ML20196K662

Proposed Tech Specs Re Rev 4 to Fermi 2 ITS Submittal

ML20196K662
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Site:	Fermi
Issue date:	04/02/1999
From:	DETROIT EDISON CO.
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ML20196K659	List: ML20196K662 NRC-99-0032, Forwards Rev 4 to Fermi 2 Its,Per Submittal .Rev Provides Responses to NRC RAI Re ITS Sections 3.4 & 3.9 Contained in Ltr
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NUDOCS 9904050069
Download: ML20196K662 (200)

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{{#Wiki_filter:F l l INSERT THIS PAGE IN FRONT OF VOLUME 6 Voluene th SECHON 3.4 L Remove Replace 3.4.1 ITS pg 3.4-1 Rev 2 3.4.1 ITS pg 3.4-1 Rev 4 3.4.1 CTS M/U (3/4 4-2) pg 3 of 6 3.4.1 CI'S M/U (3/4 4-2) pg 3 of 6 Rev 4 3.4.1 DOCS pg 4 Rev 2 3.4.1 DOCS pg 4 Rev 4 3.4.1 NUREG M/U pg 3.4-1 Rev 2 3.4.1 NUREG M/U pg 3.4-1 Rev 4 B 3.4.3 ITS pg B 3.4.3-4 Rev 0 B 3.4.3 ITS pg B 3.4.3-4 Rev 4 3.4.3 DOCS pg 2 Rev 0 3.4.3 DOCS pg 2 Rev 4 B 3.4.3 NUREG M/U pg B 3.4-15 B 3.4.3 NUREG M/U pg B 3.4-15 Rev 4 3.4.4 CTS M/U (3/4 4-10) pg 1 of 2 3.4.4 CTS M/U (3/4 4-10) pg 1 of 2 Rev 4 3.4.4 DOCS pg i Rev 0 3.4.4 DOCS pg 1 Rev 4 3.4.4 DOCS pg 2 kev 0 3.4.4 DOCS pg 2 Rev 4 { 3.4.4 DOCS pg 3 Rev 0 3.4.4 DOCS pg 3 Rev 4 3.4.5 ITS pg 3.4-11 Rev 0 3.4.5 ITS pg 3.4-11 Rev 4 3.4.5 ITS pg 3.4-12 Rev 0 3.4.5 ITS pg 3.4-12 Rev 4 3.4.5 CTS M/U (3/4 4-12) pg 3 of 3 3.4.5 CTS M/U (3/4 4-12) pg 3 of 3 Rev 4 3.4.5 DOCS pg 3 Rev 0 3.4.5 DOCS pg 3 Rev 4 3.4.5 NUREG M/U pg 3.4-9 3.4.5 NUREG M/U pg 3.4-9 Rev 4 3.4.5 NUREG M/U pg 3.4-11 3.4.5 NUREG M/U pg 3.4-11 Rev 4 B 3.4.5 NUREG M/U pg B 3.4-25 B 3.4.5 NUREG M/U pg B 3.4-25 Rev 4 3.4.5 JFD's pg 1 Rev 0 3.4.5 JFD's pg i Rev 4 3.4.6 ITS pg 3.4-13 Rev 0 3.4.6 ITS pg 3.4-13 Rev 4 3.4.6 ITS pg 3.4-14 Rev 0 3.4.6 ITS pg 3.4-14 Rev 4 B 3.4.6 ITS pg B 3.4.6-5 Rev 0 B 3.4.6 ITS pg B 3.4.6-5 Rev 4 3.4.6 DOCS pg 2 Rev 0 3.4.6 DOCS pg 2 Rev 4 3.4.6 NUREG M/U pg 3.4-12 3.4.6 NUREG M/U pg 3.4-12 Rev 4 B 3.4.6 NUREG M/U pg B 3.4-32 B 3.4.6 NUREG M/U pg B 3.4-32 Rev 4 3.4.6 JFD's pg 1 Rev 0 3.4.6 JFD's pg 1 Rev 4 3.4.7 CTS M/U (3/4 4-17) pg 2 of 3 3.4.7 CTS M/U (3/4 4-17) pg 2 of 3 Rev 4 3.4.7 CTS M/U (3/4 4-18) pg 3 of 3 3.4.7 CTS M/U (3/4 4-18) pg 3 of 3 Rev 4 3.4.7 DOCS pg 1 Rev 0 3.4.7 DOCS pg 1 Rev 4 3.4.7 DOCS pg 2 Rev 0 3.4.7 DOCS pg 2 Rev 4 3.4.7 DOCS pg 3 Rev 0 3.4.7 DOCS pg 3 Rev 4 f 3.4.7 DOCS pg 4 Rev 0 3.4.7 DOCS pg 4 Rev 4 3.4.7 NSHC pg 7 Rev 4 9904050069 990330 ~ Rev 4 04/02/99 l PDR ADOCK 05000341 l P PDR _

Volume 6: SECTION 3.4 (cont'd)L ^ ". ' ~ r / Remove Replace i ( / ' ~ " 3.4.8 DOCS pg 2 Rev 0 3.4.8 DOCS pg 2 Rev 4 3.4.9 DOCS pg i Rev 0 3.4.9 DOCS pg i Rev 4 3.4.9 DOCS pg 2 Rev 0 3.4.9 DOCS pg 2 Rev 4 3.4.9 DOCS pg 3 Rev 0 3.4.9 DOCS pg 3 Rev 4 3.4.10 ITS pg 3.4 24 Rev 0 3.4.10 ITS pg 3.4-24 Rev 4 3.4.10 CTS h1/U (3/4 4-19) pg 5 of 8 3.4.10 CTS h1/U (3/4 4-19) pg 5 of 8 Rev 4 3.4.10 DOCS pg i Rev 0 3.4.10 DOCS pg i Rev 4 3.4.10 DOCS pg 2 Rev 0 3.4.10 DOCS 08 2 Rev 4 3.4.10 DOCS pg 3 Rev 2 3.4.10 DOCS pg 3 Rev 4 3.4.10 NUREG M/U pg 3.4 24 (Insert) Rev 0 3.4.10 NUREG h1/U pg 3.4-24 (Insert) Rev 4 3.4.11 CTS M/U (3/4 4-23) pg 1 of 1 3.4.11 CTS M/U (3/4 4-23) pg 1 of 1 Rev 4 3.4.11 DOCS pg i Rev 0 3.4.11 DOCS pg i Rev 4 3.4.11 DOCS pg 2 Rev 0 3.4.11 DOCS pg 2 Rev 4 3.4.11 NSilC pg i Rev 0 3.4.11 NSHC pg 1 Rev 4 3.4.11 NSliC pg 2 Rev 4 (D (/ 4 y G Rev 4 04/02/99

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Recirculation Loops Operating 3.4.1 I 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LC0 3.4.1 The reactor core shall not exhibit core thermal-hydraulic instability or operate in the " Scram" or " Exit" Regions. I AND a. Two recirculation loo)s with matched recirculation loop jet pump flows shall )e in operation: b. One recirculation loop may be in operation provided: 1. 1C0 3.3.1.1.

• Reactor Protection System (RPS) astrumentation." Function 2.ti (Average Power Range 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. ............................N0TE-Required allowable value modification for single loop Jc operation may be delayed for up to 4 hours after transition from two recirculation loop operations to single recirculation loop operation. l 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." i '(continued) l ~ / : 4 l FERMI UNIT 2 3.4-1 Revision 4. 04/02/99

t SPEO FI CAch 0 N 3.4. l hl1o s et, Spe,cd$ceWrn 3 Y.toh ( Mio see SreelficaNon 3 *5~' I) l REACTOR COOLANT S'YSTEM SURVEftlANCE REQUIRFMFNTS [*y[4.4.1.1.1 N Each pump discharge valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during

1. 4 3'

each STARTUP* prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER. jy 4."r.l.i.i Lau w .9.1.1. With one 'eactor cools t system rectre ation loop nov m 'M operati. at least nce per,12 h rs verify that a. ERMAL POWER s less than or equal to 67. of RATED THERMAL POWE, and - b. The individ I recirculati pump flow ce roller for t operating circulation p p is in the ual mode, c The speed f the operati recirculatic pump is les tha s er equal o 75% of rate pump speed y b.4.1.1.4 With one reactor coolant system loop not in operation with THERMAL POWER less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of rated loop flow, verify the following differential temperature requirements are met within no (more than 15 minutes prior to either THERMAL POWER increase or recirculation ow increase: a. Less than or equal to 145'F between reactor vessel steam W space coolant and bottom head drain line coolant, and g;.hcofirw b. Less than or equal to 50*F between the reactor' coolant within the loop not in operation and the coolant in the N'* reactor pressure vessel", and c. Less than or equal to 50*F between the reactor coolant within the loop not in operation and the operating loop." i sea Egecificah 351

• 1f not performed within the previous 31 da'ys.

