ML20202D955
| ML20202D955 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 01/26/1999 |
| From: | DETROIT EDISON CO. |
| To: | |
| Shared Package | |
| ML20202D947 | List: |
| References | |
| NUDOCS 9902020190 | |
| Download: ML20202D955 (200) | |
Text
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SUMMARY
DISPOSITION PAATRIX FOR FERMI-2 RETAINED / - Iol I CURRENT NEW CRITERlON SASIS FOR WCLUSION/ EXCLUSION l. TS NUMBER TITLE ' TS NUMSER FOR PROPOSED NEW LOCATION FOR THE RELOCATED REQUIREMENTS MCLUSION 3/4.3.6.1** Rod Block Monitor 3.3.2.1 Yes-3 Prevents continuous withdrawal of a high worth control rod that would challenge the MCPR Safety Limit and 1 percent cladding plastic strain fuel design limit. 3/4.3.6.2" APRM Relocated No Refer to the Discussions of Change *R.1" for ITS Section 3.3.2.1, " Control Rod Block Instrumentation," for relocation justification and location. 3/4.3.6.3* Source Range Monitors Relocated No Refer to the Discussione of Change *R.2* for ITS Section 3.3.2.1 "Contraf Rod Block Instrumentation," for relocation justification and location. 3/4.3.6.4" Intermediate Range Morwtors Relocated No Refer to the Discussions of Change 'R.3* for ITS Section 3.3.2.1, " Control Rod Block Instrumentation," for relocation justification and location. 3/4.3.6.5" Scram Discharge Volume Relocated No Refer to the D=scussions of Change *R.4* for ITS Section 3.3.2.1. " Control Rod Block Instrumentation,* for relocation justification and location. 3/4.3.6.6" Reactor Coolant System Recirculation Flow Relocated No Refer to the Discussic.ss of Change *R.5" for ITS Section 3.3.2.1, " Control Rod Block Instrumentation," for relocation justification and location. 3/4.3.6.7* Reactor Mode Switch Shutdown Position 3.3.1.2 Yes 3 Ensures all control rods remain inserted when reactor is assumed to be shutdown. 3/4.3.7 Monitoring instrumentation - - 3 /4.3.7.1 " Radiation Monitoring Instrumentation - - 3/4.3.7.1.1" Control Center Normal Makeup Air Radiation 3.3.7.1 Yes-3 Actuates to maintain habitability of the control room so that operators can remain Monitor in the control room following an accident. As such, it mitigates the consequences of an accident by allowing operators to continue accident mitigation activities from the controt room. 3/4.3.7.1.2 Area Monitors Relocated No Refer to the Discussion of Change *R.1* for ITS Section 3.3.7.1, "CREF System instrumentation
- for relocation justification and location.
3/4.3.7.2 < Relocated by Amendment 115 > - - - 3/4.3.7.3 < Relocated by Amendment 115> - - - 3/4.3.7.4 Remote Shutdown Morwtonng Instrumentation 3.3.3.2 Yes-4 Retained as directed by the NRC as it is a sigruficant contributor to risk reduction. Fermi- Unit 2 5 - Revision 2, 01/18/99 l 9902020190 PDR 990126 ~ P ADOCK 05000341 l PDR -
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l 1 I I O FERMI , UNIT 2 3/4 4 6a Amendment No. JJ,87.128 PAGE 1% OF_ 57 hv1 l L
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DEFINITTONS h " " 8 ' /* * 'l ) i l TABLE 1.2 l Oat =Tiek ConniT- :: Qo0ES MODE SWITCH AVERAGE REACTOR _OITLM POSITION COOL ANT TEMPERATURE
- 1. POWER OPERATION Run- M*I Any temperature f.e M (*
!.. A(Startup/
- 2. STARTUP Hot Standby Any temperature ,
- 3. HOT SHUTDOWN I lk Shutdown *** I
> 200* F
- 4. COLD SHUTDOWNb I! Shutdown' 88.*** s 200* F h
- 5. REFUELING
- Ib Shutdown or Refuel 8
-; !?0" T h i
FEE WO (IThe reactor mode switch may be placed in the Run, Startup/ Hot Standby, or O 3,'o.L l Refuel position to test the switch interlock functions and related
% inserted in core cells containing one or more fuel assemblies by a second (licensed staff. / operator or other technically qualified member of the unit technical 9d- 88 SP44kuN The reactor mode switch may be placed in the Refuel position while a single '
control rod drive is being removed from the reactor pressure vessel per 3109 Specification 3.9.10.1. !' Lu tensioned (Q *T::1g i- _.'e the eerte* e um L- J - 2
== h. "e # Mgessel head closure
'th tE
'**S?e Sp::i:1 Test Easpi.iun= 3.10.4 ano 3.10.0. -~~
Ss1 sp;QA **The reactor mode switch may be placed in the Refuel position while a single y,m,3 control rod is being recoupled or withdrawn provided that the one-rod-out y, ,, g interlock is OPERABLE.
****ta= <nach! T::t Exsptis. 0.10.7. -- - -- /f,/7 Aop; IL bm Gnwles
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- O FERMI - UNIT 2 l 1-10 Amendment No. 75, JU l 116 ,
! PAGE OF 14 .N2
SLs O 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLS 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow: THERMAL POWER shall be s 25% RTP. 2.1.1.2 With the reactor steam dome pressure = 785 psig and core flow = 10% rated core flow: MCPR shall be = 1.11 for two recirculation loop operation or = 1.13 for single recirculation loop operation. 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be s 1325 psig. 2.2 SL Violations With any-SL violation the following actions shall be completed within 2 hours: 2.2.1 Restore compliance with all SLs: and. 2.2.2 Insert all insertable control rods. O ; l FERMI UNIT 2 2.0 1 Revision 2. 01/18/99 i
-- - . - , . . . . - . , , = , ,. . - . - -.
SPECIPt CATns) 2s0 O A .'t F.D SAFFTY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS F.1 SAFETY LIMITS THERNAL POWFR: Law Pressure or low Flow 2.,1,1,1 4rh+-THERMAL POWER shall not exceed 25E of RATED THERMAL POWER with the reactor vessel steam done pressure less than 785 psig or core flow less than 10E of rated flow.
'"~ !"I'lT7.
0." "".!;::1^._ i=iise i .no c. N - Ells: .2 f.'2 With TaraMAL roWre exceeding 25s of RATro EnMAt POWER and the reactor vessei steam " acensure less than 785 psig or, core _ flow Tess than 10% of rated O '"! E I"_a},least nui ynuyuuun m nin , _, ...;g;Z,...o -
._m..-.... _ . , . . . _ . . - . _ _ _ , ,
l.8. 7. THERMAL POWER. Hioh Pressure and Hioh Flow *
-4r+ 4 The MINIMUM CRITICAL POWER RATIO (MCPR shall not tre less than the Safety Limit MCPR of 1.11 for two recirculation loop operation and shall not '
be less than the Safety Limit MCPR of 1.13 for single loop operation with the ! mactor vessel steam done pressure ,\ _:5 > l--- '"- 785 psig and core flow gesurar-MOE of rated flow. O, ^ m:^.;;;n ; . . ;. ra t'= r=mme ' w
-t
,3 . . .
ACTION: g
- f. 1 With MCPR less than the afety Limit MC 1.11 for two recirculation loop 8 operation or less than he Safety Limit MCPR .13 for single loop operation !
and with the reac vessel steam done eressure ...;^ r ^2 r 785 psto and core flaw ~-'- "- 10E of rated flow,Lbe ' n at immer noi wur - witnin z i 42. g: " = -- ty a * " . . . . "
.. . . . ,.. ;.; .;.. i.*.:. 7 REETot_CDDLANT 5YST[M. PRESSURE A I'1 +tS4 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
""""**" "- ^":"".*:- :^ _ :- ::: 5: 1. 2, ! _ _ : .
^
EIlm: 1.2 With' the reactor coolant sys pressure, as measured in the reactor vessel 4,1 steam done. above L,325 esig, in at seast nui SHbipudN with reactor cociann (system meerence 1e 325 osis within 2 hour C r! :n y C .
=-. , _ ;s.+_=-.
...eaual . to g.
.~ .
{MCe4 Rues;in , conicai spefrication 2.y are a,,iicpie for Cycin i c j M.'l rtRMI - unir 2 2-1 Ameno nt no. Jr. JJ, J7J, 129 PAGE I 0F 03 R u 2.
l I DISCUSSION OF CHANGES
) ITS: CHAPTER 2.0 - SAFETY LIMITS j ADMINISTRATIVE ;
A.1 In the conversion of the Fermi 2 current Technical Specifications (CTS) to the proposed plant specific Improved Technical Specifications (ITS). certain wording preferences or conventions ; are adopted which do not result in technical changes (either 1 actual or interpretational). Editorial changes, reformatting, and I revised numbering are adopted to make the ITS consistent with the Boiling Water Reactor (BWR) Standard Technical Specifications NUREG 1433. Rev. 1. A.2 CTS 2.1.1. 2.1.2. and 2.1.3 Actions have been reworded in ITS 2.2.2. While meeting the intent of the CTS requirement to be in Hot Shutdown, the ITS is more direct in stating the relevant safety implication: insert all insertable control rods. For the first three CTS SLs, this action also had the effect of restoring operation to within the required Safety Limit, and the fourth CTS SL (CTS SL 2.1.4) contained a direct requirement to restore the limit. In the ITS a generic " restore compliance with all SLs" is provided to acccmplish the same intent. The intent of the ITS is v consistent with the CTS. and therefore, this change is an administrative change with no impact on safety. A.3 CTS SL 2.1.1 and SL 2.1.2 provide different safety limits based on whether steam dome pressure and core flow are above or below certain points; however, the conditions of being exactly equal to these applicability points is not explicitly addressed. l ITS 2.1.1.2 limits on steam dome pressure and core flow are now specified as " greater than or equal to." Since the difference is infinitesimal, and conservatively resolves an obvious i discontinuity, the change is considered administrative with no impact on safety. A.4 Not used. l !O l FERMI - UNIT 2 1 REVISION 2 01/18/99l ! l
q y, DISCUSSION OF CHANGES ITS: CHAPTER 2.0 - SAFETY LIMITS TECHNICAL CHANGES HORE RESTRICTIVE M.1 In CTS Section 2.1 the individual Applicabilities are specified for the applicable Operational Conditions. In ITS section 2.0, the Applicability is not specified; therefore, the requirements are applicable in all Modes. Although it is physically impossible to violate some SLs in some Modes, all SLs should be applicable at all times to ensure continued safe operation. This change imposes additional restrictions on plant operation, therefore, it is a more restrictive change. The increase in Applicability will have 1 no negative impact on safety. l H.2 CTS 2.1.2 is modified by footnote *, applied to the MCPR SLs, that states they apply only for Cycle 7 operation. The effect being l (pending any future amendment) that no MCPR SL would apply beyond Cycle 7. ITS 2.1.1.2 imposes the MCPR SL but does not exclude that limit for operation beyond Cycle 7. This results in a , conservative application of at least some MCPR SL beyond Cycle 7 i operation. Fermi-2 recognizes that an amendment is expected to be 7 ' submitted, if necessary, and approved prior to commencing (V subsequent cycle operation. This intent and understanding continues to be committed to by Fermi 2 with or without the cycle-restriction note. IECHNICAL CHANGES LESS RESTRICTIVE
" Generic" LR.1 CTS 2.1.4 provides direction on how to restore reactor water level. ITS 2.1.1.3 requires that reactor vessel water level be maintained within limits but does not define how the water level is restored, however, the safety related requirement to maintain water level at the required level is maintained. Removing these details from the Technical Specifications maintains consistency with NUREG 1433. Regulatory control of changes to these requirements (e.g., Technical Specification amendment or 10 CFR 50.59) is not necessary to provide adequate protection of the public health and safety since the safety related requirement to maintain water level at the required level is maintained.
A V FERMI - UNIT 2 2 REVISION 2, 01/18/99l
DISCUSSION OF CHANGES O ITS: CHAPTER 2.0 - SAFETY LIMITS i LR.2 CTS 2.1.1. 2.1.2. 2.1.3. and 2.1.4 provide direction on complying with the requirements of CTS 6.7.1 when a Safety Limit is not met. ' CTS 6.7.1 provides details of reports and notifications to be made after a safety limit violation. ITS 2.2 does not include this detail direction. Removing these details from Technical Specifications maintains consistency with NUREG 1433. Regulatory control of changes to these requirements (e.g.. Technical ! Specification amendment or 10 CFR 50.59) is not necessary to ] provide adequate protection of the public health and safety since ' there is no change in the requirement to maintain the Safety Limits, and 10 CFR 50.36 provides sufficient detail for post-safety limit violation actions. 1 LESS RESTRICTIVE i IECHNICAL CHANGES
Speci fic" None ]
RELOCATED SPECIFICATIONS l None l TECHNICAL SPECIFICATION BASES The CTS Bases for this Specification have been replaced by Bases that reflect , the format and applicable content of ITS 2.0 consistent with the BWR STS. NUREG 1433. Rev. 1. ; i l l 1 i , FERMI - UNIT 2 3 REVISION 2 01/18/99l
SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core Sts 2.1.1.1 With the reactor steam dome pressure < 785 psig or core [p, l, f \ flow < 10% rated core flow: / THERMAL POWER shall be s 25% RTP. 2.1.1.2 Withthereactorsteamdomepressure2785psigandcore[7,(,2 flow 210% rated cor flow: t.' / MCPR shall be 2 for two recirculation loop operation or 2 for single recirculation loop operation.
/./3 2.1.1.3 Reactor vessel water level shall be greater than the top (2,f.4 of active irradiated fuel.
\ 2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be 5 1325 psig. (2. I 3 2.2 SL Violations * With any SL violation, th;. following actions shall be complete 4: 2.2. in fy ations ac W (2. I 8C.ht o TLi> 2.2.[1 Restore compliance with all SLs; and w p 12.2.2.2 Insert all insertable control rods.
.... W[* *u," R
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1 (continued)
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2.0-1 Ru 1,- e ;/67/n O fev'2-
DISCUSSION OF CHANGES ITS: SECTION 3.0 LC0 AND SR APPLICABILITY The first sentence of CTS 4.0.3 is moved to ITS SR 3.0.1 and is stated as: " Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO, except as provided in SR 3.0.3." The stipulation " Surveillance Requirements do not have to be performed on inoperable equipment" is moved from CTS 4.0.3, to ITS SR 3.0.1. Since all LCOs do not deal exclusively with equipment Operability. a clarifying phrase is also added: "or variables outside specified limits." These changes do not change the intent of the CTS and therefore, are adH nistrative changes with no impact on safety. A.9 The CTS statement " maximum allowable extension not to exceed 25 percent of the specified surveillance interval" is clarified in ITS SR 3.0.2 which adds the paragraph, "The specified Frequency for each Surveillance Requirement is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency. as measured from the previous performance or as measured from the time a specified condition of the Frequency is O met." The ITS also establishes what constitutes meeting the specified Frequency of each Surveillance Requirement. Also, the sentence " Exceptions to this Specification are stated in the individual Specifications" is included to acknowledge the explicit use of exceptions in various Surveillances. These changes do not change the intent of the CTS 4.0.2 and therefore, are administrative changes with no impact on safety.
.A.10 CTS permits the extension of certain surveillance intervals beyond the 25 percent allowance of CTS 4.0.2 in order to accommodate the schedule for the sixth refueling outage. The affected l surveillance requirements are listed in CTS Tables 4.0.21 and 4.0.2-2. Since the ITS will be implemented after the sixth l refueling outage, this allowance will no longer be applicable and is removed. This is an administrative change because it deletes an allowance that is no longer applicable and has no impact on safety.
O FERMI UNIT 2 5 REVISION 2, 01/18/99l
CPSC IR CA110A 5. t. ') REACTIVITY EONTROL SYSTEMS [/}/so Set F/ecM*MbM S/ 2) LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) g (,3 K the inoperable csntrol rod (s) is inserted, within_1; hour hM C 2 F. disars/Ineassociaten ire m onal c trol salves == truer: a E ctrically or ydraulic y by closing he drive water and exhau ater is ation valves. e i - Otherwise, be in at least HDT SHU1DOWN witiin the next b'M E 1 12 hours.
- c. With more than 8 control rods inoperable, be in at least HOT iSHUTDOWN,within 12 hours.
M
. hith one or more scram discharge volume vent or drain lines with one
' valve inoperable, restore the ino I' 5'M"3 T
*** e.
/ within 7 days, or be in at least berable valve (s) to OPERABLE status hours.
T SHUTDOWN within the next 12 With one or more scram discharge volume vent or drain lines with both l valves inoperable 1solate the associated line within 8 hours ****, or be in at least HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated PERABLE by: g$g.g . At least once per 31 days verifying each valve to be open.* and M* I Evaluating scram discharge volume system response prior to plant RegAc4A3
' startup after each scram to verify that no abnormalities exist.
i Q s *$.t.h3 t404c. i 4.1.3.. 2-\When above the preset power level of the RWiQ all withdrawn control d :';;;.-;;;11.rI Q( rods I
.w.; ::.;;r = ;;;;:::: :::;r;; .e- I
- e. .,d.. 1A. W sha pre;;;;r;j ::ll be demonstrated OPERABLE ach control rod bygh.c at ;...
g 3,g gu,3,ta. At least once per
- - -- 3 ays, and h* .
- b. Within 24 hours when any control rod is f[ J' Raf4 d e.3 kwssue frictioyer mecipnical inteJft,rh"abJe as a A*1 -
Msuit af.f gqj l F J*These tenntenis. valves may be closed intermittently for testing under administrative c , .
- C owrYw t_ing
.-- i nt sociat tttentl under adai strativ control to pern 7 th res ing the e rol OPER stat r ,.perate~ Action entry is allowee for ea'ch FJY vent and drain line. 8
****An isolated line may be unisolated under administrative control to allow I
( draining and venting of the SDV.
!_ i O 40 k o h CA N$n11.T FERMI - UNIT 2 3/4 1-4 Amendment No. J/, NJ, 7) 120 .
3 0F 10 PAGE
.&2
,$ N ClPIC/r'T1Dtb S'I'Y REACTIVITY CONTROL SYSTEMS [Af50 hhdbbU LIMITING CONDITION FOR OPERATION (Continued) i ACTIDN: (Continued) ;
- 2. If the inoperable control rod (s) is inserted, within I hour disarm the associated directional control valves ** either:
ga a) Electrically, or W8,IO / b) Hydraulically by closing the drive water and exhaust water isolation valves. Otherwise, be in at least HDT SHl1TDOWN within the next 12 hours. M With more than 8 control rods inoperable, be in at least HOT L SHUTDOW within 12 hours. g g gf,g _
*** d. With one or more scram discharge volume vent or drain lines with one Agg4 valve inoperable, restore the inoperable valve (s) to OPERABLE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 A C'3" N C hours.
*** e. With one or more scram discharge volume vent or drain lines with both A C Uord B valves inoperable, isolate the associated line within 8 hours ****, or gg c be in at least HOT SHUTDOWN within the next 12 hours.
SURVEILLANCE REQUIREMENTS
- 1. .3.1. - The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:
l sR 3.l.8 I a. At least once per 31 days verifying each valve to be open,* and aluatingscyskdischargev e prior to pl t LR.I startuo aftgr each scram tyv)c erifyCae thatsystem no respo e orsalities t.
.3.1.2 When above the preset power level of the RW, all withdrawn control rods not required to have their dire:tional control valves disarned electrically 3 or hydraulically shall be demonstrated OPERABLE by moving each control rod at l least one notch:
g g. a. At least once per 7 days, and
)
3,l.3 b. Within 24 hours when any control rod is immovable as a result of ( excessive. friction or mechanical interference. 5%$
- controls.These valves may be closed intermittently for testing under administrative 54 ay be rearmed intermittently, under administrative control, to permit
<SPpD *= ting. associated with restoring the control rod to OPERABLE status.
@***5eparate Action entry is allowed for each 50V vent and drain line.
An isolated line may be unisolated under administrative control to allow draining and venting of the SDY. O -- MTroras Mors%gniad Ardim 6-I tJoL FERMI - UNIT 2 3/4 1-4 Amendment No. J/, JJ, JJ,120
..PAGE i 0F O'{ ,y
. . . - - _ _ - _ - ~ _ . - - . - - - - - .. - - _ _ .- _ - _ - - DISCUSSION OF CHANGES ; ITS: SECTION 3.1.8 - SDV VENT AND DRAIN VALVES
. ADMINISTRATIVE ,
i A.1 In the conversion of the Fermi 2 current Technical Specifications : (CTS) to the proposed plant specific Iniproved Technical l Specifications (ITS), certain wording preferences or conventions are adopted which do not result in technical changes (either actual or interpretational). Editorial changes, reformatting, and revised numbering are adopted to make the ITS consistent with the Boiling Water Reactor (BWR) Standard Technical Specifications ; NUREG 1433. Rev. 1. A.2 CTS 3.1.3.1 Actions and Surveillances contain requirements for SDV I I vent and drain valves, without an explicit LCO statement. ITS adds LC0 3.1.8 as an administrative presentation preference to require Operability of the SDV vent and drain valves, and to relate the associated Actions and Surveillances. A.3 CTS 4.1.3.1.4.a.1 and a.2 require the performance of a SDV vent and drain valve functional test on receipt and reset of a
si gnal . " ITS SR 3.1.8.2 permits the system functional to be O. initiated by an " actual or simulated" signal. This change allows satisfactory automatic scrams, as well as appropriately simulated scram signals, to be used to fulfill the system functional Surveillance requirement. Operability is adequately demonstrated because the SDV vent and drain valves can not discriminate between
" actual" or " simulated" scram signals. Since this is a reasonable interpretation of the existing requirement, this is considered an administrative ch::r.ge.
l l 1 l O ! FERMI - UNIT 2 1 REVISION 2, 01/18/99l
l.-s 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. , 1 CONDITION REQUIRED ACTION COMPLETION TIME
] .-3 , A. One or more required A.1 Place channel in 12 hours l channels inoperable.
(v) trip. 08 A.2 ---- - NOTE - -- - i Not applicable for l Functions 2.a 2.b, ; 2.c and 2.d. l
..................... j l
Place associated trip 12 hours ! system in trip. (continued) l. i n ( ) v l FERMI - UNIT 2 3.3 1 Revision 2 01/18/99
-q RPS Instrumentation
,RJ
) 3.3.1.1 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME 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. @ 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. b l ( ) x_ / l FERMI - UNIT 2 3.3-1(1) Revision 2, 01/18/99
-m RPS Instrumentation I 1 3.3.1.1 V
SURVEILLANCE REQUIREMENTS
.....................................N0TES-- ---- ------- -------------- ----
- 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 Recuired Actions may be delayed for up to 6 hours provided the associatec Function maintains RPS trip capability.
SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.1.2 Perform CHANNEL CHECK. 24 hours (O SR 3.3.1.1.3 ------------------NOTE \m/ l Not required to be performed until 12 I hours after THERMAL POWER = 25t RTP. J Verify the absolute difference between 7 days the average power range monitor (APRM) , channels and the calculated power is j l s 2t RTP. while operating at a 25% RTP. l (continued) I i g \ J l FERMI - UNIT 2 3.3 4 Revision 2, 01/18/99
1
?
i i RP"> Instrumentation 3.3.1.1 O- - 1 SURVEILLANCE REQUIREMENTS (continued) : SURVEILLANCE FREQUENCY
- i SR 3.3.1.1.10 Verify the trip unit setpoint. 92 days ;
SR 3.3.1.1.11 - - -- --- -- - -NOTES------------------ 1
- 1. Neutron detectors are excluded. '
l 2. For Function 1.a not required to be performed when entering MODE 2 from ! MODE 1 until 12 hours after entering ; MODE 2. j Perform CHANNEL CALIBRATION. 184 days l l SR 3.3.1.1.12 ---------- --- -NOTE----- --- --- --- - O v For Function 2.a. not required to be 3erformed when entering MODE 2 from 40DE 1 until 12 hours after entering l MODE 2. Perform CHANNEL FUNCTIONAL TEST. 184 days j l SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. 18 months l SR 3.3.1.1.14 Perform CHANNEL CALIBRATION, 18 months
-I SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months l
(continued) O i l FERMI UNIT 2 3.3-6 Revision 2, 01/18/99
- ~ _ _ ,
. _ . _ - . - _ _ _ _ _ - _ . _ _ _ . _ _ _ . ~ _ _ - - _ _ -- _ _ _ . _ _ _ _ _ _
l , l l l , l lc RPS Instrumentation 3.3.1.1 !' ' SURVEILLANCE REQUIREMENTS (continued) i l SURVEILLANCE FREQUENCY l l SR 3.3.1.1.16 Verify Turbine Stop Valve-Closure and 18 months i Turbine Control Valve Fast Closure i Functions are not bypassed when THERMAL i POWER is = 30% RTP. I SR 3.3.1.1.17 - - ---
-- -----NOTES ------ ---------- l
- 1. Neutron detectors are excluded. !
- 2. For Functions 3 and 4 channel sensor !
response times are not required to be , measured. ! l l
- 3. For Function 5 "n" equals 4 channels )
.for the purpose of determining the i STAGGERED TEST BASIS Frequency. ;
Verify the RPS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST BASIS ! [ , SR 3.3.1.1.18 ------- - - - --- NOTE----- - ---------- l Neutron detectors 'are excluded. Perform CHANNEL CALIBRATION. 24 months l SR 3.3.1.1.19 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months i O l FERMI UNIT 2 3.3 7 Revision 2, 01/18/99
L
- RPS Instrumentation !
( 3.3.1.1 ; t Table 3.3.1.1 1 (page 1 of 3) ! Reactor Protection System Instrumentation j f , APPLICABLE CONDITIONS , 1- MODES OR REQUIRED REFERENCED ' I OTER CHANNFLS FROM r SPECIFIED PER TRIP REQUIRED SlRVEILLANCE ALLOWABLE i FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREENTS VALUE j
- 1. Intermediate Range !
Monitors , j a. Neutron Flux-High 2 3 G SR 3.3.1.1.1 s 122/125 c i SR 3.3.1.1.4 divisions of a full scale SR 3.3.1.1.6 SR 3.3.1.1.7 l SR 3.3.1.1.11 ! SR 3.3.1.1.15 ! 5(a) 3 I SR 3.3.1.1.1 s 122/125 I SR 3.3.1.1.5 divisions of SR 3.3.1.1.11 full scale SR 3.3.1.1.15
- b. Inop 2 3 G SR 3.3.1.1.4 NA f SR 3.3.1.1.15 l
5(a) '3 I SR 3.3.1.1.5 NA I SR 3.3.2.2.15
- 2. Average Power Rangar l Monitors
- a. Neutron Flux-Upscale 2 3(C) G SR 3.3.1.1.2 s 20t RTP >
d (Setdown) SR 3.3.1.1.7 ] SR 3.3.1.1.8 < l SR 3.3.1.1.12 ) SR 3.3.1.1.18
- b. Simulated Thermal 1 3(C) F SR 3.3.1.1.2 s 0.63 (W 4W) l Power - Upscale SR 3.3.1.1.3 + 64.38 RTP j SR 3.3.1.1.8 and s SR 3.3.1.1.12 RTP(b)115.52 .
SR 3.3.1.1.18 i (continued) (a) With any control rod withdrawn from a core cell containing or.e or more fuel assen611es. (b) aW = 82 when reset for single loop operation per LCO 3.4.1. " Recirculation Loops Operating." Otherwise 4W = 02.
-(c) Each APRM channel provides inputs to both trip systems. /*
1 ( l-l l FERMI - UNIT 2 3.3 8 Revision 2, 01/18/99 l- 1 l .2. . -, - - . -
i l, p RPS Instrumentation Q 3.3.1.1 Table 3.3.1.11 (page 2 of 3) Reactor Protection System Instrumentation APPLICABLE CONDITIONS M00ES OR REQUIRED REFERENCED 1. OTER CHANNELS FRON SPECIFIED PER TRIP REQUIRED SlRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREENTS VALUE
- 2. Average Power Range Monitors (continued)-
I c'. Neutron - 1 3(C) F SR 3.3.1.1.2 s 1201 RTP Flux - Upscale SR 2.3.1.1.3 j SR 3.3.1.1.8 ' SR 3.3.1.1.12 , SR 3.3.1.1.18
- d. Inop 1.2 3(C) G SR 3.3.1.1.12 NA l~ e. 2-out of-4 Voter 1.2 2 G SR 3.3.1.1.2 NA l'
SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.19
- 3. Reactor Vessel Steam 1.2 2 G ER 3.3.1.1.1 s 1113 psig Dome Pressure-High SR 3.3.1.1.9 SR 3.3.1.1.10 l SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.17
-( 4 Reactor Vessel Water 1.2 2 G SR 3.3.1.1.1 e 171.9 inches A. -
Level - Low. Level 3 SR 3.3.1.1.9 SR 3.3.1.1.10 l SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.17
- 5. Main Steam Isolation 1 8 F SR 3.3.1.1.9 s 121 closed j Valve - Closure SR 3.3.1.1.14 SR 3.3.1.1.15 i SR 3.3.1.1.17
- 6. Main Steam Line 1,2 2 H SR 3.3.1.1.1 s 3.6 X full Radiation - High SR 3.3.1.1.9 power
'] SR 3.3.1.1.14 background SR 3.3.1.1.15 l, 7. . Drywell Pressure-High 1.2 2 G SR 3.3.1.1.1 s 1.88 psig
,, SR 3.3.1.1.9 1' SR 3.3.1.1.10 l' SR 3.3.1.1.14
'l.~
SR 3.3.1.1.15 (continued) (c) Each APRM channel provides inpucs to both trip systems. j f l i
. 1 l
L l l ,r
\
l r
- l- FERMI - UNIT 2- 3.3 9 Revision 2, 01/18/99 l-
t RPS Instrumentation 3.3.1.1 i . (v.) Table 3.3.1.1 1 (page 3 of 3) Reactor Protection System Instrumer;tation APPLICABLE CONDITIONS MDDES OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SLRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D 1 REQUIREMENTS VALUE
- 8. Scram Discharge Voline Water Level - High
- a. Level 1.2 2 G SR 3.3.1.1.1 s 596 ft.
Transmitter SR 3.3.1.1.9 0 inches SR 3.3.1.1.10 l SR 3.3.1.1.14 SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.1 s 596 ft. SR 3.3.1.1.9 0 inches SR 3.3.1.1.10 l SR 3.3.1.1.14 SR 3.3.1.1.15
- b. Float Switch 1,2 2 G SR 3.3.1.1.9 s 596 ft.
l SR 3.3.1.1.14 0 inches SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.9 s 596 ft. l SR 3.3.1.1.14 0 inches SR 3.3.1.1.15
- 9. Turbine Stop e 301 RTP 4 E SR 3.3.1.1.9 s 72 closed f G l Valve - Closure SR 3.3.1.1.14
( V) SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17
- 10. Turbine Cor. trol valve a 301 RTP 2 E SR 3.3.1.1.9 NA Fast Closure SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17 l 11. Reactor Mode Switch- 1.2 2 G SR 3.3.1.1.13 NA Shutdown Fasition SR 3.3.1.1.15 l 5(a) 2 I SR 3.3.1.1.13 NA SR 3.3.1.1.15
- 12. Manual Scram 1.2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.5 NA SR 3.3.1.1.15 (a) With any control rod withdrawn from a core cell containing one or mere fuel assemblies.
t ('O) l FERMI UNIT 2 3.3 10 Revision 2, 01/18/99
l r- RPS Instrumentation ( B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES. LCO. and APPLICABILITY (continued) Averaoe Power Ranoe Monitor 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 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 2-out-of-4 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 y result in a " half trip" in all four voter channels, but no trip inputs to either RPS trip system. A trip from any two unby)assM APRM channels will result in a full trip in each of t1e four voter channels, which in turn results in two trip inputs into each RPS trip logic channel (A1. A2. Bl. 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 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, are required for-each APRM channel. 2.a. Averaoe Power Ranoe Monitor Neutron Flux-UDscale (Setdown) For operation at low power (i.e.. MODE 2). the Average Power Range Monitor Neutron Flux-Upscale (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 l Average Power Range Monitor Neutron Flux-Upscale (Setdown) 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 possible that the Average Power Range Monitor Neutron O l FERMI - UNIT 2 B 3.3.1.1 - 7 Revision 2 01/18/99 1
p RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) l Flux-Upscale (Setdown) Function will provide the primary trip signal for a corewide increase in power. The Average Power Range Monitor Neutron Flux-Upscale (Setdown) Function is credited, along with the IRM neutron flux-high function, with initiating a reactor scram in the analysis of the continuous rod withdrawal during reactor startup event. This Function also 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-Upscale (Setdown) Function must be OPERABLE during MODE 2 when [-) U control rods may be withdrawn since the potential for criticality exists. In MODE 1. the Average Power Range l Monitor Neutron Flux-Upscale 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 Power-Voscale The Average Power Range Monitor Simulated Thermal Power-Upscale 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. The trip level is varied as a function of recirculation drive flow (i.e.. at lower drive flows. the setpoint is reduced proportional to the reduction in power experienced as drive 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-Upscale Function Allowable Value. The Average Power Range Monitor Simulated Thermal Power-Upscale 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 O l FERMI - UNIT 2 B 3.3.1.1 - 8 Revision 2. 01/18/99
l l RPS Instrumentation (n) v B 3.3.1.1 BASES l APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) that the MCPR SL is not exceeded. During these events, the 1 THERMAL POWER increase does not significantly lag the ; neutron flux response and, because of a lower trip setpoint. l will initiate a scram before the high neutron flux scram. For ra)id neutron flux increase events, the THERMAL POWER lags t1e neutron flux and the Average Power Range Monitor l Neutron Flux-Upscale Function will provide a scram signal I before the Average Power Range Monitor Simulated Thermal Power-Upscale Function setpoint is exceeded. The required trip setting for APRMs is dependent on whether the unit is in single recirculation loop operation or two-loop operation, as specified in Table 3.3.1.1-1 footnote (b). The setpoint variable AW is defined as the difference in indicated drive flow (in
- of rated drive flow that l produces rated core flow) between two loop and single-loop :
operation at the same core flow. i Each APRM channel uses one total drive flow signal p$ representative of total core flow. The drive flow signal at i
! rated drive flow is representative of rated core flow at O RTP. 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 Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power-Upscale Function for the mitigation of the loss of feedwater heating event. The THERMAL POWER time constant of approximately 6 seconds is based on the fuel heat transfer , dynamics and provides a signal proportional to the THERMAL l POWER. The Average Power Range Monitor Simulated Thermal Power-Upscale 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/or APRM Functions provide protection for fuel cladding integrity. w/ i l FERMI - UNIT 2 B 3.3.1.1 - 9 Revision 2 01/18/99
l l RPS Instrumentation l g) B 3.3.1.1 i l BASES ; i 1 APPLICABLE SAFETY ANALYSES, LCO. and APPLICABILI1Y (continued) l 2.c. Averaoe Power Ranoe Monitor Neutron Flux-UDscale l The Average Power Range Monitor Neutron Flux-Upscale l 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 l Average Power Range Monitor Neutron Flux-Upscale Function is assumed to terminate the main steam isolation valve (MSIV) closure event and, along with the safety relief valves (SRVs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Neutron Flux-Upscale Function to l terminate the CRDA. , The Allowable Value is based on the Analytical Limit assumed in the CRDA analyses. l The Average Power Range Monitor Neutron Flux-Upscale c Function is required to be OPERABLE in MODE 1 where the potential consequences of the analyzed transients could (s) result in the SLs (e.g., HCPR and RCS pressure) being exceeded. Although the Average Power Range Monitor Neutron Flux-Upscale Function is assumed in the CRDA analysis, which is applicable in MODE 2. the Average Power Range Monitor l Neutron Flux-Upscale (Setdown) Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average l l Power Range Monitor Neutron Flux-Upscale Function is not required in MODE 2. l l l r
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1 l l RPS Instrumentation I 7 B 3.3.1.1 i i BASES l APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) j l 2.d. Averaae Power Ranae Monitor-Inoo l This Function provides assurance that a minimum number of , APRMs are OPERABLE. For any APRM channel, any time: 1) its ! mode switch is in any position other than "0PER": 2) there ! is a loss of input power; 3) the automatic self-test system 1 detects a critical fault with the APRM channel: or 4) the firmware/ software watchdog timer has timed out, 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. i 2.e. 2-out-of-4 Voter The 2-out of-4 Voter Function provides the interface between the APRM Functions and the final RPS trip system logic. As such, it is required to be GPERABLE in the MODES where the l APRM Functions are required and is necessary to support the i safety analysis applicable to each of those Functions. i Therefore, the 2-out-of 4 Voter Function is required to be ; OPERABLE in MODES 1 and 2. Both voter channels in each trip system (all four voter channels) are required to be OPERABLE. Each voter channel i also includes self diagnostic functions. If any voter l 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. A ,D i j_ FERMI-UNIT 2 B 3.3.1.1 - 11 Revision 2, 01/18/99 i l
RPS Instrumentation (v-)' B 3.3.1.1 i BASES APPLICABLE SAFETY ANALYSES. LC0. and APPLICABILITY (continued)
- 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 results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. The overpressurization protection analysis of Reference 4 conservatively assumes scram on the Average l Power Range Monitor Neutron Flux-Upscale signal, not the Reactor Vessel Steam Dome Pressure-High signal. Along with the SRVs. the reactor scram limits the peak RPV pressure to less than the ASME Section III Code limits.
High reactor pressure signals are initiated from four tfD pressure transmitters that sense reactor pressure. The V Reactor Vessel Steam Dome Pressure-High Allowable Value is l 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 areclude a scram from this Function on a valid signal. T1e Function is required to be OPERABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists. l I N 1 l FERMI UNIT 2 B 3.3.1.1 - 12 Revision 2. 01/18/99 l l
1 ; l r- RPS Instrumentation B 3.3.1.1 1 BASES l t APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) 1 i
- 5. Main Steam Isolation Valve-Closure ;
4 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 i analysis of Reference 4 the Average Power Range Monitor i Neutron Flux-Upscale Function, along with the SRVs, limits the peak RPV pressure 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. 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 n v limits of 10 CFR 50.46. MSIV closure signals are initiated from )osition switches located on each of the eight MSIVs. Eac1 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 channel consisting of one position switch. The logic for the Main Steam Isolation Valve-Closure Function is arranged such that either the l inboard or outboard valve on three or more of the main steam ' lines must close in order for a scram to occur. 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 l Function, with eight channels in each trip system. are 1 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 high, a pressurization transient can occur if the MSIVs close. In MODE 2, the MSIV closure trip is automatically O l FERMI - UNIT 2 B 3.3.1.1- 14 Revision 2 01/18/99
l l l ! i l - - RPS Instrumentation ; l( l B 3.3.1.1 j i BASES l l' APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) i The Turbine Stop Valve-Closure Allowable Value is selected , l to be high enough to detect imminent TSV closure, thereby l l reducing the severity of the subsequent pressure transient. [ t l 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 j preclude a scram from this function if any three TSVs should close. This Function is required. consistent with analysis 3
l i assumptions, whenever THERMAL POWER is = 30t RTP. This
- Function is not required when THERMAL POWER is < 30% RTP since the Reactor Vessel Steam Dome Pressure-High and the l
l Average Power Range Monitor Neutron Flux-Upscale Functions are adequate to maintain the necessary safety margins. ! i 10. Turbine Control Valve Fast Closure ! l Fast closure of the TCVs results in the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux { ' p s transients that must be limited. Therefore, a reactor scram is initiated on TCV fast closure in anticipation of the i transients that would result from the closure of these 1 valves. The Turbine Control Valve Fast Closure Function is i the primary scram signal for the generator load rejection event analyzed in Reference 7. For this event, the reactor scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded. Turbine Control Valve Fast Closure signals are initiated by the de energization of the solenoid dump valve at each control - valve. Redundant relay signals are provided to each RPS logic channel such that fast closure of one control valve in each RPS trip system will initiate a scram. This Function must be enabled at THERMAL POWER a 30t RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure of a 161.9 psig: therefore, to consider this Function OPERABLE, the turbine bypass valves must remain shut at THERMAL POWER
= 30% RTP. ,
There is no Allowable Value for the Turbine Control Valve I Fast Closure Function since the channels are actuated solely l
- on energization of the solenoid dump valve.
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RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES. LCO. and APPLICABILITY (continued) Four channels of Turbine Control Valve Fast Closure Function with two channels in each trip system arranged in 1 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 Steam Dome Pressure-High and the Average Power Range Monitor Neutron Flux-Upscale Functions are adequate to maintain the necessary safety margins.
