ _ . - - . . . . . . - - . - - . - . _ _ - - . .
JPN-PSL-SENP-95-049 REVISION 0 PAGE 24 OF 27 8.8 ACTIGES REQUIRED NOTE: Action steps 8, 9, and 15 are provided for guidamse only and may be revised with angineering eemsurrease (i.e. Peo approval). ._
- 1. Plant procedures shall be reviewed to determine p-= man changes to wt the use of the alternative NIs escore detector arrangement. _
- 2. Prior to implementing Option 2 verify that the abort on LRD is between the junction box and the IJtD detector.
- 3. Perform a channel calibration on the control channel selected for linear power range channel D.
- 4. The appropriate Shape Annealing Factor shall be used for linear power range channel D based on the substitute detector cc2.
- 5. Perform a calorimetric in accordance with plant procedures and !
adjust linear power range channel D as required.
- 6. Place LPFCS' controle in manual for the affected steam )
generator when LPFCS is required or consider action S below ,
for use of automatic level control. l
S. If option 1 or 2 is selected and automatic control of steam Generator 1B is desired, disconnect CC2 inut to n-009-1 and ,
provide a jumper between n-009-1 & n-010-1 as per Figure 4 l and verify proper operation.
Note: During monthly calibration of the unaffected control channel, there will not be a control channel input to LPFCS.
- 9. By providing a jumper as discussed in action 8 above, both pens of the recorder JR-010 will read input fromIfthe unaffected control channel and display tha same value. it is desired that no display be produced by the pen associated with the affected channel, provide jumper at the recorder input terminals per Figure 4. l 1
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l .7PN-PSL-sENP-95-049 RsvIstoN 0 1
PAGE 25 0F 27
! 10. For Option 1 or 2, the sub channel deviation alarm is operable l
provided a penalty is applied to the alarm setpoint (see Ref 9.18). The penalty correction shall be in accordance with the i following formulas . _ _ .
l j (New setpoint = [(Old Setpoint - 1)
- 0.3622) + 1)
Units: Ratio of actual power to nminal core power
- 11. The selected control channel shell be bypassed in the power ratio calculator located in RPS ch inet D.
- l
! 12. During the monthly calibration of the unaffected control
! channel, there will not be any control channel input to RR8
' nor ASI input to Recorder JR-012.
i
, 13. The selected control channel shall have a channel calibration l performed before being placed back in service.
l l 14. Refer to Section 6 for impact on Technical Specification j surveillances. .
- 15. Implement the alternative arrangement (Option 1, or 2) per i attachment 1.
! 16. If monitoring asinuthal power tilt by excore detectors is l required, detector utD.should be assumed failed and the ion 4.3.2c of OSP-64.01, _(Wiw B
! equations given 12 l (Rat 9 24) should be utilized.
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.7PN-PSL-SENP-95-049 REVISION O PAGE 26 OF 27 9.8 REIERIEEEE 9.1 St Lucie Unit 1 FSAR, Amendment 13.
9.2 St Lucie Unit 1 Technical Specifications, Amendment 135.
9.3 Nuclear Engineering Quality InstruMions, Rev. 1, dated 12/21/94.
9.4 Engineering Evaluation Package, PSLP-90-2373, Rev. 2.
9.5 CND 8770-B-327, Sheets: 5s(RS), 55 (R11) , 61(R10) , 63 (R11) ,
399(R14), 400(R12), 403(R9), 406(R12} , 654(R2) 9.6 Siemens letter, # Pre 14=ia==7 Review of_ the CEA Rod Ejection Event and Setpoint Analysee vlth a Replacement Excore Detector '
for St. Lucie Unit 1", Dated August 21, 1993 9.7 I\M 8770-4873, Rev 0, N&S No. 776922437, P.O. No. 9901569.
9.8 A Generic Analysis of the Control Rod Ejection Transient for Pressurized Nater Reactors, EN NF-78-44(A), Exxon Nuclear Company, Richland, NA, 99352, January 1979.
9.9 St. Lucie Unit 1 Cycle 13 Safety Analysis Report, IMF-94-142, Siemens Power Corporation - Nuclear Division, Richland, NA, 99352, September 1994.
9.10 Drawing 8770-0
- SR. 4, Rev. 3 9.11 Drawing k770-G-189, SR. 1, Rev. 3 9.12 8770-B-328, Manual Cable And Conduit List Program - Report BN, 10-26-94 9.13 8770-6861, Rev 3, Gulf Ein.Eic System - Dual L4*mr Power Channel NP-6 9.14 8770-6862, Rev 0, Neutron Flux Monitoring System 9.15 8770-6970, Rev 3, Reactor Regulating Systen 9.16 8770-7120, Rev 6, Reactor Protection System, Vol 1 9.17 8770-7163, Rev 7, Reactor Protection System, Vol 2 ,
9.18 JPN Calculation PSL-1FJF-95-084, " Determination of Penalty Factor for PSL1 Sub-Channel Deviation Alara Satpoint when using CC2 Detector for LRD", Rev 0, 5/31/95.
9.19 Drawings 8770-B-328 SH. 5 Rev. 8, SR. 5A Rev. 5, SH. 9A Rev.
2, SH. 10 Rev. 5, and SH.16 Rev. 5, " Cable and conduit List Installation Notes".
I 1 JP18-PSL-SENP-95-049 1 l
REVIS100f 0 PAGE 27 0F 27
{
9.20 I/M 8770-11831, Rev. O Low Power Feedwater Regulating System.
9.21 Drawings 8770-B-325 85. D10-60 Rev. 4.
9.22 PSL-1 Appendix R Essential Equipment List, Drawing 8770-5-049,
. Rev. 1. . . _ .
9.22 Ravironmental Qualification List, 3770-A-450 Rev. 21.
9.24 087-64.01, Reactor Engineering Periodic Tests, Checks, and i calibrations, Rev s. _
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