I" / \\ ** Requirement does not apply when the recirculation' loop not in operation is \\ hl% isolated from the reactor pressure vessel. 5F FERMI - UNIT 2 3/44-2 Amendment No. #, #.57. 130 sA l PAGE 3 0F 06

f t DISCUSSION OF CHANGES ITS: SECTION 3.4.1 - RECIRCULATION LOOPS OPERATING n LA.3 CTS 3.4.10. and Figure 3.4.101. detail various power-to-flow operating regions for the reactor core: CTS Actions a, b. and c N specify requirements for operation in these regions: and l CTS 4.4.10.2 details the specific region to monitor for N, J instability when operating near the instability region. These ej ] controls are based on Generic Letter 94 02. "Long Term Solutions 3* and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors" and the June 1994 BWROG Letter (no. 94078). ITS 3.4.1 maintains the scope and intent of CTS requirements and Actions, but relocates the definition of what constitutes core thermal hydraulic instability. and the Figure that contains the definitions of the various regions, to the ITS Bases. Bases revisions require change control g in accordance with ITS 5.5.10. Bases Control Program. These j relocations continue to provide adequate protection of the public health and safety since the ITS retain sufficient requirements related to maintaining appropriate response to thermal hydraulic instabilities. TECHNICAL CHANGES LESS RESTRICTIVE "Speci fic" L.1 CTS 3.4.1.3 Applicability applies the recirculation pump speed (revised to " loop flow" in another discussion) mismatch "during two recirculation loop operation." ITS SR 3.4.1.2 also applies the mismatch criteria to "when both recirculation loops [are] in operation": but adds a Note to the SR allowing 24 hours after establishing both loops in operation before the SR is required to be performed. This allowance is necessary to avoid intentional entry into the Actions each time the second recirculation pump is started and the SR has not been performed within its required Frequency. This change, although less restrictive. will have a negligible impact on safety. 1 l l l FERMI - UNIT 2 4 REVISION 4 04/02/99l

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_.i l .y l ~.i Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) i. (' 3.4.1 Recirculatpn Loops Operating g 4g j '"LHO JLY PY" P \\ hTwo recirculation}11 oops with matched flows shall be in 'H'I*f / LCO 3.4.1 g g c,7 eg{ + og;bif-3'v 'I'3) operation, core %wmoi hyttrastic iM 0,Wlity or opetuk-in &s "Suod or '&ii%gisor }: ^ y 50ne recirculation loop may be in operation provid ,,i oHowing-Hmy e,s sppH ed-when-the-esoe+att f hpplicab%- n. L-CC 0.2.1, "A"ERAGE-Pl:ANAR41 MEAD MES.T CEFrP."-M0" ;; ATE-( HGR)," single loo operation limits pecified in eCOLRg ,e LC0 3.2.2, " MINI M CRITICAL POWER 10 (MCPR)," si 91 o_coeration_ mits {1get-ifiad in tha N D1 - =nd ( n CO 3.3.1.1, " Reactor Protection System (RPS) u 5de M.c)\\ Achm Instrumentation." Function 2.b (Average Power Ran e / MonitorsIFh" BWeimulated Thermal Power N / Allowable 7alue of Tabl

1. -1 is reset for single loop operatio g]n pppg_

--/ m _, s /i C' hj52T 3 4 /- / ) ,; j APPLICABILITY: MODES I and 2. 4 L-s,- ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Req rement of the A.1 Satisfy t 24 hou LC not'm requiren}(nts of the LCO. / / p.,. _- a...=. w._ (continued) /Msser 3 V 1-2. m I BWR/4 STS 3.4-1 Rev 1, 04/07/95 ( ~ n n a 7vV A

SRVs B 3.4.3 BASES SURVEILLANCE REQUIREMENTS (continued) The SR gives set pressures for all 15 SRVs installed. However. since only 11 SRVs are required, the SR is met if 11 SRVs are set properly. The Frequency is required by the Inservice Testing Program and is consistent with the. fact that Surveillance must be performed during shutdown conditions. SR 3.4:3.2 A manual actuation of each required SRV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to l perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the SRVs divert steam flow upon opening. Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is = 850 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by I turbine bypass valves open at least 20%. Plant startup is allowed prior to >erforming this test because valve OPERABILITY and tie setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. 'If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the SRV is considered OPERABLE. l FERMI - UNIT 2 B 3.4.3-4 Revision 4. 04/02/99

p q.., - (,.. .,,a, DISCUSSION OF CHANGES ITS: SECTION 3.4.3 - SAFETY RELIEF VALVES (SRVs) LA.2 CTS 3.4.2.1.c and 4.4.2.1.1 requires the SRV position indicators be Operable. Additionally CTS 4.4.2.1.1 requires the SRV position indicators be demonstrated Operable with the pressure setpoint of each of the tail-pipe pressure switches verified by performance of a Channel Calibration. The SRV position indicators 4 do not impact the Operability of the SRVs. ' DECO knows of no plant 1 J specific reason that these position indicators would meet the 10 1 CFR 50.36 criteria for inclusion in the ITS. ITS 3.4.3 requires the Operability of the SRVs. but does not require that the j position indicators are Operable. Therefore these requirements will be relocated to the Technical Requirements Manual (TRM). l,4 which require revisions to be controlled by 10 CFR 50.59. This 4 relocation continues to provide adequate protection of the public E health and safety since the requirement for SRV Operability 4 continues to be required by the Technical Specifications, and is consistent with the NUREG 1433. LR.1 CTS SR 4.4.2.1.2 requires 1/2 of the SRVs to be set pressure tested at least once per 18 months, such that all 15 SRVs are set pressure tested at least once per 40 months. These requirements are considered details of methods for performing the surveillance and are not included in the ITS. These details do not impact the Operability of the SRVs. ITS 3.4.3 continues to require the Operability of the SRVs and requires they be tested in accordance with the IST Program. 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 requirement for SRV Operability continues to be required by the Technical Specifications. This is consistent with the NUREG 1433. FERMI UNIT 2 2 REVISION 4 04/02/99l l

S/RVs B 3.4.3 BASES (c'ontinued) SURVEILLANCE SR 3.4.3.1 RECUIREMENTS ThisSurveillancerequiresthattheTrequired[S/RVswill open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the S/RV safe lift settings mast be performed during shutdown, since this is a bench test k, to done in accordance with the Inservice Testing Pr'ogr The lift setting pressure shall correspond to tent conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is i for OPERABILITY; however, the valves are reset t g i 15 uring the Surveillance to allow for drift. [guved by de /sre T,D/, i he 6Frequencymmeien Me3;f 0 ns Surveillancemust,beperformedduringshutdowncon.d / m 3;,y ) ,stt f ren sdc> fo rinsta $leb } Anu n -"="-v---_-n~=o="" c g) jg $g V.r hat ((osoe ver SInCt. UNIV If SR 3.4.3.1 s o<e repu/recl IAc A manual actuation of eachgrequiredKS/RV is perfomed to W l/4 e y /3 mef if // sav, verify that, mechanically, the valve is functioning properly and no blockage exists'in the valve discharge line. This te set f7/cf f f /ye can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perfom this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening. Sufficient time is therefore allowed after the required s gro pressure and flow are achieved to perform this' test. Adequate pressure at which this test is to be perfomed is 2O} psig (the pressure recommended by the valve ranufacturer). Adequate steam flow is represente:I by [et-- , -fu r hIn e j'Yfd

• b10}-1b/hrb1sas -1,25-turbine-bypass-valves-open -or-tetal-steam fl 7

r Plant startup is allowed prior to perfoming i Wlgg ogen p,,t this test because valve OPERABILITY and the setpoints for i A overpressure protection are verified, per ASME Code h 7 D 70 requirements, prior to valve installation..Therefore, this I SR is modified by a Note that states the Surveillance is not required to be perfomed until 12 hours after reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable (continued) BWR/4 sii B 3.4-15 Rev-IN4/02/35 i.