- 11. 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. l 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 i Function, with two channels in each trip system arranged in j a one out of-two logic, are available and required to be OPERABLE. The Reactor Mode Switch-Shutdown Position Function is required to be OPERABLE in MODES 1 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. O l FERMI - UNIT 2 B 3.3.1.1 - 19 Revision 2 01/18/99 i
e RPS Instrumentation I, B 3.3.1.1 BASES ACTIONS (continued) 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 l l be acceptable (Refs. 9 and 13) to permit restoration of 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 l Required Actions B.1. B.2, and C.1 Bases). If the ino>erable channel cannot be restored to OPERABLE status
- wit 11n the allowable out of service time, the channel or the l associated trip system must be placed in the tripped i
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 nV 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 I required APRM channel affects both trip systems: thus, l 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. 1 B.1 and B.2 l 1 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 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 ! rQ V , l FERMI UNIT 2 B 3.3.1.1 - 21 Revision 2, 01/18/99 l
l l ! ) RPS Instrumentation l O B 3.3.1.1 ' BASES ACTIONS (continued) failure in both trip systems (e.g., each trip system remains in a one out-of-one arrangement for a ty)ical four channel Function). The reduced reliability of t11s logic l arrangement was not evaluated in References 9 and 13 for the 12 hour Completion Time. Within the 6 hour allowance, the associated Function will have all required channels OPERABLE or in trip (or aay combination) in one trip system. Completing one of these Required Actions restores RPS to a reliability level equivalent to that evaluated in l l References 9 and 13, which justified a 12 hour allowable out l cf service time as presented in Condition A. The trip l system in the more degraded state should be placed in trip i or, alternatively, all the inoperable channels in that trip l system should be placed in trip (e.g.. a trip system with i 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 n decision of which trip system is in the more degraded state
, Q 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, it is permissible to place the other trip system or its inoperable channels in trip. The 6 hour Com)1etion Time is judged acceptable based on the remaining capa)1lity 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 i where placing the inoperable channel or associated trip system in trip would result in a scram), Condition D must be entered and its Required Action taken. l l 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 2 out of 4 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 i O l l FERMI - UNIT 2 B 3.3.1.1 - 22 Revision 2, 01/18/99 i ._
l g RPS Instrumentation 1 B 3.3.1.1 BASES ; i ACTIONS (continued) i l 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 ' and entry into Condition C, as well as entry into Condition A for each channel. Because Conditions A and C provide ' t Required Actions that are appropriate for the inoperability of APRM Functions 2.a. 2.b, 2.c and 2.d. and these i Functions are not associated with specific trip systems as ! are the APRM 2 out-of 4 voter and other non-APRM channels, Condition B does not apply. l L1 ! 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 i capability when sufficient channels are OPERABLE or in trip (or the associated trip system is in trip), such that both O trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one-out-of two taken twice logic and the IRM and APRM , Functions, this would require both trip systems to have one !' channel OPERABLE or in trip (or the associated trip system in trip). For Function 5 (Main Steam Isolation i Valve-Closure), this would require both trip systems to have each channel associated with the MSIVs in three main steam . lines (not necessarily the same main steam lines for both l trip systems) OPERABLE or in trip (or the associated trip i system in trip). j for Function 8 (Turbine Stop Valve-Closure), this would ' , require both trip systems to have three channels, each I OPERABLE or in trip (or the associated trip system in trip). The Completion Time is intended to allow the operator time to evaluate, and repair or place in trip any discovered inoperabilities that result in a loss of RPS trip operability. The 1 hour Completion Time is acceptable because it minimizes risk while allowing time for restoration or tripping of channels. O l FERMI - UNIT 2 B 3.3.1.1 - 23 Revision 2, 01/18/99 l i
- m. RPS Instrumantation ;
j
~
B 3.3.1.1 BASES ACTIONS (continued) D_l Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.1.11. 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 I 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. H.1. and H.2 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 , ) laced in a MODE or other specified condition in which the i _C0 does not apply. Alternately, for Condition H, the ' associated MSLs may be isolated (Required Action H.1), and, if allowed (i.e., plant safety analysis allows operation l with an MSL isolated), operation with that MSL isolated may continue. Isolating the affected MSL accomplishes the safety function of the inoperable channel. 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)." l O l FERMI - UNIT 2 B 3.3.1.1 - 24 Revision 2 01/18/99 l
p) RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued) Ll 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 H0DE 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 l cells containing one or more fuel assemblies. Control rods j in core cells containing no fuel assemblies do not affect 1 the reactivity of the core and are, therefore, not required i to be inserted. Action must continue until all insertable : control rods in core cells containing one or more fuel l assemblies are fully inserted. 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. For the case of the APRM Functions 2.a. 2.b. 2.c. and 2.d. RPS trip capability is maintained with any two l OPERABLE APRMs remaining. 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 >erform channel Surveillance. That analysis demonstrated tlat the 6 hour I testing allowance does not significantly reduce the l probability that the RPS will trip when necessary. l l SR 3.3.1.1.1 and SR 3.3.1.1.2 l Performance of the CHANNEL CHECK once every 12 hours and once every 24 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 ' o N) l FERMI - UNIT 2 B 3.3.1.1 - 25 Revision 2 01/18/99
n RPS Instrumentation (j B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) 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. The Frequency is based upon operating ex)erience that demonstrates channel failure is rare. T1e 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. l- SR 3.3.1.1.3 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor l power calculated from a heat balance when a 25% RTP. 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
! )rovided that requires the SR to be met only at a 25% RTP 3ecause 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, LHGR, and APLHGR). At a 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 i Frequency is not met >er SR 3.0.2. In this event, the SR must be performed wit 11n 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. m l l FERMI - UNIT 2 B 3.3.1.1 - 26 Revision 2, 01/18/99 l :
RPS Instrumentation ( B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) 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. 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 l 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. O A Frequency of 7 days provides an acceptable level of system average unavailcbility 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 References 9 and 10. (The Manual Scram Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the Reference 9 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 fully withorawing O l FERMI - UNIT 2 B 3.3.1.1 - 27 Revision 2 01/18/99 l
RPS Instrumentation {q
} B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued)
SRMs from the core since indication is being transitioned from the SRMs to the IRMs. 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 overla) is not maintained. Overlap between IRMs and APRMs exists w1en 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 fully withdrawing the SRMs from the core, 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 IRKS. maintaining overlap is not required (APRMs may be reading downscale once in MODE 2). O 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. 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 MWD /ST Frequency is based on operating experience with LPRM sensitivity changes. I FERMI UNIT 2 B 3.3.1.1 - 28 Revision 2, 01/18/99
i l p RPS Instrumentation l B 3.3.1.1 I BASES SURVEILLANCE REQUIREMENTS (continued) l l SR 3.3.1.1.9 and SR 3.3.1.1.13 : 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 un)lanned transient if the Surveillance were performed with t1e 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 This Surveillance provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setcoint methodology, but is not beyond the Allowable Value, the channel performance is still ; within the recuirements of the plant safety ! analysis. Uncer these conditions, the setpoint must be , readjusted to be equal to or more conservative than ! accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 9. j SR 3.3.1.1.11 and SR 3.3.1.1.14 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. I l FERMI - UNIT 2 B 3.3.1.1 - 29 Revision 2, 01/18/99
p RPS Instrumentation g B 3.3.1.1 I BASES SURVEILLANCE REQUIREMENTS (continued) J SR 3.3.1.1.11 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. SR 3.3.1.1.11 Note 2 is provided that requires the IRM SR to be performed within 12 hours of entering MODE 2 from MODE 1. Testing of the j MODE 2 IRH Function cannot be performed in 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 ex'perience and in consideration of providing a reasonable time in which to complete the SR. The Frequency of SR 3.3.1.1.11 is based upon a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency l of SR 3.3.1.1.14 is based upon a 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.1.1.12 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 I continuously in the APRM and voter channels. The APRM 1 CHANNEL FUNCTIONAL TEST covers the APRM channels (including for Function 2.b only, the recirculation flow input function, excluding the flow transmitter), the 2-out of 4 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 3.3.1.1.12 is based on the reliability analysis of Reference 13. (NOTE: The actual voting logic of the 2 out-of 4 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. ' O l FERMI - UNIT 2 B 3.3.1.1- 30 Revision 2 01/18/99 ! l I l l
A RPS Instrumentation () B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) l SR 3.3.1.1.15 and SR 3.3.1.1.19 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 (LC0 3.1.8), overlaps this Surveillance to provide complete testing of the assumed safety function. For the 2-out-of-4 Voter Function, the LSFT includes simulating APRM trip conditions at the APRM channel inputs to the 2 out of-4 trip voter channel to check all combinations of two tripped inputs to the 2 out-of 4 trip voter logic in the voter channels. l The 18 month Frequency of SR 3.3.1.1.15 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 a the 18 month Frequency. ('" / Additionally, the 24 month Frequency of SR 3.3.1.1.19 is based on Reference 13. SR 3.3.1.1.16 This SR ensures that scrams initiated from the Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure 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 set)oint. Additionally, consideration is given to the fact tlat main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure; where turbine first stage pressure of 161.9 psig conservatively correlates to 30% RTP), the main turbine bypass valves must remain closed at THERMAL POWER , a 30% RTP to ensure that the calibration remains valid. l 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 Functions are considered inoperable. Alternatively, the bypass channel can be placed in the A h l FERMI - UNIT 2 B 3.3.1.1 - 31 Revision 2 01/18/99
._ _ _ _ _ _ -_ _ .._ _ _ _ __ ._ .._ . _ . ~ . ._ _ _ RPS Instrumentation O B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) conservative condition (nonbypass). If placed in the nonbypass condition. this SR is met and the channel is considered OPERABLE. The Frequency of 18 months is based on engineering judgment. reliability of the components, and = 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.1.1.17 This SR ensures that the individual channel response times i 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 the APRM 2 out of-4 Voter Function includes the output relays of the voter and the associated RPS relays and contactors. (The digital portion of the APRM
' and 2 out of-4 voter channels are excluded from the RPS 1 RESPONSE TIME testing because self testing and calibration l chtcks the time base of the digital electronics.)
Confirmation of the time base is adequate to assure required response times are met. As noted. neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. In addition. Note 2 states the response time of the sensors for Functions 3 and 4 are excluded from RPS Response Time Testing. The sensors for these Functions are assumed to operate at the sensor's design response time. This l allowance is supported by Reference 12, which determined that significant degradation of the sensor channel response time can be detected during performance of other Technical Specification SR's and that the sensor response time is a small part of the overall RPS RESPONSE TIME testing. RPS RESPONSE TIME tests are conducted on an 18 month l STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS Frequency to be determined based on 4 channels per trip system. in lieu of the 8 channels specified in Table 3.3.1.1-1 for the MSIV Closure Function. This Frequency is l based on the logic interrelationships of the various l l FERMI - UNIT 2 B 3.3.1.1 - 32 Revision 2 01/18/99 I
p RPS Instrumentation , d_ B 3.3.1.1 i BASES f SURVEILLANCE REQUIREMENTS (continued) channels required to produce an RPS scram signal. The 18 month Frequency is consistent with the typical industry ! refueling cycle and is based upon plant operating ; experience, which shows that random failures of f instrumentation components causing serious response time i degradation but not channel failure. are infrequent t l occurrences. : T SR 3.3.1.1.18 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 ! range and accuracy. CHANNEL CALIBRATION leaves the channel l adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. For the APRM Simulated Thermal Power - Upscale t Function. this SR also includes calibrating the associated i recirculation loop flow channel. SR 3.3.1.1.18 is modified by a Note that 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 3.3.1.1.3) , I and the 1000 MWD /T LPRM calibration against the TIPS l (SR 3.3.1.1.8). , The Frequency of SR 3.3.1.1.18 is based upon 24 month i calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. l REFERENCES 1. UFSAR, Figure 7.2 2.
- 2. UFSAR, Section 15.4.1.2.
- 3. NED0-23842, " Continuous Control Rod Withdrawal in the Startup Range." April 18, 1978.
- 4. UFSAR, Section 5.2.2.3.
t
- 5. UFSAR, Section 15.4.9.
(vD l l FERMI - UNIT 2 B 3.3.1.1-33 Revision 2 01/18/99 l l
RPS Instrumentation j B 3.3.1.1 1 BASES REFERENCES (continued) :
- 6. UFSAR, Section 6.3.3.
)
- 7. UFSAR, Chapter 15.
- 8. P. Check (NRC) letter to G. Lainas (NRC), "BWR Scram i Discharge System Safety Evaluation." December 1, 1980. j
- 9. NED0 30851 P A . " Technical Specification Improvement Analyses for BWR Reactor Protection System," ;
March 1988. '
- 10. UFSAR, Table 7.2 4. I
- 11. NEDC 31336 " Class III. October 1986, General Electric Instrument Setpoint Methodology." i
- 12. NED0 32291, " System Analyses for Elimination of Selected Response Time Testing Requirements," January 1994: and Fermi 2 SER for Amendment 111. dated April !
c '18, 1997. i k' - 13. NEDC-32410P-A, " Nuclear Measurement Analysis and i Control Power Range Neutron Monitor (NUMAC PRNii) < Retrofit Plus Option III Stability Functions," October I 1995, and Supplement 1, May 1996 )
~
1 i l j'D %) _ l FERMI UNIT 2 B 3.3.1.1 - 34 Revision 2, 01/18/99
o 5 PECIF'ICATiod 5.3.l./ () (Nsosut S u'4casw &4.i) SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
~ l 2.2 llMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS
( c.e.y sne reactor protection syJ4em instrumentat/on setpoints shy De sy (confastent with the Trip SetpoMt values shown ffi Table 2.2.1-1ff APPLICABILITY: As shown in Table 3.3.1-1. ACTION: '@ With a reactor protection system instrumentation setpoint* less conservative i than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is rectnrarf to OPERABLE status (gs setpo/ftt aaJusy consiste;rwith the Tpfp SetpoinMalue. / i O Ylicafim b4,1 ! i
*The APRM Simulated Thermal Power - Upscale Functional Unit need not be l hl declared inoperable upon er.tering single recirculation loop operation provided the Flow Biased setpoints are changed within 4 hours per q Specification 3.4.1.1. l FERMI - UNIT 2 2-3 Amendment No. EJ. 122 PAGE / OF 11 f*/1 l
O O O i T146LE 3 '3 l.1-I TAflE 3,3,l.l-l
" CEZ22I 3' Func720^l REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS d) '
. tum,riONAL UNIT ALLOWABLE
[ ' TRIP S d lNT ; 5 .g. cl Intermediate Range Monitor, Neutron Flux - liigh s 120/125 livisions of. s 122/125 divisions VALUES _
- 2. Average Power Range Monitor: of fu scale of full scale
[ i
- 2. q #1 Neutron Flux - Upscale (Setdown) '
s .of RATED s 20% of RATED l~ L 7.b Ar. Simulated Thermal Power - Upscale lilERMAL POWER
/'HERMAL POWER l r
- l. Flow Biased Q s 0.63 (W AW)' l.4%, s 0.63 (W- AW) *464.3%,
g3 with a ma of l .
- 2. Iligh Flow Clamped with a maximum of I j rrt s 113.5% o RATED s 115.5% of RATED 1HERMA POWER { !
TilERMAL POWER i 2.C.,e. Neutron Flux - Upscale b s 11 of RATED s 120% of RATED
" T RMAL P09ER O
E Zd X Inoperative TilERMAL POWER i NA 2.E pt ! 2-out-of-4 Trip Voters NA ! NA g { 3 J; Reactor Vessel Steam Dome Pressure - High s 1093 ps - 5 1113 psig to l 9 4- Reactor Ves'sel low Water level - Level 3 173 inches *
'b a 171'.9 inches i
E z r C*See Base / Flour /R VaA-Q O E h l
'flhe verage Power ge Monitor Si lated Thermal Pow g fu tion of recir lation loop dr ve ficw (W). AW i - Upscale Flow B sed scram setpoint arfes as a M w rcent of drly flow which nro ces rated enro F1 efined as the d erence in indicat i driveatflow ( I P
i ore flow. AW = 0% for two loop operation. AW = 8% for single loop operation. hetween two lo and single loop o ration the am !{ -! b $ Tbl534I- lg E w
~ Nolf-(O I.h.1- N :
E-
~
~
{
-_ _ . _ _ . _ - _ . - _ _ _.m ______m_m____ - - _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .
-___ _,-.-t__m , . - , , --+--r%--,-..-...--e-,-.w .c.<-._r-.-++--,e--- .- *. . m.- c-eee... -w- -4s
l \ l l 1 l I 3/4.3 INSTRUMENTATION s PEC I FI CATrord 5. 3. /./ 3 /4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION tiMITING CONDITION FOR OPERATION 4*i 3-1 As a minimum, the reactor 3.31 l in Table 3.3.1-2 shall be OPERABLkrotection system instrumentation channels shown APPLICABILITY: As shown in Table 3.3.1-1. d l}DD AL71pd3 Nowk U ACTION: _ i a. With the number of OPERABLE channels less than required by the Minimum l OPERABLE channels per Trip System requirement for one trip system:t LC. )
- 3. With Functional 'JnitlwHhia t'e- I '
/fc770AlC b,inIhour,,c,ifyusel"eac1- - P"Y "' _
' " ."nuf v.3"3".....
Mff LJ),3 l"" .N.o*?r*
!pptd co-d ' 4 " g . ....
. .( "'"'Rqion Ris k;pfaf a M Q ,5*D' U.. h'""9.
u...m. , g , ACT70A/ 4
!; !I "
5!' ! 2 !b;d-l A.3 le OFET^EE atets withi hours or the ACT16N required by Table popj .D 3.3.1-1 for the affected ional Unit shall be taken. l 3. -if -ht!n; +ha 4 pr- d k ........J.' i in J edd n;; n... ; ::nn place the Inopera., t. ,hy;d ;er.dit t er.ble c annel(s) and/or t l gNgA that trip system in the tripped condition within 12 hours. b With the numbe'r of OPERABLE channels less than required by the Minimum kCTiod 6 OPERABLE Channels per Trip System requirement for both trip system . L.'Z Aatpd 4.0 place at least one trip system" in the tripped condition within and t_n kp_ t he ACTION rannired by T@h- 3. 3.2 , _ ! hour i Y - - .Mm 12 hadz kip o%er cWrseh er &ie sV SW) l p) G
- c. With one orwreTnanness requireo Dy Iaore 3.3.1 1 inoperable in one or more APRM Functional Units 2.a. 2.b 2.c or 2.d:
C-
- 1. Withiri I hour, verify sufficient ' channels remain OPERABLE or k LTON C tripped *** to maintain trip capability in the functional Unit, and ticT1Dg g 2. Within 12 hours, restore the inoperable channels to an OPERABLE I status or tripped ~~.
Otherwise, take the ACTION required by Table 3.3.1-1 for the Functional
- Unit.
1 gef& A.tNok-; Aaton B NHe) l J8 Actions a and b not applicable to APRM Functional Units 2.a. 2.b. 2.c, and l
-- 0 2.d. Action c applies only to APRM functions 2.a. 2.b, 2.c and 2.d.
*g.f.upeaupJn.nnepeuwnuyoepasc7in i cause ,p 'hae 6 n e- t r i co sso r n n n ,1 -o n wnere riu I
.1 trram n ner"r e=eatfetnsinu;cioviecnanne ia nn61 wa s fi tr[i an. i[i ' gg re s to ryu 6uvrumou , < ~ = - > < pr/t he c b e t e rWri n ed i n no i n de r$ n w af i 6. i ->abir erfthe ACTION required by f able 3.3.1-1 for that
.Ac T104f [,pFunctional ne Unit shall be taken.
out or "> : - = m ppe ca r ""
,1_
4 ini - - ,,l 1-- [cause I term tn6r mn af.n rur @'enem ne /c -h gwodi'>vn . . Uwu '$ w. ; ny a scramu to ;r og te: ur, nn b; the :-'i system wit the
. r. it ; t ,-; Va "
place most no erable chann s in the tripp d condition: if ot systems h e th b'Y sam numger of inoce hla channalt- laen either tr svttem in the trinn
.. W "!b- "
An inop rable cnannet n would ause a scram to >0 occur. not be placed In thes cases, in theif'the tripped i cerabledition where thi ] channel A.3
. not stored to OPE LE status withi the required 11 , the ACTION re red, L by ble 3.3.1-1 for the Functional it shall be tak nf-fq- FERMI - UNIT 2 3/4 3-1 Amendment No. JE, D , JEE, 122 C)
L
~
PAGE 4 0F 11 y l l !
t O
. W G C D fl f A TIO h5 5 3.l* I 3/a.3 TNSTRUMENTATION LIMITING CONDITION FOR OPERATION (Continued) i SURVEItLANcE ntuul E nEn h ge '
- I!- Each reactor protection system instrumentation channel shall be Arsr a demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL.
FUNCTIONAL TEST, and.CNANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1. 4,// s# 3 311.s f4.3.1.0 LOGIC SYSTEM FUNCTIONAL TEST 5W '-mM r'--M: -- ^ --M 6Bof all channels shall be performed at least once per la months, except "able 8 4.3.1.1-1, Items 2.a. 2.b, 2.c. 2.d. .and 2.e. Functions 2.a. 2.b. 2.c, and I i 2.d do not require separate LOGIC SYSTEM FIMCTIONAL TESTS. For in-+4- f e' i StJ.3.fl.14 tests shall be performed at least once per 24 montMhe LOG SYSTEM l I T t um. lui for r ins; ' ' 5; ieu ng APRM tr I stothe2-out.o7-4TripVeterchann to conditions a ion 4.. the RM channel in check all e inations of t trippedinputstvthe2-out-of-4}rt oter logic in
~
W ter channels. eJ
' 3 M l? ' .2. : .0 The REACTOR PROTECTION SYSTEM RESPONSE TIME of reactor trip functional unit
- shall be demonstrated to be within its t at least Ance lieutron detectors are exemot from response time testing; '
[g per wa618 months._
. . . . uw . at isast . .. .. : per tri system sucn 5 pare sted at la t once every 6 s a i ce s ines 18 mon where N is t total n rj
{ edundant ch nels in a spe fic reactor ip system. , e 3.s1.1,17 g g gg 3,5,t,j,17 i doTEl y3 L _^_ _= enq STAGGEKeo 753r8As - 4 v SR 3 3.l.l.Il AD7E 2- ,g l *The sensor response time for Reactor Vessel Steam Dome Pressu - High a Reactor Vessel Low Water Level - Level 3 need not be measured ndfrhyjce y gypto pe sne owny' sen...~ . =- j4me. g
~
FERMI - UNIT 2 3/4 3-la Amendment No. 7J, Jpp, J/J.122 PAGE T_._ OF 11 /W1
TABLE 3.3.r-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION M M1 9' T8L 3 31,j-l APPLICABLE - A'71* A : MINIMUM 5 FvNcTrord OPERATIONAL OPERABLE CllANNE tUNCT10NAL UNIT 'E CONDITIONS PER TRIP SYSTE a 7 ACTION i 5 1. Intermediate Range Monitors : [ ~ f ,4 af Neutron Flux - High. , 2 3 1 4 ' g "'ti@ ~
@ j(D 3 I f,b Jr: Inoperative i
-3, 2
' m O
3 1 c G
-< 5 3(d) 3 I i
- 2. Average Power Range Monitor. I 23 ,a. Neutron Flux - Upscale (Seldown) 2 3(k) 1 Q
- 7.b #: Simulated Thermal Power - Upscale F 1
3(k) 4
$ yj p Neutron Flux - Upscale 1 3(k) 4 F
[ g 4: Inoperative 1, 2 3(k) 1 q ; O 2.c 4 2-out-of-4 Trip Voters I, 2 2 I s 71 - 3 d' Reactor Vessel Steam Dome g Pressure - High 1, ,y 2 I G ; cf A. Reactor Veisel Low Water level - ! Level 3 1, 2 2
- 1 G q 5 5' Main Steam Line Isolation Valve -
g Closure lh b 4 F f g h h i E ! a w : U i __ .__ - _ _ ._ _ _ _ . . . - . _ . . . . _ -. _ . . . . _ . ~ _
l i l TABLE 3.3 (Continued) REACTOR PROTECTION SYSTEM INSTRUMENTATION */ TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours for i gg $k L, required surveillance /without placing the trip system in the trippea j condition provided at least one OPERABLE channel in the same trip syst]em; is monitoring that parameter. f I e e n $ . [ ICJ unie adequate s utdown mar n has been de nstrated p ' Sp ification 3. .), the "s rting links" all be rem ed from t RP c cuitry prio to and dur g the time a control ro is withdr . 1 (d) When the 's rting link are removed, he Minimum PERABLE Ch nels Pep Trio Syste is 6 IRMs d per Specifi tion 3.9.2 2 SRMs. l ICI et>t>_ _ h l Thi unction i et required t e OPERABL en the re/ctor pressuf'bf v, ,1 s .,4 4, ....a -- e - cu.u. 2 1 _f-- U [7.2 . uuu witchis[nshallbe t in the Ru omaticas osition. - ypasseo wnen Tjre reactoy vy-- h) s functi is not req d to be OPpmBLE when PRIMjKY CONTA/NPILMy EGRITY ot recuire .
- - =
T6l 33II' i) With any control rod withdrawn " M
- W 0 J ~~
N8k@ 7-- c,.ortrin+<nn 4 e in 1 . s /;a; ;--l7 -i::M Juu vi s --
.vu. . . ... . J -
Vedfie (d) This function shall be ::t:::ti :1l., bypassed when ."ii.6 L.L;..; n.yv l i 5R ul l.16 v> = a a ui = >> L .i v>ii. n ..;12-4 t; THERMAL POWER 1;; ^ 5 30% of i t RATED THERMAL POWER. Add 54 4 Itm Verificd'mk /II'3 l
. (k) psee+-each APRM channel provides input to both trip systemsfi = .. ._-
TTL rop apie cnan is specifieo in Die 3.3.1-1 are e totai APRM annels l J j M t.f-f r utred (1. ., it is not on trip system basi . The 6 hour lowed ! gg (c j est time complete a cha el surveillance t t (note a d applicabl provided at lea two OPERABLE ch nels are mo(n) i ringath' ve) is ) gramet . - - l l l.R 2 t require for co o[rodsremovedp[ Specification [9.10.1or/9.lD2. O - FERMI - UNIT 2 3/4 3-5 Amendment No. 75. E7.122 PAGE 4 0F 11 Rev 2
^
.:?.a .LI-I . -
TABLE 4.3.i.;-i A REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS i i M Y 8L 3 3 /<l"/ CllANNEL OPERATIONAL ~ F'tes/cTrad CllANNEL FUNCTIONAL CHANNEL CON 0!TIONS FOR WHICH I EWt'NCTIONAL UNIT . CHECK TEST CALIBRATION (a) SURVEILLANCE REQUIRED 3 <') <w > t
.,,, k Intermediate Range Monitors: l4 -
' . .x Neutron Flux - High S,(b) W W-O> SA-Di 2 IA SA-<lt LV
-(l) f W .<s-> 5 Ig dr. . Inoperative NA W -( 4,5 ) NA 2, 5
- 2. Average Power Range Monitor (II:
[(2) f g. ), Neutron Flux -
% - - - . Upscale (Seldown) D,(b)g7p SA .2 years [ g,gy 2
{ {'
'G3 k(m) C m (dote}-- L- M.4 2J f. Simulated Thermal Power - Upscale D -(2')
LA.9 S i 3)- W ), 2 year i g g .c. Neutron Flux - Upscale D-(2 ) M'O \3>-W(d), 2' yea (11) I [ da g #: Inoperative NA' SA-Qt.) NA 1, . 2 l M p- 2-out-oF-4 Trip Voters D-(Z) SA-(iQ NA 1, 2 3d
-"~
Reactor Vessel Steam Dome t Pressure - High 5 -(i ) k )--<t i>> R-(itf) 1, 2 N k Q(M9) N.
- 8. 4 d g"
Reactor Vessel low Water h Level - Level 3 Sd> Q(k)- < ',) x- D 'O 1, 2 U 3 M9') ' l' r6K ~~ Main Steam Line Isolation
~
Valve - Clo,sure NA Q -(4) R - < i4) I .i n
?{K Main Steam Line Radiation -
liigh 5 --d ) Q M R-( *{ } 1, 2 "7 /6coses.3.,.<.y.To Drywell Pressure - High 5 -(f) (k)M *
'R-<i9) 1. 2 7g /) DD s R3 3.1.l.lf A OTE t Y J R
[ ffp0 SA.3.3.l.t.iz.tJoTY 5 R 3 3 I'l < X ') 1 I
( ( 0 V i
\
i b 3 3.f.1-/ TABLE-4.3.i.i- H Continued) b REACTOR PROTECTI0'l SYSTEM INSTRUMENTATION SURVElttANCE PEOUIREMENTS b pew 2.s.t.u-t
-4 WNg CllANNEL OPERATIONAL.
CilANNEL FUNCTIONAL CHANNEL
" 7 FUNCT10NAL UNIT CHECK CONDITIONS FOR WHICH TEST CALIBRATION SURVElttANCE REOUIRED 3
O
- 8. Scram Discharge Volume Water m Level - liigh Q a. Float Switch NA Q
#(4 )
R -(19 1,2,5(j) '
- 7. a b. Level Transmitter 5-(t)
* .a -
Q(N80 R 48'4) 1,2,5(j)
- C
/_ 9. Turbine Stop Valve - Closure NA Q4)
Q49) R -( 84) to 10. Turbine Control Valve Fast h(s.% 6e 4,7 h - Closure NA Q.(9) NA h J l L g"rl 11. Reactor Mode Switch > g Shutdown Position HA R-61) HA 1,' 2, 3,
~ 5 N l i2.12. Manual Scrani NA LM W 4'5) NA 1, 2, . 4, 5 13 . D&hd. - .
t' 5 U 3.5l.38 4t bN'3
$/i 1.'),t.!.'1 l4oM tetw westren detectors may be escluded fran cMAnttt cAtitaallow. ff 30 I"HP' inh **'y SlM . '^ i ys
- W it) )p the IRM y 7,,,,,,,k the IRM and sRM channels shall 'be deteretned and APEN channels shall be determined to everlap o e to everlap f opeef refJ./sduring ease % o,ttetfalfduring each each startup ts,Jarr enteryg p ttistivant iman t rips 71ond i
H " ' " ^ ' ' ed shutdown, if not performed =tthin the pre,teus 7 days. tel _ this eatibration sheit consist of the adjustment er.the'"- ' ; '-- .K- L G.l 5431!!.35:s aran skannet to confor
'f1 t re to the , er valves ceie.1sted by e heat betence decine ortsArtcivAt R !
CONDillog I nhen THERMAt POWER g 25% of aAttO THERMAL POWER. Adjust the APRM chene 1 If the absolute dif ference ts greater p then 22 of aA*
.: . - . T'.- T. , -
u n.ia.h m: ,. W'.,e tri s sheii busiibreied at iesst .ece rer iceo i y Wy =. Q)
?'-"'
-.. - n
.. - m .,i, the tir 3,st-g LewoiB l
t
,,. g 4
. ;U!
- . _;. . _ ..... . - .....r _m _ m .. ... -
Wit h any cent rol red = t thdrawn * ;" -- - ": ":_ ' _....;roa .....:^^^.
. __ 4 * .
~ ' " '
gg I w -
; ? - .
.m-Jr 52 3 3 dei.lD[w p ,Inctodes_vertfleetton cf the.tr pl
._u s e t.mroint.c. 7_ _f the trip unit,
-gg -
(4 3/3.(.l.17Y itot required to be perfereed when enter ng MODE 2 frew Mott i entil 12 hours af ter er.tering M3DE 2. g % l = Q CLN z
=
sa N L
~
DISCUSSION OF CHANGES ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION ADMINISTRATIVE l A.1 In the conversion of the Fermi 2 current Technical Specifications (CTS) to the proposed plant specific Improved Technical i l Specifications (ITS), certain wording preferences or conventions l l are adopted which do not result in technical changes (either l actual or interpretational). Editorial changes, reformatting, and ) revised numbering are adopted to make the ITS consistent with the ! Boiling Water Reactor (BWR) Standard Technical Specifications NUREG-1433. Rev. 1. l l A.2 ITS LC0 3.3.1.1 Actions are modified by a Note, which provides clarification that, for the purpose of the associated LCO, i
" Separate Condition entry is allowed for each channel." This is acceptable because the Required Actions for each Condition provide i appropriate compensatory actions for each inoperable RPS channel. l Complying with the Required Actions will allow for continued ,
operation: with subsequent inoperable RPS channels governed by i subsequent Condition entry and application of associated Required ! Actions. This is an administrative change with no impact on safety because the clarifications provided by the Note are consistent with a reasonable interpretation of the CTS. A.3 CTS 3.3.1. Actions a.2, a.3, and footnotes *, **, and ***, provide l directions on options for compliance with Technical Specifications (i.e.. optional direction for not tripping channels if it would cause a scram, and optional direction to restore the inoperable , channels to Operable status). ITS 3.3.1.1 Actions do not include l detailed direction for these options: rather the optional Actions presented rely upon the guidance of LC0 3.0.2. The LC0 3.0.2 guidance allows defaulting to other Actions if one can not, or is l desired not to be complied with, and also allows that restoration
- within the time limits of the specified Required Actions. The CTS l has been revised to delete these Actions since these options always exist, and are inherent in the ITS. During this presentation reformatting, no technical changes (either actual or interpretational) were made to the TS. The change is consistent with NUREG 1433.
1 1 l
'd FERMI - UNIT 2 1 REVISION 2, 01/18/99l l
L .
- DISCUSSION OF CHANGES ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION l
l A.7 CTS Table 4.3.1.1-1. Functions 9 and 10. state the Applicability for the Surveillances as Mode 1, while the same Functions are ] listed with an Applicability for Operability of Mode 1 at ; i 2 30% RTP in CTS Table 3.3.1-1. ITS corrects the disparity with a single Applicability of 2 30% RTP. Since Surveillances are not required when the associated equipment is not required to be Operable, or is inoperable, this change reflects clarification of existing requirements, and is therefore an administrative change. J A.8 CTS Table 4.3.1.1-1 Note (f). requires an.LPRM calibration frequency of "once per 1000 EFPH." ITS SR 3.3.1.1.8 replaces this l frequency with "1000 MWD /ST." Both Frequencies consider the LPRM sensitivity changes based on neutron flux exposure, and represent ; approximately the same time interval (about 6 weeks). The units ! change allows a more convenient tracking parameter since MWD /ST is ; commonly calculated and reported by the process computer. Since the actual change in frequency of calibrations remains essentially the same, the change is considered administrative. A.9 CTS Table 3.3.1-1. Action 5, contains the requirement to "be in . STARTUP" in addition to the requirement to close the MSIVs. Since 1 MSIV closure in Mode 1 would result in an immediate automatic l scram. the need to "be in startup" to close the MSIVs is simply an operational design detail that is implicit in the requirement to close the MSIVs. This detail is therefore not included in ITS ! 3.3.1.1 Action H. Its elimination does not modify any requirements. Additionally, CTS Table 3.3.1-1, Action 5. requires this MSIV closure in 6 hours, but also provides an option to be in Mode 3 in 12 hours. Since the option is an acceptable alternative for compliance with the Action, the 12 hours is implicitly allowed for either option (close MSIVs or be in Mode 3). For example, if the option to shutdown was initially chosen, but at hour 11, the alternative action to close MSIVs was completed, the CTS Action would be satisfied without a Technical Specification violation. Therefore, to clarify the presentation, both actions are presented in ITS 3.3.1.1 Action H with a 12 hour Completion Time. Since no i technical changes are introduced, this change is considered I administrative. l LO FERMI UNIT 2 3 REVISION 2 01/18/99l 1
i l l DISCUSSION OF CHANGES , ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION A.10 CTS 2.2.1 specifies Actions for RPS channels with setpoints not within limits. These Actions state " declare the channel inoperable and apply the applicable ACTION until the channel is < restored to Operable ." ITS 3.3.1.1 contains all the requirements and Actions for RPS channels, including the setpoints. The usage rules of Technical Specifications adequately l dictate that channels be declared inoperable and Actions taken i until restored to Operable, without an RPS-specific statement to l that effect. Therefore, elimination of this statement is an administrative presentation preference only. A.11 CTS 4.3.1.2 requires an LSFT "and simulated automatic operation" of all channels. The " simulated automatic operation" is interpreted to be synonymous with the LSFT and/or Channel Functional Test. This additional detailing of the required test is unnecessary. Therefore its elimination is an administrative change. A.12 Not used. l TECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS 3.3.1 Action a.1 specifies actions in the event of a loss of scram function (potentially one entire trip system without Operable scram capability). Footnote
- to this Action provides an extension of the allowed action time from I hour to 2 hours if tripping the channel would result in a reactor scram. ITS 3.3.1.1 Action C limits this time to only one hour regardless of whether tripping the affected trip system would produce a scram (and thus eliminates the additional hour allowed in CTS). In both the CTS and the ITS, after this 1 or 2 hour allowance, provisions are made l for a controlled reactor shutdown. Therefore, eliminating the flexibility of the additional hour will not introduce any adverse consequences to safety. l n
FERMI - UNIT 2 4 REVISION 2 01/18/99l
1 (q y DISCUSSION OF CHANGES ITS: SECTION 3.3.1.1 RPS INSTRUMENTATION 1 LA.2 CTS Table 2.2.1-1 footnote #. Table 3.3.11 Action 6. and footnotes (b). (g) (j), and (k), provide design details and descriptive details for various RPS functions. ITS 3.3.1.1 addresses this information in the Bases and does not include these l details in the Technical Specifications. This change is l consistent with NUREG-1433. The information is being moved to the Bases, which requires changes to be controlled in accordance with the ITS 5.5.10. Bases Control Program. This relocation continues to provide adequate protection of the public health and safety since the requirement for instrument channel Operability continues l to be required by the Technical Specifications. LA.3 CTS 3.3.1 Action a.1 addresses the necessary requirements for a loss of RPS scram capability (one trip system with more than one inoperable channel in any Functional Unit) by detailing one option (i.e., trip inoperable channels or trip system). ITS 3.3.1.1 Action C has relocated specific details of restoration from the loss of function, by specifying " restore RPS trip capability." ITS 3.3.1.1 addresses the option of tripping the trip system in O the Bases and does not include these details in the Technical
;b Specifications. This change is consistent with NUREG 1433. The information is being moved to the Bases, which requires changes to be controlled in accordance with the ITS 5.5.10. Bases Control Program. This relocation continues to provide adequate protection of the public health and safety since the requirement for loss-of.
function restoration continues to be required by the Technical Specifications. LA.4 CTS 3.3.1 Action b is modified by a footnote stating " place the trip system with most inoperable channels in the tripped condition." which applies to situations when both trip systems have inoperable channels. The specific detailed direction for which trip system to trip is relocated from Technical Specifications to the ITS 3.3.1.1 Bases. This change is consistent with NUREG 1433. The information is being moved to the Bases, which requires changes to be controlled in accordance with the ITS 5.5.10. Bases Control Program. However, the Bases allow
" prudent judgement' in determining which trip system should be tripped. Although placing the trip system with more inoperable channels in the tripped condition may provide a small increase in reliability, since RPS is still assured of maintaining trip q capability for each Function (Action C would be entered if not Q maintained) other considerations may make it more prudent to place the other trip systems in trip. Therefore the Bases direct this i l
FERMI - UNIT 2 7 REVISION 2 01/18/99l
O DISCUSSION OF CHANGES ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION LA.6 CTS 4.3.1.2 provides LSFT performance details for the 2 out-of-4 Trip Voter logic (i.e., " simulating APRM trip conditions at the APRM channel inputs to the 2 out-of 4 Trip Voter channel to check ' all combinations of two tripped inputs to the 2 out of 4 Trip Voter logic in the Voter channels"). ITS SR 3.3.1.1.19 requires ! this same LSFT however the performance details are relocated to the Bases for ITS SR 3.3.1.1.19. which require changes to be controlled in accordance with the ITS 5.5.10. Bases Control Program. These LSFT performance details are consistent with the 1 ITS definition of LSFT. providing only the component-specific details of the tested channel. This change is consistent with the level of detail found in NUREG 1433. The relocation continues to provide adequate protection of the public health and safety since the requirement for performance of an LSFT (and the definition of an LSFT) continues to be maintained in the Technical Specifications. LA.7 CTS Table 4.3.1.11 Note (f) requires that LPRMs be calibrated "using the TIP System." ITS SR 3.3.1.1.8 requires LPRM ( calibration at the same Frequency; however, the fact that LPRM V gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System, is relocated to the Bases, which require changes to be controlled in accordance with the ITS 5.5.10. Bases Control Program. This change is consistent with NUREG-1433. The relocation continues to provide adequate protection of the public health and safety since the requirement for LPRM calibrations at the required Frequency continues to be maintained in the Technical Specifications. LA.8 Not used. l
- i. LA.9 CTS Table 4.3.1.11 Note (e) requires that the calibration for the APRM Simulated Thermal Power - Upscale trip include the " flow input function, including flow transmitters." ITS SR 3.3.1.1.18 requires the APRM calibration at the same Frequency: however, the specific detail that the flow transmitters and associated flow function is included, is relocated to the Bases, which require changes to be controlled in accordance with the ITS 5.5.10. Bases Control Program. This change is consistent with the level of detail found in NUREG-1433. The relocation continues to provide adequate protection of the public health and safety since the
(~} requirement for APRM calibrations at the required frequency (_/ continues to be maintained in the Technical Specifications. FERMI UNIT 2 9 REVISION 2 01/18/99l
DISCUSSION OF CHANGES ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION i l ; LA.10 CTS 4.3.1.3 requires Reactor Protection System Response Time testing of "each reactor trip functional unit": however, the details of the testing acceptance criteria are located in the ! Technical Requirements Manual (TRM): outside of CTS, ITS 3.3.1.1 1 provides the Reactor Protection System Response Time Surveillance ; (SR 3.3.1.1.17) and each Function on Table 3.3.1.1-1, as appropriate references the applicability of this test to that ' function. Since many of the RPS trip Functions have no specific acceptance criteria detailed in the TRM these Functions will not i have ITS SR 3.3.1.1.17 listed as a Technical Specification f required Surveillance. For these Functions. the Reactor : i O ! I I i t l FERMI - UNIT 2 9(1) REVISION 2 01/18/99l 7 y--
\ l rq DISCUSSION OF CHANGES ' ,Q ITS: SECTION 3.3.1.1 RPS INSTRUMENTATION I 1 LR.1 CTS Table 3.3.1 1 Action 9 requirement that the reactor mode l switch be locked in the shutdown position is removed from , Technical Specifications. This change is consistent with NUREG- ) 1433. The requirement to fully insert all insertable control rods ; remains as ITS 3.3.1.1 Action I. Regulatory control of changes to this requirement (e.g., Technical Specification amendment or 10 < CFR 50.59) is not necessary to provide adequate protection of the public health and safety since the requirement for all control { rods to remain inserted continues to be maintained in the i Technical Specifications. I LR.2 Not used. l LR.3 CTS Table 4.3.1.1-1 footnote (b) for Functions 1 and 2 (IRMs and i APRMs), requires that the Channel Functional Test for these j functions include verification that instrument indication overlaps i (SRM to IRH and IRM to APRM) for at least 1/2 decade. ITS SR 3.3.1.1.6 and SR 3.3.1.1.7 require verification of IRM and APRM I instrument indication overlap: but an acceptance criteria is not 1 l'O 'b specified. The acceptance criteria is removed from the Technical Specifications, consistent with NUREG 1433. Regulatory control of changes to this requirement (e.g., Technical Specification amendment or 10 CFR 50.59) is not necessary to provide adequate protection of the public health and safety since: 1) the requirement to verify overlap is maintained: 2) overlap is verified if both instruments (SRM and IRM or IRM and APRM) are onscale and tracking changes in neutron flux levels: 3) neutron monitoring channels sufficient to tolerate a single failure are required to be available and 4) once verified to be onscale, comparison of the response of the multiple IRM and APRM channels that are available provides indication that neutron monitoring instrumentation is functioning properly. This change has no impact on safety. FERMI UNIT 2 12 REVISION 2 01/18/99l 4 1
O DISCUSSION OF CHANGES ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE
" Specific" L.1 CTS Table 3.3.11 Action 6 requires that operators reduce Thermal ,
Power to less than 30% within 2 hours when the required number of ; channels of the Turbine Stop Valve-Closure or Turbine Control i Valve Fast Closure RPS Functions are not Operable. ITS 3.3.1.1 i Required Action E.1. allows 4 hours to reduce power to less than , 30%. Allowing an additional 2 hours provides the necessary time to decrease power in a controlled manner, within the capabilities ; of normal shutdown procedures for the unit. This extra time ' reduces the potential for errors that could challenge safety systems. This change is consistent with NUREG 1433. This change ; has no significant impact on safety based on the low probability ' of an event during this increased time, and based on the time being consistent with the Completion Time provided in ITS : LC0 3.2.2. "MCPR" (which the TSV and TCV closure Functions are intended to protect). L.2 CTS 3.3.1 Action b requires placing one trip system in trip within 1 hour when both trip systems have inoperable channel (s). ITS 3.3.1.1 Action B allows 6 hours to trip one trip system (note that ' ITS 3.3.1.1 Action C within 1 hour assures RPS trip capability remains). This allowance is consistent with NUREG 1433. In the event of inoperable channels in both trip systems, with RPS trip capability remaining. single failure protection may be lost in both trip systems. However. based on the remaining capability to trip, the diversity of the sensors available to pravide 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, this extension to 6 hours will not adversely affect safety. L.3 CTS Tables 3.3.1 1 and 4.3.1.1 1. require that the following RPS functions are Operable and tested in Mode 3 and/or Mode 4: l Functions 1.a (IRM flux high),1.b (IRM Inop).11 (Mode Switch l Shutdown Position), and 12 (Manual Scram). ITS 3.3.1.1 does not require the Operability of these RPS functions in Mode 3 or 4. The Actions associated with these Functions for Modes 3 and 4, (CTS Table 3.3.1-1. Actions 2. 7 and 8), are also eliminated. This is consistent with NUREG 1433. Elimination of Operability O requiremsnts for these Functions in Mode 3 and 4 is acceptable because all control rods are fully inserted and the Reactor Mode l I FERMI UNIT 2 13 REVISION 2 01/18/99l
DISCUSSION OF CHANGES O ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION L.5 Not used. l l l i 1 I l l
.(
FERMI.- UNIT 2 15 REVISION 2 01/18/99l
l l l ( DISCUSSION OF CHANGES ( ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION l L.8 CT3 Table 4.3.1.1-1 requires weekly and semi-annual Channel l Functional Tests and semi-annual ("SA") Channel Calibrations for IRMs and APRMs. ITS SR 3.3.1.1.4. SR 3.3.1.1.11. and ! SR 3.3.1.1.12 requires these same Surveillances, but are modified j with less restrictive allowances: the Note to SR 3.3.1.1.4 and i SR 3.3.1.1.12 and Note 2 to SR 3.3.1.1.11. provide a 12 hour j allowance to perform the Surveillances after entering Mode 2 l during a plant shutdown, without considering the Surveillance not l completed if the required frequency was not " current." l l The exceptions for entering Mode 2 with Surveillances not current is allowed since testing of the Mode 2 APRM and IRM Functions cannot be performed in Mode 1 without utilizing jumpers lifted l leads, or movable links. Twelve hours is based on operating experience and in consideration of providing a reasonable time in which to complete the SR. L.9 CTS Table 4.3.1.1-1 for APRM functions requires a weekly channel calibration to a heat balance when 2 25t power. ITS SR 3.3.1.1.3 l requires this same heat balance calibration, but also includes a ! [V-) Note allowing the performance to be delayed until 12 hours after l exceeding 251 power. Since it is difficult to accurately maintain APRM indication consistent with a heat balance when operating l below 25% power, some time is necessary after exceeding 25t to ; establish steady state operating conditions and complete the heat balance calibration. The 12 hour allowance is based on operating experience and in consideration of providing a reasonable time in which to complete the Surveillance. Bfj,0CATED SPECIFICATIONS l l None j i TECHNICAL SPECIFICATION BASES I l The CTS Bases for this Specification have been replaced by Bases that reflect the format and applicable content of ITS 3.3.1.1 consistent with the BWR STS. NUREG-1433. Rev. 1. FERMI - UNIT 2 17 REVISION 2. 01/18/99l t
_ _ ..._ . - _ _ - . _ . . _ _ ~ _ _ . . . _ - . _ . _ . _ . _ . _ . _ _ _ . _ . . . _ - . _ _ L RPS Instrumentation i 3.3. L I l 3.3 INSTRUMENTATION i 3 L 3.3.1.1 Reactor Protection System (RPS) Instrumentation l l i LC0.3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.