8Cth to fl S' Y lQ See Specthw fort 3.k.0 U REACTOR COOLANT SYSTEM A.I OPERATIONAL LEAKAGE llMITING CONDITION FOR OPERATION 3.0.3 F Reactor coo'lant system leakage shall be limited to: 3,4. 4.a.-ar No PRESSURE BOUNDARY LEAKAGE. 3.4 i.6 br S gpm UNIDENTIFIED LEAKAGE. 3,d.d.C c. 25 spm total leakage averaged over any 24-hour period. OcC d. Leakage specified in Table 3.4.3.2-1 at a reactor coolant system n Sp'eciE' pressure of 1045 e 10 psig from any reactor coolant system pressure = ,"j. 4 8i isolation valve specified in Table 3.4.3.2-1 4 g g y,f,4.cl s 2 gpm increase in UNIDENTIFIED LEAKAGE within 24 hour period during OPERATIONAL CONDITION 1. I L2 i /D R N N O APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 3. ^ ACTION: a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within NCts.on C 12 hours and in COLD SHUTDOWN within the next 24 hours, b. With any reactor coolant system 1.eakage greater than the limits in b MC-lion A and/or c u bove, rectuce the leakage rate to within the limits within a Lhour_s/or be in a ::ast HOT SHUTDOWN within the next 12 hours and in g,pg C COLD SHUTDOWN with. .9 following 24 hours. c. - With any reactor coolant system pressure isolation valve leakage greater GCC than the above limit, isolate the high pressure portion of the affected C[icgon system from the low pressure portion within 4 hours by use of at least SP"bM b one other closed manual, deactivated automatic, or check

• valve, or be j

in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN i within the following 24 hours. j d. With one or more of the high/ low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-2 inoperable, restore the inoperable monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours; restore the inoperable monitor (s) to OPERABLE status ~within 30 days or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. j Sf ;coh 'Which ha:; been verified not to exceed the allowable leakage limit at last refueling outage or after the last time the valve was disturbed. Sg whichever is more recent. FERMI - UNIT 2 3/4 4-10 Amendment No. E7, ES, 98 PAGE J OF 02. b4

. k h>: 'g, > f' h[, .l DISCUSSION OF CHANGES ITS: SECTION 3.4.4 RCS OPERATIONAL LEAKAGE l d-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 pref.erences 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 3.4.3.2.c snd 3.4.3.2.e allow Leakage to be averaged over "any 24 hour period." ITS 3.4.4.c and 3.4.4.d specifies that the Leakage is averaged over the " previous 24 hour period." The intent of the CTS is that the total Leakage limit applies, to the Q previous 24 hours (not any future or past 24 hour periods). This results in a " rolling average" covering "any 24 hour period." Q Therefore, changing "any" to "the previous" does not change the intent of the CTS requirement. This change is administrative with no impact on safety. TECHNICAL CHANGES MORE RESTRICTIVE l None j l TECHNICAL CHANGES LESS RESTRICTIVE J " Generic" LA.1 ' CTS 4.4.3.2.1 and footnote

• detail the specific methods for performing the Surveillance for leakage monitoring.

ITS SR 3.4.4.1 requires the performance of the Surveillance, but does not provide specific details on the ways the Surveillance can be performed. Defining the method used to determine RCS Leakage is not necessary to ensure Leakage does not exceed the LC0 limits and detect a degradation of the RCPB does not impact the requirement to maintain.the RCPB Leakage within specified limits. Therefore, these details can be relocated to the Bases. This change is consistent with NUREG-1433. The information moved to the Bases l requires changes to be controlled in accordance with the ITS l 5.5.10. Bases Control Program. This relocation continues to i provide adequate protection of the public health and safety since the requirement for determining acceptable leakage rates continues 1 to be required by the Technical Specifications. FERMI - UNIT 2 1 REVISION 4 04/02/99l u

/ '.... n y-jo. DISCUSSION OF CHANGES ITS: SECTION 3.4.4 - RCS OPERATIONAL LEAKAGE TECHNICAL CHANGES - LESS RESTRICTIVE " Specific" L.1 CTS 4.4.3.2.1 requires that leakage be demonstrated to be within limit, in part, by monitoring primary containment atmospheric gaseous radioactivity be monitored at least once per 4 hours. However, the remaining parts of CTS 4.3.2.1 require leakage o~ e demonstrated to be within limit by monitoring once per 12 hours; and this 12 hour monitoring is done on the system that actually quantifies the leakage (the atmospheric gaseous radioactivity monitor is agt utilized to quantify the leakage for comparison to the LC0 limit, as provided in CTS footnote

• to the referenced surveillance)..

ITS SR 3.4.4.1 requires verification every 12 hours that the RCS unidentified and total Leakage, and lg unidentified Leakage increase, are within limits consistent with g these latter CTS requirements. RCS Leakage is monitored by a l4 variety of instruments designed to provide alarms when excessive Leakage is indicated and to quantify the various types of Leakage. Q In conjunction with alarms and administrative controls, a 12 hour Frequency for this Surveillance is appropriate for identifying Leakage and for tracking trends. Note also that in Mode 1 this Frequency is restricted from applying the 25% extension of ITS j SR 3.0.2. This change is consistent with Generic Letter 88 01, will provide an effective means to determine any adverse trends and will have a negligible impact on safety. I b 1 q FERMI UNIT 2 2 REVISION 4, 04/02/99l

':.g- - yC,. 1:^ ~j p " r ik ' s l': 3 1 - 1. .,u ~ DISCUSSION OF CHANGES ITS: SECTION 3.4.4 RCS OPERATIONAL LEAKAGE I L.2 CTS 3.4.3.2.f. Action f. 4.4.3.2.1.b and 4.4.3.2.1.c contain requirements for Unidentified Leakage during Operational Conditions 2 and 3. IT5 3.4.4 revises the unidentified Leakage rate increase limit to be applicable only in Mode 1. instead of the CTS required Operational Conditions 1. 2 and 3. As the plant starts up and increases pressure, increasing leakage will occur due to increasing pressure. This could result in a requirement for a unit shutdown. even though there is no new leakage source with potential of rapid propagation. This change allows the limit to be applied only after Mode 1 is achieved, which is when reactor pressure has effectively reached normal operating pressure. The overall 5 gpm unidentified Leakage limit is still required to be met during the startup. This limit is much below the expected flow from a critical crack in the primary system. and provides adequate protection during the brief exception to the leakage increase limit. This change is also consistent with NUREG 1433. Therefore, this less restrictive change will have a minimal impact on safety. As discussed in 00C L.1 above, the 12 hour Frequency for monitoring for RCS Leakage is adequate. The CTS 4.4.3.2.1.b and c frequency of once per 4 hours in MODES 2 and 3 is related to h the appropriate monitoring frequency for the unidentified leakage j rate increase limit, which is eliminated for MODE 2 and 3 (as discussed in this L.2 DOC). As such, this once per 4 hour frequency can be elk nated, leaving the once per 12 hour surveillance. RELOCATED SPECIFICATIONS None TECHNICAL SPECIFICATION BASES The CTS Bases for this Specification have been replaced by Bases that reflect the format and applicable content of ITS 3.4.4 consistent with the BWR STS. NUREG 1433. Rev. 1. 4 FERMI - UNIT 2 3 REVISION 4 04/02/99l

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{, RCS PIV Leakage 3.4.5 ACTIONS ....................................-NOTES 1. Separate Condition entry is allowed for each flow path, i 2. Enter applicable Conditions and Required Actions for systems made inoperable by PIVs. l CONDITION REQUIRED ACTION COMPLETION TIME A. One or more flow paths -- -NOTE -- - - - ggl with leakage from one Each check valve used to or more RCS PIVs not" satisfy Required Action A.1 __( within limit, must have been verified to meet SR 3.4.5.1 at the last refueling outage or'after the last time the valve was disturbed, whichever is more recent. A.1 Isolate th'e high 4 hours pressure portion of a the affected system from the low pressure portion by use of one other closed manual, de activated automatic, or check valve. 1 B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. ANQ B.2 Be in MODE 4. 36 hours i l l ~ '

l. FERMI UNIT 2 3.4 11 Revision 4 04/02/99 L

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[< '. {.d., j 'e. J.! !, ' s3 f; RCS PIV Leakage 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 - --- NOTE-- - -- - --- ---- - Not required to be performed in MODE 3. Verify equivalent leakage of each RCS PIV. In accordance g at an RCS pressure a 1035 and 5 1055 psig: with the g Inservice 1 a. For PIVs other than LPCI loop A and B Testing Program injection isolation valves is s 0.5 gpm per nominal inch of valve Ll size up to a maximum of 5 gpm; b. For LPCI loop A and B outboard injection isolation valves is s 0.4 gpm through seat, and s 5 ml/ min external leakage: and C. For LPCI loop A and B inboard injection isolation testable check valves is s 10 gpm. s r j FERMI - UNIT 2 3.4-12 Revision 4, 04/02/99

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.y Spec &calwn 3.4.5 TABLE 3.4.3 2-j, REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES [,l MAXIMUM 8 i VALVE UMBER VALVE DESCRIPTION LEAKAGE faom) l I I RHR stem 3)R'g'g,[LO. 4"' 'PCI Lo.o A Injection I ation Valve E 1-F015A L 0.4"' 8 I 11-F015B LPCI Ltop B Injectioylsolation Valve E11-F050 LPC VLoop A Inje In Line Testable Check Valve

1. S Ell-SOB PCI Loop B Inj tion Line Testable N.5% 10 l

Check Val,v6 / 11-F008 Shutdown < Cooling RPV Suction Oufboard l Iso 1'ation Valve / Ell-F 9 Shutdown Cooling RPV Suction Inboard JY j / Isolation Valve / I E -F608 Shutdown Cooling Suction g l / Isolation Valve' / . Core Spray System / /,2 - / ./ E21-F005A Loop A,Iriboard Isolati alve E214005B Loop B Inboard Isolat,1 fin Valve f E21-F006A Loop,A ContainmentjCheck Valve SR [3AM E21-F006B Ltiop B Containm9nt Check Valve '.HighPressure[ Coolant / / Injection, System /- / E41-F007 Pump,D'Ischarge Outbo r j / / Isolation Valv l/ [ j E4'l-F006 , Pump Discharge ard Isolation ve

4. Reactor Core Isolation Cooling' System

/ 1 \\ / / E51-F012 Pum Discharge lation Valve E51-F013 Bdinp Dischar to Feedwater Header

s. N lsolation Valve g,y (a) External Leakage from this valve shall be limited to 5 ml/ min.

g TABLE 3.4.3.2 2 REACTOR COOLANT SYSTEM-INTERFACE VALVf5 LEAKAGE PRESSVRE MONITORS i ALARM SETPOINT VALVE NUMB R SYjilDi fosfa) g,[ Ell-F035A & B. El.14050A & RHR I 449 E11.,F008, F009/7608 R Shutdown ling s 135 EM -F005A 1 X, E21 F0 &B ore Spray s4 ,E41-T006,/F007 HPCI ESI-FD1f. F013 RCIC s 71 '^ FERMI -' UNIT 2 3/4 4-12 Amendment No. J/, SJ, SJ, 98 I PAGE 3 0F 03

lf .L' - ' 5,.. m,. * ',. [ c3[, p l DISCUSSION OF CHANGES l ITS: SECTION 3.4.5 - RCS PRESSURE ISOLATION VALVE (PIV) LEAKAGE 1 i 3 LR.1 CTS 3.4.3.2 Action d 4.4.3.2.3, and Table 3.4.3.2-2 contain Actions, Surveillance details, and a specific list of PIV leakage d4 pressure monitors related to alarm only functions, which are removed from Technical Specifications. These alarm only functions lg are not assumed in any accident analysis. Alarm only functions do not relate directly to the Operability requirements for the Reactor Coolant System. ITS does not specify indication only, alarm-only, or test equipment to be Operable to support Operability of a system or component. Additionally, DECO knows of 9-no Fermi-specific reason these monitors would be required to be 1 retained. 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 alarm-only functions do not relate directly to the Operability requirements for the system or analysis assumptions and the requirement for RCS and PIV leakage limits continues to be required by the Technical Specifications. LR.2 CTS 4.4.3.2.2.b requires that any time the leakage rate of a PIV is affected by maintenance, repair, or replacement, post niaintenance testing is required to demonstrate Operability of the PIV. ITS requires the PIV Operability, but does not direct the performance of testing when reoair activities have been performed. The requirement to perform post msintenance testing is applicable to all plant equipment. The majority of CTS LCOs do not contain this requirement to perform post maintenance testing although this requirement is applicable. Therefore, these post maintenance 4 i testing requirement details can be removed from Technical 6 Specifications consistent with NUREG 1433. Regulatory control of g 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 requirement for PIV Operability continues to be required by the Technical Specifications. FERMI UNIT 2 3 REVISION 4, 04/02/99l

J 9 Vi 2., L - 9 ~ ' 1;. 3 y'. +.

5.,.,.:,

T-RCS PIV Leakage 3.4.5 , ~ ;- 3.4 REACTORCOOLANTSYSTEM(RCS) g RCS Pressur' Isolation Valve (PIV) Leakage 3.4.5 e The leahp from each RCS PIV shall be within limit. [3*i 3.2. LCO 3.4.5 APPLICABILITY: MODES I and 2. MODE 3, except valves in the residual heat removal (RHR) gpoc shutdown cooling flow path when in, or during the transition to or from, the shutdown cooling mode of operation. ACTIONS NOTES - 2. Separate Condition entry is allowed for each flow path. 2. Enter applicable Conditions and Required Actions for systems made OK $

• inoperable by PIVs.

CONDITION REQUIRED ACTION COMPLETION TIME ,chur W A. One or more flow paths NOT E--------- with leakage from one Each alve used to satis 6_qguired Action A.I g,f,3,2 %y or more PCS PIVs not R within limit. 1*ar m and have been verified to meet Ac_IIon C SR 3.4.5.1 l}e -en to te h r nd t 7 ci r po t he s em. .I \\ (continued) Whe fasi nGeling ookgt). orallu -lhe las/ Hme, the VMot was 4is/vrbd, Whichever is mwe t:euni, h m

"7,/.i 5i5 3.4-9

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RCS PIV Leakage 3.4.5 'f SURVEILLt.NCE REQUIREMENTS SURVEILLANCE FREQUENCY <4 A.s. 2n '] SR 3.4.5.1 NOTE----- $_5.*)$[$ $5 $. b" 4 O

• 2-

) Verify equivalent leakage of each RCS PIh In accordance INSE,lt I _- --+ 4e s 0.5 gpm per nominal inch of val _ye size with the

f. jj.6-l -

6 _up to a maximum _of 5 apt at an RCS Inservice }, Testing y 'O [ pressure 2 4 ) and s 4 4.psig ; y Proeiram r 3 s /0 3f /065 " 5 s P./ 4 / in.senT ' 3A.5-2 w.e e BWR/4 STS 3.4-11 Rev 1, 04/07/95

k '.$h_.$ [h Kj h, .C, .,1 M RCS PIV Leakage B 3.4.5 i BASES ACTIONS A.1 -d *.P (continued) c,hf[Q

• 3

. Required Action A.I ' "--"'--d odified by Tr a Note stating that valves used for isolation must meet R I the same leakage requirraents as the PIVs -rf ;;;;i k u, th /AfSER'T~ W ?_t?! ;- _th: H;h ;;r;;;;7: port!= :f th: :;:te g ) b 3'W~ Four hours provides time to reduce leakage in excess of the llowable limit and to isolate the flow path if leakage cannot be reduced while corrective actions to ressat the leaking PIVs are taken. The 4 hours allows time for these actions and restricts the time of operation with leaking valves. u _ Action A.2 specifies that the double i on barrier o lves be restored by cle nother valve qualified for iso or restor e leaking PIV. The p.3 72 hour Completion Time c t e time required to complete the action e Tow pr ity of a second valve failing durt ime period, and probability of a pr ndary rupture of the low press CS pipin vernmeenenart en n -t: ;;r;;:; : (n-r v} B.1 and B.2 If leakage cannot be reduced or the system isolated, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours and MODE 4 within 36 hours. This action may reduce the leakage and also reduces the potential for a LOCA outside the containment. The Completion Times are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without cha11enging plant syster SURVEILLANCE SR 3.4.5.1 REQUIREMENTS Performance of leakage testing on each RCS PIV is requ' ired to verify that leakage is below the specified limit and to identify each leaking valve. h:- k ik.i,e i tit ;f 0.5,i u,h - af a a " a m - - - ^ ^ ' @MEM S M leakage testin[ requife In sh ble ~ pressure condition. For the two PIVs in series, the leakage ] ) (continued) nunf8 STS-B 3.4-25 Rei I, 04/07/ W ^ I m i