'*f t
APPLICABILITY: According to Table 3.3.1.1-1. ACTIONS i l NOTE--- - - - - 5eparate Condition entry is allowed for each channel. Doc 4.7,\
/
\ CONDITION- REQUIRED ACTION COMPLETION TIME ' A. One or more required A.1 Place channel in 12 hours /lchs 4.t;o.3; pl channels inoperable. trip.
--WL--
bid 'l E r Not arYlicahte fa , gg l . L fu 2Aj7.6,183 i A.2 Place associated trip 12 hours L Q__ __f systen in trip.. g ;
- Y J heohnk. A)
B.( One or more functions Place channel-in one with one or more required channels B.3 trip system in trip. 6 hours
/jj,45 T g/i ;
i inoparable in both 3-trip systems. B.2 Place one trip system 6 hours in trip, y C. One or more Functions C.1 Restore RPS trip I hour with RPS trip capability. (4chs sl 3c.l} l capability not maintained. (continued) BWR/4 STS 3.3-1 Rev 1, 04/07/95 l
*d
!Q L
._ _ . _ _ . _ _ . . _ _ _ . . ~ _ - . _ . _ _ .- .._ . _ . _ . ._. _ _ .._ _ . _. _ . .
RPS Instrumentation , 3.3.1.1 l C75 SURVEILLANCE REQUIREMENTS I NOT E S --------------------------------- j
- 1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function. (y,3, j,f) :
)
1
- 2. When a channel is placed in an inoperable status solely for performance of l required Surveillances, entry into associated Conditions and Required 73453/-1, ,
Actions may be delayed for up to 6 hours provided the associated Function 8) l maintains RPS trip capability. ; i l SURVEILLANCE FREQUENCY j 1 SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours P - - , _. -- (tl.3.1.1 -1) S R 5 3.I./.2. Pe.c4wrc C HAMUEl CHemk. 1 A b e v r.5 SR 3.3.1.1@ ------------ NOT E---- Not required to be performed until 12 G)~ ) hours after THERMAL POWER 2 25% RTP. Verify the absolute difference between 7 days the average power range monitor (APRM) O P.I channels and the calculated power is s 2% Ryl.. ..., si;. 4;..: ..: ,
/E S 5 j at'25%
operating 2 RTPJd64';dp[hne s S 3.3.1.1. Adjust he channel conform a days f.I calibrat flow signa SR 3.3.1.1.4 ------------- NOT E------------------ 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 b (continued) BWR/4 STS 3.3-3 Rev 1, 04/07/95 twL 9
1 RPS Instrumentation p 3.3.1.1 5 SURVEILLANCE REQUIREMENTS (continued) (C75) SURVEILLANCE FREQUENCY SR 3.3.1.1.11 - - - - - - - - -
- 1. Neutron dete
-NOTES-are excluded.
D,/J-lf
- 2. For Function not required to be performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2.
Perform CHANNEL CALIBRATION. 184 days S 3.3.1.1 Perform CHANNEL FUNCTIONAL TEST. yl8pmonths SR 3.3.1.1 -
-------NOT . - / --
84 (1. Ne ron detector are exclude .
- 2. or Function , not requir to be performed n entering 2 from O MODE 1 unt 12 hours aft enterin
, ..)
Perform CHANNEL CALIBRATION. (18knonths SR 3.3.1.1.14 Verify the A Flow Biased mulated 18] months Thermal P r-High time co stant is s T71 sec ds. V . SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONAL TEST. V347. f18) months N -- ,
' N- - " - -
continued) O Tgg 3.3.3,3,n ror _ _. _. pots - - - - - - - - - - 1 nu% 2,a , no+ reg *idlo be P**%d kiben en4-*(ind Mp06 2. Orn AADOE 1. L4n Hl. l2. kovrs aW Cnkering _ - - - lADDE *2--
' ~ . _
BWR/4 STS Pwh CHAtMEL. FUNcMM4L T57
=- -
1 .J-n _ Mhd _ Rev J., 04/07/95
? \
Ra2.
l RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued) [ CTS > I SURVEILLANCE FREQUENCY SR 3.3.1.1.16 Verify Turbine Stop Valve-Closure and K18gmonths Turbine Control Valve Fast Closure A
-99%:styssgepunges. Functions are not t 7,,, g 1
4 ON bypassed when THERMAL POWER is h@ kg3gRTP. SR 3.3.1.1.17 ---
---------NOTES ----------
- 1. Neutron detectors are excluded. Q 3,l,3)
For Function 5 "n" equals 4 channels d3 for the purpose of determining M PA the STAGGERED TEST BASIS Frequency. Verify the RPS RESPONSE TIME is within (187monthson limits. a STAGGERED r TEST BASIS [
,.- L For Tvnc%s 3and 4 c%rd 54+tsof f:45Per1piinte.s art noF 3 If4M b Su fJ
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f A y Ygi RPS Instrumentation 3.3.1.1
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!HMrCD (continued)
(a) With any centrol red withdrawn from a core cett centelnire one or more fuel essemblies. (b) - -
^ 1C^^" b reset for sinste loop operetten per Lco 3.4.1, Meetreutetten Leops T ,
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. -... ._ . . ~ - . . . . - . . . . . . ~
l l RPS Instrumentation
#g[69
- O f Tebte 3.3.1.1 1 (page 2 of 3)
@I hek 1 Desctor Protection System Instrumentation l MWIEED OTMA CMNNELS FRGI SPECIFIS PER TRIP REWIggp $URVE!LLANCE ALLCmgLE 1. 2. e l~ !
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RPS Instrumentation [ 3.3.1.1 0\ febte 3.3.1.1 1 (pose 3 of 3) { 44 Beector Protection systee Instru entatten N7 fiA NCTlDN APPLICABLE tas!TIONS mees sa u sulass u P Restso 3. 3. / - / OfhER CNAMELs PtWI 4 SPECIFI S Ptt TRIP MallteD m RVEILLANCE REspiteMENis ALLOWASLE 2 2./- 1 PtsICTION CONDITICMs sYs194 ACTION 9.1 WALUE
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. Reactor leede switch - 1,2 G st 3.3.1.1. NA lI shutdoen Positten 2 st 3.3.1.1.15 g(s) st 3.3.1.1. NA 9
2 .T st 3.3.1.1.15 nanuel Scree 1,2 3 s st 3.3.1.1.5 st 3.3.1.1.15 RA (l'l h 5(*) st 3.3.1.1.5 NA st 3.3.1.1.15 o (a) With eriy control rod wIthdrain from a core cett centelnins one er more fuel assembtles. WR/4 STS 3.3-9 Rev 1, 04/07/95 O, g
_ _._ _ _ _ _ . .- ___... _ _ _ . _ - _ _ _ _ . _ _ _ _ __ ~ . - _ - _ . - _ 4 RPS Instrumentation
, B 3.3.3.1
( BASES APPLICABLE 1.a. Intermediate Rance Monitor Neutron Flux-Hioh SAFETY ANALYSES, (continued) LCO, and , APPLICABILITY System and the RWM provide protection against control r
]gg[Qgggwithdrawal error events and the IRMs are not reouired3
(% [gh / 1.b. Intemediate Ranoe Monitor-Inon This trip signal provides assurance that a minimum number of IRMs are OPERABLE. Anytime an IRM mode switch is moved to f,2- any position other than ' Operate," the detector voltage drops below a preset level, or when a module is not plugged in, an inoperative trip signal will be received by the RPS unless the IRM is bypassed. Since only one IRM in each trip system may be bypassed, only one IRM in each RPS trip system may be inoperable without resulting in an RPS trip signal. This Function was not specifically credited in the accident 4 analysis but it is retained for the overall redundancy and I diversity of the RPS as required by the NRC approved i licensing basis. Six channels of Intermediate Range Monitor-Inop with three channels in each trip system are required to be OPERABLE to ensure that no single instrument failure will preclude a .A g h scram from this Function on a valid signal.
! Since this Function is not assumed in the safety analysis, there is no Allowable Value for this Function.
This Function is required to be OPERABLE when the Intermediate Range Monitor Neutron Flux-High Function is h required.
] ggt- Averaoe Power Ranoe Monitor 93 4 ;
3 3.l,l- 8 J 3.a. Averaoe Power Ranoe Monitor Neutron Flux _ 9_ ' {Setdown')
"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. For operation at (continued)
BWR/4 STS B 3.3-6 Rev I, 04/07/95
%t
/N RPS Instrumentation b-B 3.3.1.1 INSERT B 3.3.1.1-8 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 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 1 a continuous indication of average reactor power from a few ; percent to greater than RTP. l l The APRM System is divided into 4 APRM channels and 4 2-out-of 4 ' 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 l 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 ! f3 voter channels, but no trip inputs to either RPS trip system. A V 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. Bl. and B2). Three of the four APRM channels and all four of the voter channels are required to be OPERABLE to ensure t%t 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 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, are required for-each APRM channel. l l O FERMI - UNIT 2 B 3.3-6 (Insert) REVISION 2 01/18/99l
e. I RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Averaoe Power Ranoe Monitor Neutron Flux M
~
f SAFETY ANALYSES, (Setdown) (continued) LCO, and APPLICABILITY low power (i.e. 2), the Average Power Range Monitor Neutron Flu , etdown Function is. capable of generating a trip si nal th)at prevents fuel damage resulting from abnomal operating transients in this power range. For Of.2, qE3g most operation at low levels the Average Power Range Monitor Neutron Flux $etdowhfunction will provide a ' secondary scram to th reediate Range Monitor Neutron Flux-High Function because of the relative setpoints. With IRMs at Ranoe 9 or 10._it__is oossille that the Average Power Range Monitor Neutron Fluxq.(Setdown) Function will provide the primary trip signal Tor a corewide increase in power.
" No-spee44e-s:fety =1y= t:ke-d4*ect ved"- f r th:
T h Average Power Range Monitor Neutron Flux (Setdo 15 CE67/TED, AltdG K//7g A Functions 4 Pn:=,'This Functiondndirect y ensures that 7N6 ffki before the reactor mode switch is pliiced in the run (pd(<T/Os)//64TB0d FWY-Mice
, Wint /#lT/A7/MG A # position, reactor power does not exceed 25% RTP (SL 2.1.1.1)
FC4cr0R f, GAM /d TB6 AN4u/5IS when operating at low reactor pressure and low core flow. of 7#6 t'MT/A/v#v5 400 W/;WPEAvML Therefore, it indirectly prevents fuel damage during A DURIAK EFACTOR 7 STARTvr EWA/T; f nt reactivity increases with THERMAL POWER k " f2 fTheAPRMgystemisdivide into two groups of ch nnels witn i fthree RM channel input to each trip system. he system is de gned to allow o channel in each trip ystem to be i byp sed. Any one AP channel in a trip sy em can cause I th associated trip ystem to trip. Four annels of erage Power Ran Monitor Neutron Flux digh, Setdown with wo channels in ach trip system are re ired to be OPERAB to ensure tha no single failure will reclude a scram fr this Functio on a valid signal. I addition, to provi adequate c erage of the entire co , at least 11 LP inputs ar required for each AP channel, with at 1 st twt LPRM in ts from each of the fo axial levels at ichthej UPRMs, re located. s The Allowab'le Value is based on preventing significant increases in power when THERMAL POWER is < 25% RTP. gP g The Average Power Range Monitor Neutron Flux'M(Setdown) Function must be OPERABLE during MODE 2 when control rods i may be withdrawn since the potential for criticality exists. I (continued) BWR/4 STS B 3.3-7 Rev 1, 04/07/95 m IM
RPS Instrumentation l [ B 3.3.1.1 BASES P.2. APPLICABLE 2.a. Averaoe Power Ranoe Monitor Neutron Flux [Hte - SAFETY ANALYSES, (Setdown) (continued) LCO, and APPLICABILITY In MODE 1 the Average Power Range Monitor Neutron j Flux,#35$\ Function provides protection against reactivity trans'ients and the RWM and rod block monitor protect agairist r p {g control rod withdrawal error events. ' 2.b. Averaoe Power Ranoe Monitor S h "i " M Simulated JAgraal Power The Average Pow Range Monitor Flow Biased Simulated Thermal Power _ AFunction monitors neutron flux to approumue sne 7HERMAL 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. The trip level is varied as a function of recirculation drive flow (i.e., at lower OP.1 car.W6 flows, the setpoint is reduced proportional to the reduction in power evnerienced as.nease flow is reduced with ___a fixed control rod pattern) but is clamped at an upper limit that is always lower .h the Average Power Range
'7 gfA MonitorWNeutron Flux- Function Allowable Value.
bs) (/ The Average Power Range Mon' torb ... ...._ Simulated Thermal PowerM Function provides protection against transients whire IHtRMAL 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 a lower trip setpoint, will initiate a scram I before the high neutron flux scram. For rapid neutron flux (ggj$gy/y 6 53'lel-b ; increase events, the THERMAL POWER la s the neutron g and h the Average Power Range Monitor Neutron Flux- 3 Functionwillprovi%ascramsignalbeforetheAvera qf[e, Q> Power Ran Monitord - a;eeedisimulated Thermal Power unction setpoint is exceeded. l
The APRM ystem is divi d into two groups of hannels with~ i four A inputs to e trip system. The stem is I desi ed to allow on channel in each tri system to be - l byp ssed. Any one PRM channel in a tr system can se t associated tr p system to trip. F r channels o i l V (continued) l BWR/4 STS B 3.3-8 Rev 1, 04/07/95 l'h V
del
i RPS Instrumentation ( B 3.3.1.1 l INSERT B 3.3.1.1 6 The required trip setting for APRMs is dependent on whether the unit is in single recirculation loop operation or two-loop operation. As specified in Table 3.3.1.11 footnote (b). The setpoint variable AW is defined as the difference in indicated drive flow (in i of drive flow that produces rated core flow) between two-loop and single loop operation at the same core flow. l l l l O 1 i l I l 0 l FERMI UNIT 2 B 3.3-8 (Insert) REVISION 2, 01/18/99 I
- . ,. . _ - - _ - . .1
1 RPS Instrumentation /^ B 3.3.1.1 \ BASES ( h [c) APPLICABLE 2.b. Averace Power Rance MonitorIF1MSimulated SAFETY ANALYSES, Therinal Power-4 (continued) '
~
LCO, and _ APPLICABILITY verage r er nange rioMT.or r low sia 'o simulated i rwas x Power- gh with two iannels in ea trip system rranged ) in a o e-out-of-two ogic are req red to be OPE BLE to ensu that no si le instrument ailure will p clude a ser from this nction on a v id signal. addition, to p vide adequa coverage of e entire core at least 1 LPRM inpu are required r each APRM c nnel, with least two L inputs from /ach e' t ' 'as 5 9 ? h ;; L et o which the WRMe are lar=* Each APRM channelw-- : 2 I Wtotal drive ct?? dr*nflow
!? signals
= :'; =? representative
.;^; ,......
of total core flow. WS M ' , iv . f? = Jnits, two which suppl signals to the rip system A 0^6 l [ g g$eTLT I LPRMs, whi the other t supply signal to the trip I system B RNs. Each ow unit signal provided by 6 5.3.bl- 9 6unning p the flow s nals from the rectreulation loops. To obtain t most conservat e reference signa , Op,g the tal flow sig 1s from the tw flow units (assoc ed rit a trip syst as described a ve) are routed to low nu ion circuit ssociated with ach APRM. Each AP 's i tion circui selects the 1 r of the two flow nit L gnals for e as the scram ip reference for at
> articular RM. Each requ ed Average Power nge Monitor low Bias Simulated The 1 Power-High cha el only )
equires n input from o OPERABLE flow uni , since the l Indifid 1 APRM channel ill perform the i ended function ! vith o y one OPERABLE low unit input. wever, in order : to ntain single f ure criteria for e Function, at l lea one required erage Power Range nitor Flow Bia d si lated Thermal wer-High channel each trip sy em o t be capable maintaining an OPE BLE flow unit ignal - in the event of failure of an auct n circuit, or a flow mit, in the a ociated trip system (e.g., if a f unit i
.noperable, o e of the two requir Average Powe Range Konitor F1 Biased Simulated Th al Power-Hi channel l n the ass lated trin m t.m - + k. eam, u- _ j -
nnparah g The :? r;;f Allowable Value is based on analyses that take credit for the Average Powe ange Monitor D o. Ci n-d) Simulated Thermal Power Function for the mitigation of the loss of fe dwater heating event. The THERMAL POWER time constant o 7 seconds is based on the fuel heat transfer S(W (continued) BWR/4 STS B 3.3-9 Rev 1, 04/07/95 wz
RPS Instrumentation O:' B 3.3.1.1 ;
~
INSERT B 3.3.1.1-9 The drive flow signal 6t rated drive flow is representative of rated core flow at RTP. 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. , r i . O l l i 'O FERMI - UNIT 2- B 3.3 9 (Insert) REVISION 2, 01/18/99 l-l
RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.b. Averaae Power Ranae Monitor (FTom Bt1rilGTSimulated ETY ANALYSES, Thermal Power 4 ontinued) APPLICABILITY dynamics POWER. and provides a signal proportional to tkERMA pgcql,, ange Monitor [, low-BtustrdTSimulated
~
The Average Powe Thermal Power- unction is requtred to be OPERABLE in MODE 1 when there is the possibility of generating excessive THERMAL POWER and potentially exceeding the SL appitcable to ! During high pressure and coreIRM aflowPRM conditions Functions(MCPR SL),ide MODES 2 and 5, other prov i protection for fuel cladd n integrity. 7 l 2.c. Averaae Power Ranae Monitor G ' Neutron Flux T:.. a en......n y. .. .e u.. y, i-r, in6is.u.. .c ..ao ry ;;;te th; g,,
;,;;,;;;;;;g,;;;;;r;=:gtp
.. fl.a in:r:::;::. The Average Power Range Monitor Neutron Flux- Function is capable of generating a Qpscd6 signal to preven uel damage or excessive RCS pressure. For the overpressurization protec" on analysis of Reference 4,theAveragePowerRangeMoniterV Neutron Flux Function is assumed to terminate the main steam b p,y safet n valve (MSIV) losure event and, along with the y} elief valves ( Vs), limits the peak reactor pressure vessel (RPV) pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis Ref 5) takes cred' " the Average Power Range Monitor f3 eutron Flux - Function to terminate the CRDA.
rene ar system 1s divide into two groups r cn.nne n w un three RM channels inp ting to each tri system. The syste is designed to low one channel n each trip sys a to bypassed. Any ne APRM channel < a trip system an f,l ca e the associate trip system to t p. Four chann s of erage Power Rang Monitor Fixed Ne ron Flux-High ith wo channels in ch trip system ar anged in a one- t-of-two logic are r utred to be OPE LE to ensure th t no single inst t failure will p clude a scram f on this Function on valid signal. In addition, to pr ide adequate co rage of the entir core, at least 1 LPRM inputs are equired for each PRM channel, wi at least wo LPRM inpu from each of th four axial leve s at which he) LPRMs ar located. r / (continued) BWR/4 STS B 3.3-10 Rev 1, 04/07/95 O en
RPS Instrumentation B 3.3.1.1 BASES 1 ( APPLICABLE 2.c. Averaoe Power Ranoe Monitor M Neutron Flux l SAFETY ANALYSES, (continued) ~ l LCO, and l APPLICABILITY The Allowable Value is based on the Analytical Limit assumed l in the CRDA analyses. The Average Power Range Monitor (MNeutron Flux Function is required to be OPERAILE in MODE 1 where t e
,, ' potential consequences of the analyzed transients could
+
l A result in the SLs (e.g., MCPR and RCS pressure) bei exceeded. Al h the Average Power Range Monitor Neutron Fluxa Function is assumed in the CRDA ana sis, x wnicn is appl' in MODE 2 the Average Power Range Mnitor Neutron FluxmM(Se,tdowrQFunction conservatively bounas sne assumea tr p and, together with the assumed IRM trips, provides ade uste rotection. Therefore, the Averege Power Range Monitor Neutron Flux Function is not quired in MODE 2. i d. Avernoe Power Ranoe Monitor-Downscale This ignal ensures that there is adequate Neutron Monito System protection if the reactor mode s tch is placed in he run position prior to the APRMs co ng on i scale. Wit the reactor mode switch in run, APRM l downscale sig 1 coincident with an associa Intermediate Range Monitor N ron Flux-High or Inop nal generates a
- trip signal. This unction was not spe fically credited in the accident analysi >ut it is reta d for the overall redundancy and diversis of the RP s required by the NRC approved licensing basis.
b' The APRM System it divided i two groups of channels with three inputs into each tr sys . The system is designed to allow one channel in ach trip stem to be bypassed. Four channels of Aver e Power Range nitor-Downscale with two channels in ea trip system arran in a one-out-of-two logic are re red to be OPERABLE to ure that no single failure 11 preclude a scram from t Function on a valid signal The Intermediate Range Monitor _utron Flux-High nd Inop Functions are also part of t OPERABIL of the Average Power Range Monitor-Do cale Functi (i.e., if either of these IRM Functions can send as al to the Average Power Range Monitor-Downscale F tion, the associated Average Power Range nitor-Downscale channel is considered inoperable). (continued) BWR/4 STS B 3.3-11 Rev 1, 04/07/95 l ees2_
_ _. .m _ _ _ _ _ _ _ _ _ _ . . . _ . _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ . _ _ _ _ _ . . _ _ _ _ RPS Instrumentation i B 3.3.1.1 BASES APPLICABLE' Cveraoe Power Ranoe Monitor-Downscale (con SAFETY ANALYSES, LCO, and The Allowable based upon ensu at the APRMs APPLICABILITY are in the linear scale wh ansfers are made between APRMs and IRMs. This Function quired to be OPERABLE in e he APRMs are the primary indicators of rea 2.en Averaae Power Ranoe Monitor-Inoo Th r s assurance that a minimum number of F I APRMs are OPERABLE W ;- ,'^;e.,e L . 4. an " 4; en ^ % _ i .. a w i .
-.- 6 Ig g "Operat ," an APRM mod e is unplug , the el tronic oper ing voltage i Iow, or t 8 3 3,l l-10 b APRM s too few PRM inputs 11), an ino ative tri i
i i j sig will be ceived by RPS, unless e APRM i l b ssed. Sin only one in each tr system
+
b assed on
.u, . +i.~.+...i one APRM i each trip sys may be
{< r one __ r =40=* . This Function was not specifica1Iy 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. O)
~
I Four e nnels of Avera chan Power Range s in each tri system are re red to be OPE itor-Inop w
~
to en e that no si e failure will reclude a scram om ( s function on valid signal. f There is no Allowable Value for this Function. This Function is required to be OPERABLE in the MODES where 7 the APRM Functions are required. l SSERT . , m 1 I b 3'N'I J 3. Reactor Vessel Steam Dome Pressure-Hioh An increase in the RPV pressure during reactor operation ) compresses the steam voids and results in a positive i 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 (continued) BWR/4 STS B 3.3-12 Rev I, 04/07/95
. . _ - _ - - . . - _ - _. . . _ - - - =-
l l (
\
RPS Instrumentation B 3.3.1.1 INSERT B 3.3.1.1 10 For any APRM channel, any time: 1) its mode switch is in any position other than "0PER": 2) there is a loss of input power: 3) the automatic self-test system detects a critical fault with the APRM channel; or 4) the firnware/ software watchdog timer has timed out, 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. INSERT B 3.3.1.1-11 2.e. 2-out-of-4 Voter The 2 out-of-4 Voter Function provides the interface between the (g) APRM Functions and the final RPS trip system logic. As such, it V 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 2-out-of 4 Voter Function is required to be OPERABLE in MODES 1 and 2. Both voter channels in each trip system (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 2 out-of 4 trip voter includes separate outputs to RPS for the independently voted sets of functions, each of which is redundant (four total outputs). The 2-out of-4 Trip Voter Function is inoperable if any of its functionality is inoperable. Due to the independent voting of APRM trips and the redundancy of outputs, there may be conditions where the trip voter function is inoperable, but trip capability for one or more of the other APRM functions through that Trip Voter is still maintained. This may be considered when l: determining the condition of the other APRM functions resulting l from partial inoperability of the trip voter function. f] There is no Allowable Value for this Function. V r FERMI - UNIT 2 B 3.3-12 (Insert) REVISION 2, 01/18/99 l
RPS Instrumentation [ B 3.3.1.1 BASES APPLICABLE 3. Reactor Yessel Steam Dome Pressure-Hioh (continued) SAFETY ANALYSES, LCO, and Vessel Steam Dome Pressure-High Function initiates a scram APPLICABILITY for transients that results in a pressure increase, counteractin the pressure increase by rapidly reducing core
< 2. power. 4e* e overpressurization protection analysis of-
, Reference 4. n ::t r :: = 'th ; ; 6 conser y,2 g$g> assume $ scram on Neutron Flux
! Average Power Range Monitor sign ", not the Reactor Vesse eam Dome Pressure-Hig gnal klong with the Ys,dtlimits the peak RPV pressure to ess than the ASM g ectio~n Ill Code gg,g,g limits.
w 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 g scram from this Function on a valid signal. The Function is V) , required to be OPERABLE in MODES ] and 2 when the RCS is pressurized and the potential for pressure increase exists.
- 4. Reactor Vessel Water Level-tow. Level 3 Low RPV water level indicates the capability to cool the fuel may be threatened. Should RPY water level decrease too far, fuel camage 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. 6). The reactor scram redeces 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 differenc's between the pressure due to a constant column of (continued) BWR/4 STS B 3.3-D Rev 1, 04/07/95 r i
\
l RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 4. SAFETY ANALYSES, Reactor vessel Water Level-tow. Level 3 (continued) LCO, and APPLICABILITY water (reference leg) a,nd 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 1 ensure that no single instrument failure will preclude a scram from this Function on a valid signal.
.p;g nvJcytod WILL.FCffo44 A5 The Reactor Vessel Water Level-Low, level 3 A11'owable Value h PEEPETEP 84 7HE R6cIRoutA710ij is selected to ensure thattduring normal operation the
'l-Id6 pg6R Alm /5IS. /r A/So separator skirts are not uncovered (this protects available l CAJSvfLG5 77&Prp recirculation pump net positive suction head (NPSH) from i sig'iificant carryunder) and, for transients involving loss l of all normal feedwater flow, initiation of the low pressure l hAtlosable, volat .
ECCS subsystems at Reactor Vessel Water-Low Low Low, LevelIwillnotberequired.5 h CO'*c*d OOP 6b l- h i jlf The Function is required in MODES I and 2 where considerable energy exists in the RCS resulting in the limiting l transients and accidents. ECCS initiations at Reactor l( l i Vessel Water Level-Low Low, Level 2 and Low Low Low, Level 1 provide sufficient protection for level transients in all other MODES.
- 5.
i Main Steam Isolation Valve-Closure j 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 PSq,(c' of the norwa1 heat sink and subsequent overpressurization transient. Howeve or the overpressurization protection i ana is of Referene the Avera E Neutron Flux Function,ge Power along withRange the S Monitor T ts the peak RPV pressure to less than the ASME odes, limits. That is the direct scram on position switches for l MSIV closure even,ts is not assumed in the overpressurization Analysis. JAdditionall , MSIV closure s assumed in t l transi nts analyzed 1 f Reference 7 ( g., low steam ine l
@ prespere, manual c sure of MSIVs, gh steam line ow).J (continued) i BWR/4 STS B 3.3-14 Rev 1, 04/07/95 1
h Ad L
-- .. . ._ - -. . - . . - _ . - - - _ . . . . - . - . = - - - - .
l RPS Instrumentation l h() B 3.3.1.1 BASES 5@] O'I APPLICABLE # Turbine stoo. Valve-Closurg (continued) SAFETY ANALYSES, / LCO, and stage pressure, therefore, to consider this Function APPLICABILITY OPERABLE, the turbine bypass valves must remain shut at THERMAL POWER 2 30% RTP. The Turbine Stop Valve-Closure Allowable Value is selected to be high enough to detect inmiinent TSV closure, thereby reducing the severity of the subsequent pressure transient. 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 , preclude a scram from this Function if any three 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 St Dome Pressure-Hi nd the Average Power Range Monitor eutron Flux psce Functions are adequate to a the necessary safet margins.
V N @ f. Turbine Control Valve Fast ClosureArt: O!b f,'2 WJ 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 Tri; Sii d ~ "n _ . . L: Function is the primary scram sign 51 for the generator load rejection event analyzed in Reference 7. For this event, the reactor scram reduces the amount of energy required to be absorbed and " - - " " " - - " - - ' " -
-EOC "."T Sy;t;; ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast ClosurerTrip Oil Presp;r; L:; P&f#E#ffA7108J OF W signals are initiated by thej:le:tr:hy&: li: :::tr:1 (E"C) SOLGvolp DUMP VALVE 9 M prc::gr: at each control valve. A 0r.e ,,.eas re 7:nx : u scu.e e. ::e. :::tr:1 v:17:, d the
, _ _ -ai;;..! 'se :nh tr;.r.nitt:r i; e;;igr.et, te e se,,. .b 17";
KOVAJMdr* KELA Y SIGAIALS AAC PfoVlp6D *10 CACII g/5 LO&lC CIIAA/AIEL
& {h}p,Th u t n t THERNAL i
SucH THAT VAST Cl.DSUAC of OAll COOy4L vat.VC IAI CAcil US TAlf 6'/$1TM WILLINITIA TE A SCRAMy { continued} i i BWR/4 STS B 3.3-18 Rev 1, 04/07/95 i . ,3 6 r u Z-
l n RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9. Turbine Control Valve Fast Closure. Trio 011 SAFETY ANALYSES, Pres sure-tow l LCO, and (continued)- h 21614 Psn automatically APPLICABILITY pressure transmitters sensing turbine first stage pressur therefore, to consider this Function OPERABLE, the urbine bypass valves must remain shut at THERMAL POWER 2 30% RTP. IN5 tar w - _ T:M _ _ _ x...: s ,u,_;1
. 22
".'.; r;.:..:1. =3...a u_,.._ = Trip ^ n___..a t,_u ., 4 t
6 3 3.l.l- 7 Z. _ C_ _ _ ' O.IT .=.....
. . 2 :J '. ;. Z - - - - '
' " ~ ~ ' " ~
l 0p,2 Four channels of Turbine Control Valve Fast Closure e THp g
^il T..;;..;- W Function with two channels in each trip system arranged in a one-out-of-two logic are required t 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 t 30% RTP. This Function is not required when THERMAL POWER is < 30% RTP, since the Reactor Vessel St y Pressure-Hi h nd th Average Power Range Monitor '
sco A' Functions are adequate to main [Jeutron Flux-{Wety' n the necessary -- margins. II M. Reactor Mode Switch-Shutdown Position i The Reactor Mode Switch-Shutdown Position Function provides l signals, via the manual scram logic channels, to each of the i 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 j as required by the NRC approved licensing basis. l The reactor mode switch is a single switch with four channels, each of which provides input into one of the RPS
,Mic channels.
1 There is no Allowable Value for this Function, since the i channels are mechanically actuated based solely on reactor mode switch position. (continued) BWR/4 STS B J.3-19 Rev 1, 04/07/95
%z
l p RPS Instrumentation B 3.3.1.1 BASES ACTIONS A Note has been provided to modify the ACTIONS related to (continued) 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 i not within limits, will not result in separate entry into- I 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 i 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 Condition entry for each inoperable RPS instrumentation channel. l A.1 and A.2 l Because of the diversity of sensors available to provide g'{5 q #g g 3 trio signals and theservice redundancy of the RPS design, anh allowable out of time of 12 hours has been shown to l be acceptable (Nr#=6) to permit restoration of any l inoperable channel to OPERABLE status. However, this out of O 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 i Required Actions B.1, 8.2, 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 trip) would conservatively compensate for the inoperability, restore capability to acconnodate 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 ! p i and its Required Action taken. 195EfLT 2 y 6 33 M-6 , s.1 and s.2 Condition B exists when, for any one or more Functions, at least one required channel is inoperable in each trip (continued) BWR/4 STS B 3.3-21 Rev 1, 04/07/95 C (-
O RPS Instrumentation B 3.3.1.1 INSERT B 3.3.1.1-12 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. O 4 1 j O ! FERH1 - UNIT 2 B 3.3 21 (Insert) REVISION 2, 01/18/99l
RPS Instrumentation B 3.3.1.1 BASES ACTIONS B.1 and B.2 (continued) system. In this condition, provided at least one channel per trip system is OPERABLE, the RPS still maintains trip capability for that functici, 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., = ::t- ' rr -' MCg,f 5Ty one-out-of-one arrangement for a typical four channel f femaMi tw A Function). The reduced reliability of thi i'c arrangement was not evaluated in Referenc 9. or the 12 hou completion Time. Within the 6 hour allowance, the *4 I3 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 Referenced 9$ which justified a 12 hour allowable out of service une 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 O- 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). pb If this action would result in a scramtii4iM1 it is I 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. 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 (continued) BWR/4 STS B 3.3-22 Rev 1, 04/07/95 w
i i RPS Instrumentation B 3.3.I.1 l BASES ACTIONS B.1 and B.2 (continued) system in trip would result in a scram ".. Z },, Condition D aust be entered and its Required Action taken. 1MsefLT N
$ J,3.l.) -13 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 tapability. 3 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. For the typical Function with one-out-of-two taken twice logic and the IRM and APRM Functions, this would require both trip systems to have one channel DPERABLE or in trip (or the associated trip system in tri Valve p). For Function Closure), 5 (Main this would Steam require Isolation both trip systems to have each channel associated with the MSIVs in three main steam lines (not necessarily the same main steam lines for O both trip systems) OPERABLE or in trip (or the associated trip system in trip).
For Function 8 (Turbine Stop Valve-Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated lrio system in trip).
'dr pCoce HL terD 2 he Completion Time is intended h the operator time 1 o evaluatgland repai any discovered inoperabilitie . The l 1 hour Completion Ti is acceptable because it minim zes I risk while allowing time for restoration or tripping of Q resuli 15 & Io.1T of RPS frip 0PEEABILITY _
Required Action D.1 directs entry into the appropriate Condition referenced in Table 3.3.2.1-1. 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 (continued) BWR/4 STS B 3.3-23 Rev 1, 04/07/95 RwL
O./ RPS Instrumentation B 3.3.1.1 MSERT B 3.3.1.1-13 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 2 out-of-4 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 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 p 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 2-out-of 4 voter and other non APRM channels. Condition B J;es not appiy. O FERMI - UNIT 2 B 3.3 23 (Insert) REVISION 2 01/18/99l
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE As noted at the beginning of the SRs, the SRs for each RPS REQUIREMENTS instrumentation Function are located in the SRs column of (continued) Table 3.3.1.1-1. >
; The Surveillances are modified by a Note to indicate that F
when a channel is placed in an inoperable status solely for lgg performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to
'$y.3.11-)? " s hours, provided the associated runction maintains RPS trip
.canability.f'Upon completion of the Surveillance, or expiration of the 6 hour allowance, the channel must be i
' returned to OPERABLE status or the applicable Condition
% entered and Required Actions taken. This Note is based on gg 9 Ine reliability analysis (Ref.'F'9 time required to perform channel) Surveillance. That assumption of the average analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that the RPS will trip when necessary.