si .V g

l:.-

y.;, A, y a .e p JUSTIFICATION FOR DIFFERENCES FROM NUREG 1433 ITS: SECTION 3.4.5 - RCS PIV LEAKAGE FON BRACKETED PLANT SPECIFIC CHANGES P.1 These changes are made to NUREG-1433 to reflect Fermi 2 current licensing basis: including design features, existing license requirements and commitments. 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.~ Additional rewording. reformatting, and revised numbering is made to incorporate these changes consistent with Writar's Guide conventions. Specifically: a. ISTS 3.4.5 Action A provides a generic Note detailing limitations on valves used for isolating a flow path with an inoperable PIV. s j The Fermi 2 CTS provides a Fermi specific current licensing basis note (3.4.3.2 footnote *) with the applicable limitations. These Fermi-specific limitations are incorporated in to the ITS. k replacing the generic limitations. P.2 Bases changes are made to reflect plar,t specific design details. equipment terminology. and analyses. P.3 Bases changes are made to reflect changes made to the Specification. Refer to the Specification change (and associated JFD) for additional detail. P.4 Change made for editorial preference or clarity. v.5 The Bases "def.inition" of RCS PIVs is not an accurate presentation. First, some PIV pairs consist of one normally closed valve and one interlocked-to-close (but normally open) valve. Second, many pairs of in-series, normally closed. valves are within the RCPB. but are no_t PIVs. This inaccurate statement can be eliminated without any loss of appropriate detail. P.6 The reference to the NRC Policy Statement has been rep aced with a more appropriate reference to the Improved Technical Specification " split" criteria found in 10 CFR 50.36(c)(2)(ii). m i3 FERMI - UNIT 2 1 REVISION 4 04/02/99l

e- -RCS Leakage Detection Instrumentation 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Leakage Detection Instrumentation LCG 3.4.6 The following RCS leakage detection instrumentation shall be OPERABLE: a. Drywell floor drain sump flow monitoring system: 2 4l b. The primary containment atmosphere gaseous radioactivity monitoring system; and .k c. Drywell floor drain sump level monitoring system. APPLICABILITY: MODES 1, 2, and 3. ACTIONS ..................................N0TE--- LC0 3.0.4 is not applicable. F^,y CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell floor drain A.1 Restore drywell floor 30 days sump flow monitoring drain sump flow system inoperable. monitoring system to OPERABLE status. B. Required primary B.1 Analyze grab samples Once per containment atmosphere of primary 24 hours gaseous radioactivity containment monitoring system atmosphere. inoperable. (continued) 7: l FERMI UNIT 2 3.4-13 Revision 4, 04/02/99 _4 w

.l

7..

,-g: ,y RCS Leakage Detection Instrumentation 3.4.6 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Drywell floor drain C.1 - ------ NOTE--- - -- 4 sump level monitoring Not applicable when {l system inoperable. primary containment atmosphere gaseous Q-radioactivity monitoring system is inoperable. Perform SR 3.4.6.1. Once per 8 hours D. Primary containment D.1 Restore primary 30 days atmosphere gaseous containment radioactivity atmosphere gaseous monitoring system radioactivity inoperable. monitoring system to OPERABLE status. AND 08 Drywell floor drain sump level monitoring D.2 Restore drywell floor 30 days system inoperable. drain sump level monitoring system to OPERABLE status. E., Required Action and E.1 Be in MODE 3. 12 hours associated Completion Time of Condition A. ANQ

8. C or D not met.

E.2 Be in MODE 4. 36 hours F. All required leakage F.1 Enter LC0 3.0.3. Immediately detection systems inoperable. ,n s l FERMI - UNIT 2 3.4 14 Revision 4 04/02/99

RCS Leakage Detection Instrumentation B 3.4.6 BASES ACTIONS (continued) Action is to restore either of the inoperable monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a degraded configu. ration for a lengthy time period. E.1 and E.2 If any Required Action of Condition A. 5, C. or D cannot be met within the associated Completion Time, the plant must be brought to a MODE in which the LC0 does not apply. To 1 achieve this status, the plant must be brought to at least MODE 3 within 12 hours and MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating-experience, to perform the actions in an orderly manner and without challenging plant systems. El With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and immediate i plant shutdown in accordance with LC0 3.0.3 is required. SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR is for the performance of a CHANNEL CHECK of the recuired primary containment atmosphere gaseous racioactivity monitoring system. The check gives reasonable confidence that'the channel is operating properly. The frequency of 12 hours is based en instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.6.2 This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The test ensures that the monitors can perform their function in I the desired manner. The Frequency of 31 days considers instrument reliability, and operating experience has shown it proper for detecting degradation. I I l FERMI UNIT 2 B 3.4.6-5 Revision 4 04/02/99

NY i 'H l DISCUSSION OF CHANGES ITS: SECTION 3.4.6 - RCS LEAKAGE DETECTION INSTRUMENTATION NUREG-1433. The information moved to the Bases 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 RCS leakage detection system Operability continues to be required by the Technical Specifications. LA.2 CTS 3.4.3.1.b.2 requires operationai requirements for The drywell equipment drain sump level, flow and pump-run time system." which is utilized for monitoring identified leakage rates. ITS 3.4.6 contains only requirements related to meeting Regulatory Guide 1.45 for unidentified leakage. Based on comparison to the ? Technical Specification " split" criteria of 10 CFR 50.36(c)(2)(ii) 9 the drywell equipment drain system will be relocated to the l[ Technical "equirements Manual (TRM), which require revisions to be Q controlled by 10 CFR 50.59. The drywell equi'pment drain sump level, flow and pump run time system is not a part of a primary success path in the mitigation of a DBA or transient. and is a non-significant risk contributor to core damage frequency and offsite releases. This relocation continue to provide adequate protection of the public health and safety since the ITS retain sufficient requirements related to controlling the magnitude of the identified leakage rate. TECHNICAL CHANGES - LESS RESTRICTIVE Speci fi c" L.1 CTS 3.4.3.1 Action requires two leakage detection systems to remain Operable when allowing a 30-day restoration time. ITS Action statements are provided whigh allow unlimited continued operation with either required primary containment atmosphere gaseous radioactivity monitor or drywell floor drain sump level monitoring system inoperable. In either case the primary system for identifying and quantifying unidentified leakage in the i containment (i.e., the drywell floor drain sump flow monitoring system) remains OPERABLE. Since the inoperable system (s) provide actions for more frequent leakage measurements (ITS 3.4.6 Required h Actions B.1 and/or C.1), operation is alloed to continue with the system (s) inoperable. However. with both backup systems inormble, this continued operation is limited to 30 days since the diversity is significantly reduced. t FERMI UNIT 2 2 REVISION 4 04/02/99l

[. c v RCS Leakage Detection Instrumentation 3.4.6 i 3.4 REACTOR COOLANT SYSTEM (RCS) lC ~15 \\ 3.4.6 RCS Leakass Detection Instrumentation N / S LCO 3.4.6 The following RCS leakage detection instrumentation shall be y OPERABLE. Drywell floor drain sumpfmo o ing system; 44adF(7.4.J.l.b a. b. 0 r " ---'

---

primary containment 41mospher13 j 4 3-l-

?' @d~ tTeslith ats:osphe gaseous monitoring _ system; (3.

jan e.

ra ioacjivily) c. rimyrf co jainoen 1r cooler Sprftfensafrflpriate]/ 4.3.l.CD n n1torjaf.,syst r 7 \\, / ' Or f" 'O floor drain sang "j 538b' APPLICABILITY: MODES 1, 2, and 3. E#

1. 1. 4 ACTIONS <

N \\ CONDITION REQUIRED ACTION COMPLETION TIME \\ A. Drywell floor drain

---

NOT E-------- 3 3, 4,3, l sumptmonitoring system LCO 3.0.4 is not applicable.

inopera e. c----------------------, 4c-{., ort i -flow A.1 Restore drywell floor 30 days drain sungmonitoring p,l system tof0PERABLE status. (continued) 6.e-- 3.4-12 R, ;, ^4/sifs5

s..,..

y ,m. -..-..w

RCS Leakage Detection Instrume-tation B 3.4.6 BASES (continued) _e sasee>5 SURVEILLANCE SR 3.4.6.1 r o dlorac h.VI fV j REQUIREMENTS This SR is for the performance of a CHAllNE CHECK of the required primary containment atmosphe onitoring system. The check gives reasonable confidence at the channel is operating properly. The Frequency of 12 hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.6.2 This SR is for the performance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The test ensures that the monitors can perform their fun the desired y nn.tr 4 - -.==>- -....-...- - -..ction in O h.1 G~edoint-and-rehtiw ecrureci ef th: fr.:tr =r.t :trir,- L/ b The Frequency of 31 dayiconsiders instrument rel

ity, and operating experience has shown it proper for detecting degradation.

3R 3.4.6.3 This SR is for the performance of a CHANNEL CALIBRATION of required leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. The Frequency of118kmonths is a typical refueling cycle and f considers channel reliability. Operating experience has proven this Frequency is acceptable. E ??2NCES 1. 10 CFR 50, Appendix A, GDC 30. 2. Regulatory Guide 1.45 1973. 3.hSAR,Section45.2.7. 4. GEAP-5620, April 1968. 5. NUREG-75/067, October 1975.