_ _ _ - b caw m 3.3.1.1.1Qvf SR 3.3././, 2) ? h*"3 SR 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 t ' 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. l 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. The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHAN:lEL CHECK supplements less formal, but more frequent, checks of (continued)
- BWR/4 STS B 3.3-25 Rev 1, 04/07/95
( 1
l l l RPS Instrumentation B 3.3.1.1 INSERT B 3.3.1.1-17 l l For the case of the APRM Functions 2.a. 2.b 2.c, and 2.d RPS trip capability is maintained with any two OPERABLE APRMs remaining. 1 I I l l l l i i l l l i l l l FERMI UNIT 2 B 3.3 25 (Insert) REVISION 2 01/18/99l
l RPS Instrumentation B 3.3.1.1 BASES S4 3 J ./* 2. SURVEILLANCE SR 3.3.1.1.10(continued) REQUIREMENTS channels during normal operational use of the displays , associated with the channels required by the LCO. ! SR 3.3.1.1g en y 25% RPT To ensure that the APP.Ms are accu lindicating the true i core average power, the APRMs are rated to the reactor power calculated from a heat balanc gwj5.2.4, " Average l rFvwur n r F j o..n .m. My ints," allows th l APRMs h a nun.6vre reading g ater than actua HERMAL POWER t ! comp sate for loca ed power peakin . When this adj tment is mad the requirement or the APRMs t i icate within RTP of calcula d power is mod" edto) I quire the AP s to indicate wi in 2% RTP of g/ culated/ FLPD. / The Frequency of once per 7 days is caseo 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 i indication of core THERMAL POWER consistent with a heat l balance when < 25% RTP. At low power levels, a high degree N of accuracy is unnecessary because of the large, inherent , , LHGR, i marain to thermal limits (MCPRfand APLHGR). At 2 25% RTP, I the Surveillance is requirea 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 l 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. 9 3.3.1.1 The Avera Power Range Mo tor Flow Biased Simula d ! Thermal ower-High Funct n uses the recirculat n loop l f.I drive ows to vary the ip setpoint. This S ensures that the t ;al loop drive fl signals from the fl units used to y ry the setpoint i appropriately compar d to a cal brated flow signa and, therefore, the PRM Function o (continued) l
^
BWR/4 STS B 3.3-26 Rev 1, 04/07/95
~ __ _
Me d _
, . . - _ --y
l RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE T SR 3.3.1.1.3 (continued) REQUIREMENTS accurately eflects the requir d setpoint as a nction of flow. En flow signal from he respective f1 unit must be s 105 of the calibrated low signal. If e flow unit signal s not within the 1 it, one require APRM that f,) receiv s an input from t deci ed inoperable. inoperable flow nit must be Th Frequency of 7 da is based on en neeringjud t, o rating experience and the reliabil ty of this strumentation. _ 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. As noted, SR 3.3.1.1.4 is not required to be performed when (Q
~y .
entering MODE 2 fram MBE 1, since testing of the MODE 2 required IRM (arid.AMtlGunctions cannot be performed in MODE 1 without ut111 zing 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 (continued) BhR/4 STS B 3.3-27 Rev 1, 04/07/95 O m
RPS Instrumentation B 3.3.1.1 BASES R L SR 3.3.1.1.5 (continued) 9g[ Frequency and is based onvth -Ilt.t Et, __-J ,.:. ..' 0.
":fr: :: (The Manual Scram Function's CHANNEL FUNCTIONAL TEST Frequency was credited in thefiinalysis ) totQ,ge 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 suberitical 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 bII3 increased into a neutron flux region without adeouste indication. This is required prior toir ithdrawing w 3RMs from P. the fe l; t :: n:e ::iti: since indication is being transitioned from s to the IMs. The overlap between s and APRMs is of concern when reducing power into the IRM range. On power increases, the system design will prevent further increases (by initiating O a rod block) if adequate overlap is not maintained. Overlap between IRMs and APRMs exists when sufficient IRMs and APMs concurrently have onscale readings such that the transition between M00E I and MODE 2 can be made without either APRM downscale rod block, or IRM upscale rod block. Overlap fe, o between SRMs and IRMs similarly exists when, prior to F' .fgll withdrawing the SRMs from the f;.11, L.;;de ' roiii , IRMs are above mid-scale on range I 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 downstale once in MDDE 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. (continued) BWR/4 STS B 3.3-28 Rev 1, 04/07/95 1 0 - Rex 2-1 .
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 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 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 j fer appropriate representative input to the APRM System. i The 1000 MWD /
- P. withtpRnsen$sitivitychanges, Frequency is based on operating experi i
SR 3.3.1.1.9 and SR 3.3.1.1 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. l The 18 month Frequency is based on the need to perform this I j 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. l 7his hweillancL) Op3 m e c , e s: =t e provides a check of the actuai trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.1-1. If the trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be (continued) l BWR/4 STS B 3.3-29 Rev 1, 04/07/95 1 h. AvZ_ i
l l RPS Instrumentation 8 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.10 (continued) REQUIREMENTS readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability analysis of Reference 9. SR 3.3.1.1.11 and SR I'l 3.3.1.1. 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. SR 3.3.11. ll WNote 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. pr;r ' ; .;;.. 2;tx t;[ ;{x 't' p;
. .. . . - w r v P'>
s..... y
= ..- u . ; = :.:.;i .;. r. a.....mya we 2 mi a ..,
(o) f.l .Q:n . b--**~~ =:=*~** * % M M (T* ?.? ' l.", p* = = . . 2 wh gog t;w.ng : l 1s provided that requires the IRMSRjftobe ! performed within 12 hgurs of antering MODE ' Prom MODE 1. 1 Testing of the MODE 2)APBlL.antTJIRM Functio cannot be performed in MODE 1 witnout utilizing jump rs, lifted leads, ' or movable links. This Note allows entry into MODE 2 from MODE I 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. j f*d The Frequency of SR 3.3.1.1.11 is based upon ti; ;n x;t's
'l se6 a 184 day calibration interval in the detemination of
.the magnitude of equipment dri t in the setpoint analysis.
The Frequency of SR 3.3.1.1 is based upon ' ' x x;t' n e6@l8 month calibration, nterval in the determination of the magnitude of equipment drift in the setpoint analysis. pi IN5 M ~ [ f8</el'/f (continued) BWR/4 STS B 3.3-30 Rev 1, 04/07/95 (t
l i 1 l 7'N RPS Instrumentation U B 3.3.1.1 , INSERT B 3.3.1.1-14 SR 3.3.1.1.12 : i A CHANNEL FUNCTIONAL TEST is performed on each required channel ! to ensure that the entire channel will perform the intended j 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 for Function 2.b only, the recirculation flow input function excluding the flow ; transmitter) the 2 out of-4 voter channels, and the interface l 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 3.3.1.1.12 is based on the reliability analysis ) of Reference 13. (NOTE: The actual voting logic of the 2 out-of-4 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 3 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. l t i O a FERMI - UNIT 2 B 3.3-30 (Insert) REVISION 2. 01/18/99l
T l ? I ? l RPS Instrumentation : B 3.3.1.1 l % BASES l I SURVEILLANCE-REQUIREMENTS
' SR 3.3.1.1 h I (continued) The Ave e Power Range tor Fic;, Bissed Simul ed i
The Power-High Fun on uses an electronic Iter ! n ci it to generate a gnal proportional to t core i fe l RMAL POWER from APRM neutron flux s1 . This 11ter circuit is presentative of the f
- heat transfer l i
dynamics that pr uce the relationship ween the neutron 1 flux and the THERMAL POWER. The srveillance filter j time constan must be verified to s 7 seconds to ensure
- that the c nnel is accurately re cting the desired j parameter
- LT =he dering
+
F quency of 18 months i sed on engineering dgmenti ! l the reliabilit f the components. J 4 l SR 3.3.1.1.15 54 3,34. The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the 3 OPERABILITY of the required trip logic for a specific d channel. The functional testing of control rods (LCO 3.1.3), and SDV vent and drain valves (LCO 3.1.8), 1 gggyLp overlaps this Surveillance to provide complete testing of 4 f 3,3,Ialello]a. theA assumed safety function j L J The 18 month Frequency s st 2.3+t. & en two need to perform this i a
.' Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Survalliance were performed with the reactor at power.
y- 1 Operating experience has shown that these components usually j- - g pass the Surveillance when performed at the 18 month
$ 3 3.l.1-lk -
SR 3.3.1.1.16 This SR ensures that scrans initiated from the Turbine Stop. Valve-Closure and Turbine Control Valve Fast ClosureM 9644pecespo%een Functions will not be inadvertently bypassed when THERMAL POWER is 2 30% RTP, This involves g; elly calibration of the bypass channels. Adequate margins for Mg,g is - - - the instrument setpoint methodologies are incorporated into the anno setpoint) Seeense main turbine bypass flow can j fivm O N affect this setpoint nonconservatively (T RMAL POWER is M #ef derived f e first staae pressur , the main turbine J WM br hW. Mrsf daqA- fu sswa. of /6/,9/sT.9 fori5yd-@ (continued)
\comoisks #30 % RTP BWR/4 STS s T.3-31 / ev 1 ,04/07/95 R
l l j
.. -. . - -- -. .- . _~ -- . -
I I l l y L RPS Instrumentation B 3.3.1.1 l l l l l INSERT B 3.3.1.1-16a l For the 2-out-of 4 Voter Function. the LSFT includes simulating APRM trip conditions at the APRM channel inputs to the 2-out of 4 trip voter channel to check all combinaticns of two tripped inputs to the 2-out-of 4 trip voter logic in the voter channels. l i l l INSERT B 3.3.1.1 16b Additionally, the 24 month Frequency of SR 3.3.1.1.19 is based on l( I Reference 13. l l
)
1 i I q iNu/ '
- FERMI - UNIT 2 B 3.3 31 (Insert) REVISION 2 01/18/99{
RPS Instrumentation B 3.3.1.1 BASES SR 3.3.1.1.16 (continued) SURVEILLANCE REQUIREMENTS bypass valves must remain closed at THERMAL POWER 2 30% RTP I to ensure that the calibration remains valid. If any bypass chanrel's setpoint is nonconservative (i.et, 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 Functions are considered inoperable. Alternatively, the bypass channel i can be placed in the conservative condition (nonbypass). If placed in the nonbypass condition, this SR is met and the channel is considered OPERABLE. The Frequency of 18 months is based on engineering judgmen MQ h IB MDHnl CouteAnod re11abilityofthecomponents,f- ! I la762 VAL. Ist THE DETERMINA17%I \
' of J}l2 MA(,NITUDE of SR 3.3.1.1.17 EGotPMtMT DAlfr IAI 7dE 6ET70/VT A//Aty.5/5. . 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 l
- measurement or in overlapping segments, with verification that all components are tested. The RPS RESPONSE TIME l '3 /M66RT ) acceptance criteria are included in Reference 10.
! lO3'3*['!' As noted, neutron detectors are excluded from RPS RESPONSE f, j TIME testing because the principles of detector operation l virtually ensure an instantan response time. i F $g WT- RPS RESPONSE TIME tests are cted on an 18 month STAGGERED TEST BASIS. Note requires STAGGERED TEST BASIS 33.3.l!~Tq Frequency to be determined ased on 4 channels per trip system, in lieu of the 8 channels specified in Table 3.3.1.1-1 for the MSIV Closure Function. This frequency is based on the logic interrelationships of the various channels required to produce an RPS scram signal. The 18 month 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 I lNSERT occurrences. ( 833 I l-15 , y (continued) i BWR/4 STS B 3.3-32 Rev 1, 04/07/95 ; lO
& 2.
\
i !c RPS Instrumentation B 3.3.1.1 i l INSERT B 3.3.1.1 4a RPS RESPONSE TIME for the APRM 2-out-of 4 Voter Function includes the output relays of the voter and the associated RPS relays and i contactors. (The digital portion of the APRM and 2-out-of 4 voter ! I channels are excluded from the RPS RESPONSE TIME testing because self testing and calibration checks the time base of the digital , electronics. Confirmation of the time base is adequate to assure l required response times are met. ! INSERT B 3.3.1.1 4b l In addition. Note 2 states the response time of the sensors for ! Functions 3 and 4 are excluded from RPS RESPONSE TIME testing, i The sensors for these Functions are assumed to operate at the ; l sensor's design response time. This allowance is supported by Reference 12, which determined that significant degradation of the l sensor channel response time can be detected during performance of other Technical Specification SRs and that the sensor response ! time is a small part of the overall RPS RESPONSE TIME testing. , ! O V INSERT B 3.3.1.1-15 i SR 3.3.1.1.18 ; l A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to l 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 the APRM Simulated Thermal Power - Upscale Function this SR also includes ! calibrating the associated recirculation loop flow channel. l SR 3.3.1.1.18 is modified by a Note that states that neutron 3 i_ detectors are excluded from CHANNEL CALIBRATION because they are l 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 3.3.1.1.3) and the 1000 MWD /T LPRM calibration against the TIPS (SR 3.3.1.1.8). g The Frequency of SR 3.3.1.1.18 is based upon 24 month calibration Q interval in the determination of the magnitude of equipment drift in the setpoint analysis. FERMI - UNIT 2 B 3.3 32 (Insert) REVISION 2 01/18/99l L
1 RPS Instrumentation B 3.3.1.1 1 l BASES (continued)
~
l j REFERENCES 1. U FSAR, figure k.--7,2-2.
- 2. UFSAR, Section f.h(2}. / F.4. l.f..
- 3. NEDO-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.
- 4. UFSAR, Section -[D4). 5.2.2.3.
- 5. UFSAR, Section [MM^).15. 4. 9.
i 6. UFSAR, Section@.;J;. lo.3.3.
- 7. UFSAR, Chapter-{b 16. j
- 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. j
- 10. UFSAR, Table i7Y.((}. 7, 2- 4.
2 O
, , _ , l
\ \. NEDC- S I9% , " CL ASS JLC, Oc.706EK 19 6to, GENELAL. ELECTRIC lb)SRVMEM7 65TPDn)T MG7110 ppt.%f"
)
L 12. MEDO-322H,"5ysh Adalyses for gi,,,;ndian h get.eeled Respoase ne Terl:ng repare,,,ents," 7pnaq (99 4; and fw&-t seR te no,oesmi /n, do4ed ApriI tg J 13, A/EFOc - 32.4:o P-A, " /J u char M casurvatsuf tys&sts and Cekm/ /%*r kp dea % lYlm iIw' l (rJvMM. PRtJM) Re+ro&f PIu.s opHm 'Dr Sk bili /y Trip funclim, " Dehdc M ff~, ad Qupplernw+/,fyl B 3.3-33 199(p . Rev 1, 04/07/95 BWR/4 STS IO 9' fail
lgm JUSTIFICATION FOR DIFFERENCES FROM NUREG 1433 l( ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION I NON BRACKETED PLANT SPECIFIC CHANGES l P.1 These changes are made to NUREG-1433 to reflect Fermi 2 current i licensing basis: including design features, existing license requirements and commitments. Additional rewording, reformatting, I and revised numbering is made to incorporate these changes consistent with Writer's Guide conventions. Refer to CTS Discussion Of Changes to the related requirement for a detailed justification of changes made to the current licensing basis which are also reflected in the ITS as presented. Specifically, some of these changes are discussed ' below:
- a. The Channel Calibration requirements for the IRM RPS Function is l indicated in the ISTS as SR 3.3.1.1.13 at 18 months. However, l the CTS Frequency for this calibration is "SA" (i.e.,184 days).
l In adopting the existing CTS requirement, the ISTS is modified to eliminate SR 3.3.1.1.13 and combine the IRM Channel Calibration requirement with SR 3.3.1.1.11. As such, the ISTS Note in SR l l 3.3.1.1.13 is deleted and its equivalent is added to ISTS SR ! () 3.3.1.1.11. '%) l b. In accordance with Fermi CTS Amendment #122 (to incorporate the l PRNM System), the ITS and Bases are modified utilizing NEDC 32410 as guidance. P.2 Bases changes are made to reflect plant specific design details, equipment terminology, and analyses. l P.3 Bases changes are made to reflect changes made to the Specification. E Refer to the Specification, and associated JFD if applicable, for additional detail. P.4 Change made for editorial preference or clarity. Specifically, one change revises " MWD /T" to " MWD /ST" to clarify "short" ton versus metric ton. ! I r i FERMI - UNIT 2 1 REVISION 2. 01/18/99l l l
q N0 SIGNIFICANT HAZARDS EVALUATION l Q ITS: SECTION 3.3.1.1 RPS INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (SDecification 3.3.1.1 "L.3" Labeled Comments / Discussions) Detroit Edison has evaluated the proposed Technical Specification change ! identified as "Less Restrictive" in accordance with the criteria specified by l 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration. l The bases for the determination that the proposed change does not involve a significant hazards consideration is an evaluation of these changes against each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the i evaluation are presented below. j
- 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?
The proposed change eliminates requirements for Operability of the IRM. Reactor Mode Switch, and Manual RPS Functions in Modes 3 and 4. This (] change will not result in a significant increase in the probability of l k./ an accident previously evaluated because RPS Functions are not considered as initiators for any accidents previously analyzed. This change will not result in a significant increase in the consequences of an accident previously evaluated because control rod withdrawal is not allowed in Mode 3 or 4 (except for Special Operations allowed by LCOs 3.10.3 and 3.10.4. which require the necessary protection systems). Under these conditions the RPS system has no safety function.
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
This proposed change will not involve any physical changes to plant systems, structures, or components (SSC). or changes in normal plant operation. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated. t c
~
l FERMI.- UNIT 2 5 REVISION 2 01/18/99l I i
. - ~ . - - - . . - - . . - - - - . _ - - .
I NO SIGNIFICANT HAZARDS EVALUATION
'ITS: SECTION 3.3.1.1'- RPS INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (Specification 3.3.1.1 "L.5" Labeled Comments / Discussions) '
Not used. l p I i ( I q L l l-O I-I i i l
- l. j l.
l lO FERMI UNIT 2 9- . REVISION 2 01/18/99l
NO SIGNIFICANT HAZARDS EVALUATION ! O. ITS: SECTION 3.3.1.1 - RPS INSTRUMENTATION l TECHNICAL CHANGES - LESS RESTRICTIVE ; (Soecification 3.3.1.1 "L.5" Labeled Comments / Discussions) ; Not used.' l' i I i
- f. i l
E h O : { , O I - WIT 2 10 REVISION 2 01/18/99l 1
.a;
. h. - n _ . , . . . . . , -e.. -. -. --- -- -- ,,-. --3 . y.,- .,r-- 2---
-- - - ~ . . _ . . - - . _- -- . .- .-
NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.3.1.1 RPS INSTRUMENTATION , TECHNICAL CHANGES - LESS RESTRICTIVE l (Specification 3.3.1.1 "L.8" Labeled Comments / Discussions)
~
l Detroit Edison has evaluated the proposed Technical Specification change identified as "Less Restrictive" in accordance with the criteria specified by 10 CFR 50.92 and has determined that the proposed change does not involve a significant hazards consideration. The bases for the determination that the proposed change does not involve a l _ significant hazards consideration is an evaluation of these changes against i each of the criteria in 10 CFR 50.92. The criteria and the conclusions of the ! evaluation are presented below. i
- 1. Does the change involve a significant increase in the probability or l consequences of an accident previously evaluated?
The proposed change' to the weekly and semi annual Channel Functional l Tests and semi annual Channel Calibrations for IRMs and APRMs provides a Q V 12 hour allowance to perform the Surveillances after entering Mode 2 during a plant shutdown. The proposed change does not involve a l significant increase in the probability of an accident previously evaluated because the potential for a reactor transient caused by the performance of these tests utilizing jumpers or lifted leads is avoided. Additionally. delayed performance of Channel Calibrations for these , Functions is not considered an initiator for any accidents previously analyzed. The proposed change does not involve a significant increase in the consequences of an accident previously evaluated because the additional time allowed to perform the Surveillance is not expected to adversely affect the performance of any credited equipment. Therefore. the consequences remain unchanged from those that would apply utilizing the existing CTS requirements. l
- 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?
-This proposed change will not involve any physical changes to plant :
systems, structures or components (SSC). or changes in normal plant ' operation. Therefore, this change will not create the possibility of a , new or different kind of accident from any accident previously l evaluated. ' .O FERMI - UNIT 2 15 REVISION 2 01/18/99l
rw Control Rod Block Instrumentation ( ) 3.3.2.1 v 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 RBH 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.
SURVEILLANCE FREQUENCY l SR 3.3.2.1.1 - - - -- -- - -- NOTE-- ----- - - -- Not required to be performed until I hour after any control rod is withdrawn at s 10% RTP in MODE 2. Perform CHANNEL FUNCTIONAL TEST. 92 days L) l l l SR 3.3.2.1.2 ------ ---- -.-- NOTE - --- -- -- Not required to be performed until I bour , after THERMAL POWER is s 10% RTP in l I MODE 1. l Perform CHANNEL FUNCTIONAL TEST. 92 days l l SR 3.3.2.1.3 Perform CHANNEL FUNCTIONAL TEST. 184 days 1 (continued) r3 G l FERMI - UNIT 2 3.3 19 Revision 2 01/18/99
i Control Rod Block Instrumentation ( ' ff<~-)s - 3.3.2.1
- SURVEILLANCE REQUIREMENTS (continued) :
SURVEILLANCE FREQUENCY l .SR. 3.3.2.1.4 - ------
-. -- NOTE -.--- --- ------
Not required to be performed until I hour ! after reactor mode switch is in the shutdown position. i Perform CHANNEL FUNCTIONAL TEST. 18 months :
-l SR 3.3.2.1.5 Verify the RBM is not bypassed when 24 months THERMAL POWER is = 30t RTP. ,
l l SR 3.3.2.1.6 ------- ------ -- NOTE- --- - - - - ----- Neutron detectors are excluded. j l Perform CHANNEL CALIBRATION. 24 months 1 l I SR 3.3.2.1.7 Verify control rod sequences input to the Prior to i RWM are in conformance with the declaring RWM ' prescribed withdrawal sequence. OPERABLE following loading of secuence into RWh I i l 1 l l r**'% 1' k) w i i l ' FERM1 UNIT 2 3.3 20 Revision 2, 01/18/99
Control Rod Block Instrumentation (o) N_/ 3.3.2.1 Table 3.3.2.1 1 (page 1 of 1) Control Rod Block Instrtnentation APPLICABLE MODES OR OTER SPECIFIED REQUIRED SlRVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE
- 1. Rod Block Monitor
- a. Upscale (a) 2 SR 3.3.2.1.3 As specified in SR 3.3.2.1.5 the COLR SR 3.3.2.1.6 l b. Inop (a) 2 SR 3.3.2.1.3 NA
- c. Downscale (a) 2 SR 3.3.2.1.3 a 92.3! of SR 3.3.2.1.6 reference level
- 2. Rod Worth Hinimiter 1(b) 2(b)
. 1 SR 3.3.2.1.1 NA SR 3.3.2.1.2 SR 3.3.2.1.7 l 3. Reactor Mode Switch- Shutdown (c) 2 SR 3.3.2.1.4 NA Position /~N
{ b N ,/ (a) THERMAL POWER = 303. (b) With T}ERMAL POWER s 10! R1P. (c) Reactor node switch in the shutdown position. i l l l i i \ \ /
%/
l FERMI - UNIT 2 3.3 21 Revision 2, 01/18/99
I 1 l l ex Control Rod Block Instrumentation '(v ) B 3.3.2.1 l B 3.3 INSTRUMENTATION 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 l accidents. During high power operation. the rod block monitor (RBM) provides protection for controi rod withdrawal error events. During low power operatinns, control rod l 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 l Switch-Shutdown Position Function ensure that all control l rods remain inserted to prevent inadvertent criticalities. l n The purpose of the RBM is to limit control rod withdrawal if ; localized neutron flux exceeds a predetermined setpoint (d) 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 l the 3 reset power level. The RBM has two channels, either of ; whic1 can initiate a control rod block when the channel l output exceeds the control rod block setpoint. One RBM l channel inputs into one RMCS rod block circuit and the other l RBM channel inputs into the second RMCS rod block circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A i signal from one of the four average power range monitor l (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 preset power level, the RBM is automatically by)assed. The RBM is also automatically bypassed if a peripleral control rod is selected (Ref.1). ' ( c') v' l \ l l FERMI - UNIT 2 B 3.3.2.1 - 1 Revision 2. 01/18/99 i
. - _. _ ~ - . -. _ - - . ~ - . - - - - - - - . . -. .. - - Control Rod Block Instrumentation .O B 3.3.2.1 BASES I SURVEILLANCE REQUIREMENTS (continued) 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 a control rod block will be initiated when necessary. l SR 3.3.2.1.1 and SR 3.3.2.1.2 A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. l The SR 3.3.2.1.1 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 selection error is indicated and a control rod block occurs.
SR 3.3.2.1.1 is performed during startup. As noted in the SRs, SR 3.3.2.1.1 is not recuired to be performed until I hour after any control roc is withdrawn at s 10% RTP in l MODE 2. The SR 3.3.2.1.2 CHANNEL FUNCTIONAL TEST is performed by attempting to insert and withdraw a control rod (A") not in compliance with the prescribed sequence and verifying a selection error is indicated and a control rod insert and withdraw block (respectively) occur. SR 3.3.2.1.2 is performed during a shutdown. As noted, SR 3.3.2.1.2 is not required to be performed until I hour after THERMAL POWER is s 10% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.1, and THERMAL POWER reduction to s 10t RTP when in MODE 1 for SR 3.3.2.1.2, to perform the required Surveillance 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 Frequencies are based on reliability analysis (Ref. 8). l SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel l to ensure that the entire channel will perform the intended
~
function. i 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.10). O. V l l FERMI UNIT 2 ' B 3.3.2.1- 8 Revision 2, 01/18/99 l
i Control Rod Block Instrumentation (m) B 3.3.2.1 ; i l BASES SURVEILLANCE REQUIREMENTS (continued) l SR 3.3.2.1.4 j l \ A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch-Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL l l FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position Function is performed by attem) ting to withdraw any control rod with the reactor mode switc1 in the shutdown j l position and verifying a control rod block is present. As noted in the SR the Surveillance is not required to be performed until I hour after the reactor mode switch is in ; the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be : performed without using jumpers, lifted leads, or movable i links. This allows entry into MODES 3 and 4 if the 18 month i 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, r~ ( 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 un)lanned transient if the Surveillance were performed with tie reactor at power 0)eratingexperiencehasshownthesecomponentsusualiypass t1e Surveillance when performed at the 18 month Frequency. SR 3.3.2.1.5 The power at which the RBM is automatically by)assed is based on the APRM signal's in)ut to each RBM clannel. Below the minimum )ower setpoint, t1e RBM is automatically bypassed. T11s >ower Allowable Value must be verified periodically to >e less than 30% RTP. If this setpoint is nonconservative, then the affected RBM channel is considered inoperable. Alternatively. the power range channel can be
> laced in the conservative condition (i.e., enabling the RBM
- unction). If placed in this condition, the SR is met and l the RBM channel is not considered inoperable. The 24 month Frequency is based on the actual trip setpoint methodology utilized for these channels, i
I l FERMI - UNIT 2 B 3.3.2.1- 9 Revision 2 01/18/99 i l
4 Control Rod Block Instrumentation O B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) l SR 3.3.2.1.6 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies 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. As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices. with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately l surveilled in SR 3.3.1.1.1 and SR 3.3.1.1.7. l The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equ ment drift in the setpoint analysis. The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring the RWM OPERABLE following loading of the prescribed withdrawal sequence into the RWM since this is when rod sequence input t errors are possible. REFERENCES 1. UFSAR. Section 7.6.2.13.5. 2 .' UFSAR. Section 7.6.1.20.
- 3. General Electric Energy, " Maximum Extended Operating Domain Analysis for Detroit Edison Company Enrico Fermi Energy Center Unit 2 " NEDC - 31B43P, July 1990.
- 4. NEDE-24011-P A-10 US. " General Electric Standard Application for Reload Fuel," Supplement for United States. March 1991.
l FERMI - UNIT 2 B 3.3.2.1-10 Revision 2, 01/18/99
- . . - . . ~ - - . - - - - . . - .
f i Control Rod Block Instrumentation ! B 3.3.2.1 ,mf) ( ; BASES l REFERENCES (continued) i
, 5. " Modifications to the Requirements for Control Rod Drop ;
Accident Mitigating Systems," BWR Owners' Group, l July 1986.
- 6. NED0 21231. " Banked Position Withdrawal Sequence." ,
January 1977. ,
- 7. NRC SER, " Acceptance of Referencing of Licensing !
Topical Report NEDE-24011 P A." " General Electric 1 Standard Application for Reactor Fuel, Revision 8, , Amendment 17," December 27. 1987. ! i
- 8. NEDC-30851-P A. " Technical Specification Improvement l Analysis for BWR Control Rod Block Instrumentation." 1 October 1988.
- 9. GENE 770 06 1 A. " Bases for Changes to Surveillance Test Intervals and Allowed Out-of Service Times for Selected Instrumentation Technical Specifications."
m December 1992.
- 10. NEDC-32410P A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Functions," October 1995, and Supplement 1. May 19%. l l l . FERMI - UNIT 2 B 3.3.2.1-11 Revision 2 01/18/99 l i
W ECIF104TT00 $ 31 d INSTRUMENTATION i 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION N.[ [1MITING CONDTTION FOR OPERATION LC41.LLI -ft.t. The control rod block instrumentation channels shown in Table 3.3.6-1 shall be OPERABLEjwitn tnyir trip sm points set copsistent witnj tne valuiv1 3 (snownjin tne fr4 Setpoint column f Table 3.3.6/. f APPLICABitTTY: As shown in Table 3.3.6-1. ACTION:
- a. With a control rod block instrumentation channel trip setpoint* 1esst gNg conservative than the value shown in the Allowable Values column of AS Table 3.3.6-2, declare the channel inoperable until the channel is ;(
restored to OPERAglE statupirith trip point aaa yceo ; [ con)#htent wi njme irip setooi , value. .3
- b. With the number of OPERABLE channels less than, required by the Acw A,B,E Minimum OPERABLE Channels per Trip Function requirement, take the !
" ACTION required by Table 3.3.6-1. I h.I i
SURVEfttANCE RE0VIREMENTS
.3R tkn!? I
-4wb6 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations I for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1.
@oosw zb @ ;
i l l I i k.l The APRM mulated Ihermal ower - Upscale Fun ional Unit need not be f* declared noperable upon tering single rea or recirculation loop o rat] ion l provid the Flow Blase setpoints are chan d within 4 hours per l , Qeci cation 3.4.1.1. # l V FERMI.- UNIT 2 l 3/4 3 41 Amendment No. 53. pp,121 ! PAGE 3 0F 09 Av1
O O rABLE 2.2.5 ! 3 3 2.I-I O CONTROL R00 BLOCK INSTRUMENTATION A N*
- HINIMUM A M'M APPLICABLE SE OPERABLE CilANNELS OPEPAT10HAL
~* M RIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION
- 1. 800 BLOCK MONITOR (a)
E la a. Upscale 2 1* p /. b
~ 1 c-
- b. Inoperative 2 1* 60 60 }g 33,2./ b A,8
- c. Downstale 2 1* 60 ,i (2. AVERAGE POWER RANGE MONITOR Y l
- a. Si ated Thermal Power - U scale 3 1 6 8
- b. I perative 3 1, 2 I
- c. utron Flux - Downstale 3 1 61 L
Imulated Thermal Power - Upscale (Seldown) , g \.d.o/ rinw - Upscale 3 3 2 1 61 61 i G) - IT3 '3. SOURCE RANGE MONITORS \ w a. Dete or not full in(b) 2 4 1 61 3(f) 2 5 61 w b. Up ale (C) 2 61 1-
~ 3(f) 2 5 61
- c. operative (C) 0 2 61 Downscale(d) 3(f 2
3 5 6/ o !
- 4. INTERMEDIATE RANGE MONITORS g f a. 4tector not full in 6 -
Z, 5 b./1pscale - 6 2, 5 1 ( g. Inoperati nnwntrale ) 6 c 2, 5 61 %
? 4 61 1 y j h. SCPdH DISCitARGE Water Level -H VOLUMF( : i, 2, 5 62
~
y . Scram Trio Broa / 2 / 2, 5** 62 g K
- a ng .,,a g
.l {
g ,= 7. REACTOR MODE SWITCH SHUTOOWN POSITION 2 (4,-+] 63 } J.3.z I 4<T.'*3 E { hN "
'c3 p h. I-T8L3rio4(N 3.7 l-l L
O 4
.O O .
I 3 3,2./-l TABLE 3.3.6-2 l TA6te 53Z.1-1 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS I b R lPNFUNCTIONC'00 0 fTRIPSET/0 INT ALLOWA8LE VALUE !
. .& - ROD BLOCK MONITOR I
,a Upscale .) As ecified in the As specified in the e IA 5 '
C OPERATING - CORE OPERATING
*' MITS REPORT LIMITS REPORT
$ +:- Inoperative / NA NA 1.C .e Downscale 1 94% of eference L el 192.3% of Reference Level .
J ! 3 g i r--
- 2. AVERAGE POWEl(RANGE MONITOR l !
m a. Simula Thermal Power - Upscale s0.63(W8W),+55.6%, 0.63(W-AW)*+58.5%, 3 j l I) F1 Biased t with a maximum of with a maximum of i
@ 2) Igh flow Clamped 108% of RATED THERMAL 110% of RATED THERMAL POWER
^ k
- b. operative NA NA O a c. eutron Flux - Downscale a 5% of RATED THERMAL ER a 3% of RATED THERMAL POWE !
m a d. Simulated Thermal Power - Upscale (Setdown) s 12% of RATED THE . POWER s 14% of RATED THERMAL ~ER t Flow-liner =1= s 110% of rated f1 s 113% of rated flow i O
+
RI
= F. 3 SOURCE RAN(1 MONilutox W ;
@ 'E a. Deteci4r not full in A i
I b. Ups 1.0 x 105 cps s 1.6 x 105 cps ! E c. e In erative NA NA a 2 cps ** k2 Tg g g d. D scale a 3 cps ** j
. g.1 >
U+na. n . .A n , n 7 can aravidad o ' '; " a- - '-- st- . se_ " y wa werey, ru=vr nenge-(W). 4W is defleed or stanslates une i ro==r - upscais es the difference issea uso n ect set int variv. .. . . . . . indicated drive flow I percent of drive f which produces ret core floul bet _ . ,ivop drive two loop and { i { nele loop operetton a he sese core ft 4W
- 05 for two I retten. 4W = 01 r single loop operati g b - - w ;
Db b
.. L i U U l
f
O O . O ; i
;;i TABLE 3.3.6-2 E
i
, CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS (Continued)- i
, 2 *
.4 TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE l 4.'
INTERMEDIATI RANGE MONITORS \ k a. Detec r not fu'l in NA NA
- b. Ups le s108/125 divisions of s110/125 divisions of full scale full scale 3G) c. operative t
E NA NA t rrt ' Downscale a 5/125 divisions of a 3/125 divisions f gA I M ull trale full scale i M ,
*d SLHsi U450HARGE VOLUMt
, 5.
h => a Water Level - liigh s 589'11 .s 59 " C3
.( b. Scram Trip Bypass NA
- 5. Deleted j g 7. REACTOR MODE SWITCH SilUTDOWN POSITION NA NA q kE ;
$. - TBL 3 3,2.l-1 Fvnenou 3 E D i
= S
- e. [ :.
y D N u . 8 M M E ' i
O O f'\ V V . () TABLE l.3.5 ' CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS
~
TMW E3'M'I CHANHEL WMcN CIIANNEL OPERAT10NAL
- FUNCTIONAL CilANNEL CONDITIONS FOR WHICil IRIP FUNCTION CHECK TEST CAllBRAT10N(a)
EI 1. ROD BLOCK HONITOR SURVEILLANCE REQUIRED w to h a. b. Upscale Inoperative NA NA 5 2 years 1* 1 c. Downscale NA 5 SA NA (14 3 3 LI 32 years (sR 1328G 1* 1* '
- 2. AVERAGE 10WER RANGE MONITOR
- a. Si lated thermal Power -
U scale NA SA 2 years M o b. noperative NA SA NA 1, 2
- c. Neutron Flux - Downse e N h '
Simulated Thermal Po r-NA NA SA 5 2 years 2 years 1 l Upscale (Setdown)[ 2 q* Flow - Upscale NA _ 2 years I ' C4 w !m3. 50VR9E RANGE MONITORS h a. elector not full in NA S/U$ W NA 2***, 5 ! b Upscale 5 S
,W SA 2***, 5
. Inoperative NA S ,W NA 2***, 5
; i
( . Downscale t 5 ( .W SA 7"* 'i > O f' 4. INTERWEDIATE RANGE MONIIUH5 4 a. elector not full in
- b. Upscale NA S
S/U , NA 2,5] , S/U SA 2, 5 Inoperative NA S/U ( NA 2, 5 Downscale 5 S/U g E .W SA 2, 1 E 5. SCRAM 1IISCHARGEVOLUNE k g
- a. %ter Level - High NA Q / R / 1,.z, 5"-)
s 3
- b. / Scram Trip Bypass NA /
R NA / 2, 5** / 5 -E. "Whi lc h3Z REACTOR MODE SWITCH g# R geog rE 7 ~ SilUIDOWN POSITION NA m.u.l.s NA h g y - t e i
d GPEL LFIC&T10r) 3 3 2 o1 4 min .r . ;c=ted) CONTR01 R0D BLOCK INSTRLMENTATION SURVEf tt ANCE RE0VIREMENTI TABLE NOTATIONS SR#hbM(*) Neutron detectors may be excluded from. CHANNEL CALIBRATION. wig hours pjetir to startup, if,m6t performedynnin une prevp> R ."L 3,[ .h-l
- With THERMAL POWER greater than or equal to 30% of RATED THERMAL POWER.
l
%foi" "4' ".;"! 'm2 I'i"!RM'i.it!/* """%
- e. _ ... _ ...:. . - mgu g 2.
m U FERMI - UNIT 2 3/4 3 46 A- Amendment No. JJ, 212 d. PAGE- 9 0F 09
%L
DISCUSSION OF CHANGES ITS: SECTION 3.3.2.1 - CONTROL R0D BLOCK INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M.1 Not used. H.2 CTS Tables 3.3.6-1 and 4.3.6 1 require the reactor mode switch shutdown position rod block to be Operable in Modes 3 and 4. ITS Table 3.3.2.1-1 Note (c) requires this rod block function to be Operable when the mode switch is in the shutdown position. This change adds rod block function Operability requirements while in Mode 5 with the mode switch in shutdown position. Mode 5 operation with the mode switch in shutdown credits this rod block function. Therefore, this change will not introduce any negative impact on safety. M.3 CTS Table 3.3.6-1 footnote (a) states that "the RBM shall be automatically bypassed when < 30% RTP." However, CTS contains no specific surveillance of this automatic enable / bypass function for the RBM. Furthermore, the footnote (a) wording is misleading in its intent to assure an Operable function at 2 30% RTP: it ( N specifically would only enforce that the automatic trip be bypassed below 30% RTP, leaving silent any require 0ents at 2 30% RTP. ITS 3.3.2.1 includes SR 3.3.2.1.5 that requires the RBM to be not bypassed at 2 30% RTP. This imposes a periodic verification and enhances the presentation to avoid potential misinterpretation. H.4 CTS 4.1.4.1.a and b, and footnote *, allow the withdrawal of control rods to perform the RWM functional test anytime. and for any duration, prior to withdrawal of control rods "for the purpose of [ making] the reactor [ critical)." The ITS allowance in Notes to SR 3.3.2.1.1 and SR 3.3.2.1.2 imposes a time frame (1 hour) to l complete the test in Mode 2: not allowing unlimited rod withdrawals. These additional restrictions do not impose any negative impact on safety. TECHNICAL CHANGES LESS RESTRICTIVE
" Generic" LA.1 CTS 3.3.6 requires that control rod block instrumentation trip setpoints to be within Trip Setpoint column values of CTS A Table 3.3.6 2. Additionally, CTS 3.3.6 Action a requires that an h inoperable control rod block instrument channel setpoint be FERMI UNIT 2 2 REVISION 2, 01/18/99l
i l l l S DISCUSSION OF CHANGES
)
ITS: SECTION 3.3.2.1 -' CONTROL R0D BLOCK INSTRUMENTATION j i i LB.1 Not used. l ! LC.1 . CTS 4.1.4.1.a and b require RWM Channel Functional Tests within 8 hours prior to withdrawing control rods for startup, and CTS ! 4.1.4.1.c. requires RWM Channel Functional Test within 1 hour J of reaching RWM automatic initiation during a reactor shutdown. l These surveillances are required to be performed every reactor l
.startup and shutdown, regardless of the frequency of these l events. ITS SR 3.3.2.1.1 and ITS SR 3.3.2.1.2 only require l I that the Channel Functional Test be performed every reactor startup and shutdown if it was not performed in the previous 92 days (i.e., the ITS Frequency is 92 days: therefore. if j performed in the previous 92 days, the SR is considered -
performed as required for startup).
. CTS 4.3.6 requires Surveillances be performed on the RBMs; but contains no provisions for avoiding Technical Specification Actions when the surveillance causes the RBM to be inoperable.
ITS 3.3.2.1 Surveillance Requirements are provided with Note 2 allowing RBM surveillances that result in an inoperable RBM to d be conducted for up to 6 hours without entering the associated Conditions and Required Actions for the inoperable RBM, provided the other RSM is Operable. l
)
l l I O FERMI UNIT 2 4 REVISION 2 01/18/99l
' It i
I
- DISCUSSION OF CHANGES :
ITS: SECTION 3.3.2.1 CONTROL R0D BLOCK INSTRUMENTATION l 1
- s .,
R.5 Not used. , i Y b i i i t i 1
'k i
t i l' l l l l l l l 4 I .' i f, i f i-
' FERMI -~ UNIT 2 12- REVISION 2 01/18/99l 1
- h, .
l I,. ,,mv.m..I6.,-.,,..,w.- _w._,,,...m.....,,..~,wr ....,_....-._,-._,_.,,,mm--..,-.. -
.. _,..m.w...---,_,,,
Control Rod Block Instrumentation p 3.3.2.1 ACTIONS (continued)
& TS}
l CONDITION REQUIRED ACTION COMPLETION TIME l l
~
E. One or more Reactor E.1 uspend ntrol Immediately 6L 3e3. W, Mode Switch-Shutdown Position channels thdrawal .- Ac4;g3, i inoperable. ; E.2 Initiate action to
-de>14r insert QH ttS}
.\ -inse soie c rol in co cells c tainin one or j g
i re fu ass I h CmW toi %ck'., \ v w ! l SURVEILLANCE REQUIREMENTS
--------------------------------NOTES---------------------
- 1. Refer to Table 3.3.2.1-1 to determine which SRs apply for each Control R
- h*W kk Block Function. '3*b O 2. When an RBM channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions [poc tt,gh and Required Actions may be delayed for up to 6 hours provided the X /
associated Function maintains control rod block capability. SURVEILLANCE FREQUENCY SR 3.3.2.1) Perform CHANNEL FUNCTIONAL TEST. days b 3 h e
- e &fg l
BWR/4 STS 3.3-17 Rev 1, 04/07/95 aa , l
Control Rod Block Instrumentation 3.3.2.2 dp { SURVEILLANCE REQUIREMENTS (continued) (crs) SURVEILLANCE FREQUENCY SR 3.3.2.1 ---- -- NOTE------------ l Not required to be perfomed until I hour after any control rod is withdrawn at sp0gRTPinMODE2.