6. hSAR, Section [L2v7-r5:23-{

BWR/4 STS B 3.4-32 Rev 1, 04/07/95

w ' JUSTIFICATION FOR DIFFERENCES FROM NUREG 1433 ITS: SECTION 3.4.6 RCS LEAKAGE DETECTION INSTRUMENTATION. NON BRACKETED PLANT SPECIFIC CHANGES ~ P.1~. These changes are made;to NUREG 1433 to reflect Fermi 2 current licensing basis: including design features, existing license requirements and commitments. Additional rewording. reformatting, and revised numbering is made to incorporate these changes consistent with Writer's Guide conventions. Refer to CTS Discussion Of Changes to the relat'ed requirement for a detailed justification of' changes made.to the current licensing basis which are also reflected in the ITS as presented. Some of these changes are specifically discussed below:

a. The ISTS bracketed options reflect designs that include leakage detection from containment air cooler condensate flow rate. Fermi-2 design does not include containment air cooler condensate flow rate as leakage detection instrumentation. Revi; ions reflect the ISTS intended Actions for removal of this option.

P.2 Bases changes are made to reflect plant specific design details, equipment terminology, and analyses. Some of these changes are j specifically discussed below: l a. The CHANNEL FUNCTIONAL TEST (ITS SR 3.4.6.2) ensures that the channel performs its intended function, but does not verify the setpoint of all alarms, timers, and switches. These setpoints h would be the subject of the CHANNEL CALIBRATION. As such, the Bases sentence referencing verification of the alarm setpoint and string accuracy is deleted. This reflects the existing Fermi-2 implementation of the CHANNEL FUNCTIONAL TEST and is consistent with the Bases description for other instrument channels. 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 ~ The reference to the NRC Policy Statement has been replaced with a more appropriate reference to the Imoroved Technical Specification " split" criteria found in 10 CFR 50.36(c)(2)(ii). GENERIC CHANGES C.1' TSTF-60: NRC approved change to NUREG 1433. ' FERMI . UNIT 2 1 REVISION 4 04/02/99l I

.y }.{ Q: $Y Y.b'n- ,v( { CClfCQ. A s i REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION (Continb?d) ( j [ ( { ACTION: (Continued) c. In OPERATIO L CONDITION or 2, with: 1 1. T L POWER nged by more an 15% of RA THERMAL P WER n I hour *, The o gas level, t6edelaypip.e nereased b re than se rates les 'f our during. ady-state f 10 microcuri r second in eration at r an 75,000 ni ocuries per i /second,or / 3. The of a level, at. e delay pipe, i reased by more an 15% ne hour dur g' steady-state o ration at rele rates r than 75,0 microcuries per ~ econd,' / orm the sampi and analysis r, trements of em 4b of htle 4.4.5-1 until' the specific activit' of the pri y coolant is restored to within its limit. SURVEILLANCE RE0VIREMENTS k. %R -4r4 4 The spt:ific activity of the reactor coolant shall be demonstrated to 3.A.7.l be within the limits by performance of the sampling and analysis program of i Table 4.4.5-1. E LM Gat,RpurhrewohAS) 4 h:j FERMI - UNIT 2 3/4 4-17 Amendment No. 6 PAGE c9 0F 03

[ 5 P s N Q;,. d~ i yM', s m N. +N '$ g f ia n' p t MI %m ME' D@k3 wg m 1 g 88'd I j, f ( SN D 3 b O E h 4 r I R o TEI t ILU DPQ 4 ca NME OAR e CS r LHl 3 e ACS cn NIY k, i 0HL s s 1WA t T N h d i ANA J 2 e m RI s i E S p l P I 1 1 a O A k s N e t \\ A i R e G v n O a i R h h P t n N i h S ee O w I e rd I e, S w oe T o s Y r t. t r A t L ib Ri R A e m nEu E d N s h hciN iWq P e A y ilO ,wO O r % M eoP o a 4f I D i aT ll R t N rc C plL s E s A eesA mo a W e 2 ppd af O r E S 7 sey s P M I L r r e eb s l s S P e e cec er e. f i Y 1 M p p nhxd nu vc o L A ot ee o e m AY 5 S s e e e r nlN fd NC O y t t r yi t i AN c b Y 4 set u s sI a s n E T y a vi q eat DU o c 4 I eeve dgg s NQ V AE 0 l nir nn-t. a E I 2 t et aa R s s L T thcs t y n EF a a B C Awaa A2. ob d a A A L e e Q n l l I P z T a o M C o A t ) A a 0 c I S ea F P ~ I y F C r E E 1 P 2r m S 4 ei fa r T on p n N o 3 o M A A 3 ml e { M L 1 u h n O mr t o e 8 e io O i n X n f a C -t t is o Ea i d. Y m-. d Sr R a

a. y I

Ot e o A Dn t I u a it a Q h M e e A r r v I M rc R o e8 i l on P t t4 t m fo f ' a f c t A m C s ar a a s t i r8 o o i1 s i o8 nf c s3 i y f cf - e i T y1 l n r klf [o N l l l I sK aa i E al a a M n c n t t c c e ES AT i A c en i e A l ; et p h RI N US 'e i c M bi s r SY l l i oc e d p AL oA EA oV o o d 3 tbh >1 u t e I MN tI t e o a - es r A oU F sQ a s r ee l l i OD IE I MX pti u N 9 m t P rr aW*n q EA [ U e e Y R 3 4 { (3 T j ( ?E hzM 7 b, ) a6 R= ( / f47 3 'e 3 f K 3R c c' G M O'T Oy 1 i i

r-DISCUSSION OF CHANGES ITS: SECTION 3.4.7 - RCS SPECIFIC ACTIVITY i ADMINISTRATIVE A.1 In the conver'sion of the Fermi 2 current Technical Specifications (CTS) to the proposed plant specific Improved Technical Specific'ations (ITS), certain wording preferences or conventions are adopted which do not result in technical changes (either actual or. interpretational). Editorial changes, ref.ormatting; and. revised numbering are adopted to make the ITS consistent with the Boiling Water Reactor (BWR) Standard Technical Specifications NUREG 1433. Rev. 1. in li A.2 (Not used.) g A.' 3 CTS Table 4.4.51 Item 5. requires " Isotopic Analysis of an Offgas Sample. " ITS 3.4.7 does not retain this requirement. This is acceptable because the CTS requirement is the same as ITS SR 3.7.5.1 which requires the same sample to be.taken every 31 days. Therefore, this is an administrative change with no impact on i safety because it only eliminates a duplicate requirement contained in CTS. TECHNICAL CHANGES MORE RESTRICTIVE M.1 CTS Table 4.4.51 Item 2 requires that isotopic analysis for Dose Equivalent I 131 concentration be made at least once per 31 days. ITS SR 3.4.7.1 requires that this sample be taken once every 7 days. The increased surveillance frequency provides a compensatory measure for the removal of the requirement that gross specific activity remain less than or equal to 100/E bar gCi/ gram. This more restrictive change will have no negative impact on. safety, based on the fact that the increased sampling will provide an earlier detection of a degraded condition. m: FERMI - UNIT 2 1 REVISION 4 04/02/99 l 3

n .,,g..gyjgt..;#,. 's .my .L DISCUSSION OF CHANGES ITS: SECTION 3.4.7 RCS SPECIFIC ACTIVITY U TECHNICAL' CHANGES - LESS RESTRICTIVE " Generic" None TECHNICAL CHANGES -'LESS RESTRICTIVE " Specific" L.1 ' CTS 3.4.5 requirement to maintain specific activity s 100/E-bar g pCi/gm and Surveillance for gross Beta and Gamma activity, as well as the radiochemical analysis for E bar (Table 4.4.51. Items 1 'Sk and 3) have been deleted. The CTS Bases state that the intent of the requirement to limit the specific activity of the reactor coolant is to ensure that whole body and thyroid doses at the site boundary would not exceed a small fraction of the limits stated in 10 CFR 100 (i.e.,10% of 25 rem and 300 rem. respectively) in the event of a main steam line failure outside containment or an instrument line break. To ensure that offsite thyroid doses do not exceed 30' rem. reactor coolant dose equivalent I 131-(DEI) is limited to less than or equal to 0.2 pCi/gm. CTS also limits reactor coolant gross specific activity to less than or equal to 100/E bar pCi/gm to ensure that whole body doses do not exceed 2.5 rem. CTS LC0 3.11.2.7 (ITS LCO 3.7.5) associated with radioactive - effluents requires that the gross gamma radioactivity rate of the i noble gases Xe 133. Xe 135. Xe 138. Kr-85m. Kr 87. and Kr 88 measured at the main-condenser evacuation system pretreatment monitor station be limited to less than or equal to 340 mci /second. The CTS Bases for LC0 3.11.2.7 state that ~ ~ restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total-body exposure to an individual at the exclusion area boundary will not exceed a small fraction (10%) of the limits of 10 CFR 100 in the event this effluent is inadvertently discharged without treatment directly to the environment. The Main Condenser Offgas Treatment System.' as re gired by CTS LC0 3.11.2.7 (ITS LC0 3.7.5) provides reasonable assurance.the reactor. coolant gross specific activity is maintained at a sufficiently ) low level-to preclude offsite. doses from exceeding a small fraction.of the limits of 10'CFR 100 in the event o.f a main steam j line. failure. Therefore. CTS LC0 3.4.5.b is redundant and places an unnecessary burden on the licensee without a commensurate l FERMI -l UNIT 2 '2 REVISION 4 04/02/99l

x

.t A.w S

t. .:; e p", i 6 .de 9 L., i-Q 13. - DISCUSSION OF CHANGES ITS: SECTION 3.4.7 RCS SPECIFIC ACTIVITY s. increase in safety. Elimination of CTS LC0 3.4.5.b and i Surveillance Items 1 and 3 of Table 4.4.5-1 will allow plant g personnel to focus attention on efficient. safe operation of the plant without the unnecessary distraction of the redundant Surveillance Requirement. Additional assurance that the offsite doses will not exceed a small fraction of the 10 CFR 100 limits is provided by increasing the frequency of sampling and analysis of the reactor coolant for DEI from at least once per 31 days to at least once per 7 days. Since (1) the reactor coolant limit on DEI adequately assures that offsite doses will not exceed small fractio.ns of the limits of 10 CFR 100 in the event of a main steam line failure outside containment and (2) gross gamma radioactivity rate of the noble gases measured at the condenser evacuation system pretreatment monitor station is limited by ITS 3.7.5 to a value that provides reasonable assurance the reactor coolant gross specific activity is maintained at a sufficiently low level to preclude offsite doses from exceeding a small fraction of the limits of 10 CFR 100, t the requirements associated with LC0 3.4.5.b and Surveillance 4 Items 1 and 3 of Table 4.4.5-1 are unnecessary. The associated Actions and Surveillance Requirements are also being deleted. consistent with the LC0 requirement deletion. L.2 CTS 3.4.5 Applicability includes Operational Condition 2. 3. and 4. ITS 3.4.7 Applicability is limited to only those conditions which represent a potential for release of significant quantities of radioactive coolant to the environment. Mode 4 is omitted since the reactor is not pressurized and the potential for leakage is significantly reduced. In Modes 2 and 3, with the main steam lines isolated, no escape path exists.for significant releases and requiremencs for limiting the specific activity are not required. The Required Actions are also modified to reflect the new Applicabi.ity, and an option for exiting the applicable Modes is provided for cases where isolation is not desired. Based on the fact that the ITS Applicability is consistent with plant conditions where event consequences are significant, this less restrictive change will have a negligible impact on safety. 7 FERMI - UNIT 2 3 REVISION 4 04/02/99l

DISCUSSION OF CHANGES ITS: SECTION 3.4.7 - RCS SPECIFIC ACTIVITY L.3 CTS 3.4.5 Actions do not allow Mode changes while Dose Equivalent I-131 is not within limit (CTS 3.0.4). ITS 3.4.7 Action A provides a Note to the Required Actions to indicate that LC0 3.0.4 is not applicable. Entry into the applicable Modes should not be restricted due to the significant conservatism incorporated into the specific activity limit the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient specific activity excursions while the plant remains at, or proceeds to power operation. Therefore, operation during the allowed time frame would not represent a significant impact to the health and safety of the public. This less restrictive change will have a negligible impact on safety. L.4 CTS 3.4.5 Action c requires increased sampling of Item 4b of CTS Table 4.4.5-1. CTS 3.4.5 Action c associates this sampling with 8 changing plant conditions. However, as required by CTS 3.0.1 i this Action applies only when the LC0 is not met. Similarly. CTS s 3.4.5 Act.on b requires the sampling of Item 4a of Table 4.4.5-1 when the LC0 is not met. Since sampling of Item 4a provides D sampling requirements equivalent to Item 4b. the Action c requirement (and associated Item 4b) are unnecessary. Therefore, s elimination of the duplicative requirements of CTS Action c (including associated footnote *) involves no impact on safety. RELOCATED SPECIFICATIONS None TECHNICAL SPECIFICATION BASES The CTS Bases for this Specification have been replaced by Bases that reflect the format and applicable content of ITS 3.4.7 consistent with the BWR STS. NUREG-1433. Rev. 1. FERMI - UNIT 2 4 REVISION 4 04/02/99 v e Y.

,J - .;. g $f S[ f.] I,'. J..]: [" 9 NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.7 - RCS SPECIFIC ACTIVITY i TECHNICAL CHANGES LESS RESTRICTIVE f (Soecification 3.4.7 "L.4" Labeled Comments / Discussions) 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 significart hazards consideration. The bases fe 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 j evaluation are presented below. 1. Does the change involve a significant increase in 'the probability or consequences of an accident previously evaluated? The proposed change eliminates the requirements for increased sampling of isotopic analysis for iodine (which is stated in terms of changing plant conditions. but required only by CTS 3.4.5 Action c and thus applicable only when the associated LC0 is not met). Equivalent sampling requirements, also applicable when the LC0 is not met, are g included in CTS Action b. The removal of an action covered by an equivalent action is not considered as an initiator of any previously Q evaluated accident, nor will it have any impact on the consequences of N any previously evaluated accident. Therefore, the proposed change will not increase the probability of any accident previously evaluated and will not increase the consequences of any accident previously evaluated. 2. Does the change create the possibility of a new or different kind of j accident from any accident previously evaluated? i The proposed change does not introduce a mode of plant operation, not previously analyzed, and does not involve a physical modification to the plant. Therefore it does not create the possibility of a new or different kind of accident from any accident previously evaluated. 3. Does this change involve a significant reduction in a margin of safety? i This change does not involve a significant reduction in a margin of safety si' ice no limitation, operating parameter, or analysis assumption is modified. Therefore, no question of safety is involved and these changes do not involve any reduction in a margin of safety. FERMI - UNIT 2 7 REVISION 4 04/02/99

Q

l

. y: y: Q ?." - l_ ,n DISCUSSION OF CHANGES ITS: SECTION 3.4.8 - RHR SHUTDOWN COOLING SYSTEM - HOT SHUTDOWN .3 d TECHNICAL CHANGES - MORE RESTRICTIVE None TECHNICAL CHANGES - LESS RESTRICTIVE Generic" LA.1 CTS 3.4.9.1 LCO details system design. ITS 3.4.8 requires that RHR SDC be Operable, but does not detail the system design. This is acceptable because the system design does not impact the requirement to have the RHR SDC system Operable. Therefore. these details can be relocated to the Bases. This change is consistent with NUREG 1433. The information moved to the Bases 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 RHR SDC Operability continues to be required by the Technical Specifications. l LR.1 CTS SR 4.4.9.1.2 requires that "At least once per 12 hours verify the required RHR shutdown cooling mode loop (s) are capable of taking suction from the reEctor vessel through the RHR heat exchanger (s) with their associated cooling water available." This is essentially equivalent to verify the required RHR subsystem is Operable." ISTS surveillances prescribe specific acceptance ~ criteria or require verification of a specific feature: but ISTS T surveillances do not re' quire non-specific verification of syste.m

t-Operability without specifying details for this' verification.