*l' M'I* "k h 1
Perfom CHANNEL FUNCTIONAL TEST. g2Kdays g l SR 3.3.2.1 ----------NOTE - - - - - - - - - - Not required to be performed until I hour / i afterTHERMALPOWERiss(101%RTPin \4.1,%), 4,c MODE 1. N _____ _ __ ____________ I
# " Perfom EL FUNCTIONAL TEST.
gS2gdays w A <(f 'HSW Ffom NEyr PAGC ' SR 3.3.2.1. -- E------------- g Ne on detectors re excluded. O , _________ _____ Verify the months Q. ;. , 6.. _.. _- , - 4
;....mN; - 's is 78 L 3
- 3' 6 'I not
, m_D.ypas. sed when THERMAL POW 6 4mrte.(n)
--~ RTP.
(),\ -t. -...... ..... <. . .,...... s g..... Fun ion is not b assed when THE D R is > 64% a 5 84% RTP.
- c. Igh Power Ran .-Upscale Func .on i not bypassed w en THFRMA! pmr te a
> 84% RTP. F fI S 3.3.2.1.5 Verify the RWM THERMAL POWER is not bypassed Il01% RTP.
en I.18 nths (continued) BWR/4 STS 3.3-18' Rev 1, 04/07/95 Q n1
l 1 Control Rod Block Instrumentation 3.3.2.1 j SURVEILLANCE REQUIREMENTS (continued)
<as>
SURVEILLANCE FREQUENCY l l
$ SR 3.3.2.1 -----------
NOTE--- Not required to be performed until I hour (*y after reactor mode switch is in the L 9 3. G -1 3 g shutdown position. pu,4cw 7 Perform CHANNEL FUNCTIONAL TEST. [18gmonths l I SR 3.3.2.1.h ---- ------NOTE------ ------- , T6L 4.3.b-1 l f Neutron detectors are excluded. s wta(a) 1 Perform CHANNEL CALIBRATION. W anti: 79 L 4 3,G-l j j FuuHa L l SR 3.3.2.1 Verify control rod sequences input to the Prior to
} ! RWM are in conformance with-{ Bis-6 declaring RWM [q,g,qq, OPERABLE P,5 N* P""O'd followins l[ - u>&dvawa{ loading of
\
5 gm sequence into RWM BWR/4 STS 3.3-19 Rev 1, 04/07/95 O v ,v gL
T' Control Rod Block Instrumentation ( 3.3.2.1 C.'?dt'f,it'a i::.::.1L G rs > (( T8Ls s 2.3. 6-lj L3. 6 -2.J
,m , e,,a noors en u. ss-1))
8,01MR ECIFIS REGUIRED RRVEILLANCE ALLOW 4BLE FUNCTIC,1 COW ITIOut CNAMMELS REGUIR$ FEN 18 VALW
- 1. a.d st.et m. nit.,
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r ( ! !ti g s c(3.i.u.3)) L
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. . - . . ...... , ,, 7,, ,
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sa 3.3. 1.7 fett .c.
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- s. n s it.n - em - s.s...'
4 * (TBts 16 7) (314 ~b . A > t.)
@ 8'C =- = 'O TennnL Pawa e 1e,t:
w 1 % , . . i.c x A =>x it. .i.nl. [(T8Lf.3.(,-1,D ret v. 3. c -i , .v ) p't u,4- . . (, . ,,, . ,.n. J y/ 1. ,., /. 1, -. 14. / ' u, L . e ,, - , - 1, - i. w
) With THERIEL POE R g g RT.,
, 9n . .., -iui,in *, np m .
.3.l.9.I,Arg/.aidih) 73L 3 3.t,-/ /h 7 BWR/4 STS 3.3-20 Rev 1, 04/07/95 O- g
i I Control Rod Block Instrumentation B 3.3.2.1 8 3.3 INSTRUMENTATION 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) provHes protection for control r6d 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 predetemined setpoint during control rod manipulations. It is assumed to function to block further control rod withdrawal to preclude a MCPR O _ _ _ _ GPsed pa y 7 Safety Limit (SL) violation. The RBM supplies a trip signal to the Reactor Manual Control System (RMCS) to appropriately inhgit control rod withdrawal during power operation above e the .- _ . . The R8M has two channels, lwd 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 circuit. The R8N channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various o k k b T* _ core heights surrounding the control rod being withdrawn. A O. s}gp}youonpayeragepowerrangemonitorgRM annep I )$EftT jh - supplies a'refE'ence'EIgnU"foY5EE r E"-A +L 6 g 3,Q,j- \ ca.) 7 .. ;r ,,.^. - . This reference signal is used to detemine L which R8M range setpoint (low, intermediate, or high) is enabled. If the APRM is indicating less than the 4es=poseer pg p _ - :-" . the RBM is automatically bypassed. The RBM g p +is also automatically bypassed if a peripheral control rod g is selected (Ref.1). L _OU 1I- 0 (contiaued) BWR/4 STS B 3.3-44 Rev 1, Da/07/95 kW 1
l~ l Control Rod Block Instrumentation B 3.3.2.1 l Insert B 3.3.2.2.1-la One of the RBM channels and a signal from another of the ! , APRM channels supplies the reference signal to the second RBM . channel. i 1 I I l Insert B 3.3.2.1 1b l ) l l A rod block signal is also generated if an RBM downscale trip i j or an inoperable trip occurs, since this could indicate a l ! malfunctioning RBM channel. The downscale trip will occur if l , the RBM channel signal decreases below the downscale trip set point after the RBM channel signal has been normalized. The inoperable trip will occur if too few LPRM inputs are available, if a module is not plugged in, or the function switch is in any position other than " operate." O !~
-l l
l l i d i U I. l FERMI UNIT 2 Page B 3.3 44 (Insert) REVISION 2, 01/18/99l t
i Control Rod Block Instrumentation J B 3.3.2.1 i BASES (con}inued)
- - - w m, SURVEILLANCE ' Note: certain Frequencies are bas REQUIREMENTS topical repo . r fo o use these l Frequencies Frequencies as !
y the staff SER for the topica r por..
~
l s noted at the beginning of the SRs, the SRs for each Control Rod Block instrumentation Function are found in the i , SRs column of Table 3.3.2.1-1. l ! 1 l The Surveillances are modified by a Note to indicate that l 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 1 l control rod block capability. Upon completion of the I l Surveillance, or expiration of the 6 hour allowance, the channel must be returned to OPERABLE status or the appilcable Condition entered and Required Actions taken. This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to perfom channel Surveillance. That analysis demonstrated that the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when 'n necessary. SR 3.3.2.1 ~ A CHANNEL FUNCTIONAL TEST is performed for each RBM channel I to ensure that the entire channel will perform the intended function. -It ira!rd:: th: Pa nt;r Pan =! C: tr;i Faltiplexii., ";g t. ir.p.t. Of.1 Any setpoint adjustment shall be consistent with the l assumptions of the current plant specific setpoint , methodology. The Frequency of days is based on ! l reliability analyses (Ref. J). gg i l fl SR 3.3.2.1 m and SR 3.3.2.1 A CHANNEL FUNCTIONAL TEST is performed for the RWH to ensure that the entire system will perform the intended function. g yg3jg.g TheSCHANNEL FUNCTIONAL TEST for the RWM is performed by attempting to withdraw a control rod not in compliance with l (continued) BWR/4 STS B 3.3-51 Rev 1, 04/07/95 k ' v Rw l l
Control Rod Block Instrumentation O BASES B 3.3.2.1 set u b m g is g indica Af ed et SURVEILLANCE SR 3.3.2.1/ RTE'nd SR 3.3.2.1 ornUIREME " (continued) I l#ME/1T -,'- the prescribed sequence and verifying a ontrol rod block g N'EI _,, g occurs.J As noted in the SRs, SR 3.3.2.) not required
'to be performed until I hour after any control rod is withdrawn i :^^:: E- As noted, SR 3.3.2.1 is not required to be performed until I hour after THERMAL WER is
(/ M 56124' s 10% RTP in E 1. This allows entry into MODE 2 for 633.2l-2.h SR 3.3.2.1 d antry i=+WE 2 %:: THERMAL POWER 4e- (duchtn 5 10% RTBjfor SR 3.3.2.1 o perform the required .h> r surveillance if the 92 da requency is not met per gb 6 / SR 3.0.2. The 1 hour allowance is based on operating S006 1 experience and in consideration of providing a reasonable time in which to complete the SRs. The frequencies are based on reliability analysis (Ref. 8). N SE(LT,N eThe1tBM d5y3'f*SR sttp:f-t: :r: 3.3.2.1.,& f,3 y)C} ;;nr. Thrn ^.ile etiv;;.t;;;.eticelly isives are y.d.ri;d T;;dni-; f;f-*i:n Of L 3 i_Li. t.t.". M e nh .;ithis, e e recific p - r 1.n .. 1he p3M )3auhmaby 3 power E ter tiat ally which ch n thej;.str;J
, ere basedred i onb!eck the APRM ?!!rd!? v= "aeinput to W signal's O bifA5 " g 'S z -_ -
each RBM channel. Below the is automatically bypassed. Thesepower Allowable valuek minimum must be verified periodically to be less than er :;=1 te power setpoint./the RBM 3f/oRTP) "h; ;PCifi d ' iu. . If m y P - ~ - setpoint is i nonconservative,thentheaffectedRBMcIIannelisconsidered __ inoperable. Alternatively, the power range channel can be olated in the conservative condition (i.e., enabling the {ua&n] ~ troper RB h etp;ir,t). If placed in this condition, the SR
-+isari met neutron and the RBM channel is not considered inoperable g;r detertare Amer n ;g;j ,,, y,,;;;; ;,,;cee, .;ith .,in; ie e=ind:d m i h ,f-ano +'e Er;;ill. .i 4 i
,N N N ,N'I_$$ b "!$' 5 M N5).E5 5. U 'dE b ; I f ~3 .c.; r,T;pf," ""QgyreilueU[IslasAd3nMet$1 trip setpoint methodology utilized for these channels.
SR 3.3.2.1.'5k n The RWM i automatically byp sed when power is abov a i f- specifi value. The power evel is determined fr feedw er flow and steam ow signals. The auto ic bypas s L (continued) BWR/4 STS B 3.3-52 Rev 1, 04/07/95 ea
. . . . . _ - . . - . _ ~ . - ~ . - . - . . - . - - . . - . . - . . - ~ . - - . - - . . . . . . _
1
' Control Rod Block Instrumentation O B 3.3.2.1 i i
Insert B 3.3.2.1-2a f s SR 3.3.2.1.1 is performed during startup. l ) l Insert' B 3.3.2.1 2b
. at 5 10% RTP in MODE 2. The SR 3.3.2.1.2 CHANNEL FUNCTIONAL l TEST is performed by attempting to insert and withdraw a control rod not in compliance with the prescribed sequence and verifying a selection error is indicated and a control rod insert and A withdraw block (respectively) occur. SR 3.3.2.1.2 is performed l 3 during a shutdown.
1 I J l~ t. i FERMI - UNIT 2 Page B 3.3 52- (Insert)- REVISION 2 01/18/99l l 1
- l. . .. , .. . . . - _
. . ~ . .
Control Rod Block Instrumentation O B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS f C. 3.2.f 5 (continued) ( setpoi must be verifi periodically to s [10]% RTP. If th RWM low power s point is nonconser tive, then the RWM considered in erable. Alternatel , the low power se oint channel ca be placed in the e servative conditio ( nbypass). If aced in the nonbypa ed condition, the R ' met and the is not considered operable. The requency is b ed on the trip setp nt methodology ut iz far the low a r retnnint channel 7 SR 3.3.2.1.$ A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch-Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position Function is performed by attempting to withdraw any a control rod with the reactor mode switch in the shutdown i position and verifying a control rod block accur Q pryje h As noted in the SR, the Surveillance is not required to be ! performed until I hour after the reactor mode switch is in the shutdown position, since testing of this interlock with O. the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 18 month 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 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 these components usually pass the Surveillance when performed at the 18 month Frequency. SR 3.3.2.1 5 A CHANNEL CALIBRATION is a complete check of the instrument ge do[rewos/5 loop and the sensor. This test verifies the channel pg. 46 [ responds to the measured parameter within the necessary 1 J range and accuracy. CHANNEL CALIBRATION leaves the channel (continued) BWR/4 STS B 3.3-53 Rev 1, 04/07/95 Rsn L
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1 continued) REQUIREMENTS adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology. As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with mini drift, and because of the dif,ficulty of simulating a (meilled Of,2. meaningful si 1. Neutron detectors are adequately + in SR 3.3.1.1 ndSR3.3.1.f.f The Frequency based upon the assumption of an B month I P*2/ calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. SR 3.3.2.1 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the i RWM so that it can perform its intended function. The Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into , since this y' p.5 errors are possi L 11 when rod seauence innQtfe}Lf%ed WEhdraws r ~ REFERENCES 1. V FSAR, Section [?.S.2.2.;). 7 4.2.15. 5'}
- 2. U FSAR, Section -U.0.3.2.Gt. 7, /,, f. 2o /
~
- 3. E -3u 6r, mr r ru - nung nl}'er, B1 1
it ,a ech al S cifica n nt A ) Pr ran f Edw I. Ha iN[4earrov n ,"
%, aber 10A1 ~_
- 4. NEDE-240ll-P-A-[US, " General Electried Standard Application for Reload Fuel," Supplement for United States, E--if r ! 2.2.?.;, !:;t:-t:, ;^^0.
- 5. " Modifications to the Requirements for Control Rod
( Drop Accident Mitigating Systems," BWR Owners' Group, I f July 1986. - _ _ __ l
" MHIMuu exTsab60 OPERAOly "PCWid GEMECAL. ELECTRtC.
h14Q5LS foA DETAbtT* EDIS00 GJ6AGf,CoviPAM'/ EHRJC0 FSRMI 9 ENEA&4 CfM1EA NEDC,- 31 B43 f, Jin.'/ 1990- - _ M continued) l BWR/4 STS B 3.3-54 Rev 1, 04/07/95 Q NJ [
Control Rod Block Instrumentation B 3.3.2.1 BASES REFERENCES 6. NEDO-21231, " Banked Position Withdrawal sequence," (continued) January 1977.
- 7. NRC SER, " Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A," " General Electric Standard Application for Reactor Fuel, Revision B, Amendment 17 " December 27, 1987.
- 8. NEDC-30851-P-A, " Technical Specification Improvement Analysis for BWR Co rol Rod Block Instrumentation,"
October 1988. -
- 9. GENE-770-06-1 "Addendmate-Bases for Changes t urveillance Test Intervals and Allowed Out-of-Service Times for Selected Instrumentation Technic _a pecifications, ........, ..... ^ M c N e M92 m-pJS w T
, i 68 ; g,332.t-3
\
4 f BWR/4 STS B 3.3-55 Rev 1, 04/07/95 m Sg L
~
Control Rod Block Instrumentation 'O B 3.3.2.1 Insert B 3.3.2.1 3
- 10. NEDC 32410 P A, " Nuclear Measurement Analysis and Control Power Range Neutron Monitoring (NUMAC PRNM) Retrofit Plus Option III Stability Trip Function," October 1995, and Supplement 1 May 1996.
O O FERMI - UNIT 2 Page B 3.3-55 (Insert) REVISION 2, 01/18/99l
_ ~ - - - - - . , - . - _ _ . - .. -- - . - - - . s JUSTIFICATION FOR DIFFERENCES FROM NUREG - 1433 l ITS: SECTION 3.3.2.1 CONTROL R0D BLOCK INSTRUMENTATION NON-BRACKETED PLANT SPECIFIC CHANGES P.1 These changes are made to NUREG-1433 to reflect Fermi 2 current l licensing basis: including design features, existing license requirements and commitments. Refer to CTS Discussion Of Changes to i the related requirement for a detailed justification of changes made to the current licensing basis which are also reflected in the ITS as presented. Additional rewording, reformatting, and revised numbering i is made to incorporate these changes consistent with Writer's Guide conventions. P.2 Bases changes are made to reflect plant specific design details, equipment terminology, and analyses. P.3 Bases changes are made to reflect changes made to the Specification. Refer to the Specification, and associated JFD if applicable, for additional detail. P.4 "Calendt . ar" revised to ~12 months." Other editorial p) w corrections made to more clearly and accurately reflect the requirement as detailed in the Bases. P.5 "BPWS" is revised to more generally refer to " prescribed withdrawal sequence." Plant / cycle specific analysis may require ! certain deviations from BPWS, which are utilized as initial assumptions for CRDA analyses. Therefore, this essentially reflects a terminology change with no technical change in practice. P.6 Clarifying details of the performance of the Channel Functional Tests required by SR 3.3.2.1.1 and SR 3.3.2.1.2 is added to the l Bases. This detail is provided to enhance operator understanding of the intent of and differences in these two Channel Functional Tests. . P.7 The reference to the NRC Policy Statement has been replaced with a l more appropriate reference to the Improved Technical Specification
" split" criteria found in 10 CFR 50.36(c)(2)(ii).
O l FERMI - UNIT 2 1 REVISION 2, 01/18/99
l gy PAM Instrumentation B 3.3.3.1 i (_) BASES LC0 (continued) Only two Category I thermocouple channels are needed for l post-accident monitoring of sup)ression pool water l temperature (Refs. 3 and 4). T1e outputs for the PAM l sensors T50N404A and T50N405B are recorded on two l independent recorders in the control room (channel A is ! redundant to channel B). Both of these recorders must be ! OPERABLE to furnish two channels of PAM indication. These l recorders are the primary indication used by the operator during an accident. Therefore, the PAM Specification deals i specifically with this portion of the instrument channels.
- 6. Drywell Pressure Drywell pressure is a Type A. Category I variable provided ,
to detect a breach of the RCPB and to verify ECCS functions l that operate to maintain RCS integrity. Two wide range ) drywell pressure signals are transmitted from separate pressure transmitters and are continuously recorded and displayed on two control room recorders. These recorders l s are the primary indication used by the operator during an ( accident. Therefore, the PAM Specification deals
\s) specifically with this portion of the instrument channel.
l 7. 8. Primary Containment Hydroaen and OXYoen Concentration j Primary containment hydrogen and oxygen analyzers are Type C, Category I instruments provided to detect high hydrogen or oxygen concentration conditions that represent a potential for containment breach. This variable is also important in verifying the adequacy of mitigating actions.
- 9. Primary Containment Hiah Ranoe Radiation Monitor Primary containment area radiation (high range) is a Type E.
Category I variable, and is provided to monitor the potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans. The instrumentation ! provided for this function consists of redundant sensors. microprocessors and indicators. A common 2-pen recorder in the control room continuously records signals from both channels. The redundant indicators in the relay room and the common recorder in the control room are the primary indication used by the operator during an accident. 73 l FERMI - UNIT 2 B 3.3.3.1 - 5 Revision 2 01/18/99 - l l
- -- - .. = --_ -- -- .-
l l l f PAM Instrumentation B 3.3.3.1 t]m BASES SURVEILLANCE REQUIREMENTS (continued) The 18 month Frequency for all channels except the primary containment oxygen and hydrogen analyzers (per Note 1 to SR 3.3.3.1.3) is based on operating experience and consistency with the typical industry refueling cycles. The 92 day Frequency for the primary containment oxygen and hydrogen analyzers (per Note 1 to SR 3.3.3.1.2) is based upon vendor i recommendations and instrument accuracy requirements. j SR 3.3.3.1.2 is modified by Note 2 stating that performance ) of the calibration of the oxygen and hydrogen monitors may ' be delayed until after exceeding 15% RTP (i.e., the power at which LC0 3.6.3.2 requires the primary containment to be inerted). This delay is allowed for up to 72 hours for one l oxygen and one hydrogen monitor, and for 7 days for the l second oxygen and hydrogen monitor. These delays facilitate i more accurate calibration methods, which can be employed l with the primary containment inerted. ' SR 3.3.3.1.3 is also modified by Note 2 stating that radiation detectors are excluded from calibration requirements. ( REFERENCES 1. Regulatory Guide 1.97. " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident " Rev. 2, December 1980.
- 2. Detroit Edison Letter NRC 89-0148, " Additional Clarification to Fermi 2 Compliance to Regulatory Guide 1.97, Revision 2," dated June 19, 1989.
- 3. Detroit Edison Letter NRC 89 201, " Regulatory Guide 1.97 Revision 2 Design Review," dated September 12, 1989.
- 4. NRC Letter, " Emergency Response Capability-Conformance to Regulatory Guide 1.97, Revision 2 (TAC No. 59620) "
l dated May 2, 1990.
- 5. Detroit Edison Letter NRC-93-0105. " Fermi 2 Review of Neutron Monitoring System Against Criteria of NED0-31558A," dated September 28, 1993.
%.)
l FERMI.- UNIT _2 B 3.3.3.1- 11 Revision 2. 01/18/99 l
3.33. /-/ TABLE-3.3.7.L ; (Continued) E ACCIDENT NONITORING INSTRilNENTATION 7 T6L 3.3 3 I-/ E M MON NiNi m mat y INSTRilM[Ni REQUIRED MUfGER CHAISIELS OPERATIONAL. 0F CHAISIE OPEqAOLE CONDITIONS ACTION J3 Si n dt, i, 3 i vai-..t system Radiation onttors
- a. SGIS - oble Gas (Low-range)' 1/0PERA8tE I/0PERABLE I,2,3 SG15 subsystem 81 pe SGTS subsystem I"'
b. S 5 . Noble Gas (Mid-ran ) I/0PERA8l I/0PERA8LE I. 2, 3 l' SGTS subs les SGis subsystem
,M c. SG15 - AXM-Moble Gas Id-range) 1/0 LE I/0PERA8tt 1, 2, 3 81 M SGT subsystem SGIS subsystem Q= w t sctt - ann.na-ki. c = fyt?-rannel I/0PERABLE I/0PERA8tt 1 ,3 8
( GIS subsystem SGIS subsystem E na noi . e.,e l
- 15. S :':d
/O Rr. Primary Containment Isolation Valve Position 1, 2[ - 2 f /0/..}v=] I'/nkck t 82 g 2jenetmun g
x
/6 (a)(b) ,3 ,
p E 3 F Q b _ 3 Al cluded in th FSITE DOSE CALCUL ON MAN y ./ g Dh t N W i l
S&ctncAnd 3 331 TABLE 5 (Continued) ( iso SM- he$cdd.N) ACCIDENT MONITORING INSTRUMENTATION ACTION STATEMENTS m_ d4DO: AcuoM 6)
- a.
- With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in able 3.3.7.5-1, restore U MTtodh the inoperable channel (s) to OPERABLE status wit sys c' h "'74 I.!Opiig@ = e qpy,3 [f::-3,,f 3o g
_ _ - . ., Y $
/_---Withb. the number of OPERABLE accclent monitoring instrumentation nels A ND less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status withinK T.e.c.k be in A M F-y' , ,.
u,_at least HOT SHUTDOWN ___,_ ~ - within the next 12 hours.
- - - - -- - rthas AGTHM&S4-~ With the number of OPERABLE than required by the minimum Lg,} i A6 TIM C channels OPERABLE requirementsh ="ejne g+; 2.f.o- ._..-.,...u g
{2f monjidhng the sonrborsie omranistarfaMwithin 72 hs0rsfe,G. (one less &JRquet: AM Ac.d4 44 6 L.4
', 1) either restore the inoperable channel (s) to OPERABLE status within 7 days of the event, or M 2) prepare and submit a Special Report to the Commission pursuant to Stachab SM7 s;m,;T=.eu6 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
^C"^M M - With the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, withir3" Prdis- i rdSC W A'[3oA.N
) a. Restore the inoperable channel (s) to OPERABLE status, or L*F 1 Aalg)C: @ !
-==- 7 gg., y,'5 3.1,1 b.'
%,. ___are the affected isolation valve inoperapie ano u - une m i surs) g
' ~
., MorsLa.h specified by Speerfication 3.6.3 ACTION a.#
,9
,-, -? - ACT1oM F (Be,in rno0E 3)
ACTION 83 ' -Achi d 1(so rb r.GM }
- a. With the number ofOPE LE accident monitoring instrumentation channels ACT6oh) A less than the Required Number of Channels shown in Table 3.3.7.5-1 restore the ino2erable channel (s],to OPERABLE status within 30 days,%r sNTa~
~~
ALnm 8 report to the 6ommTs~sion pursiianfTo SjELM6.E2 within the following 1 4 set W 6 0 7) and schedule for restoring ?.he instrument channeI(s) to OPERA!
- b. With the number of OPERABLE accident monitoring instrumentation channels 86Md O less than the Minimum Channels OPERABLE requirements of ble3.3.7.5-1, t
restore the inoperable channel (s) to OPERABLE status within hours or be in g p at least HOT SHUTDOWN within the next 12 hours.
- 12. L,'l O FERMI- UNIT 2 3/4 3-62 Amendment No. 28.58,117
' PAGE U
OF 06 84/ 2-
~~
O O O
, """1 ' . 2. ' 5 ' (r- . ;;;
ACCIDENT MON 110 RING INSTRUMENTATION SURVEILLANCE RED'JIREMENTS ' e APPLICA8LE t
=
CHANNEL CHAISIEL OPERATIONAL 4 INSTRUM[NT CHECK
~ CALIMATION C0lelil0NS
- 13. Stand Gas treatment Syst ediation Monitors .
- a. SETS - Noble Gas (L .-range M R 1, 2, 3 8./
b SGTS Noble Gas I d range)) M R 1, 2, f
. SGTS - AXti Noble as (Mid range) M R 1, 2, 2 ccre avis.u-hi c.,
a u t ok . - - ) g , n 3, 3 3 g ,,. .., m y ,_,_,,, , jg I W p A Primary Containment Isolation Valve Position to (f) M (3) R 1, 2 O E C l.7 i 64 .3,3-3 l - ( h O M m i i g @- ' a 3 3 ! A
?
u z.
- Ww 5
(.O N L t i I
e' DISCUSSION OF CHANGES l ( ITS: SECTION 3.3.3.1 - PAM INSTRUMENTATION R.1 (continued) Conclusion Since the screening criteria have not satisfied for non-Regulatory Guide 1.97 Type A or Category 1 variable instruments, their associated LC0 and Surveillances may be relocated to other plant controlled documents outside the Technical Specifications. The instruments to be relocated are as follows (numbers reflect the CTS Table 3.3.7.5 1 Instrument Functions): l
- 5. Suppression Chamber Air Temperature:
- 6. Suppression Chamber Pressure:
- 8. Drywell Air Temperature:
- 11. Safety / Relief Valve Position Indicators:
- 13. SGTS Radiation Monitors; and
)
rx
'% J l
( y FERMI UNIT 2 7 REVISION 2 01/18/99l
Recirculation Loops Operating 3.4.1 T 3.4 REACTOR COOLANT SYSTEM (RCS) ., 3.4.1 Recirculation Loops Operating LC0 3.4.1 a.1 Two recirculation loo)s with matched recirculation loop jet pump flows shall >e in operation: , AND a.2 The reactor core shall not exhibit core thermal-hydraulic instability or operate in the " Scram" or
" Exit" Regions.
DB j b.1 One recirculation loop may be in operation and the reactor core not operating in " Scram" or " Exit" regions as specified in administrative controls provided j LC0 3.3.1.1, " Reactor Protection System (RPS) ' Instrumentation," Function 2.b (Average Power Range l l Monitors Simulated Thermal Power-Upscale) Allowable ! Value of Table 3.3.1.1-1 is reset for single loop operation, when in MODE 1. Q ............................N0TE--- ------ ------- --- -- - Required allowable value modification for single loop operation may be delayed for up to 4 hours after transition from two recirculation loop operations to single recirculation loop operation. APPLICABILITY: MODES I and 2. ACTIONS ! CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation jet pump A.1 Declare recirculation 2 hours loop flow mismatch not loop with lower flow: within limits. "not in operation." (continued) d. l FERMI UNIT 2 3.4 1 Revision 2 01/18/99
7- Recirculation Loops Operating i 3.4.1 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. . Reactor core operating ------------NOTE------- ----- in the " Exit" Region. Restart of an idle recirculation loop or resetting a recirculation flow limiter is not allowed. B.1 Initiate action to Immediately insert control rods l or increase core flow to restore operation outside the " Exit"
-l- Region.
C. No recirculation loops C.1 Be in MODE 3. 6 hours operating while in MODE 2. D. No recirculation loops D.1 Place the reactor Immediately operating while in mode switch in the MODE 1. shutdown position. E. Reactor core operating
-l in the " Scram" Region.
Core thermal hydraulic l- instability evidenced.
\
l FERMI UNIT 2 3.4 2 Revision 2 01/18/99
r'N Recirculation Loops Operating Q 3.4.1 SURVEILLANCE REQUIREMENTS
- SURVEILLANCE FREQUENCY
,SR '3.4.1.1 -- ---- ------ - NOTE- - ---- --- -- ---
Only re uired to be performed when-operati g in the " Stability Awareness" l Region. Verify the reactor core is not exhibiting 1 hour core thermal-hydraulic instability. SR' 3.4.1.2 ---- --------- -- --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
- O' operation is:
- 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 = 70% of rated core flow.
10: l' FERMI.- UNIT 2 3.4 3 Revision 2 01/18/99
l l e3 Recirculation Loops Operating ; (
- B 3.4.1
%)
1 BASES BACKGROUND (continued) l begins to boil, creating steam voids within the fuel channel ! that continue until the coolant exits the core. Because of l reduced moderation, the steam voiding introduces negative l reactivity that must be compensated for to maintain or tc i increase reactor power. The recirculation flow control l system allows o>erators to increase recirculation flow and sweep some of tie voids from the fuel channel, overcoming the negative reactivity void effect. Thus, the reason for i having variable recirculation flow is to compensate for l l reactivity effects of boiling over a wide range of power generation without having to move control rods and disturb desirable flux patterns. l' l Each recirculation loop is manually started frun the control room. The MG set provides regulation of individual recirculation loop drive flows. The flow in each loop is manually controlled within limits established by the recirculation speed control system. GDC 12 of 10 CFR 50 Appendix A (Ref. 4) states that the reactor core and associated coolant, control, and protection (V~h systems shall be designed to assure that power oscillations which can result in exceeding specified fuel design limits are not possible or can be reliably detected and suppressed.
- i. BWR cores typically operate with the presence of global flux l l
noise in a stable mode which is due to random boiling and flow noise. As the power / flow conditions are changed, along l with other system parameters (xenon. subcooling. power J distribution, etc.) the thermal-hydraulic / reactor kinetic ' feedback mechanism can be enhanced such that perturbations may result in sustained limit cycle or divergent oscillations in power and flow. , Two major modes of oscillations have been observed in BWRs. The first mode is the fundamental or core-wide oscillation mode in which the entire core oscillates in phase in a given axial plane. The second mode involves regional oscillation in which one half of the core oscillates 180 degrees out of 1 phase with the other half. Studies have indicated that i adequate margin to the Safety Limit MCPR may not exist I during oscillations. (^N G l FERMI - UNIT 2 B 3.4.1 - 2 Revision 2 01/18/99 i
1 l l h Recirculation Loops Operating B 3.4.1 ' BASES I APPLICABLE The operation of the Reactor Coolant Recirculation System is SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref.1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to 4 , provide coolant flow during the first few seconds of the l accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump , in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered. The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis l was reviewed for the case with a flow mismatch between the l l two loops, with the pipe break assumed to be in the loop . l with the higher flow. While the flow coastdown and core i ! response are potentially more severe in this assumed case : i (since the intact loop starts at a lower flow rate and the ' I core response is the same as if both loops were 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 L coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the UFSAR. 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 caus'ed by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling (Ref. 3). The transient analyses of Chapter 15 of the UFSAR have also been performed for single recirculation loop operation , (Ref. 3) and demonstrate sufficient flow coastdown i characteristics to maintain fuel thermal margins during the l abnormal operational transients analyzed. During single recirculation loop operation, modification to the Reactor l Protection System (RPS) average power range monitor (APRM) ; instrument setpoints is also required to account for the I different relationships between recirculation drive flow and reactor core flow. The APRM Simulated Thermal Power - Upscale setpoint is in LC0 3.3.1.1. " Reactor Protection i System (RPS) Instrumentation." i O ! l l FERMI - UNIT 2 B 3.4.1 - 3 Revision 2 01/18/99 l
l 1 7.s Recirculation Loops Operating f i B 3.4.1 V BASES APPLICABLE SAFETY ANALYSES (continued) Thermal hydraulic stability analysis (Ref. 5) has concluded that procedures for detecting and suppressing power oscillations that might be induced by a thermal hydraulic instability are necessary to provide reasonable assurance that the requirements of Reference 4 are satisfied. Recirculation loops operating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LC0 Two recirculation loops are required to be in operation with their flows matched within the limits specified in SR 3.4.1.2 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. With the limits specified in SR 3.4.1.2 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the APRM Simulated Thermal Power-Upscale setpoint (LC0 3.3.1.1) may be applied to allow continued operation consistent with the assumptions of the safety analysis. Operations that exhibit core thermal hydraulic instability are not permitted. Additionally. in order to avoid potential power oscillations due to thermal-hydraulic instability, operation at certain combinations of power and flow are not permitted. These restricted power and flow regions are referred to as the " Scram" and " Exit" regions and are defined by Bases Figure B 3.4.1-1. A Note is provided to allow 4 hours following the transition to single loop operation from two loop operation to l establish the APRM Simulated Thermal Power - Upscale setpoint in accordance with the single loop allowable value, l which is specified in Table 3.3.1.1-1. The 4 hour period is sufficient to make the adjustment given the relatively small change required. This transition only results in applying the new single-loop allowable values to APRM OPERABILITY. Any ARPM non compliance with the required allowable value after this 4 hour allowance, results in ACTIONS of LC0 3.3.1.1 being entered no ACTION of LC0 3.4.1 would apply. O l FERMI UNIT 2 B 3.4.1- 4 Revision 2, 01/18/99
I l Recirculation Loops Operating B 3.4.1 BASES l ACTIONS (continued) i L1 l When operating in the " Exit" region (refer to Figure L B 3.4.11), the potential for thermal-hydraulic I instabilities is increased and sufficient margin may not be
- available for operator response to sup)ress potential power oscillations. Therefore, action must ye initiated immediately to restore operation outside of the " Exit" region. Control rod insertion and/or core flow increases are designated as the means to accomplish this objective.
Required Action B.1 is modified by a Note that precludes core flow increases by restart of an idle recirculation I loop, or by resetting a recirculation flow limiter. Core I flow increases by these means would not support timely I completion of the action to restore operation outside the
" Exit" Region.
M With no recirculation loops in operation in MODE 2 the lk plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours. In this condition, the recirculation loops are not required to be operating because of the ! reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from MODE 2 conditions in an orderly manner and without challenging plant systems. L1 If operating with no recirculation pumps in operation in MODE 1 or operating in the " Scram" region (refer to Bases Figure B 3.4.1-1), or if core thermal hydraulic instability is detected, then unacceptable power oscillations may result. Therefore, the reactor mode switch must be immediately placed in the shutdown position to terminate the i potential for unacceptable power oscillations. l .l Thermal-hydraulic instability is evidenced by a sustained increase in APRM or LPRM peak to peak noise level reaching 2 or more times its initial level and occurring with a characteristic period of less than 3 seconds. C\ , LJ l l FERMI - UNIT 2 B 3.4.1 - 6 Revision 2, 01/18/99 l 1
.- - . .-- - -- . _ _ _ - - ~ . --..- -_- - - _ -
Recirculation Loops Operating [ B 3.4.1 BASES j ACTIONS (continued) ! If entry into this condition is an unavoidable and well : known consequence of an event, early initiation of the t Required Action is appropriate. Also i+ is recognized that ' during certain abnormal conditions, it muy become i operationally necessary to enter the " Scram" or " Exit" region for the purpose of: 1) protecting plant equipment, which if it were to fail could impact plant safety, or
- 2) protecting a safety or fuel operating limit. In these i cases, the appropriate actions for the region entered would be performed as required.
These requirements are consistent with References 5 and 6. SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR provides frequent periodic monitoring for core thermal-hydraulic instability by monitoring APRM and LPRM . signals for a sustained increase in APRM or LPRM peak to : peak noise level reaching 2 or more times its initial level L and occurring with a characteristic period of less than 3 seconds. The 1 hour Frequency is based on the small potential for core thermal hydraulic oscillations to occur outside the " Scram" or " Exit" regions. Therefore, frequent monitoring of the APRM and LPRM signals is appropriate when-operating in the " Stability Awareness" region. This SR is modified by a Note that states performance is only required when operating in the " Stability Awareness" region (refer to Bases Figure B 3.4.1-1) (i.e., in the power to flow region that is near regions of higher probability for core thermal hydraulic instabilities). This is acceptable because outside the " Stability Awareness" region, power and flow conditions are such that sufficient margin exists to the potential for core thermal-hydraulic instability to allow routine core monitoring. Any unanticipated entry into the " Stability Awareness" region would require immediate verification of core stability since the Surveillance would not be current. l FERMI UNIT 2 B 3.4.1- 7 Revision 2 01/18/99
'q Recirculation Loops Operating B 3.4.1 ; BASES
]
SURVEILLANCE REQUIREMENTS (continued) ! SR 3.4.1.2 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 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.
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 O' loo)s'are in operation. The 24 hour Frequency is consistent wit 1 the Surveillance Frequency for jet pump OPERABILITY l verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely raanner, l l l i i P I
- D )-
l FERMI - UNIT 2 B 3.4.1- 8 Revision 2 01/18/99 1
1 Recirculation Loops Operating i i B 3.4.1 i s BASES REFERENCES 1. UFSAR. Section 6.3.3.
- 2. NEDE 23785-P A. " SAFER /GESTR Models for the Evaluation of the Loss of Coolant Accident." Revision 1. October 1984. ;
- 3. M)E-56 0386. " Fermi 2 Single Loop 0)eration Analysis."
Rev. 1. April 1987, and NEDC-32313 P. "Enrico Fermi . Energy Center Unit 2 Single-Loop Operation." September . 1994. : i
- 4. 10 CFR 50. Appendix A. GDC 12. ;
- 5. NRC Generic Letter 94 02 "Long Term Solutions and I U) grade of Interim Operating Recommendations for 4 Tiermal Hydraulic Instabilities in Boiling Water I Reactors " July 1994.
- 6. BWROG Letter 94078. "BWR Owners' Group Guidelines for ,
Interim Corrective Action." June 1994. ! d i
?
i { 9
, (v)
J
, 1 FERMI.- UNIT 2 B 3.4.1- 9 Revision 2 01/18/99 l
t l L. 1 i
' Recirculation Loops Operating.
I B 3.4.1 l
. )
BASES j l 5 i i 9 I
'I a
s l 3 9 h l 1
?
t
. I 1
i 1( L.ATER 1 i i i I l i l
- - i THERMAL POWER vs CORE FLOW' s
Figure B 3.4.1-1 r. L l.. FERMI'-' UNIT 2 B 3.4.1- 10 . Revision 2, 01/18/99
l SPectFscanon) 3.4.1 (Afw see Spe4Scakan 3.3.1 I) l l SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.3-1. l 6d gpghg APPLICABILITY: As shown in Table 3.3.1-1. 3.%bl ACTION: With a reactor protection system instrumentation setpoint* less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value. C l 1 { & he l i
*The APRM Simulated Thermal Power - Upscale Functional Unit need not be h,\ l N#5 declared inoperable upon entering single recirculation loop operation provided the Flow Biased setpoints are changed within 4 hours per
'q Specification 3.4.1.1. l () FERMI - UNIT 2 2-3 Amendment No. JJ, 122 i PAGE / OF 06 g l
S PW R MwrJ 3 4.t (3 3 /4. 4 REACTOR COOLANT SY' STEM {M80SteS/GoikC4Ncm34/0) 3/4.4.1 RECIRCULATION SYSTEM RECIRCUtAT10N LOOPS , t IMITING- CONDITION FOR OPERATION Leo 3,4,i 3.t.1.1- Two reactor coolant system recirculation loops shall be in operation. - APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*. ACTION:
- a. With one reactor coolant system recirculation loop not in operation:
Llo M l 1. Within 4 hours: LAil tJtTT ' a) Pla the individual re culation pump flow co roller'for the 7 op ating recirculatio pump in the Manual mod . b) educe THERMAL POWE to less than or equal 67.2% of RATED THERMAL POWER.
) Limit the speed f the operating recire ation pump to les than or'eaual to 7 of rated pump speed s
'd) An ase Ine rilNI LKITICAL PDWLR Mall p%VK) baTEIy L1 11 value for sing loop operation req red by Specifica on 1.2. f O, e) Change the Average Power Range Monitor (APRM) Simulated Thermal, 3
Power - Upscale Flow Bissed Scram L-d Ld ;"en - sr et-+s/ I
@ M*! WA110wable Values to those applicable for single recirculation
' i loop operation per Specifications 2.2.1 and 3.3.6.
l f) Perform Surveille.nce Requirement 4.4.1.1.4 if THERMAL POWER is f,gg fa h ! g,4 ,,, f equal less than or equal to 30% of RATED THERMAL POWER or therecirc i to 50% of rated loop flow. v
-2. 00.:r t- ka i-it 1:::t ".0T IJW,.;n mnin sne um n .. m . . i
- b. With no reactor coolant system recirculation loop in operation while in
{: bDD OPERATIONAL CONDITION 1, immediately place the Reactor Mode Switch in the SHUTDOWN position.
- c. With no reactor coolant system recirculation loops in operation, while in OPERATIONAL CONDITION 2, initiate measures to place the unit in at least i htmDtd C HDT SHUTDOWN within the next 6 hours.
i see ecialTe3rException3.fD.4.] r FERMI - UNIT 2 3/4 4 Amendment No. JJ,64,JJ,EJ, E7.JES, 122 PAGE 2- 0F 06 AL 4
CPectFicxtlm) M .I REACTOR COOLANT SYSTEM 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY 8! LIMITING CONDITION FOR OPERATION Lco 3 tl I The iteactor core shall not exhibit core thermal-hydraulic instability
-0.0.10 8 nr be operated in the Scram or Exit Regions' = ;;;;; ::5 ' e i;s3 ~ l (4..w-a l APPLICABILITY: OPERATIONAL CONDITION 1 and 2.
.3 A,Gl.Ql[: Ode
- 4. With the Reactor operating in the Scram Regionj= ===;i'f rd tu) '
gpg Q gi;;r: 2.' in tJ immeciately place the Reactor Mode Switch in the Shutdown position. l
- b. With the Reactor operating in the Exit RegionJ: ::::i'i:d ',. "i: --
.e.10 Irimmediately Intuate action to leave the Exit Region by I' fcT10310 inserting control rods or by increasing core flow.* l
- c. If core thermal hydraulic instability occurdas u nceu uy .
st eo increa peak-to-peak ise level re ing I' O n a n m o r i. 2o re times s initial 1 1, and occurri with w f l MM A ::t -i r t M ,_-i ~' e f '-- Reactor Mode Switch in the Shutdown position.
+han 3 second C amediately place the I l
SURVEILLANCE REQUIREMENTS g*.iv.4 ine uminons vi oym;r;,eu;r. 4.0.: r: .=.;;1i;d:. g pp g 1.0.20.2 When operating within the Stability Awarenett Rooic e : :::: m :: , j
'i; n ; . - - II verify that the reactor core is not exhibiting core I
Inermal hydraulic instability [#y - it:P : " r n d ;.._. : 7 "# (:$i!:!y:gatleastonceeverynour. l [CTid* Restarting an Idle Recirculation Loop or resetting a Recirculation 8 NDE Flow Limiter are not acceptable methods of immediately increasing core I flow to leave the Exit Region. ! i O FERMI - UNIT 2 3/4 4 30 Amendment No. JJ.128 l PAGE F 0F 06 gw 1
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CP&~cIfrCADetJ 5.I.'3 REACTIVITY CONTROL SYSTEMS [ Abo S#4 6/(cI$'cdb M 8 l*f) LIMITING CONDITION FOR DPERATION (Continued) ACTION: (Continued) g (,,3 K the inoperable control rod (s) is inserted, within 1; hour OpM O2t. disarm /Ine associawa ir m onal c trol valves = itner: a E ctrically or i ydraulic y by closing he drive water and exhau ater is ation valves. f-- < - bNE Otherwise, be in at least HDT SHUTDOWN within the next 1 12 hours,
- c. With more than 8 control rods inoperable, be in at least HOT SHUTDOWN,within 12 hours.
*di
. hithoneormorescramdischargevolumeventordrainlineswithone
. valve inoperable, restore the ino btMM'^ '
within 7 days, or be in at least berable valve (s) to OPERABLE status I 3 T
*** e.
/ hours. T SHUTDOWN within the next 12 With one or more scram discharge volume vent or drain lines with both '
valves inoperable, isolate the associated line within 8 hours ****, or 3 in at least HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REOUIREMENTS f4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated QPERABLEby: M *Y h I At least once per 31 days verifying each valve to be open,* and Evaluating scram discharge volume system response prior to plant neg Ac4A3
' startup after each scram to verify that no abnormalities exist.
i g s.t.i.t u ok. 4.1.{.ppn rods ... . above the preset power level of the @all,, withdrawn control , , QJ f ,....f .. .....
- ...... ......._.._. __..... ...._, .... - . ......,,,, m
..,1..L.W shall be demonstrated OPERABLE bytumuaigeach control rod at g g, - - - - -
j g u,3,ta. At least once per ays, and
- b. Within 24 hours when any control rod is Li ab)e as a rwsun er ]! N1 h Act4 3 6.ssiyt frictioyor mecipntcal Inteprer[ee.f g fg ~
PLcantenis eJ*These valves may be closed intermittently for testing under administrative ou r--
~
ting ds .ociate( insprutttent1# under admististrativorcontrolfto perm with res)6 ring the ep6 trol ro@(o OPE 8 =___E r--- wrerate Action entry 1s allowee for each sur vent and drain line. stat 8 (
****An isolated line may be unisolated under administrative centrol to allow draining and venting of the SDV.
I { FERMI - UNIT 2 44 M hCA Ndhlt. 3/4 1-4 Amendment No. J/, U , M ,120 . PAGE 3 0F 10 . Ry/ 2
l
,$@EClFICAcT10N 3I'Y REACTIVITY CONTROL SYSTEMS f50 Tu b/ d b N*' N' LIMITING CONDITION FOR OPERATION (Continued)
ACTION: (Continued)
- 2. If the inoperable control rod (s) is inserted, within I hour disarm the associated directional control valves ** either:
544 a) Electrically, or : g b) Hydraulically by closing the drive water and exhaust 8'N water isolation valves. Otherwise, be in at least HOT SHUTDOW within the next ' 12 hours,
- p. With more than 8 control rods inoperable, be in at least HOT LCo 318 _ t SHUTDOW within 12 hours. Add: Loo 3/,Ek d 2-
*** d. With one or more scram discharge volume vent or drain lines with one Agg4 valve ino erable, restore the inoperable valve (s) to OPERABLE status within 7 ays, or be in at least HOT SHUTDOW within the next 12 4 CT M C hours.
*** e With one or more scram discharge volume vent or drain lines with both AcrkrJ B. valves inoperable, isolate the associated line within 8 hours ****, or gg c be in at least HOT SHUTDOW within the next 12 hours.
SURVEILLANCE REQUIREMENTS f.I.0.2. - The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by: sR3.18ia. At least once per 31 days verifying each valve to be open,* and
. aluating se system respo e prior to p1 t LR.I startup aftgr,r,ah discharge each scram t yo f erify that no ormalities e t. 1 3.1.2 When above the preset power level of the RW, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:
- h. a. At least once per 7 days, and l 3*I3 b. Within 24 hours when any control rod is famovable as a result of i
( excessive friction or mechanical interference. $R 3 %
- controls.These valves may be closed intemittently for testing under administrative Su ay be rearmed intermittently, under administrative control, to pernit Pf'['3 ta= ting. associated with restoring the control rod to OPERABLE status.
***5eparate Action entry is allowed for each 50V vent and drain line.
An isolated line may be unisolated under administrative control to allow I draining and venting of the SDY. O - 86170tJj MOT 6heguired Atdim 8-l No4L. FERMI - UNIT 2 3/4 1-4 Amendment No. J/, JJ, D,120 1
.PAGE i 0F 02
.Ra 2
3/4.3 INSTRUMENTATION S PECi Fi c4Tited 5. 3. /./ 3 /4 . 3 .1 REACTOR PROTECTION SYSTEM INSTRUMENTATION tIMITING CONDITION FOR OPERATION As a minimum, the reactor in Table 3.3.1-1 shall be OPERABLkrotection system instrumentation channels shown 3.3 Id 3-1 . APPLICABILITY: As shown in Table ACTION: 3.3.1-1. [ /}00 4677PM4 NOTZih _
- a. With the number of OPERABLE channels less than required by the Minimum OPERABLE channels per Trip System requirement for one trip system:#
LC.
- 1. Within I hour, istify ih ifeach Functional Unitlu+hia t' - I 6ffecteu b iv -usiem
- - - - m ,c o n u i n s n o mo re . I. n a n. ane ino aerapie channeT) [g,3 hc.770/J C ,,. , , , , , . - . _ _ , _ . - - _ _ _ . . . . _ _ _
EWW@M.QW;"iiijr;'"%diIisjY u -, .,_.
' '" .m'"'I
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- withi fiours
- - - - - or
- - -the
- - - -Act
- - - bH i required by Table gc,Tiod b 3.3.1-1 for the affected ional Unit shall be taken. i
- 3. +ha . ;47 = .--d k -..;...oJ '. in Ja t. .
-If -p;4.;; ;;..;; place the inoperable chr;J ndit:;,..
gNgA r .d that trip system in the tripped condition within 12 annel(s hour)s. and/or j 1 b With the numbe'r of OPERABLE channels less than required by the Minimum I kCT1od 6 OPERABLE Channels per Trip System requirement for both trip system , L,2. l
/)cTipd 4.0 placeatleastonetripsystem**inthetrippedconditionwithin and takp_the ACT10N rannired hv Tshia 3.3.1 ._!
hour i Y-~ in IZ hoght fr'ip cher CWnsis or drip _sV 5 N&- g !
- c. With one or morhannels requirea oy laore 3.3.1-1 inoperable in one
\
d- ' or more APRM Functional Units 2.a. 2.b. 2.c, or 2.d:
- 1. Within I hour, verify sufficient ' channels remain OPERABLE or k CSO /J C- tripped *** to maintain trip capability in the Functional Unit, and A CT1D g g 2. Within 12 hours, e ore the inoperable channels to an OPERABLE status or tripped
.e Otherwise, take the ACTION required by Table 3.3.1 1 for the Functional Unit.
RegA ct A 2Nokj A4710n B N& utActions a and b not applicable to APRM Functional Units 2.a. 2.b 2.c, and 112.d. Action c applies only to APRM functions 2.a. 2.b, 2.c and 2.d. l n vgui . aonne eus no
- gun,gdenutpeaser>mpnneene7uepiace7envoer en une t r wo en r n n co i -n ww re mi t
'- +ker i n anun taus e unain.eesan d besto$rpd 6v vr tn Nod laiu orfthe fiu. - ACTION - -: .her the c iann(1 was fir /t
- eter inndrable required by Iable 3.3.1-1 for that Functional Unit shall be taken.
Ac TIOv M,pne 6 rip system neeg out -n'm - = m ppe: :=- - ,1 1nis, 4
- u rr f
- f. cause i ter2m tn n/eur/ p en . L ;p ;j-u n n M -1;; > ja "h; trip-i" pund,'iwu wi Uavd wou 'nw ascramtooeur,placetbe i system wit the most noperable chann s in the tripp d condition; if ot systems h e th sam number of inone hia "h>nnpit- lace either tr h*9 svetem in the trinn W eth. # '
" An inop raole channel n >0 not be placed in the tripped c would ause a scram to occur. In thes cases, if'the i erable ditionchannel where thi ]
not stored to OPE E status withi the required ti A* L by , the ACTION re red;) ble 3.3.1-1 for the Functional it shall be tak nf FERMI - UNIT 2 3/4 3-1 Amendment No. JE, p , J $ , 122 PAGE 4 0F 11 hv 2.
_ _ . _ _ _ ~ _ _ _ . . . __ _ _ _ _ . . . , . - . . . _ _ _ _ _ _ . _ _ . _ _ _ . . _ - _ . . _ . _ SPEts FI Skrrohl 13 t.( 314.3 INSTRUMENTATION LIMTTING CONDITTON FOR OPFRATION (Continued) SURVEf tt ANCE REDUIRE~riEv=i5 - j ge M Li- Each reactor protection system instrumentation channel shall be g,rsrs demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNIL. FUNCTIONAL TEST,' and.CNANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.3-1. 4.// sf 3 311.15'4.3.1." LOGIC SYSTEM FUNCTIONAL TEST 5P" :*-mt'"' r'--ti: -- :ui:Bof all channels shall be performed at least once per 18 months, except "able 8 i 4.3.1.1-1 Items 2.a. 2.b, 2.c 2.d..and 2.e. Functions 2.a, 2.b. 2.c, and I 2.d do not require separate LOGIC SY5 TEM FUNCTIONAL TESTS. For FuncHaa 9.e' I l StJ.31.l.84 tests shall be performed at least once per 24 month M e LOG SYSTEM l I f tunci .L ns for r cuan c.= ie.sia ' ,; i nng APRM tr conditions a I the RM channel in s to the 2-out- 4 Trip Voter chann to check all I j co inations of t tripped inputs t he 2-out-of-4 Tri oter logic in heJ~ ' i Qfterchannels. !
531417'.:.;.: The REACTOR PROTECTION SYSTEM RESPONSE TIME of reactor trip functional unit
- shall be demonstrated to be within its a t at least pace i
- )(per18 t e n. months.Jeutrondetectorsareexemotfromresponsetimetestings ass inwimos at usast vies- 6 unsi per tri system sucn to l
sai chan s Jare ested at le t once every imes 18 mon where N is t total n { edundant ch nels in a spe fic reactor ip systers. , r( E S21 l.lf /100 SR 3.3./.ld7 ' \ Mowl. y3
% _^_
=
enq STAGGEKeD TESr8A - a SR 3 3.l.I.I7 ADW'l- [g,Q ' .- High a
*The Reactorsensor Vessel Low response Water Leve)time
- Levelfor Reactor 3 need not be measured Vessel Steam Dome Pressuhndjr gspumeopo oe une owny sena. .m --~ j 4me. f
~
FERMI . UNIT 2 3/4 3-la Amendment No. 75, Jpp. JJJ.122 PAGE E_0F 11 M
TABLE'3.3 REACTOR PROTECTION SYSTEM INSTRUMENTATION. MM c.
.a T64.3 3.l.1 -/ APPLICABLE MINIMUM E Fv N cT1ord-c
.: UNCTIONAL UNIT '
. OPERATIONAL CONDITIONS OPERABLE CilANNE PER TRIP SYSTE a ACTION
[ 5 1. Intermediate Range Monitors : [ g,4 et Neutron Flux - High. 2 3 __
'l d.
f,g .ir: Inoperative
'd ~
@ jD '5 E 2_ 3 1 G
-( 5 3(d) $ r-
- 2. Average Power Range Monitor.'
23 a. Neutron Flux - Upscale (Seldown)~ 2 3(k) 'I 4
- 7.b R Simulated Thermal Power - Upscale ~ F l 3(k) 4 p 1 g ,Aa Neutron Flux - Upscale 1 3(k) 4 5:
[ gA Inoperative 1, 2 3(k) 1 q. O 2.c t. 2-out-or-4 Trip Voters I , 2_ 2 (, m s 1 3 p.' Reactor Vessel Steam Dome u Pressure - High 1, 2 i G t/ A. Reactor Vessel low Water level - Level 3 'l,' 2 2 1 4 g _E s'. Ma steam Line Isolation Valve - ia O 1 W a F
.h . Y i
I O Q S Peca 1 canon 3.5.I.I TABLE 3.3. -1 (Continued) REACTOR PROTECTION SYSTEM INSTRUMENTATION 'l TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 6 hours for SR t36lc Z. required surveillance /without placing the trip system in the trippea condition provided at least one OPERABLE channel in the same trip syste is monitoring that parameter. f 4 c Run . 7 59 Unle adequate s utdown mar n has been de nstrated p ' l Spe ification 3. .I, the "s rting links" all be rem ed from i RP I c' cuitry prio to and dur' g the time a control ro is withdr n.* . (d) When the "s rting link are removed, he Minimum PERABLE Ch nels Pey Trio Syste is 6 IRMs d per Specifi tion 3.9.2 2 SRMs. # l l d' g , . . . :. - l This/Tunction is/not required toAe OPERABLE >fhen the re/ctor oressef9f
- v. 61 h. d u 4- e --- e-- Ir4--+<-- 2 %1 e
~~ i a iuna omatica ypassea wnen Tjve reactoy muy itch is [ sna11 in thebeRua osition.
s funct is not requ eo to be OPymBLE when PRIMjKY CONTWtny M EGRITY not require . f N8 M (i) With any control rod withdrawn / ;t r-H:%.1 Gnu vi
--- c a r < n em + 4 nn t o in 1 n 1 An C
.vus .. ~..J -
l (d) This function shall be : t::: tic;il,, VuiAth
~
bypassed when tsi.yr 6
.M .; 'i.L 5R 33.I l.16 v> =>> us = >> > 101.s v ais, .c.a.;i;;.t te THERMAL POWER hr "9 30% of
( RATED THERMAL POWER. gam SR ( Itrno VeriOchRmk *
. (k) f,*se-each APRM channel provides input to both trip systemsyti c "..... q TYL fop acie cnan as specifieo in Die 3.3.1-1 are e tocas APRM annels 3,M.t -f r utred (i. ., it is not on trip system basi '. The 6 hour lowed f,g (c) est time complete a cha el surveillance t t (note a e applicabl provided at lea two OPERABLE ch nels are mo(n) i ringa th ve)is gramet . - -
l LR.1 t require or co [ rods removed p[ Specification 3/9.10.1 or[9.10.2.) O' FERMI - UNIT 2 3/4 3-5 Amendment No. 75, E7,122 1 4 0F 11 PAGE [av1
4 s
. 9.B .l.1-[ ~
TABLE 4. 3. i . ; -; _A REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS I 'I~8L 3 3 l<l-1 N@ad CilANNEL ~ CIIANNEL . FUNCT10NAL CilANNEL ' OPERAT10NAL CONDITIONS FOR WHICH- - E W UNCTIONAL UNIT CHECK- TEST Call 8 RATION (a) a .
<8) g ,> SURVEILLANCE REQUIRED
..k Intermediate Range Monitors:
.m
.x; . ' Neutron Flux'- High' f4 S,(b) M W-*>
IA LSl SA-(ll[ - 2 L'I i S-(t) fW% SA-<llJ 5~
- g. tr. Inoperative NA W -( alg) NA 2, 5
- 2. Average [ Power Range MonitorIII: (2) j 2,g p Neutron Flux - .
% - Upscale (Setdown). ,>D[,(b)g7~ - A(m) 2 year
<isy .
O O OJoTE>--- L. 2
-[
FT1 2J 5~. Simulated Thermal L/f.4 ' Power ~- Upscale 0 42) ta.9 S 3)- W(d), 2 year ; g .c , Neutron Flux - Upscale D -D > M'O . 3/ W(d), 2 yea it) 1 't de A- Inoperative NA SA--Qr.) NA 1, 2 L 71 y y;- 2-out-of-4 Trip Voters D -42.) ' SA-(sz,) NA 1, 2 W ' 3 4~ Reactor Vessel Steam Dome
- E Pressure - High 5-(l) k) 4 D R-(18f) I, 2.
14 /( Reactor Vessel low Water Q(M9) . 3- Level - Level 3 SM> Q(k)-g,,,y R- ("l> 1, 2 ]W !
.?6K z.
Hain Steam Line Isolation-M9') - 4 ? " Valve - Clo,sure NA Q -(9 ') R - < ii) I
$ f6 Main Steam Line Radiation -
y liigh S --d ) Q M) R-(s*{ } 1, 2 -
/ Drywell Pressure - High S-(t) Q,(k) M - R-(W) 1, 2 gN =7 % ses.,.<.i.n,T6 @ L 4 DD s R3 3 f l.Ir /JoTe v J ^
k ' As 54. 3.3,l.i.iz. Noir sR 3 3 l 3 < x> m_.-__m__.mm. _ -___2-_._____..m..._ ___m__ .._m-_.__m_ _ . _ ____ m.____ _ __.___ ___m___.-_=,__a._.,.,.u - 2_m _ _ - . . . .-w., e .- ._.w.m . ,.i..nm.,.-e m- -m.-- _. .a ,_e._-m m__ r..,,m
i . 5 3,3.f.H y TABLE-4.3.i.i- H Continued) E REACTOR PROTECTION SYSTEH INSTRUMENTATION SURVEILLANCE REOUIREMENTS S TABLG O - 3 3.1.I-l *
" Ft/M CT10t) CllANNEL OPERATIONAL CHANNEL. FUNCTIONAL CHANNEL
" 7 UNCT10NAL UNIT CH(CK TEST CAllBRATION CONDITIONS FOR WHICll O
- 8. Scram Discharge Volume Water SURVElltANCE REOUIRED m Level - liigh Q a. Float Switch NA Q
#(9) R -(14) 7.a b. Level Transmitter 5-(t) I,2,5(j)
' -a C
Q[kki$NC R 4:4) 1,2,5(j) l 9. Turbine Stop Valve - Closure NA Q49) R -( t'Q to 10. Turbine Control Valve Fast h(so%Firr A.'l h ~ Closure NA Qf9) NA h fr-tl 11. Reactor Mode Switch
~~
g Shutdown Position NA R-bO NA 1,' 2, 3, 5 H Lh, y r2.12. Manual Scram HA W -<(5) NA 1, 2, , 4. 5> 1+r D& k4,-- h
$A$3.IJ.St AnQQ 5A 1.1.4.5.s1 floid litet% tsentron detectors enay be escluded from CHANN(( (Al[Entt0N. [fdef-lo_-Ugis enwR$ $& V g gs.t. t (b) f - - - - - --
g,g,p..h .g TheAPRM and IRMchannels aed sRM shall channels be detennined shall to overlap be tieterminede a to overlap i aj/tesp Y.jJfecaduring each startup erp entery, ortrgprimt waviiUPt Dand the I eas( f, detachrtIduring each Y J'~-y'...pw .' -p _ -"-"; =---- s '- w L O .I ed shutdown. If not perforwed =tthtn the pre,tous 7 days.
.... .. n WI344 hs -") g This calibration shall consist of the adjustment of the AFEM thennel to conforse to the M to
.er valves celestated by a heat balance during CrtRAllCut h
$2 3 "$.I.L 3
? W.
cow 0lil0N I = hen THERMAL POVER 3 ist of RAlt0 THERMAL POWER. 5h ~ Adjust the APRM chanc I If the absolute difference-is greater than 2% of RAlt0 Intact,
, i.e.
~
p
-s
'm e trkk[sheth be.calibratedSt keast ente perruhr 1000(abarE.e co-cr h = m s nq esten the TIP system som-h-
, 2,17 g s gg U hj ..__ . __. -. ..,..,e ,. t. c e r e , e, r A- e s ... . - - ---
- L b e- -
it ).blel.lD[ T N-45
-(4.L _
With any centro! rod =lthdrawn * ":' laclodes verificatlon of the,tr r"-d'r
- po nt of the trip unit. . . ,. . r r-' :m - -:-'^'. :*'"' e, ,m, .;- g f gp)
. ..e __
(R 3 /3.l.l.l?Y ' ' ' Not required to be perfor+ed
- ben enter cg MODE 2 frce MODE l entil 12 hours af ter er.tering M30t 2.
,...m
......3 .r., u.ein ..
l . l - a ek CLU h~ U _ _ _ . _ _ _ _ _ _ _ _ . . . . _ _ _ _ _ - . _ - - - - _ - - - - -- - - - - - ~ ~'" ~ ~ ~ ~
_- . - ~ . - . . .- . - - . - - . - _ - . . - - . . _ - . - . - . _ . t l 0
%/
WEClF10%T100 372d ( INSTRUMENTATION 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION [,[ tIMITING CONDITION FOR OPERATION LC811,1I-t:$-t. The control rod block instrumentation channels shown in Table 3.3.61 shall be OPERABLEjwirn Infir trip smpoints set cofststent witnj tne valug 3 (snownfin Inc fr.p Setpottit column f" Table 3.3.6f. 7
'I APPLTCABitTTY: As shown in Table 3.3.6-1.
ACTION: L a. With a control rod block instrumentation channel trip setpoint* 1ess g*" g t l: conservative than the value shown in the Allowable Values column of , l Table 3.3.6-2, declare the channel inoperable until the channel is ;( l restored to OPERAgLE statusnrTin yn trip sypoint adatyceo;
~
[conpfttent witu p e arip 5etpoint value. & [A,( ! b. With the number of OPERABLE channels less than. required by the Acnoo A,B,E Minimum OPERABLE Channels per Trip Function requirement, take the
' ACTION required by Table 3.3.6-1.
p.I l l SURVEILLANCE REOUTREMENTS l
.3R PJ6E #
l -4 r3,4 Each of the above required control rod block trip systems and instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.6-1. ffl00 SAM 2-l l l R.I f*TheAPRM mulated Thermal ower - Upscale Fun 61onal Unit need not be l declared noperable upon tering single rea or recirculation loop o rat] ion 5 provid the flow Biase setpoints are chan d within 4 hours per l Qeci cation 3.4.1.1.
"V FERMI - UNIT 2 3/4 3-41 Amendment No. D , S.121 PAGE .3 0F 09 ov1
rABLE 3.3.5 ? 'I'7*I-/ CONTROL R00 BLOCK IN3TRUMENTATION A N# '
'l MINIMUM APPLICABLE 5 f~-V W M OPERABLE CilANNELS OPERATIONAL
~* M RIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION
- 1. ROD BLOCK HONITOR(a)
- a. Upscale E (l+A q .b b. Inoperative 2 2
1* 60 }g 33,2./ b 4,8 1* 60 m f . c- c. Downscale 2 1* 60 i J2. AVERAGE POWER RANGE MONITOR i l ,
- a. Si ated Thermal Power - U scale 3 1 6 8
- b. I perative 3 1, 2 I I
- c. utron Flux - Downscale (d.a, imulated Thermal Power - Upscale (Setdown) 3 3
1 61
& I 3
C) _ ring . Uoscale 3 2 1 61 61 ; , .
+
rTl '3. SOURCE RAMGE MONITORS A w a. Dete or not full in(b) 2 4 1 3(I) 2 5 6 61 w b. Up ale (C) 2 61
~
=
3(f) 2 5 Cl g c. operative (C) 3( 2 6i 8.1 ri 2 5 e Downscale(d) 3 x f) o M NTERMEDIATE RANGE MONITOP ' 4 f .a. 4tector not full in 6 z, 5 b./Jpscale - 6 2, 5 1-f.Inoperatig) 6 2, 5 61 % nawnerale , K ?- 4 61). f5.SCidMDISCHARGEVOLUMif -
! Cf"#:"si 's;af m
/ i / i: i d 7 !!) O 4u s nai dad .
-l $
g 7. REACTOR MODE SWITCil SifUTDOWN POSITION 2 {3r-+d 63 } Jtz I W E h ' T p" @. L nAGO 8L3-33 l-I
'N
_ _ _ _ _ _ _ ____._.________________________s________. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- __.-m - - -_ _ _ _ _ _
O4 O O i 3.*3.2,1-I
- TABLE 3. .e-2 l i TD6te 532.1-1 'ONTROL ROD BLOCK INSTRUMENTATION SETPOINTS i
- RN C-Tio O !
b lP FUNCTION ITRIPSEI/ DINT ALLOWABLE VALUE
. .L - ROD BLOCK MONITOR c .a Upscale .I '
As ecified in the As specified in the 5
* 'I'E C OPERATING MITS REPORT
- CORE OPERATING LIMITS REPORT N
$ 4rt- Inoperative / NA NA 1.c c Downscale 194% of eference L e1 192.3% of Reference Level ,
J g -
- 2. '
g l AVERAGE POWEI/ RANGE MONITOR l rT1 a. Simula Thermal Power - Upscale 3 g,[ : J 1) fl Blased s0.63(W-AW),+55.6%, 0.63(W-AW),+58.5%, I with a maximum of with a maximum of
@ s
, 2) Igh Flow Clamped 108% of RATED THERMAL , 110% of RATED THERMAL POWER i
- b. operative NA NA o [ c. eutron Flux - Downscale a 5% of RATED THERMAL ER a 3% of RATED THERMAL POWE
*yy a d. Simulated Thermal Power - t Upscale (Setdown) s 12% of RATED THE . POWER s 14% of RATED THERMAL ER Flow flaer=1= s 110% of rated f1 s 113% of rated flow
@ kF.
E 3 SOURCE RANn NONiluto x
- a. Detec r not full in 4 V
@N '
I b. Ups e 1.0 x 105 cps s 1.6 x 105 cps 52 5 c. In erative NA NA k,2, m g d. D nscale a 3 cps ** a 2 cps ** g
. R.I h C LA . . L .... n. .,.. m .m e- - ~ ..... 7 5 1 y ['T .criyu rs=vr menge- or slaulates i i rower - upscene e s ieses nos unoct ser Int vor m .. . ^ __ r' __
_ . ioop drive ( m f [W). AW is define the difference Indicated drive flow ( percent of drive f whicts produces ret core fle=] bet two loop ecd ngle toep operstlee a he se== core il AW = 01 for two I ellon. 8W = OE r single loop operall . Q w D
** ~
9 y . M . i
I A TABLE 3.3.6-2' E
, CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS (Continued)
', E 1 M TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE : I 4. INTERMEDIATI RANGE MONITQB1 \ k a. Detec r not full in NA NA
- b. Ups le s108/125 divisions of s110/125 divisions of [
full scale full scale r 3 C) c. operative NA NA rT1
. Downscale a 5/125 divisions of a 3/125 divisions f i w ( x rall m ia full scale Q1 w
- 5. 5tHiri UI5LHARGE VOLUML h a Water level - liigh s 589'11 s 59 "
ev O ( b. Scram Trip Bypass NA
- 5. Delvied j p j7.
a REACTOR MODE SWITCH Sil0TDOWN POSITION NA NA g g
,Y. - TEL s.7.2.I-I Funenon 3 g E
.E e
Q . 4 N O n E t i
O O . O-i
" '#*I'I TABLE t. .! :
, CONTROL ROD BLOCK INSTRUMENTATION SURVElllANCE REQUIREMENTS
~
7 96LE 33* 4 8 CllANNEL puncpoN OPERATIONAL i tilANNEL FUNCTIONAL CllANNEL CONDITIONS FOR WHICll
' ~ iRIP FUNCTION CHECK 1EST CAllBRATION(a)
EI 1. ROD BLOCK _NONITOR SURVEILLANCE REQUIRED e m h a. b. Upscale Inoperative NA NA 5 S 2 years 1* g c. Downscale NA SA {sst 3 3.2.i.3NA 2 years s a 1 3 2-8 G 1* i 1*
- 2. AVERAGE JUWER RANGE MONITOR i
- a. !
Si lated thermal Power - U scale NA SA 2 years 4
- cy b. noperative NA
% c.
SA NA 1, 2 Neutron Flux - Downst e NA SA 2 years N-[ i h Simulated Thermal P Upscale (Seldown) r-NA 5 2 years 1 t 2 g- Flow - Upscale NA 2 years I ! CA w I 3. SOURPE RANGE MONITORS : h a. Jetector not full in NA S/U$ ,W NA 2***, 5 t b Upscale 5 S/ ,W SA 2***, 5
. Inoperative NA S ,W NA 2***, 5 )
( , Downscale t S .W SA 7'**- E 1 O I4, INTERNEDIATE RANGE MONIIUH5 4 a. b. etector not full in Upscale NA S S/U S/U
, NA SA 2,
2, 5 5} g' Inoperative NA S/U d NA 2, 5 Downstale S S/U .W g SA 2,
- 5. SCRAMIISCHARGEVOLUME g a. ter Level - High NA Q R
~
s
/1 - z, 3=- } 3 g b. cram Trip Bypass NA R NA / 2, 5** / ,
f 4. "ikf l3
$ 3 A'. REACTOR MODE SWITCH g# SHU1DOWN POSIT 10N NA R '(geon.m NA 7w Sg 3 3.2.14 g
y i ML F
-____ .. ___- . . _ - -- _ _ _ ___ _ _______ _ __)
O -i m .r. r (cent!r r 4 GPECiPIC&T10tJ 3.3 2of i l 1 CONTROL ROD BLOCK INSTRLMENTATION SURVEILLANCE REOUIREMENTS TAatt NOTATIONS s R S g (a) Neutron detectors may be excluded from. CHANNEL CALIBRATION. v1gg hours pyh6r to startup, ifAt performedynhin tne prevp. R .2. 3,[ .k-l
- With THERMAL POWER greater than or equal to 30% of RATED THERMAL POWER.
l
%@ "Ea'"e;" .:nntrti" IT!ci"gi'i.it!"/" """% ,
c.
,,,,,__..._,_,,..,k - f,2 O
l 't i 1 l l l 1 FERMI - UNIT 2 3/4 3-46 Amendment No. JJ. 212
~ !
PAGE- 9 _ 0F 09 l Pa 2.
o ' O O i 3.33.1-I
, TABLE' .3.7.' ; (Continmed)~
h ACCIDENT NONITORING INSTRUNENTA110N 7 T6L 3. 3 3. I-/ E MTION REQUIRE 0 NUMBER
,m,catg M y INSTRUMENT CHAletELS OPERA 110NAL i
- OF EHAlgliLL._ OPER$0LE ColWITIONS ACTION l
11 SL:f'., en i esw..t syste.n Radiation -~.. stors - i
- a. SGIS - oble Gas (Low-range)I / l/0PERABLE I/0PERABLE I,2,3 81 SG15 subsystem SGIS subsystem g*l !
- b. .S 5 - Noble Gas (Mid-ran ) I/0PERA8l i I/0PERA8LE 1, 2, 3 I g SGTS subs tee SGIS subsysten M c. SG15 AXM-Noble Gas Id-range) 1/0 SGT subsystem LE l/0PERABLE SGIS subsystes 1, 2, 3 81 l
i QMy a-t _ _. scit . Anm.u-u - c . ruaf -ranael I/0PERA8LE C GTS subsystem I/0PERABLE 1 ,3 8 ; e SGTS subsystem g, o .. i 1
- 15. S':ted '
/O Mr. Primary Containment isolation Valve Position '/..; w]
- w (M 1, 2[ -82 g I a L.6 fzfenetm% y ;
- ca> ls> .3 t.z -
7 E . E } ! L _ 3 Also ncluded in th FSITE DOSE CALCUL ON MAN
./
[ - g m . I p . N 4
. . . - . ._ .--...__~-_--_.-___...--___.__.-_--...-.--_.--__---__-__---___-_-_-___._.--__--_________-.--------,-_-___--_.e -
WECtficAT1& 3 33,l TABLE h (Continued) (AISO 3"' S P'#lNC#N'^ 5 lh ACCIDENT MONITORING INSTRUMENTATION ACTION STATEMENTS w_ k4Do t AcTroA) 8 h
- a.
- With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore l'3 ,
tQoperable channe)(s) to OPERABLE status wit ' _ ays W H W-* ' r , , N _ _ t~ '_ =.~ u~ .:p r-r .~ "::: :,- '.::~:r ' 3 0 r _ _ - ys
- b. the number of OPERABLE accident monitoring instrumentation s'_-With nels
' 4 @y D less than the Minimum Channels OPERABLE requirements of Table .3.7.5-1, restore the inoperable channel (s) to OPERABLE status withinGO : . M be in Adtpy F aheapHOT SH,,@,__N ,
W within the next 12 hours. A e H O N 4 4---- With the number of OPERABLE gog c than required by the minimum Lg,) channels OPERABLE rec uirementsjinrusi.Jne ,,, ,= .p .........y......g. Qf mon)cinns the soorbor ate omraketerfaWithin 72 hs0r:Ac.c
# 1) either restore the ino(perable channel (s) to OPERABLEL.9 status within days of the event, or
%w h 2) prepara and submit a Special Report to the Commission pursuant to SPMdo'b 54 7 Sp+d" ^--. 6.9.2 within 14 days following the event outlining the action
( taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
^C".OM S - With the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, withinl%Aewabither:
)rld'MN %3o d.I3 a. Restore the inoperable channel (s) to OPERABLE status, or L*r W'$
I 4CT1od
===- C!
i - 7 6t 3,3.4,l-l b." are the affected isolation valve inoperapie ano == Ine m HUN) g ( nors lam specified by Speerfication 3.6.3 ACTION a#
' ~ ~ 4
, , P - AcT1ow F (Be.4 rnoor 3)
ACTION B3 ' -ncium; e mo rm .r,V1 '
- a. With the number ofOPE LE accident monitoring instrumentation channels 4Cnod A less than the Required Number of Channels shown in Table 3.3.7.5-1 restore the inoperable channel (sho OPERABLE status within 30 da_ys,j5r su6mTa---
4 tug 8
~
report to the 6o~mm'ission purTs aniTo SpadOG5ir6.9'2 witnin the following f 14 days outlining the action taken, the cause of the inoperability, and the plans Sec.t S et 5 6 7)and schedule for restoring the instrument channel (s) to OPERABLE status.
- b. With the number of OPERABLE accident monitoring instrumentation channels d CU bid O less than the Minimum Channels OPERABLE requirements of ble3.3.7.5-1, restore the inoperable channel (s) to OPERABLE status within hours or be in gg p at least HOT SHUTDOWN within the next 12 hours.
72 LN O FERM1 - UNIT 2 3/4 3-62 Amendment No. 28.56,117 PAGE k 0F 06 hi2-
O
~
O O
,. '"C'l ". .' 5 ' 'r 1- d ; i ACCIDENT MON 110 RING INSTRtMENTATION SURVEILLME R[GulREMEtts E APPLICA8tE U CHANNEL CHAIflEL OPERAllollAL .i INSTRUN(NT CHECK
~ CALIMIATION C0lEllil0NS !
- 13. Stand Gas Treatment Syst adiation Monitors
- a. SGIS - Noble Gas (L frange).. M R 1,2,3 R./ i b SGTS - Noble Gas I d-range) l M A 1, 2 ;
. SGTS - AIti-Noble as (Mid-range) M R 1, 2.
2 cerc _ a vu i c ., su s A. --- ) M R ], 3 ' 3 9
,. -.~.
e
- m. 1, ._,_._, .
e - U* ,
.Jk Primary Containment Isolation Valve Position p
- (f)M (3) R 1, Zh E {O L l.7.
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- .6.C 3,3-3.I - ( x > .
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i S PEco Ft cA-nas) S *f ! l 3 /4. 4 REACTOR COOLANT SY' STEM I 3/A.4.1 RECIRCULATION SYSTEM RECIRCUtATION LOOPS , tIMITING CONDITION FOR OPERATf0N Lco gj 2.'.i.1-- Two reactor coolant system recirculation loops shall be in operation. ApPLICABf LfTY: OPERATIONAL CONDITIONS I and 2*. A[IlM:
- a. With one reactor coolant system recirculation loop not in operation:
LLo M1 1. Within 4 hours: LAil 14tTIT' l
'a) Pla the individual re culation pump flow co roller' for the l l op ating recirculatio pump in the Manual mod .
b) educe THERMAL POWE to less than or equal 67.2% of RATED THERMAL POWER.
) Limit the speed f the operating recirc ation pump to les than or'eaual to 7 of rated pump speed e
^
d) inc ase Ine Minin LRITICAL PDWLK KAI th value for sing loopoperationreq[Ared (nwx) barety List nt ]to by Specifica i (D a e) Change the Average Power Range Monitor (APRM) Simulated Thermal, 8 Power - Upscale Flow Biased Scram far.d M LG "tr seirrtrtu I
@ N*I WAllowable Values to those applicable for single recirculation
~
loop operation per Specifications 2.2.1 and 3.3.6. l f) Terform Surveille.nce Requirement 4.4.1.1.4 if THERMAL POWER is g ggn f;ake less than or equal to 30% of RATED THERMAL POWER or the g,4 ,, recirculation loop flow in the operating loop is less than or equal to 50% of rated loop flow.
- 2. cer e- x- " :t 1::n = ::=;;,m ..m m uie - n .._ .
3
- b. With no reactor coolant system recirculation loop in operation while in MMD OPERATIONAL CONDITION 1, immediately place the Reactor Mode Switch in the SHUTDOWN position.
- c. With no reactor coolant system recirculation loops in operation, while in OPERATIONAL CONDITION 2, initiate measures to place the unit in at least h CT10bj C, HOT SHUTDOWN within the next 6 hours. ,
{SeefecialTe7 Exception 3.f0.4.]
,I FERMI - UNIT 2- 3/4 4-1 Amendment No. JJ EA,JJ,EJ, E7,JEJ,
- 122 PAGE 2- 0F 06 g
SPE C Ipac A T1o a 3.q,i o 3 /4. A REACTOR COOLANT SY' STEM h 50 b M b 8.4*I 3/4.4.1 RECIRCULATION SYSTEM RECTRCULATION LOOPS ' LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.
' APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2*.
ACTION:
- a. With one reactor coolant system recirculation loop not in operation:
- 1. Within 4 hours:
a) Place the individual recirculation pump flow controller' for the operating recirculation pump in the Manual mede. g b) Reduce THERMAL POWER to less than or equal to 67.2% of RATED
.g. THERMAL POWER.
3 4.l c) Limit the speed of the operating recirculatien pump to less than or equal to 75% of rated pump speed. d) Increase the MINIMUM CRITICA!. POWER RATIO (HCPR) Safety Limit to the value for single loop operation required by Specification 2.1.2. e) Change the Average. Power Range Monitor (APRM) Simulated Thermal ' Power - Upscale Flow Biased Scram and Rod Block Trip Setpoints ! and Allowable Values to those applicable for single recirculation loop operation per Specifications 2.2.1 and 3.3.6. l f) Perform Surveille.nce Requirement 4.4.1.1.4 if THERMAL POWER is S WO,( . less than or equal to 30% of RATED THERMAL POWER or the recirculation loop flow in the operating loop is less than or f/23AJ9 4 equal to 50% of rated loop flow. r
- 2. Otherwis[beinatleastHOTSHUTDOWNwithinthenext12 hours,
- b. With no reactor coelant system recirculation loop in operation while in M-YN" OPERATIONAL CONDITION 1, imediately place the Reactor Mode Switch in the SHUTDOWN position.
3,( .) c.f With no reactor coolant system recirculation loops in operation, while in OPERATIONAL CONDITION 2, initiate measures to place the unit in at least HOT SHUTDOWN within the next 6 hours. ('See Special Test Exception 3.10.4. l e f\
/ FERMI - UNIT 2 3/4 4-1 Amendment No. JJ,H ,JJ,JJ, 57,JJJ, 122
/ 08 PAGE OF g 2.
SPECIFicAnord pt,io i l RFACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STARTUp *i LIMITING CONDITION Fort OPERATION 1.4.1.4- An idle recirculation loop shall not be started unless the temperature l SR3,lo,(,3 differential between the reactor pressure vessel steam space coolant and the- t bottom head drain line coolant is less than or equal to 145'F, and: (When both loops have been idlh unless the temperature differential SE 3.4,to/f a. . tween tne reactor cooiant within the idle loop to be started up and the coolant in the reactor ressure vessel is less than or equal to 50*F, or 5, b. hen only one loop has been idh unless the temperature
- f ciiierenttai netween tne reactor coolant within the idle and g san rm m g r.rirculan an ia- y is less than or equal to 50*F l
l 543,V.to,3 Mok S U .+.to.'l "O k-APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4 ACTION: With temperature differences exceeding the above limits, suspend startup of an l idle recirculatio - SURVFittANCE REOUTREMENTS 5/3A.l0,3
-4.t.;.? - The temperature differentials shall be determined to be within the '
SR3.4 to.411mits within 15 minutes prior to startup of an idle recirculation loop. e. FERMI - UNIT 2 3/4 4 6 Amendment No.JJ,128 PAGE 8 0F 08 92.
... ..-.,..--~ _ .- ... - n - -- - -- - - - - _ . . . - . . . - - - --
i O oeterza I I ,* I i . I i i I l l 1 ) l I I : l l. I I I i l 1 1 1' l l l n GURE 3.4.1.4 1 - DELETED I l i I l-O l 1 I I- ;
.)
l: I l l
- 8 l I
< ,. g .. I-l l l l l l l l i l l I i I l ' l l L I , , I I J l
,~ .
O . FERMI - UNIT 2 3/4 4-6a Amendment No. JJ 57.128 PAGE 19af0F 57 hv2-
CPEC8AcArim) 3N / REACTOR COOLANT SYSTEM 3/4.4.10 CORE THERMAL HYDRAULIC STABILITY '! tIMITING CONDITION FOR OPERATION Lco 34.I
-3.0.;0 The Reactor core shall not exhibit core thermal-hydraulic instability 8
_nr be operated in the Scram or Exit Regions '= secu :2d ri M- I [0. : . w m -. l APPLICABILITY: OPERATIONAL CONDITION 1 cLnd '/_ LA.3 A ACTION: 0 fa
- a. With the React __or operating in the Scram Regionj : :=rif t:d ir) '
ggQ y t;r: 3.LlD=LT immeoiately place the Reactor Mode Switch in the Shutdown position. l
- b. With the Reactor operating in the Exit RegionJ;: ::::i'i:d = "i m '
3.f.10 .'.Timmediately Initiate action to leave the Exit Region by I Ac.170310 inserting control rods or by increasing core flow.* l
- c. If core thermal hydraulic instability occur das av oenceo u,y . '
sta ec increas in arnn or t peak-to-peak ise level re ing I 2o re times s initial 1 1, and occurri wi t h a- --- f I [CT7M A . ct ri tic -i e ' e f 6 than 3 secondV'immediately place th? I Reactor Mode Switch in the Shutdown position. l SURVElltANCE REQUIREMENTS p.iv.4 in. grv isions of 5pm ;r;m.ti;a :.0.: :na;t;;pinb:. g gpyg d.0.10.2 When operating _within the Stability Awareness Recio m : :::: " : -i
)l' rigr: ;.0. W u verify that the reactor core is not exhibiting core I Jnermal hydraulic instabilityLby -iter n; - =d dix : ;-sFM I 1
Q dirt:!y : g at least once every nour. [60DPlb
- Restarting an Idle Recirculation Loop or resetting a Recirculation '
NON Flow limiter are not acceptable methods of immediately increasing core I flow to leave the Exit Region. ! O, FERMI - UNIT 2 3/4 4 30 Amendment No. JJ.128 PAGE 5~ OF 06 gu 1
O O o
~ -
85 N * -80 - E 80 Stability Reg i on De ptions p / 7 N' 70E
"~
m 70-65- E FA/W" ele. Elm
' 85 60
>96%RodUne,<40% CoreFlow em Regbn
>67% Rod , <40% Core Flow 55- , 55 in ,
>77% Rod Line, 45% Core Flow O P
s
s Q- 50- - 005 lMlk \ ' 50
>103% Rod Uno, < Core Flow Staldq Awareness Region:
45- 45 Notin Scram or Enit Regions
~
O E - mus - n F 40- >62% Rod Une. <45% Core Flow
>72% RW Une, <50% Core Flow
- 35 ' ~ 35
' >98% Rod Une, <55% Core Flow v
25 '\ , , ,, , ,,, , , ,, y , ,, ,,,, 25 30 35 40 45 50 55 60 f Core Flow (% Rated)
= ?I D i w
THu mal. POWER VE 5 CORE FLOW g ( FIGURE 3.4.10-tx , 2 b s
- _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __________.________ y, P
- f
l l tipscIFtCkTroM 3.8.I 3/4.8 ELECTRICAL POWER SYSTEMS [ Al50 5tt h NC# k*1 #"I'3 ) 3/4.8.1 A.C. SOURCES A.C. SOURCES - OPERATING f,f LIMITING CONDITION FOR OPERATION Ltc 3.8 As a minimum, the following A.C. electrical power sources shall be OPE :
- a. Two #:;/:.cLD 15f:p;nd:nt-circuits between the offsite l transmission network and the onsite Class IE distribution system, -
and LA.I
- b. Twolse ateand/hdepende onsite A.f4 electr/ cal powerjsoun_5y, Div sion I and Elvision ,[each consisting of two emergency diesel generatorsleac diesel genentor wiw ,
SR3,8.lM 1. A separate day fuel tank containing a minimum of 210 gallons of fuel,
,gy 2. A separate fuel storage system containing a minimum of 34,3 35,280 gallons of fuel, and A sep; ret; f;:1 tr:n:f:r p r;. _ ((
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ACTION:
- a. With one or both offsite circuits of the above required A.C.
4c.w C electrical power sources inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours; . demonst te the OPERABI Y of the remain A.t.. sources oy , perfo ing Surveillan mat ast once per 8 oursRequirement thereafter 4.8 ds
.l.1.withinone/fourand, ' -
UN A kb. With one or both diesel generators in one of the above required onsite A.C. electrical power divisions inoperable; Demonstrate the OPERABILITY of the remaining A.C. sources by P /}el- A ,) 1. '_ performing Surveillance Requirement 4.8.1.1.1 within one 1, hour and at least once per 8 hours thereafterfana if he l (die generator (s ecame inoperao oue to any ca eoth]er, th an inoperab support system, independent 1 testable' mponent, or o lanned preventi maintenance testina.; I performing Surveillance Requirement 4.8.1.1.2.a.4 for one Re9 Ach4,3.2.4 diesel generator at a time within 24 hours, unless the R, Ac-l- absence of any potential common mode failure for the f 4.2. 31-f remaining diesel generators is determined, and i i t .- FERMI UNIT 2 3/4 8-1 Amendment No. JJ,119 PAGE C 0F 08 M
I
. I O . .3/4.8 ELECTRICAL POWER SYSTEMS S PEclFt CA17orJ 19 3 3/4.8.1 A.C. SOURCES [4l50 5et SpecIhsakm 8.8./ )
A.C. SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be l PERA8LE:
- Two physically independent circuits between the offsite transmission network and the onsite Class IE distribution system,
_ sa.c. and
'kcekm by y*y*g Two separate and independent onsite A.C. electrical power sources, Division I and Division II, each consisting of two emergency diesel' generators, each diesel generator wjth:
- 1. A separate day fuel tank containing a minimum of 210 gallons C of fue,j,, '
gR 3.g,3,l 2. A sersarate fuel stora e system containing a minimum of 35,780 gallons of fue , and (*l hab uo 3.8,3
^
- 3. Aseparaterueltransrerpumj
/)pplicabllly AtjiGnS, d -
PPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3.
.5A 3,s.3 2_ l CTION:
. With one or both offsite circuits of the above required A.C. i electrical- power sources inoperable, be in at least HOT SHUTDOWN l
within 12 hours and in COLD SHUTDOWN within the next 24 hours; ' demonstrate the OPERABILITY of the remaining A.C. sources by ; pr.rforming Surveillance Requirement 4.8.1.1.1. within one hour and at least once per 8 hours thereafter and, g . With one or both diesel generators in one of the above required g onsi_te A.C electrical power divisions inoperable; J,$.l 1. Demonstrate the OPERABILITY of the' remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1 within one hour and at least once per 8 hours thereafter, and if the l diesel generator (s) became inoperable due to_any cause other than an inoperable support system, an independently testable { j component,'or preplanned preventive maintenance or testing,
' by performing Surveillance Requirement 4.8.1.1.2.a.4 for one diesel generator at a time within 24 hours, unless the i absence of any potential common mode failure for the !
L remaining diesel generators is determined, and lO l
'/ERMI UNIT 2 3/4 8-1 Amendment No JJ,119 PAGE / OF 05 gL ;
i i i S'pEcIFlchT10 5S'I [ ELECTRICAL POWER SYSTEMS l LIMITING CONDITION FOR OPERATION (Continued) ACTION (Continued) LR,d l
- 2. !
gd fy r!!"- S hnr; :rd :t 1;.;t m,7 n.
.hereaf te. , thet CTC 11 1 H OPE"a?LE O hu s '
Restore the g)Mcy-4,5' inoperable division to OPERABLE status within 7 days or be , in at least HOT SHUTDOWN within the next 12 hours and in I ' g c.770td C COLD SHUTDOWN within the following 24 hours. 4,/l I
- 3. fif e requirements f ACTION b.2. ab/ve for CTG ll-ljnnot be et, either re ore the inocerabM division to OPF)mBLEj}
4 atus within 7 ourf (not to exce'ed 7 days from the time = the division became inoperable); or, satisfy the ' g ggdl requirements of ACTION b.2 above withjn 72 hours and res g%gp 4,y.w' the inoperable division to OPERABLE status within 7 days from the time the division became inoperable; or, be in at ACT10tl O'lIleast HOT SHUTOOWN l SHUTDOWN within within the following 24' thers. next 12 hours and in COLD : q ._ ___ _ ._ !
- c. With one or both diesel generators i one of the above required 1 onsite A.C. electrical power divi i n inoperable, in addition to gag bt A,2. ACTION b, above, verify within hours that all required systems, 0
subsystems, trains, components and devices that depend on the remaining onsite A.C. electrical power division as a source of I1 l emergency power are also OPERABLE: ttherwise),e in at seast T UTD next u noups and in C015 SHUTDOWN with within t the oil ing 24 hou
.r g g __ _ __
d With both of the above required onsite A.C. electrical power I 4cT1obJ 6 divisions inoperable; I
- 1. Demonstrate the OPERABILITY of the remaining A.C. sources by
/2,. / performing Surveillance Requirement 4.8.1.1.1 within one 'l )cf- /} I hour and at least once per 8 hours thereafter; and i l
- 2. Restore at least one of the above required inoperable g g,( divisions to OPERABLE status within 2 hours or be in at
! 4[ b o?J C least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours; and
- 3. Restore the second of the above required divisions to g g}-4,g" OPERABLE status within the time required by Action b above from the time of initial loss or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the
/k,not)C- following 24 hours.
.j i
4 FERMI - UNIT 2 3/4 8-2 Amendment No.ll
t
.51%cIFICA 770A) 8.[,t/
ELECTpfCAL POWER SYSTEMS (Aiso su spe;%% r.g.6.) i SURVE1LLANCE REOUIREMENTS (Continued) - m
- 2. There is no visible corrosion at either terminals or SA M'N'1 connectors, or the connection resistance of these items is less than 150 x 10 6 ohm, and 5% p. The average electrolyte temperature of ten of the connected Sp c& dim 3.54g\ cells is above 60*F.
- c. At least once per 18 months by verifying that:
gg 3,y,d,3 1. The cells, cell plates and battery ratks show no visual l indication of physical damage or abnormal deterioration, g g .5.9 8l.g 2. The cell-to-cell and terminal connections are e'--- ng free of corrosion and coated with anticorrosion material, 3. A 3'#'Y' The resistance of each cell-to-cell and terminal connection is less than or equal to 150 x 10 ' ohm, and
- 4. The battery charger will supply at least 100 amperes at a M 3' N* b minimum of 129 volts for Division I and at a minimum of 8 124.7 volts for Division II for at least 4 hours. l
- d. At least once per 18 months by verifying that either-5t,ma laled
- 1. The battery capacity is adequate o suppiy and~ maintain in M 1. FN.'7 OPERABLE status all of the actual emergency loads for the u
design a battery duty cycle test, service f 'r n or
)] when
' the battery is subjected K O .
ine o tery capacity is quequate su suppiy a a y loaa of \ the 11owing profile w le maintaining the b tery terminali vol age greater than o equal to 105 or 210 its, as ap icable: ,
) Batteries 2P and 2PB greater than r equal to 710 !
amperes dur' g the initial 6 seco s of the test.
~
b) Batteries A and 2PB greater th 182 amptres dur' g the next 2 seconds of the test c) Batteri 2PA and 2P8 greater an or equal to ampere during the next 4 hou s of the test. d) Batte ies 2PA and 2PB greate than or equal t 480 amne er durina the last 6 ennde af + ka +ae sR N *TE ~ ~ ~ ~ ~ ~ - s 'l-
- e. ,r,- tA least once per 60 months (durino snutcowriby verifying that the g 3* gg* g i battery capacity is at least 80% of the manufacturer's rating when t
subjected to a performance discharge test. [At thTs once per 60-g3,1.4~7 g*g month .in lieu intervas, of-theInts performance battery service test. cisenarge test may be performed f. At least once per 18 months performance discharge tests of battery
'g '9,q, g capacity shall be given to any battery that shows signs of degradation or has_ reached 85% of the service life ernetted for the anolication. /Degrad ton is indicated when he battery )
capac' prev' usyperformance drops more tha 10% of rated capacit from its averag on i sts, or is below 90% grat' g. j the manufactur 's J FERMI - UNIT 2 3/4 8-13. Amendment No. 95,121 PAGE A 0F 02 &2~
Spec.LFleTtord 6 6 ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Co inued)
- f. cEtrTo 4, /
g,g g" Primary Containment Leakage Rate Testing Program f.5, Y2. 4 A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option 8 as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based containment Leak-Test Program," dated September 1995. 5.017-.b The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 56.5 psig. 3,s.n.,c. The maximum allowable primary containment leakage rate, L., shall be 0.5% of primary containment air weight per day at P.. 5.L G c The provisions of specification 4.0.2 do not apply to the test frequencies Testing specified in the Primary Containment Leakage Rate Program. 6 5JZ. f The provisions of Specification 4.0.3 are applicable to the Primary Containment Leakage Rate Testing Program. 56A
-h-Confiouration Risk Manacement Procram The Configuration Risk Management Program (CRMP) provides a proceduralized risk informed assessment to manage the risk ]
C " fg y associated with equipment inoperability. The program applies to I D / which'vahas been grantedtechnical a risk-informed %'l=:d =t= specification structures, systems, or Yb$ Specification 3.8.1.1.b.2@. ollowing **** Pro: [pr . , program The shall include the L EEN11 Provisions for control and implementation of a level 1, at ' I
. power, m-wh2" internal be capable events PRA-informed of evaluating the a Y.
methodology'p;plicable plant configuration. %.[),((.
- g. f. 4. g.2. 2 Provisions for performina an assessment prior to entering I
the ' N""" " m for preplanned activitie u Wapelstalo tt r;s k -in(maA GospleK ATs' tat E' b *
- N -4r- {EC ACTIC" STATC".O'T3 for unplanned ent" intaProvisionstfor oerrnian~
STATD C I. + h # LC0 ".CT PA' 6,6,/Y,b,N Provisions for assessing _ __ D I additional actions } after the discovery of Jf"en2L equipment out of service I l conditions while in the [C0 SCH ON 5!^!tMu m R l l FERMI - UNIT 2 6-16b Amendment No. EJ, JpE, JJJ,119 PAGE 23 0F 24 M
1 TPECLFl d4 Tion) 6, 2-J ( Als o Su- Welfnahm Sar) \ ADMINISTRATIVE CONTROLS ($l50 M f dd/ f N Ca b 6 6) PROCEDURES AND PROGRAMS (Continued) ! M 5. Provisions for considering oth6r applicable risk significant ' Y b" contributions such as Level 2 PRA issues and external I 66 events, qualitatively, or quantitatively. ! ! k b i dministrative controls shall be developed and implemented to limit the working hours of personnel who perform safety-related functions (e.g., senior reactor and operators, reactor key maintenance personnel). operators, auxiliary The controls shalloperators, hc.ith ebc:it;Y include guidelines on $ working hours that ensure that adequate shift coverage is maintained without _ routine heavy use of ove individuals. /.radi& H en profe<r> f . %. J u kni6 m $ Any deviation from/$ e w
~
our guidelines shall be authorized in aava'nce' by the@fa'fn jnfgerjor his designee, in accordance with approved ' administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that-individual overtime shall be reviewed monthly by the Plant Manager or his designee to ensure that excessive hours have not been assigned. Routine deviation from the above guidelines shall not be authorized, n '6.9 REPORTING RE0VIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted. ' STARTUP REPORT
/ 6. 9.1.1 A summary report of plant startup ar.d power escalation testing shall V /,1 be submitted following (1) receipt of an Operating License, (2) amendment to ,
the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel l supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. 6.9.1.2 The startup report shall address each of the tests identified in < Subsection 14.1.4.8 of the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions hatwererequiredto p) Sec Spsikk GG \_ FERMI - UNIT 2 6-16c Amendment No. JS/U,119 PAGE 5 0F 05 M
i l l I O SPEcnvuemu3 S.5 d j ADMINISTRATIVE CONTROLS m sawL% SA) Am> 54 Spec} 6 cagsw O(,, PROCEDURES AND PROGRAMS (Continued) g,5,lY.h.ffdh Provisions for considering othsr applicable risk significant contributions such as Level 2 PRA issues and external l events, qualitatively, or quantitatively. f.8.6 6 Administrative controls shall be developed and implemented to limit the working hours of personnel who perform safety-related functions (e.g., senior reactor operators, reactor operators, auxiliary operators, health physicists, ) and key maintenance personnel). The controls shall include guidelines on ! 1 [g,t k pt working hours that ensure that adequate shift coverage is maintained without routine heavy use of overtime for individuals. l
\6't . Any deviation from the working hour guidelines shall be authorized in advance by the Plant Manager or his designee, in accordance with approved administrative procedures, or by higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or his designee to ensure that excessive hours have not been assigned. Routine l heviation from the above guidelines shall not be authorized.
p I 6.9 REPORTING RE0VIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise no,ted, i h g STARTUP REPORT 6.9.1.1
, h Iy be submitted following (1) receipt of an Operating License, (2) amendment toA s the license involving a planned increase in power level, (3) installation of N fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit.
6.9.1.2 The startup report shall address each of the tests identified in Subsection 14.1.4.8 of the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions were required to O FERMI - UNIT 2 6 16c Amendment No. JEE/U,119 PAGE 234 0F 24 &1
. _ ~ . _ - . - - - . - - - . - - - . - . - - . - - . - _ - - . . . - -
1 l h 6 Ci PIC kTioAl S o G (( % o sart S1ect k cafi4h sos-)M ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) s e^- SWM% 5. Provisions for considering other applicable risk significant 8 g,f contributions such as Level 2 PRA issues and external I events, qualitatively, or quantitatively. l 6.8.6 Administrative controls shall be developed and implemented to limit the Gd wsrking hours of personnel who perform safety-related functions (e.g., senior 6 andreactor operators, reactor operators, auxiliary operators, health physicists, key maintenance personnel). The controls shall include guidelines on gf routine heavy use of overtime for individuals. working hours that ensure that adequate shift coverage is maintained without Any deviation from the working hour guidelines shall be authorized in advance ! N l by the Plant Manager or his designee, in accordance with approved d administrative procedures, or by higher levels of management, in accordance g with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that-d individual overtime shall be reviewed monthly by the Plant Manager or his' l W deviationfromtheaboveguidelinesshallnotbeauthorized. designee to ensure that excessivl 4-+ REPORTING REOUIREMENTS ROUTINE REPORTS
' 6*
0.q. ( In addition to the applicable reporting requirements of Title 10, Code ! cf Federal Regulations, the following reports shall be submitted to the R2gional Administrator of the Regional Office of the NRC unless otherwise g'{ noted. , ISTARTUP RE#bRT b 6.9.1.1 A summary report of ant startup and power scalation testing shal be su i tied following (1) r ceipt of an Operating cense, (2) amendment t the 1 ense involving a pl ned increase in power evel, (3) installation f fuel hat has a differen design'or has been man actured by a different uel su lier, and 4 modif ations that may have clear, therma (l),or draulic performance of ;he gnificantly altered the unit.
.9.1.2 The start report shall address ch of the tests identif d in Subsection 14.1.4 of the Final Safety alysis Report and shall nelude a d2scription of e measured values of t operating conditions or characteristic obtained during the te program and a compariso of these values with de ign predictions and spe ifications. Any correct e actions
- 4. hat were reg tred to - -
O l L FERMI - UNIT 2 6-16c Amendment No. JJJ/U,119 PAGE 1 0F- 05
() 0 2.0 SAFETY LIMITS (SLs) l 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow: l l THERMAL POWER shall be s 25% RTP. ! I 2.1.1.2 With the reactor steam dome pressure a 785 psig and core flow = 10% rated core flow- ' l MCPR shall be = 1.11 for two recirculation loop operation l or a 1.13 for single recirculation loop operation. 2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel. 2.1.2 Reactor Coolant System Pressure SL ey (
\ /
, Reactor steam dome pressure shall be s 1325 psig.
I 2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hours: 2.2.1 Restore compliance with all SLs: and 2.2.2 Insert all insertable control rods.
,/- 'wl l . FERMI - UNIT 2 2.0 1 Revision 2 01/18/99 l
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 3.3.1.11 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 O channels inoperable. trip.
.Q8 A.2 --- --
NOTE-- ----- Not applicable for Functions 2.a. 2.b. 2.c. and 2.d. Place associated trip 12 hours system in trip. (continued) O
] FERMI UNIT 2' 3.3 1 Revision 2 01/18/99 p
l
l I _ RPS Instrumentation 3.3.1.1 l ACTIONS' (continued) , CONDITION REQUIRED ACTION COMPLETION TIME !
----NOTE -------- B.1 Place channel in one 6 hours
~
B. . - i Not applicable for trip. system in trip. i Funct-lons 2..a. 2.b. ; 2.c, and 2.d. IE One or more Functions in trip. 'j with one or more l required channels i inoperable in both ,
. trip systems. i I
i l l A U O
'l FERMI UNIT 2 3.3-1(1) Revision 2. 01/18/99
i ; l RPS Instrumentation l 3.3.1.1 [O - i SURVEILLANCE REQUIREMENTS 1 l ...................................- NOTES ------ - ------ ----------- ----- i 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.
SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours SR 3.3.1.1.2 Perform CHANNEL CHECK. 24 hours l SR 3.3.1.1.3 - --- - ----
--NOTE --- -- --- -
Not required to be performed until 12 hours after THERMAL POWER = 25% RTP. Verify the absolute difference between 7 days the average power range monitor (APRM) channels and the calculated power is j s 2t RTP while operating at = 25t RTP. l (continued) O l FERMI - UNIT 2 3.3 4 Revision 2 01/18/99
r~' I l RPS Instrumentation I ( ,)g 3.3.1.1 , t i SURVEILLANCE REQUIREMENTS (continued) 1 i SURVEILLANCE FREQUENCY ! r SR 3.3.1.1.10 Verify the trip unit setpoint. 92 days l s 1 SR 3.3.1.1.11 ----- ------ -----NOTES--- - - -- --- --
- 1. Neutron detectors are excluded.
l 2. For Function 1.a not required to be performed when entering MODE 2 from ; MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL CALIBRATION. 184 days SR 3.3.1.1.12 --- - --
--------NOTE--- ---- -------- -
(~ ) For Function 2.a. not required to be , (m> performed when entering MODE 2 from MODE 1 until 12 hours after entering MODE 2. Perform CHANNEL FUNCTIONAL TEST. 184 days , l SR 3.3.1.1.13 Perform CHANNEL FUNCTIONAL TEST. 18 months l SR 3.3.1.1.14 Perform CHANNEL CALIBRATION. 18 months I SR 3.3.1.1.1S Perform LOGIC SYSTEM FUNCTIONAL TEST. 18 months > (continued) (f ( N l' FERMI UNIT 2 3.3 6 Revision 2 01/18/99
P i l RPS Instrumentation 7T 3.3.1.1 (LJ i SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.1.1.16 Verify Turbine Stop Valve-Closure and 18 months ! Turbine Control Valve Fast Closure i Functions are not bypassed when THERMAL ! POWER is a: 30% RTP. l SR 3.3.1.1^.17 ----
- ------ NOTES-- --- - -- -- --- l
- 1. Neutron detectors are excluded. !
- 2. For Functions 3 and 4 channel sensor response times'are not required to be measured.
- 3. For Function S "n" equals 4 channels for the purpose of determining the !
STAGGERED TEST BASIS Frequency. i 7-<) ( Verify the RPS RESPONSE TIME is within 18 months on a limits. STAGGERED TEST BASIS SR 3.3.1.1.18 - - --
- -- - -.. NOTE- -------- - -----
Neutron detectors are excluded. Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.19 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months t, l l
~l FERMI - UNIT 2 3.3 7 Revision 2 01/18/99
s RPS Instrumentation ( ) 3.3.1.1
%)
Table 3.3.1.1 1 (page 1 of 3) Reactor Protection System Instrunentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE ! FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 1. Intermediate Range Monitors
- a. Neutron Flux-High 2 3 G SR 3.3.1.1.1 s 122/125 SR 3.3.1.1.4 divisions of SR 3.3.1.1.6 full scale '
SR 3.3.1.1.7 SR 3.3.1.1.11 SR 3.3.1.1.15 5(a) 3 I SR 3.3.1.1.1 s 122/125 SR 3.3.1.1.5 divisions of SR 3.3.1.1.11 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 I SR 3.3.1.1.5 NA SR 3.3.2.2.15
- 2. Average Power Range Monitors (q
L/ 4 a. Neutron Flux - Upscale (Setdown) 2 3(C) G SR 3.3.1.1.2 SR 3.3.1.1.7 s 202 RTP SR 3.3.1.1.8 SR 3.3.1.1.12 SR 3.3.1.1.18
- b. Simulated Thermal 1 3(C) F SR 3.3.1.1.2 s 0.63 (W aW)
Power - Upscale SR 3.3.1.1.3 + 64.3t RTP SR 3.3.1.1.8 and s l SR 3.3.1.1.12 RTP(b)115.58 j SR 3.3.1.1.18 1 1 1 (continued) (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies. (b) 4W = 81 when reset for single loop operation per LCO 3.4.1. " Recirculation Loops Operating." Otherwise aW = 08. (c) Each APRM channel provides inputs to both trip systems. l l l l l l (~\ l L.) l FERMI - UNIT 2 3.3-8 Revision 2, 01/18/99 l l l
p 1 l
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RPS Instrumentation l !! 3.3.1.1 l i Table 3.3.1.1 1 (page 2 of 3) Reactor Protection System Instrtmentation j l l i APPLICABLE CONDITIONS . MODES OR REQUIRED REFERENCED j l OTHER CHANNELS FROM , SPECIFIED PER TRIP REQUIRED StRVEILLANCE ' ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREENTS VALUE
- 2. Average Power Range' !
Monitors (continued) ; c.' Neutron 1 3(C) F SR 3.3.1.1.2 s 1201 RTP Flux - Upscale SR 3.3.1.1.3 'j SR 3.3.1.1.8 i SR 3.3.1.1.12 ; SR-3.3.1.1.18-
- d. Inop 1.2 3(C) G' SR 3.3.1.1.12 NA
- e. 2-out of 4 Voter 1.2 2 G SR 3.3.1.1.2 NA SR 3.3.1.1.12 SR 3.3.1.1.17 SR 3.3.1.1.19 r
' 3. Reactor Vessel Steam 1.2 2 G SR 3.3.1.1.1 s 1113 psig i Dome Pressure-High SR 3.3.1.1.9 l SR 3.3.1.1.10 l SR 3.3.1.1.14 SR 3.3.1.1.15
- m. SR 3.3.1.1.17 4 Reactor Vessel Water 1.2 2 SR 3.3.1.1.1 e 171.9 inches V) . Level - Low Level 3 G
SR 3.3.1.1.9 SR'3.3.1.1.10 l.. SR 3.3.1.1.14 l SR 3.3.1.1.15 - : SR 3.3.1.1.17 l
- 5. Main Steam Isolation 1 8 F SR 3.3.1.1.9 s 121 closed f
. l- Valve - Closure SR 3.3.1.1.14 -
SR 3.3.1.1.15 SR 3.3.1.1.17 j
. 6. Main Steam Line ' 1,2 2 H SR 3.3.1.1.1 s 3.6 X full l Radiation - High SR 3.3.1.1.9 power l l' SR 3.3.1.1.14 SR 3.3.1.1.15 background i
.{
- 7. Drywell Pressure-High 1.2 2 G SR 3.3.1.1.1 s 1.88 psig ;
SR 3.3.1.1.9 SR 3.3.1.1.10
-l- SR 3.3.1.1.14 SR 3,3.1.1.15 (continued)
.(c) Each APRM channel provides inputs to both trip systems.
O
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-l: FERMI UNIT 2- 3.3 9 Revision 2, 01/18/99
4 i i es RPS Instrumentation (
%J
) 3.3.1.1 i i
Table 3.3.1.1-1 (page 3 of 3) ) Reactor Protection System Instrunentation i APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTKR CHANNELS FROM SPECIFIED PER TRIP REQUIRED SLRVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE
- 8. Scram Discharge Volune Water Level- High
- a. Level 1.2 2 G SR 3.3.1.1.1 s $96 ft.
Transmitter SR 3.3.1.1.9 0 inches SR 3.3.1.1.10 l SR 3.3.1.1.14 SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.1 s 596 ft. SR 3.3.1.1.9 0 inches SR 3.3.1.1.10 l SR 3.3.1.1.14 SR 3.3.1.1.15
- b. Float Switch 1,2 2 G SR 3.3.1.1.9 s 596 ft.
l SR 3.3.1.1.14 0 inches SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.9 s 5% ft. . l SR 3.3.1.1.14 0 inches SR 3.3.1.1.15
- 9. Turbine Stop e 302 RTP 4 E SR 3.3.1.1.9 s 72 closed '
[^\ l Valwe - Closure SR 3.3.1.1.14 i SR 3.3.1.1.15 k/ ) SR 3.3.1.1.16 SR 3.3.1.1.17
- 10. Turbine Control Valve = 303 RTP 2 E SR 3.3.1.1.9 NA i Fast Closure SR 3.3.1.1.15 SR 3.3.1.1.16 SR 3.3.1.1.17 l 11. Reactor Mode Switch- 1.2 2 G SR 3.3.1.1.13 NA Shutdown Position SR 3.3.1.1.15 ,
l 5(a) 2 I SR 3.3.1.1.13 NA SR 3.3.1.1.15
- 12. Manual Scram 1.2 2 G SR 3.3.1.1.5 NA SR 3.3.1.1.15 5(a) 2 I SR 3.3.1.1.5 NA SR 3.3.1.1.15 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.
1 l l l gy
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.l FERMI - UNIT 2 3.3 10 Revision 2, 01/18/99 l
l !
I l l gi Control Rod Block Instrumentation 3.3.2.1 i I SURVEILLANCE REQUIREMENTS l t
..................................... NOTES - ------- - --------- - ----
l
- 1. Refer to Table 3.3.2.1 1 to determine which SRs apply for each Control Rod l Block Function. i i
- 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. i SURVEILLANCE FREQUENCY ! l SR 3.3.2.1.1 - --
----- -NOTE - ----- - ------
}
Not required to be performed until I hour i after a"1y control rod is withdrawn at i s 10t RTP in MODE 2. l I p Perform CHANNEL FUNCTIONAL TEST. 92 days U i l 'SR '3.3.2.1.2 -- --- - --
----NOTE--- ---- ----- - - I Not required to be performed until I hour l after THERMAL POWER is s 10% RTP in ;
N0DE 1. l Perform CHANNEL FUNCTIONAL TEST. 92 days I, I l SR 3.3.2.1.3 Perform CHANNEL FUNCTIONAL TEST. 184 days l i (continued) i l l i l FERMI - UNIT 2 3.3-19 Revision 2 01/18/99 l: l
.~. _. __ _ _ _ _ _ . . ~ . . _ . . _ _ . . . _. _ _ . . - _ . . _ .__.___
I 1 Control Rod Block Instrumentation i--) V 3.3.2.1 ! SURVEILLANCE REQUIREMENTS (continued) , SURVEILLANCE FREQUENCY ! ____ i
'l SR 3.3.2.1.4' ------
- --- - NOTE- ---- ---- ---- - I Not required to be performed until I hour after reactor mode switch is in the shutdown position.
Perform CHANNEL FUNCTIONAL TEST. 18 months l l
-- i
-l SR 3.3.2.1.5 Verify the RBM is not bypassed when 24 months THERMAL POWER is = 30% RTP. !
i l SR 3.3.2.1.6 -- ------
- ---NOTE -- -- -- ---- - .
Neutron detectors are excluded. O b l Perform CHANNEL CALIBRATION. 24 months SR 3.3.2.1.7 Verify control rod sequences input to the Prior to RWM are in conformance with the declaring RWM prescribed withdrawal sequence. OPERABLE I following I loading of sequence into RWM i
/N .
(J ! l 1 I l FERMI UNIT 2. 3.3 20 Revision 2 01/18/99 ; l l
l 1 l Control Rod Block Instrumentation ) s 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1) Control Rod Block Instrumentation APPLICABLE MODES OR ( OTER SPECIFIED REQUIRED SlRVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREENTS VALUE
- 1. Rod Block Monitor
- a. Upscale (a) 2 SR 3.3.2.1.3 As specified in SR 3.3.2.1.5 the COLR SR 3.3.2.1.6
.l b. Inop (a) 2 SR 3.3.2.1.3 NA l c.. Downscale (a) 2 SR 3.3.2.1.3 a 92.3* of SR 3.3.2.1.6 reference level l
- 2. Rod Worth Minimizer 1(b) 2(b) . 1 SR 3.3.2.1.1 NA SR 3.3.2.1.2 SR 3.3.2.1.7 l 3. . Reactor Mode Switch- Shutdown (c) 2 SR 3.3.2.1.4 NA Position.
\I (a) TERMAL POWER = 30%.
.(b) With TIERMAL POWER s 104 RTP, (c) Reactor mode switch in the shutdown position.
l l l 1 L i..
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.. FERMI' UNIT.2 Revision 2. 01/18/99 !
m .d y .- g- w 9 --'7 y. 7 T'
l l l l Recirculation Loops Operating 7- 3.4.1 g 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating i LC0 3.4.1 a.1 Two recirculation loops with matched recirculation loop jet pump flows shall be in operation: 6HD a.2 The reactor core shall not exhibit core thermal-hydraulic instability or operate in the " Scram" or
" Exit" Regions.
08 l b.1 One recirculation loop may be in operation and the reactor core not operating in " Scram" or " Exit" regions as specified in administrative controls provided LC0 3.3.1.1 " Reactor Protection System (RPS) Instrumentation." Function 2.b (Average Power Range l Monitors Simulated Thermal Power-Upscale) Allowable , Value of Table 3.3.1.1-1 is reset for single loop !
,S operation, when in MODE 1. j
............................N0TE--- ------ -- - -----------
l Required allowable value modification for single loop i operation may be delayed for up to 4 hours after transition from two recirculation loop operations to single recirculation loop operation. APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation jet pump A.1 Declare recirculation 2 hours loop flow mismatch not loop with lower flow: within limits. "not in operation." (continued) O O l FERMI UNIT 2 3.4 1 Revision 2 01/18/99
Recirculation Loops Operating
--O 3.4.1 v
ACTIONS (continued)- CONDITION REQUIRED ACTION ~ COMPLETION TIME B. Reactor core operating --- - - NOTE-- - ----- -- in the." Exit" Region. Restart of an idle recirculation loop or resetting a recirculation-flow limiter is not allowed. B.1 Initiate action to Immediately insert control rods
,l' or increase core flow to restore operation outside the " Exit" l Region.
C. No recirculation loops C.1 Be in MODE 3. 6 hours operating while in
,O MODE 2.
D. No recirculation loo D.1 Place the reactor Immediately operating while in ps -mode switch in the
' MODE 1. shutdown position.
08
' Reactor core operating
- l: in the " Scram" Region.
08: Core thermal hydraulic l instability evidenced. N j. U
/l-l FERMI . - UNIT 2 - 3.4 2- Revision 2. 01/18/99
1
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i I Recirculation Loops Operating 3.4.1 1 SURVEILLANCE REQUIREMENTS SURVEILUWCE FREQUENCY ! Sr.:3.4.1.1 - - -- - ------ NOTE - ----- - - --
!Only required to be performed when l operating in the " Stability Awareness" l l Region, 1 Verify the reactor core is not exhibiting i hour '
core. thermal hydraulic instability. SR 3.4.1.2 -- ---- NOTE - --- - -- ---- Not re uired to be erformed until 24 hours after oth recircul tion loops are in operation. Verify recirculation loop , jet pump flow 24 hours I
- c. mismatch with both recirculation loops in '
-t operation is: \s)
- a. s 10% of rated core flow when operating at < 70% of rated core flow; and l b. s 5% of rated core flow when operating
! at = 70% of rated core flow. l l l I I i O-l FERMI - UNIT 2 3.4 3 Revision 2, 01/18/99
l l ! Battery Cell Parameters- ! l 3.8.6 j 1 SURVEILLANCE REQUIREMENTS (continued) ! l SURVEILLANCE FREQUENCY j l SR 3.8.6.2 Verify battery cell parameters meet 92 days l Table 3.8.6 1 Category B limits. - l~ E { ! Once within , 24 hours after
- battery :
discharge l
< 105 V ;
M Once within l 24 hours after i battery ; overcharge
> 150 V for
Division I and -
> 145 V for O Division II l SR 3,8.6.3 Verify average electrolyte temperature of 92 days representative cells is > 60*F.
l l O-l FERMI' UNIT 2 3.8 24 Revision 2, 01/18/99 - i l v
I l (~ q Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Primary Containment Leakaae Rate Testina Procram (continued)
- f. The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.
5.5.13 Hiah Density Soent Fuel Racks A program shall be provided which will assure that any ! unanticipated degradation of the high density spent fuel racks l will be detected and will not compromise the integrity of the j racks. l 5.5.14 Confiauration Risk Manaaement Procram The Configuration Risk Management Program (CRMP) provides a proceduralized risk informed assessment to manage the risk associated with equipment inoperability. The program applies to Technical 5)ecification structures. systems, or components for p which a risc-infonned Completion Time has been granted.
- a. Specifically. the risk-informed Completion Times include:
- 1. From a 72 hours after entering Condition A of LC0 3.8.1. "AC Sources 0perating."
- b. The program shall include the following:
- 1. Provisions for control and implementation of a level-
- 1. at power. internal events PRA informed methodology capable of evaluating the applicable plant configuration:
- 2. Provisions for performing an assessment prior to operating within the applicable risk-informed Completion Time for preplanned activities:
- 3. Provisions for performing an assessment after operating within the applicable risk-informed Completion Time for unplanned events:
,a l FERMI UNIT 2 5.0-19 Revision 2 01/18/99
Programs and Manuals O. 5.5
- . l 5.5 Programs and Manuals )
5.5.14 Confiauration Risk Manaoement Proaram (continued) ; 1
- 4. Provisions for assessing the need for additional ,
actions after the discovery of subsequent equipment- ; out of service conditions while operating within the 1 applicable risk informed Completion Time: and '
- 5. Provisions for considering other applicable risk i significant contributions, such as level-2 PRA issues i and external events - qualitatively or quantitatively. J l
l I i 1 O i- , l 1 i l l i I l( . (continued) l FERMI.- UNIT 2 5.0-19(1) Revision 2, 01/18/99
r RPS Instrumentation ( B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) ayeraoe Power Ranoe Monitor 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 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 2-out-of-4 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 (-) U trip inputs to either RPS trip system. A trip from any two unby)assed APRM channels will result in a full trip in each of tie four voter channels, which in turn results in two trip inputs into each RPS trip logic channel (A1, A2, Bl. 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 20 LPRM inputs, with at least three LPRM inputs from each of the four axial levels at which the LPRMs are located, are required for each APRM channel. 2.a. Averaoe Power Ranoe Monitor Neutron Flux-Voscale (Setdown) For operation at low power (i.e. MODE 2), the Average Power Range Monitor Neutron Flux-Upscale (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 l Average Power Range Monitor Neutron Flux-Upscale (Setdown) Function will provide a secondary scram to the Intermediate 4 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 l FERMI - UNIT 2 B 3.3.1.1 - 7 Revision 2 01/18/99 1
RPS Instrumentation Q,o B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) l Flux-Upscale (Setdown) Function will provide the primary trip signal for a corewide increase in power. The Average Power Range Monitor Neutron Flux-Upscale (Setdown) Function is credited, along with the IRM neutron flux high function, with initiating a reactor scram in the analysis of the continuous rod withdrawal during reactor startup event. This Function also 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-Upscale n (Setdown) Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for (V) criticality exists. In MODE 1, the Average Power Range l Monitor Neutron Flux-Upscale Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events.
2.b. Averaoe Power Ranoe Monitor Simulated Thermal l Power-UDscale The Average Power Range Monitor Simulated Thermal Power-Upscale 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 roportional to the THERMAL POWER in the reactor. The tri level is varied as a function of recirculation drive f ow (i.e., at lower drive flows, the setpoint is reduced proportional to the reduction in power experienced as drive 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-Upscale Function Allowable Value. The Average Power Range Monitor Simulated Thermal Power-Upscale 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 ' O( > l-FERMI-UNIT 2 B 3.3.1.1 - 8 Revision 2, 01/18/99
i RPS Instrumentation B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) 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 before the high neutron flux scram. For ra)id neutron flux increase events, the THERMAL POWER lags t1e neutron flux and the Average Power Range Monitor Neutron Flux-Upscale Function will provide a scram signal before the Average Power Range Monitor Simulated Thermal Power-Upscale Function setpoint is exceeded. The required trip setting for APRMs is dependent on whether the unit is in single recirculation loop operation or two-loop operation, as specified in Table 3.3.1.1-1 footnote (b). The setpoint variable AW is defined as the difference in indicated drive flow (in i of rated drive flow that produces rated core flow) between two loop and single loop operation at the same core flow. Each APRM channel uses one total drive flow signal representative of total core flow. The drive flow signal at (n rated drive flow is representative of rated core flow at RTP. 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 Allowable Value is based on analyses that take credit for the Average Power Range Monitor Simulated Thermal Power-Upscale Function for the mitigation of the loss of feedwater heating event. The THERMAL POWER time constant of approximately 6 seconds 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-Upscale 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/or APRM Functions provide l protection for fuel cladding integrity. 1_ i
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l FERMI - UNIT 2 B 3.3.1.1 - 9 Revision 2, 01/18/99 i'
3
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I 4
- n. RPS Instrumentation
() B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES LCO, and APPLICABILITY (continued) Pl. c. Averaoe Power Ranoe Monitor Neutron Flux-Voscale The Average Power Range Monitor Neutron Flux-Upscale 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 l Average Power Range Monitor Neutron Flux-Upscale Function is assumed to terminate the main steam isolation valve (MSIV) , closure event and, along with the safety relief valves i (SRVs). limits the peak reactor pressure vessel (RPV) , pressure to less than the ASME Code limits. The control rod drop accident (CRDA) analysis (Ref. 5) takes credit for the Average Power Range Monitor Neutron Flux-Upscale Function to I terminate the CRDA. l l The Allowable Value is based on the Analytical Limit assumed , in the CRDA analyses. l 1 The Average Power Range Monitor Neutron Flux-Upscale Function is required to be OPERABLE in MODE 1 where the (n v'
) potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) being 4
exceeded. Although the Average Power Range Monitor Neutron ! Flux-Upscale Function is assumed in the CRDA analysis, which l is applicable in MODE 2. the Average Power Range Monitor l Neutron Flux-Upscale (Setdown) Function conservatively bounds the assumed trip and. together with the assumed IRM trips, provides adequate protection. Therefore, the Average l Power Range Monitor Neutron Flux-Upscale Function is not required in MODE 2. (3l
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n RPS Instrumentation ;
- g B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) j 2.d. Averaoe Power Ranoe Monitor-Inoo l This Function provides assurance that a minimum nut er of APRMs are OPERABLE. For any APRM channel, any time: 1) its mode switch is in any position other than "0PER": 2) there is a loss of input power; 3) the automatic self test system i detects a critical fault with the APRM channel; or 4) the ,
firmware/ software watchdog timer has timed out, 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 q the APRM Functions are required. Le. 2-out of 4 Votgr l The 2 out of-4 Voter Function provides the interface between ! the APRM Functions and the final RPS trip system logic. As I such. it is required to be OPERABLE in the MODES where the l APRM Functions are required and is necessary to support the ! safety analysis applicable to each of those Functions. Therefore, the 2 out of 4 Voter Function is required to be OPERABLE in MODES 1 and 2. Both voter channels in each trip system (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 l FERMI - UNIT 2 B 3.3.1.1 - 11 Revision 2, 01/18/99 l
RPS Instrumentation ( B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES. LCO. and APPLICABILITY (continued)
- 3. Reactor Vessel 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 results in a pressure increase, counteracting the pressure increase by rapidly reducing core power. The overpressurization protection analysis of Reference 4 conservatively assumes scram on the Average l Power Range Monitor Neutron Flux-Upscale signal, not the Reactor Vessel Steam Dome Pressure-High signal. Along with the SRVs, the reactor scram 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 areclude a scram from this Function on a valid signal. T1e Function is required to be OPEP.ABLE in MODES 1 and 2 when the RCS is pressurized and the potential for pressure increase exists. 7
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n RPS Instrumentation ! ! B 3.3.1.1 V BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)
- 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 l transient. However, for the overpressurization protection analysis of Reference 4. the Average Power Range Monitor Neutron Flux-Upscale Function, along with the SRVs, limits the peak RPV pressure 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. 1 The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of the ECCS. ensures e that the fuel peak cladding temperature remains below the 1 limits of 10 CFR 50.46. j MSIV closure signals are initiated from 30sition switches l located on each of the eight MSIVs. Eac1 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 channel 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. 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 l'e 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 high, a pressurization transient can occur if the MSIVs close. In MODE 2, the MSIV closure trip is automatically G Q,1 l FERMI - UNIT 2 B 3.3.1.1 - 14 Revision 2 01/18/99
p RPS Instrumentation Q B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES. LCO, and APPLICABILITY (continued) The Turbine Stop Valve-Closure Allowable Value is selected to be high en) ugh to detect imminent TSV closure, thereby reducing the severity of the subsequent pressure transient. 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 preclude a scram from this Function if any three TSVs should close. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is = 30% RTP. This Function is not required when THERMAL POWER is < 30t RTP since the Reactor Vessel Steam Dome Pressure-High and the l Average Power Range Monitor Neutron Flux-Upscale Functions are adequate to maintain the necessary safety margins.
- 10. Turbine Control Valve Fast Closure 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 0 is initiated on TCV fast closure in anticipation of the transients that would result from the closure of these valves. The Turbine Centrol Valve Fast Closure Function is the primary scram signal for the generator load rejection event analyzed in Reference 7. For this event, the reactor scram reduces the amount of energy required to be absorbed and ensures that the MCPR SL is not exceeded.
Turbine Control Valve Fast Closure signals are initiated by the de energization of the solenoid dump valve at each control valve. Redundant relay signals are provided to each RPS logic channel such that fast closure of one control valve in each RPS trip system will initiate a scram. This Function must be enabled at THERMAL POWER a 30t RTP. This is normally accomplished automatically by pressure transmitters sensing turbine first stage pressure of I a 161.9 psig: therefore, to consider this Function OPERABLE. ' the turbine bypass valves must remain shut at THERMAL POWER a 30% RTP. l There is no Allowable Value for the Turbine Control Valve Fast Closure Function since the channels are actuated solely i on energization of the solenoid dump valve. ' l i O G l FERMI - UNIT 2 B 3.3.1.1 - 18 Revision 2, 01/18/99
RPS Instrumentation i B 3.3.1.1 BASES APPLICABLE SAFETY ANALYSES, LCO. and APPLICABILITY (continued) Four channels of Turbine Control Valve Fast' Closure Function
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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 < 30t RTP, since the Reactor Vessel Steam Dome Pressure-High and the Average Power Range Monitor Neutron Flux-Upscale Functions are adequate to maintain the necessary safety margins.
- 11. 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 O 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 arranged in a one-out-of-two logic, are available and required to be OPERABLE. The Reactor Mode Switch-Shutdown Position Function is required to be OPERABLE in MODES 1 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. O j FERMI - UNIT 2 B 3.3.1.1 - 19 Revision 2, 01/18/99
q RPS Instrumentation B 3.3.1.1 BASES ACTIONS (continued) 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 l be acce) table (Refs. 9 and 13) to permit restoration of any inopera)1e 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, 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, i Required Action A.1 must be satisfied. This is the only j 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. B.1 and B.2 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 accominodate single Q V l FERMI - UNIT 2 B 3.3.1.1 -21 Revision 2, 01/18/99
~_ . __ -- _ ._ .- - - - - - - .
p RPS Instrumentation , Q B 3.3.1.1 j BASES ACTIONS (continued) failure in both trip systems (e.g., each trip system remains ! in a one out of one arrangement for a ty)ical four channel Function). The reduced reliability of t11s logic l arrangement was not evaluated in References 9 and 13 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 I reliability level equivalent to that evaluated in l References 9 and 13, which justified a 12 hour allowable 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 channels are all in different Functions). The p 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).
If this action would result in a scram, it is permissible to place the other trip system or its inoperable channels in trip. The 6 hour Com)letion Time is judged acceptable based on the , remaining capa)ility 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 system in trip would result in a scram), 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 2 out-of-4 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 O l FERMI - UNIT 2 B 3.3.1.1 - 22 Revision 2, 01/18/99
l n () RPS Instrumentation B 3.3.1.1 l BASES ACTIONS (continued) OPERABILITY) that will restore caoability to accommodate a single failure. Inoperability of a Function in more than l 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 not associated with specific trip systems as i are the APRM 2-out of-4 voter and other non APRM channels. - Condition B does not apply. El Required Action C.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped i channels within the same trip system for the same Function ; result in the Function not maintaining RPS trip capability. l A Function is considered to be maintaining RPS trip capability when sufficient channels are OPERABLE or in trip n (or the associated trip system is in trip), such that both : (j trip systems will generate a trip signal from the given Function on a valid signal. For the typical Function with one-out-of two taken twice logic and the IRM and APRM > Functions, this would require both trip systems to have one 1 channel OPERABLE or in trip (or the associated trip system ) in trip). For Function 5 (Main Steam Isolation > Valve-Closure), this would require both tri) systems to have each channel associated with the MSIVs in t1ree main steam lines (not necessarily the same main steam lines for both . trip systems)-0PERABLE or in trip (or the associated trip system in trip). l For Function 8 (Turbine Stop Valve-Closure), this would require both trip systems to have three channels, each OPERABLE or in trip (or the associated trip system in trip). The Completion Time is intended to allow the operator time to evaluate, and repair or place in trip any discovered I inoperabilities that result in a loss of RPS trip i operability. The 1 hour Completion Time is acceptable I because it minimizes risk while allowing time for i restoration or tripping of channels. l h
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. RPS Instrumentation ( B 3.3.1.1 BASES ACTIONS (continued) [L1 l Required Action D.1 directs entry into the appropriate : Condition referenced in Table 3.3.1.11. 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 of Condition A, B, or C and the associated Com)letion Time I has expired, Condition D will be entered for t1at channel : and provides for transfer to the appropriate subsequent Condition. E.1. F.1. G.I. H.1. and H.2 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 p) s*
) laced in a MODE or other specified condition in which the LCO does not apply. Alternately, for Condition H. the )
associated MSLs may be isolated (Required Action H.1), and, l if allowed (i.e., plant safety analysis allows operation with an MSL isolated), operation with that MSL isolated may continue. Isolating the affected MSL accomplishes the safety function of the inoperable channel. 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 1 plant systems. In eddition, the Comp'letion Time of Required Action E.1 is consistent with the Completion Time provided in LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)." l i l FERMI UNIT 2 B 3.3.1.1 - 24 Revision 2. 01/18/99 l l
RPS Instrumentation B 3.3.1.1 ( BASES-ACTIONS (continued) 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 placed in a MODE or other specified condition in which the _C0 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. I l 1 l SURVEILLANCE As noted at the beginning of the SRs. the Srs for each RPS REQUIREMENTS instrumentation Function are located in the SRs column of ) s Table 3.3.1.1-1. l 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. For the case of the APRM Functions 2.a. 2.b. 2.c. and 2.d. RPS trip capability is maintained with any two OPERABLE APtMs remaining. Upon completion of the l Surveillance, or expiration of the 6 hour allowance. the l channel must be returned to OPERABLE status or the a)plicable Condition entered and Required Actions taken. T11s Note is based on the reliability analysis (Ref. 9) assumption of the average time required to )erform channel Surveillance. That analysis demonstrated tlat the 6 hour testing allowance does not significantly reduce the probability that the RPS will trip when necessary. 1
'l SR 3,.3 1.1.1 and SR 3.3.1.1.2 l Performance of the CHANNEL CHECK once every 12 hours and once every 24 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 ;
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f () n RPS Instrumentation - B 3.3.1.1 { BASES SURVEILLANCE REQUIREMENTS (continued) 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. The Frequency is based upon operating ex)erience that demonstrates channel failure is rare. T1e CHANNEL CHECK supplements ~1ess formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO. . (Oj l SR 3.3.1.1.3 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 when a 25% RTP. 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 3rovided that requires the SR to be met only at a 25% RTP
)erause 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. LHGR. and APLHGR). At a 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 3er SR 3.0.2. In this event, the SR must be performed wit 11n 12 hours after reaching or exceeding 25% RTP. Twelve hours is based on operating experience and in cmsideration of providing a reasonable time in which to complete the SR.
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'A RPS Instrumentation :
Q B 3.3.1.1 ! BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.3.1.1.4 i l A CHANNEL FUNCTIONAL TEST is performed on each required : channel to ensure that the entire channel will perform the l intended function. 1 Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology. 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 l 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. j p 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 intended function. A Frequency of 7 days provides an acceptable level of system average availability over the l Frequency and is based on References 9 and 10. (The Manual Scram Function's CHANNEL FUNCTIONAL TEST Frequency was credited in the Reference 9 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 IMs 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 fully withdrawing O l FERMI - UNIT.2 B 3.3.1.1- 27 Revision 2. 01/18/99 i
- RPS Instrumentation I B 3.3.1.1 l IV) i BASES 1 SURVEILLANCE REQUIREMENTS (continued)
SRMs from the core since indication is being transitioned from the SRMs to the IRMs. l 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 overla) is not maintained. Overlap between IRMs and APRMs exists w1en sufficient IRMs and APRMs concurrently have onscale readings such that the transition i 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 fully i withdrawing the SRMs from the core. 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 n reading downscale once in MODE 2). s O If overlap for a group of channels is not demonstrated l (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 l should be declared inoperable. I A Frequency of 7 days is reasonable based on engineering judgment and the reliability of the IRMs and APRMs. S_R 3.3.1.1.8 LPRM gain settings are determined from the local flux , profiles measured by the Traversing Incore Probe (TIP) l System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 MWD /ST Frequency is based on operating experience with LPRM sensitivity changes. i i b) V l FERMI - UNIT 2 B 3.3.1.1 - 28 Revision 2 01/18/99
l 1 o RPS Instrumentation Q B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) l SR 3.3.1.1.9 and SR 3.3.1.1.13 , 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 un)lanned transient if the Surveillance were performed with tie 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 t% ('~'/ This Surveillance provides a check of the actual trip , setpoints. The channel must be declared inoperable if the 4 trip setting is discovered to be less conservative than the I Allowable Value specified in Table 3.3.1.11. If the trip i setting is discovered to be less conservative than accounted i for in the appropriate setpoint methodology, but is not ! beyond the Allowable Value, the channel performance is still within the recuirements of the plant safety analysis. Uncer these conditions. the setpoint must be j readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology. The Frequency of 92 days is based on the reliability i analysis of Reference 9. ' l SR 3.3.1.1.11 and SR 3.3.1.1.14 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. D l FERMI UNIT 2 B 3.3.1.1 - 29 Revision 2 01/18/99
A RPS Instrumentation B 3.3.1.1 V BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.3.1.1.11 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. SR 3.3.1.1.11 Note 2 is provided that requires the IRM SR to be performed within 12 hours of entering MODE 2 from MODE 1. Testing of the l MODE 2 IRM Function cannot be performed in MODE 1 without utilizing jumpers, lifted leads, or movable 1 inks. This i 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 3.3.1.1.11 is based upon a 184 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. The Frequency l of SR 3.3.1.1.14 is based upon a 18 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. A U SR 3.3.1.1.12 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 i supplements the automatic self test functions that operate l continuously in the APRM and voter channels. The APRM l CHANNEL FUNCTIONAL TEST covers the APRM channels (including l for Function 2.b only, the recirculation flow input l function, excluding the flow transmitter), the 2 out-of-4 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 3.3.1.1.12 is based on the reliability analysis of Reference 13. (NOTE: The actual voting logic of the 2 out-of 4 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 HODE 1 if the associated Frequency is not met per SR 3.0.2. (3 \.) l FERMI UNIT 2 B 3.3.1.1- 30 Revision 2 01/18/99
i q RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) , l SR 3.3.1.1.15 and SR 3.3.1.1.19 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required trip logic for a specific l 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. For the 2-out of 4 Voter - Function. the LSFT includes simulating APRM trip conditions , at the APRM channel inputs to the 2-out of 4 trip voter channel to check all combinations of two tripped inputs to the 2 out-of 4 trip voter logic in the voter channels. ; l The 18 month Frequency of SR 3.3.1.1.15 is based on the need to perform this Surveillance under the conditions that apply i 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 i components usually pass the Surveillance when performed at the 18 month Frequency. Additionally, the 24 month Frequency of SR 3.3.1.1.19 is l based on Reference 13. I SR 3.3.1.1.16 This SR ensures that scrams initiated from the Turbine Stop Valve-Closure and Turbine Control Valve Fast Closure Functions will not be inadvertently bypassed when THERMAL POWER is = 30t RTP. This involves calibration of the bypass channels. Adequate margins for the instrument setpoint methodologies are incorporated into the actual set >oint. Additionally, consideration is given to the fact tlat main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure; where turbine first stage pressure of 161.9 psig conservatively correlates to 30% RTP), the main turbine bypass valves must remain closed at THERMAL POWER
= 30% RTP to ensure that the calibration remains 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 Functions are considered inoperable. Alternatively, the bypass channel can be placed in the 3
. (G l FERMI UNIT 2 B 3.3.1.1- 31 Revision 2, 01/18/99 l
\
n RPS Instrumentation (] B 3.3.1.1 BASES l SURVEILLANCE REQUIREMENTS (continued)
' conservative condition (nonbypass). -If placed in the i nonbypass condition, this SR is met and the channel is considered OPERABLE. ,
The Frequency of 18 months is based on engineering judgment. l reliability of the components, and = 18 month calibration , interval in the determination of the magnitude of equipment ! drift in the setpoint analysis. ' SR 3.3.1.1.17 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 the APRM 2-out-of 4 Voter Function includes the output relays of the voter and the associated e RPS relays and contactors. (The digital portion of the APRM and 2 out of 4 voter channels are excluded from the RPS RESPONSE TIME testing because self testing and calibration checks the time base of the digital electronics.) { Confirmation of the time base is adequate to assure required i response times are met. ; As noted, neutron detectors are excluded from RPS RESPONSE TIME testing because the principles of detector operation virtually ensure an instantaneous response time. In addition. Note 2 states the response time of the sensors for Functions 3 and 4 are excluded from RPS Response Time Testing. The sensors for these Functions are assumed to ! operate at the sensor's design response time. This j allowance is supported by Reference 12, which determined that significant degradation of the sensor channel response , time can be detected during performance of other Technical Specification SR's and that the sensor response time is a small part of the overall RPS RESPONSE TIME testing. 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 4 channels per trip system, in lieu of the 8 channels specified in Table 3.3.1.1 1 for the HSIV Closure Function. This Frequency is based on the logic interrelationships of the various l FERMI UNIT 2 B 3.3.1.1-32 Revision 2 01/18/99
RPS Instrumentation O i B 3.3.1.1 BASES-SURVEILLANCE REQUIREHENTS (continued) 1 channels required to produce an RPS scram signal. The 3 18 month Frequency is consistent with the typical industry i refueling cycle and is based upon plant operating i experience, which shows that random failures of : instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences. 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 i calibrations consistent with the plant specific setpoint methodology. For the APRM Simulated Thermal Power - Upscale i Function, this SR also-includes calibrating the associated recirculation loop flow channel. : SR 3.3.1.1.18 is modified by a Note that 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 3.3.1.1.3)
-and the 1000 MWD /T LPRM calibration against the TIPS (SR 3.3.1.1.8).
The Frequency of SR 3.3.1.1.18 is based upon 24 mont.. calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis. REFERENCES 1. UFSAR. Figure 7.2 2.
- 2. UFSAR. Section 15.4.1.2.
- 3. NED0 23842 " Continuous Control Rod Withdrawal in the Startup Range." April 18. 1978.
- 4. UFSAR. Section 5.2.2.3.
. 5. UFSAR. Section 15.4.9. , (O%) r FERMI' UNIT 2 B 3.3.1.1-33 Revision 2 01/18/99 l
l
- (m RPS Instrumentation
( B 3.3.1.1 ; BASES : REFERENCES (continued) . 1
- 6. UFSAR, Section 6.3.3. ;
1
- 7. UFSAR, Chapter 15.
- 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 i Analyses for BWR Reactor Protection System." !
March 1988.
- 10. UFSAR, Table 7.2 4. )
i
- 11. NEDC 31336, " Class III. October 1986. General Electric Instrts.ent Setpoint Methodology."
- 12. NED0 32291. " System Analyses for Elimination of Selected Response Time Testing Requirements." January 1994: and Fermi-2 SER for Amendment 111. dated April 18, 1997.
- 13. NEDC-32410P A. " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Functions." October , 1995, and Supplement 1. May 1996. ; O) . l FERMI - UNIT 2 B 3.3.1.1-34 Revision 2 01/18/99
l m' Control Rod Block Instrumentation ,'L) B 3.3.2.1 B 3.3 INSTRUMENTATION 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. O' The purpose of the RBM is to limit control rod withdrawal if (v) 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 ) reset power level. The RBM hcs two channels, either of whic1 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 circuit. The RBM channel signal is generated by averaging a set of local power range monitor (LPRM) signals at various core heights surrounding the control rod being withdrawn. A signal from one of the four 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 preset power level, the RBM is automatically by)assed. The RBM is also automatically bypassed if a peripleral control rod is selected (Ref.1). G l FERMI - UNIT 2 B 3.3.2.1 - 1 Revision 2 01/18/99
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) This Note is based on the reliability analysis (Ref. 9) assumption of the average time required to >erform channel Surveillance. That analysis demonstrated tlat the 6 hour testing allowance does not significantly reduce the probability that a control rod block will be initiated when necessary. l SR 3.3.2.1.1 and SR 3.3.2.1.2 l A CHANNEL FUNCTIONAL TEST is performed for the RWM to ensure that the entire system will perform the intended function. l The SR 3.3.2.1.1 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 selection error is indicated and a control rod block occurs. SR 3.3.2.1.1 is performed during startup. As noted in the SRs. SR 3.3.2.1.1 is not recuired to be performed until I hour after any control roc is withdrawn at s 10% RTP in l MODE 2. -The SR 3.3.2.1.2 CHANNEL FUNCTIONAL TEST is performed by attempting to insert and withdraw a control rod O not in compliance with the prescribed sequence and verifying a selection error is indicated and a control rod insert and withdraw block (respectively) occur. SR 3.3.2.1.2 is performed during'a shutdown. As noted. SR 3.3.2.1.2 is not required to be performed until I hour after THERMAL POWER is s 10% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.1.-and THERMAL POWER reduction to s 10t RTP when in MODE 1 for SR 3.3.2.1.2. to perform the required Surveillance 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 Frequencies are based on reliability analysis (Ref. 8). I SR 3.3.2.1.3 A CHANNEL FUNCTIONAL TEST is performed for each RBM channel to ensure that the entire channel will perform the intended I function. { Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint i methodology. -The Frequency of 184 days is based on l reliability analyses (Ref.10). O
'l FERMI UNIT 2 B 3.3.2.1 - 8 Revision 2 01/18/99
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s Control Rod Block Instrumentation B 3.3.2.1 BASES I SURVEILLANCE REQUIREMENTS (continued) l SR 3.3.2.1.4 A CHANNEL FUNCTIONAL TEST is performed for the Reactor Mode Switch-Shutdown Position Function to ensure that the entire channel will perform the intended function. The CHANNEL FUNCTIONAL TEST for the Reactor Mode Switch-Shutdown Position Function is performed by attem) ting to withdraw any control rod with the reactor mode switc1 in the shutdown position and verifying a control rod block is present. j As noted in the SR the Surveillance is not required to be performed until 1 hour after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be performed without using jumpers, lifted leads, or movable links. This allows entry into MODES 3 and 4 if the 18 month 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. (v 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 un)lanned transient if the Surveillance were performed with tie reactor at power. 0)erating experience has shown these components usually pass t1e Surveillance when performed at the 18 month Frequency. SR 3.3.2.1.5 , l The power at which the RBM is automatically by)assed is ! based on the APRM signal's in>ut to each RBM clannel. Below l the minimum >ower setpoint, t1e RBM is automatically I bypassed. T11s >ower Allowable Value must be verified - periodically to >e less than 30% RTP. If this setpoint is nonconservative, then the affected RBM channel is considered 1 inoperable. Alternatively, the power range channel can be 1 placed in the conservative condition (i.e. enabling the RBM I Function). If placed in this condition, the SR is met and , l the RBM channel is not considered inoperable. The 24 month Frequency is based on the actual trip setpoint methodology utilized for these channels. p Q) l FERMI - UNIT 2 B 3.3.2.1- 9 Revision 2. 01/18/99
- .- .. -. _ - ~ . - -- -_. - - -. .-.
rg Control Rod Block Instrumentation (y B 3.3.2.1 BASES SURVEILLANCE REQUIREMENTS (continued) l SR 3.3.2.1.6 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies 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. As noted, neutron detectors are excluded from the CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Neutron detectors are adequately l surveilled in SR 3.3.1.1.1 and SR 3.3.1.1.7. l The Frequency is based upon the assumption of a 24 month calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis, i SR 3.3.2.1.7 The RWM will only enforce the proper control rod sequence if the rod sequence is properly input into the RWM computer. This SR ensures that the proper sequence is loaded into the RWM so that it can perform its intended function. The Surveillance is performed once arior to declaring the RWM OPERABLE following loading of tie prescribed withdrawal sequence into the RWM since this is when rod sequence input errors are possible. REFERENCES 1. UFSAR, Section 7.6.2.13.5.
- 2. UFSAR. Section 7.6.1.20.
- 3. General Electric Energy, " Maximum Extended Operating Domain Analysis for Detroit Edison Company Enrico Fermi Energy Center Unit 2," NEDC 31843P, July 1990.
- 4. NEDE 24011 P A-10-US, " General Electric Standard Application for Reload Fuel," Supplement for United States March 1991.
s l FERMI UNIT 2 B 3.3.2.1 - 10 Revision 2, 01/18/99
l Control Rod Block Instrumentation B 3.3.2.1 BASES l , i REFERENCES (continued)
- 5. " Modifications to the Requirements for Control Rod Drop Accident Mitigating Systems." BWR Owners' Group.
July 1986.
- 6. NED0-21231. " Banked Position Withdrawal Sequence."
January 1977.
- 7. NRC SER. " Acceptance of Referencing of Licensing i 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-770 061 A. " Bases for Changes to Surveillance Test Intervals and Allowed Out-of Service Times for Selected Instrumentation Technical Specifications."
,s December 1992.
- 10. NEDC 32410P A. " Nuclear Measurement Analysis and Control Power Range Neutron Monitor (NUMAC PRNM)
Retrofit Plus Option III Stability Functions." October 1995, and Supplement 1. May 19%. i i i I i i l lD i l FERMI - UNIT 2 B 3.3.2.1- 11 Revision 2 01/18/99 1' h - , _ ,
l f- PM Instrumentation B 3.3.3.1 l BASES LCO (continued) ; Only two Category I thermocouple channels are needed for i post accident monitoring of sup)ression pool water j temperature (Refs. 3 and 4). Tw outputs for the PAM sensors T50N404A and T50N405B are recorded on two I independent recorders in the control room (channel A is ! redundant to channel B). Both of these recorders must be ; OPERABLE to furnish two channels of PAM indication. These I recorders are the primary indication used by the operator l l during an accident. Therefore, the PAM Specification deals < specifically with this portion of the instrument channels.
- 6. Drywell Pressure Drywell pressure is a Type A. Category I variable provided to detect a breach of the RCPB and to verify ECCS functions that operate to maintain RCS integrity. Two wide range drywell pressure signals are transmitted from separate pressure transmitters and are continuously recorded and displayed on two control room recorders. These recorders n are the primary indication used by the operator during an accident. Therefore. the PAM Specification deals V) t specifically with this portion of the instrument channel.
l 7. 8. Primary Containment Hydroaen and 0xvoen Concentration l l Primary containment hydrogen and oxygen analyzers are Type C. Category I instruments provided to detect high hydrogen or oxygen concentration conditions that represent a potential for containment breach. This variable is also important in verifying the adequacy of mitigating actions.
- 9. Primary Containment Hiah Ranoe Radiation Monitor i Primary containment area radiation (high range) is a Type E.
Category I variable, and is provided to monitor the l potential of significant radiation releases and to provide release assessment for use by operators in determining the need to invoke site emergency plans. The instrumentation provided for this function consists of redundant sensors. l microprocessors and indicators. A common 2-pen recorder in the control room continuously records signals from both channels. The redundant indicators in the relay room and the common recorder in the control room are the primary indication used by the operator during an accident. l FERMI UNIT 2 B 3.3.3.1 - 5 Revision 2 01/18/99 - 1
PAM Instrumentation O B 3.3.3.1 BASES SURVEILLANCE REQUIREMENTS (continued) The 18 month Frequency for all channels except the primary containment oxygen and hydrogen analyzers (per Note 1 to SR 3.3.3.1.3) is based on operating experience and consistency l with the typical industry refueling cycles. The 92 day Frequency for the primary containment oxygen r 1 hydrogen analyzers (per Note 1 to SR 3.3.3.1.2) is baso upon vendor recommendations and instrument accuracy requironents. SR 3.3.3.1.2 is modified by Note 2 stating that performance of the calibration of the oxygen and hydrogen monitors may be delayed until after exceeding 15% RTP (i.e., the power at l which LC0 3.6.3.2 requires the primary containment to be inerted). This delay is allowed for up to 72 hours for one oxygen and one hydrogen monitor, and for 7 days for the l second oxygen and hydrogen monitor. These delays facilitate more accurate calibration methods, which can be employed with the primary containment inerted. SR 3.3.3.1.3 is also modified by Note 2 stating that radiation detectors are excluded from calibration
/c requirements.
REFERENCES 1. Regulatory Guide 1.97. " Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," Rev. 2, December 1980.
- 2. Detroit Edison Letter NRC-89-0148, " Additional i Clarification to Fermi 2 Compliance to Regulatory Guide 1.97. Revision 2." dated June 19, 1989.
- 3. Detroit Edison Letter NRC 89 201, " Regulatory Guide 1.97 Revision 2 Design Review," dated September 12, 1989.
- 4. NRC Letter, " Emergency Response Capability Conformance
, to Regulatory Guide 1.97, Revision 2 (TAC No. 59620)." dated May 2, 1990.
- 5. Detroit Edison Letter NRC 93 0105, " Fermi 2 Review of Neutron Monitoring System Against Criteria of NED0 31558A," dated September 28, 1993.
O l FERMI UNIT 2 B 3.3.3.1 - 11 Revision 2, 01/18/99 1
Recirculation Loops Operating i ( B 3.4.1 ' BASES BACKGROUND (continued) begins to boil, creating steam voids within the fuel channel that continue until the coolant exits the core. Because of reduced moderation the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control system allows o)erators to increase recirculation flow and , sweep some of tie voids from the fuel channel, overcoming the negative reactivity void effect. Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation without having to move control rods and disturb desirable flux patterns. Each recirculation loop is manually started from the control room. The MG set provides regulation of individual recirculation loop drive flows. The flow in each loop is manually controlled within limits established by the recirculation speed control system, GDC 12 of 10 CFR 50 Appendix A (Ref. 4) states that the /n V) reactor core and associated coolant, control, and protection systems shall be designed to assure that power oscillations which can result in exceeding specified fuel design limits are not possible or can be reliably detected and suppressed. BWR cores typically operate with the presence of global flux noise in a stable mode which is due to random boiling and flow noise. As the power / flow conditions are changed. along l with other system parameters (xenon, subcooling, power distribution etc.) the thermal hydraulic / reactor kinetic feedback mechanism can be enhanced.such that perturbations may result in sustained limit cycle or divergent oscillations in power and flow. Two major modes of oscillations have been observed in BWRs. i The first mode is the fundamental or core wide oscillation ) mode in which the entire core oscillates in phase in a given axial plane. The second mode involves regional oscillation ; in which one half of the core oscillates 180 degrees out of phase with the other half. Studies have indicated that adequate margin to the Safety Limit MCPR may not exist i l during oscillations. ! l O ' G' l FERMI UNIT 2 B 3.4.1 - 2 Revision 2 01/18/99
p) Recirculation Loops Operating B 3.4.1 BASES APPLICABLE The operation of the Reactor Coolant Recirculation System is SAFETY ANALYSES an initial condition assumed in the design basis loss of coolant accident (LOCA) (Ref.1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered. The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pipe break assumed to be in the loop with the higher flow. While the flow coastdown and core response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate), a small mismatch has been determined to o) (d be acceptable based on engineering judgement. The recirculation system is also assumed to have ufficient flow coastdown characteristics to maintain fuel thermal margins i during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the UFSAR. l 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 (Ref. 3). The transient analyses of Chapter 15 of the UFSAR 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. 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 l different relationshi)s between recirculation drive flow and i reactor core flow. T1e APRM Simulated Thermal Power - I Upscale setpoint is in LCO 3.3.1.1, " Reactor Protection . System (RPS) Instrumentation." %) l FERMI UNIT 2 B 3.4.1 - 3 Revision 2, 01/18/99
3 ! l i Recirculation Loops Operating B 3.4.1
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_ BASES APPLICABLE SAFETY ANALYSES (continued)- Thermal-hydraulic stability analysis (Ref. 5) has concluded i that procedures for detecting and suppressing power i oscillations that might be induced by a thermal-hydraulic 1 instability are necessary to provide reasonable assurance I l that the requirements of Reference 4 are satisfied. ' Recirculation loops operating satisfies Criterion 2 of 10 , CFR 50.36(c)(2)(ii). '
.LCO. Two recirculation loops are required to be in operation with their flows matched within the limits specified in ,
SR 3.4.1.2 to ensure that during a LOCA caused by a break of i the piping of one recirculation loop the assumptions of the I LOCA analysis are satisfied. With the limits specified in l SR 3.4.1.2 not met, the recirculation loop with the lower ! flow must be considered not in operation. With only one , recirculation loop in operation, modifications to the APRM 'l i Simulated Thermal Power-Upscale setpoint (LCO 3.3.1.1) may 1 ,O be applied to allow continued operation consistent with the. assumptions of the safety analysis. Operations that exhibit core thermal hydraulic instability are not permitted. Additionally, in order to avoid potential power oscillations due to thermal hydraulic instability, operation at certain combinations of power and flow are not permitted. These restricted power and flow regions are referred to as the " Scram" and " Exit" regions and are defined by Bases Figure B 3.4.1-1. A Note is provided to allow 4 hours following the transition to single loop operation from two loop operation to l l establish the APRM Simulated Thermal Power - Upscale setpoint in accordance with the single loop allowable value, l which is specified in Table 3.3.1.11. The 4 hour period is sufficient to make the adjustment given the relatively small change required. This transition only results in applying the new single-loop allowable values to APRM OPERABILITY. Any ARPM non compliance with the required allowable value after this 4 hour allowance, results in ACTIONS of LC0 3.3.1.1 being entered: no ACTION of LC0 3.4.1 would L apply. LO l t l _ FERMI UNIT 2- B 3.4.1- 4 Revision 2, 01/18/99 j ! _ - _ . . . ._ . . . - . ._ - ._ . - - - . __ - . - ~
Recirculation Loops Operating ] B 3.4.1 BASES ACTIONS (continued) ' IL1 When operating in the " Exit" region (refer to Figure B 3.4.11), the potential for thermal hydraulic instabilities is increased and sufficient margin may not be available for operator response to sup3ress potential power oscillations. Therefore, action must ye initiated immediately to restore operation outside of the " Exit" region. Control rod insertion and/or core flow increases are designated as the means to accomplish this objective. Required Action B.1 is modified by a Note that precludes core flow increases by restart of an idle recirculation loop, or by resetting a recirculation flow limiter. Core flow increases by these means would not support timely completion of the action to restore operation outside the
" Exit" Region.
C. I h d With no recirculation loops in operation in MODE 2, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 6 hours is reasonable, based on operating experience, to reach MODE 3 from NODE 2 conditions in an orderly manner and without challenging plant systems. D_d If operating with no recirculation pumps in operation in MODE 1 or operating in the " Scram" region (refer to Bases Figure B 3.4.1-1), or if core thermal-hydraulic instability
-is detected, then unacceptable power oscillations may result. Therefore,.the reactor mode switch must be i immediately placed in the shutdown position to terminate the ,
l potential for unacceptable power oscillations. l l Thermal hydraulic instability is evidenced by a sustained increase in APRM or LPRM peak to peak noise level reaching 2 or more times its initial level and occurring with a characteristic period of less than 3 seconds. L l FERMI - UNIT 2 B 3.4.1 - 6 Revision 2, 01/18/99
l r' Recirculation Loops Operating ( B 3.4.1 I I BASES ACTIONS (continued) l If entry into this condition is an unavoidable and well known consequence of an event, early initiation of the Required Action is appropriate. Also it is recognized that , during certain abnormal conditions, it may become l operationally necessary to enter the " Scram" or " Exit" i region for the purpose of: 1) protecting plant equipment, which if it were to fail could impact plant safety, or
- 2) protecting a safety or fuel operating limit. In these i cases, the appropriate actions for the region entered would be performed as required.
, These requirements are consistent with References 5 and 6. 1 SURVEILLANCE SR 3.4.1.1 REQUIREMENTS l This SR provides frequent periodic monitoring for core thermal-hydraulic instability by monitoring APRM and LPRM x signals for a sustained increase in APRM or LPRM peak to , peak noise level reaching 2 or more times its initial level 4 and occurring with a characteristic period of less than 3 seconds. The 1 hour Frequency is based on the small potential for core thermal-hydraulic oscillations to occur outside the " Scram" or " Exit" regions. Therefore, frequent monitoring of the APRM and LPRM signals is appropriate when
- operating in the " Stability Awareness" region.
This SR is modified by a Note that states performance is only required when operating in the " Stability Awareness" region (refer to Bases Figure B 3.4.1-1) (i.e.. in the power to-flow region that is near regions of higher probability for core thermal hydraulic instabilities). This is acceptable because outside the " Stability Awareness" region, power and flow conditions are such that sufficient margin exists to the potential for core thermal-hydraulic instability to allow routine core monitoring. Any unanticipated entry into the " Stability Awareness" region would require immediate verification of core stability since the Surveillance would not be current. l l FERMI - UNIT 2 B 3.4.1 - 7 Revision 2. 01/18/99
Recirculation Loops Operating v) B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.4.1.2 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 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.
The mismatch is measured in terms of >ercent of rated core flow. If the flow mismatch exceeds t1e 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 gs Surveillance must be performed within 24 hours after both loo)s are in operation. The 24 hour Frequency is consistent wit 1 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. I l 1 i U , i l FERMI - UNIT 2 B 3.4.1-8 Revision 2 01/18/99 l
- ... - - - - _ .. - . - . _ . . . . . . - .= . . - - _ . _ . . . _ .
l t Recirculation Loops Operating {q} B 3.4.1 i BASES REFERENCES 1. UFSAR, Section 6.3.3.
- 2. NEDE-23785 P A, " SAFER /GESTR Models for the Evaluation 7 of the Loss of-Coolant Accident," Revision 1, October f 1984.
- 3. EE-56-0386, " Fermi 2 Single Loop Operation Analysis,"
Rev. 1. April 1987, and NEDC-32313-P, "Enrico Fermi Energy Center Unit 2 Single-Loop Operation." September 1994.
- 4. 10 CFR 50, Appendix A. GDC 12.- ;
- 5. NRC Generic Letter 94-02, "Long Term Solutions and U) grade of Interim Operating Recommendations for T1ermal Hydraulic Insthilities in Boiling Water Reactors," July 1994.
- 6. BWROG Letter 94078, "BWR Owners' Group Guidelines for Interim Corrective Action," June 1994.
' s.. f')T : I fl FERMI - UNIT 2 B 3.4.1- 9 Revision 2, 01/18/99
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l Recirculation Loops Operating O B 3.4.1-BASES 1 l L l 1 f O un r THERMAL POWER vs CORE FLOW Figure B 3.4.1-1 l
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- - 11 FERMI ~. UNIT.2 B 3.4.1 -10 Revision 2 01/18/99 !
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fm Battery Cell Parameters (j B 3.8.6 BASES ACTIONS (continued) Continued operation is only permitted for 31 days before battery cell parameters must be restored to within Category A and B limits. Taking into consideration that, while battery capacity is degraded, sufficient capacity exists to perform the intended function and to allow time to fully restore the battery cell parameters to normal limits, this time is acceptable for operation prior to declaring the DC batteries inoperable. BJ When any battery parameter is outside the Category C limit for any connected cell. sufficient capacity to supply the maximum expected load requirement is not ensured and the corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potentially extreme conditions, such as not completing the Required Actions of Condition A within the required Completion Time or average electrolyte temperature of representative cells falling below 60*F also are cause for immediately declaring the (a) associated DC electrical power subsystem inoperable. SURVEILLANCE SR 3.8.6.1 ! REQUIREMENTS ! This SR verifies that Category A battery cell parameters are i consistent with IEEE 450 (Ref. 3), which recommends regular battery inspections (at least one per month) including voltage, specific gravity, and electrolyte temperature of pilot cells. SR 3.8.6.2 The quarterly inspection of specific gravity and voltage is consistent with IEEE-450 (Ref. 3). In addition. within 24 hours of a battery discharge < 105 V or a battery overcharge > 150 V for Division I and > 145 V for Division II, the battery must be demonstrated to meet Category B limits. Transients, such as motor starting transients, which may momentarily cause battery voltage to drop to s 105 V, do not constitute a battery discharge provided the battery terminal voltage and float current {} %y' l FERMI UNIT 2 B 3.8.6 -3 Revision 2 01/18/99 - _ _ _ _ _ _ _ _ _ _ _ _ _ _}}