M Tracking the status of system Operability is an ongoing activity. 1 Therefore, the requirements of CTS SR 4.4.9.1.2 are removed from lg Technical Specifications consistent with the 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 requirement for RHR Shutdown Cooling system Operability continues to be required by the Technical Specifications. FERMI - UNIT 2 2 REVISION 4 04/02/99 l

DISCUSSION OF CHANGES ITS: SECTION 3.4.9 - RHR SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN U ADMINISTRATIVE A.1 In the conversion of the Fermi 2 current Technical Specificialons (CTS) to the proposed plant specific Improved Technical . Specific'ations (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 ITS 3.4.9 adds a Note to the Actions that " Separate Condition entry is allowed for each RHR shutdown cooling subsystem." This Note provides explicit instructions for proper application of the Actions -for Technical Specification compliance. In conjunction with the ITS 1.3 " Completion Times." this Note provides direction consistent with the intent of the CTS Actions. Therefore. this is an administrative change with no impact on safety. a A.3 CTS 3.4.9.2 Applicability is stated as Operational Condition 4 (i.e.. Cold Shutdown), but is further modified with " irradiated fuel is in the reactor vessel and the water level is less than g

20. feet 6 inches above-the top of the reactor pressure vessel

~ fl ange... " This additional modifier is eliminated with no-k resultant change in requirement or interpretation. The Cold Q Shutdown condition is defined such that there is fuel in the vessel and the reactor vessel head is bolted in place. With the head on, the reactor cavity is not flooded and water level. is not (and could not be) 20' feet 6 inches above the vessel flange. Therefore, eliminating this explicit modifier will not change the requirements, and is considered adrainistrative. TECHNICAL CHANGES - HORE RESTRICTIVE f l M.1 CTS 3.4.9.2 Action d requires "The provisions of Specification 3.0.4 are not applicable for up to 4 hours for the purpose of 1 establishing the RHR system in the shutdown cooling mode once the reactor vessel pressure is less than the RHR cut in permissive setpoint." Since no benefit or rationale could be determined for this Action, it has been deleted. FERMI - UNIT 2 1 REVISPy! %, 04/02/99l 1 )

h N )); h 4 yi k,D e i U DISCUSSION OF CHANGES ITS: SECTION 3.4.9 - RHR SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN TECHNICAL CHANGES - LESS RESTRICTIE " Generic" LA.1 CTS LC0 3.4.9.2 details the system design. ITS 3.4.9 requires the RHR Shutdown Cooling System to be Operable, but does not provide details as to the system design. This is acceptable because the system design details do not impact the requirement to maintain the RHR Shutdown Cooling System Operable. This change is consistent with NUREG 1433. The information is being moved to the l 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 RHR Shutdown Cooling System Operability continues to be required by the Technical Specifications. LA.2 CTS 3.4.9.2 and 4.4.9.2.2 during Mode 4 operation with the reactor cavity not completely flooded requires reactor water level to be maintained 2 214 inches in addition to a means of forced circulation. CTS 3.9.11.2 A.ction c requires two means of forced circulation from either two recirculation pumps or two RHR SDC loops in the event level is not maintained 2 214 inches. ITS 3.4.9 does not retain these restrictions -- only requiring one means of forced circulation regardless of reactor water level. T These will be relocated to the Technical Requirements Manual 0 d i (TRM). This is consistent with the NUREG 1433. This water level requirement can be adequately defined and controlled in the TRM. g which requires revisions to be controlled by 10 CFR 50.59. This relocation continues to provide adequate protection of the public health and safety since the requirement for forced circulation. and alternate means of reactor coolant circulation on loss of 4 forced circulation. continues to be required by the Technical Specifications. e. FERMI UNIT 2 7. REVISION 4 04/02/99l

j y

9).p....

h. ? t.j @.' . k i DISCUSSION OF CHANGES ITS: SECTION 3.4.9 - RHR SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN LR.1 CTS SR 4.4.9.2.3 requires that "At least once per 12 hours verify the required RHR shutdown cooling mode loop (s) are capable of taking suction from the reactor vessel through the RHR heat exchanger (s) with their associated cooling water available." This is essentially equivalent to " verify the required RHR subsystem is - Operable." ISTS surveillances prescribe specific acceptance criteria or require verificat. ion of a specific feature: but ISTS s surveillances do not require non specific verification of system g Operability without specifying details for this verification. s Tracking the status of system Operability is an ongoing activity. Therefore, the requirements of CTS SR 4.4.9.1.2 are removed from l3' Technical Specifications, consistent with the 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 requirement for RHR Shutdown Cooling system Operability continues to be required by the Technical Specifications. '7 TECHNICAL CHANGES - LESS RESTRIGTIVE " Specific" None RELOCATED SPECIFICATIONS None TECHNICAL SPECIFICATION BASES The CTS Bases for this Specification have been replaced by Bases that reflect the format and applicable content of ITS 3.4.9 consistent with the BWR STS, NUREG 1433. Rev. 1. t FERMI UNIT 2 3 REVISION 4, 04/02/99l~

Q'.. m.'b.a. ;?y '!: f. I ^ .: O.( p.. i. $." [ RCS P/T Limits 3.4.10 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. --NOTE --- C.1 Initiate action to Immediately Required Action C.2 restore parameter: shall be completed if to within limits. this Condition is' entered. AND C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LC0 not met in other operation, or 3 than MODE! 1, 2, and 3. SURVEILLANCE REQUIREMENTS SURVEILLANCE FRE0VENCY SR 3.4.10.1 -- -NOTE -- -- ------- Only required to be performed as applicable during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing. Verify: 30 minutes a. RCS pressure and RCS temperature are 3j to the right of the limits specified in Figure 3.4.101: and k b. RCS heatup and cooldown rates are limited to: 1. s 100"F in any 1 hour period: and 2. s 20*F in any 1 hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. i (continued) o I FERMI UNIT 2 3.4-24 Revision 4, 04/02/99

i n., Y( [ i.I 3- ]lA $6 , [- g Specsnesma 3.y.1o ..t REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITTON FOR OPERATION t_C0 3.4.10 -2.'.'_1-The reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1-1 (1) Curve A for N M 0*I A 1 hydrostatic or leak testing;-(2) Curve B for heatup by non-nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS: and (3) Curve C for operations with a critical core other than low power PHYSICS SR 3.4 lo.'2-ITESTS, with: q a. A maximum heatup of 100*F in any 1-hour period. 9 3 4.io.L.b.) hb. A maximum cooldown of 100'F in any 1-hour period, c. A maximum temperature change of less than or equal to 20*F in any 4e

0. \\.b,1 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, and 10 W

d. The reactor vessel flange and head flange temperature greater than 4 $R 3.v.'D. 7 or equal to 71'F when reactor vessel head bolting studs are under tension. .tPP' IC ABILITY: At all times, ...( R.k. C I) yy 'N With an, )f the above limits exceeded. ore the temperature and/or pressure bl 9 C'I to within the limits within(30 minute $: Dertorm an engineering evaluation ?,cangco /tcn.a determine the ettects of tne out-or-timit ccnoition on tne structurai g,l A.t 4 c.2-integrity of the reactor coolant system; determine that the reactor coolant o system remains acceptable for continued operationsI51' Be in at least HOT j}cg/gJ 8-SHUTDOWN within 12 hours and in COLD 5HUIDOWN within the following 24 hours. g ADD ActoN A' dote \\ N SURVEfttANCE REOUTREMENTS

• Acted C M DTE /

i <r

1. . 4.C.I.1 - During system heatup, cooldown and inservice leak and hydrostatic 6A 3M.to l testing operations, the reactor coolant system temperature and pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1-1 Curves A, B, or C, as applicable, at least once per 30 minutes.

..~ FERMI - UNIT 2 3/4 4-19 Amendment No. 77 PAGE b 0F 08 6" Y

1 f.; :n:,.s 'se .,'.r. 4.; DISCUSSION OF CHANGES ITS: SECTION 3.4.10 RCS P/T LIMITS 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 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 3.4.1.4.a. 3.4.1.4.b. and 4.4.1.1.4.b and c address a 50*F limitation between the idle. loop and either the " reactor coolant" or an " operating recirculation loop." ITS SR 3.4.10.4 and SR 3.4.10.6 presents these limitations together with the idle loop compared to "RPV coolant temperature." Since the operating loop temperature is cnnsidered synonymous with RPV coolant temperature. this change is considered an editorial presentation preference. As such, the stated requirements for "both loops have been idle" Q (CTS 3.4.1.4.a) and for "only one loop has been idle" (CTS 3.4.1.4.b) will be the same and separate discussion is no longer Q necessary. Th.ese will be combined in ITS SR 3.4.10.4. A.3 CTS 3.4.1.4 Action requires " suspend startup of any idle recirculation loop" if any required differential temperature is not within limit. ITS SRs 3.4.10.3 and 3.4.10.4 present the same required differential temperatures. however, with the requirement that SRs be met (ITS SR 3.0.1) the action to preclude startup with unacceptable temperatures is inherent without an explicit action. Therefore, elimination of this CTS Action is an administrative presentation preference. A.4 CTS 3.4.6.1 Action for any out of limits condition requires " restore...within 30 minutes." ITS 3.4.10 Required Action C.1 for Modes 4 and 5 states " initiate action to restore... Immediately." The CTS Action provides 30 minutes in which pressure and temperature requirements could exceed the limits, and if the parameters are incapable of being restored to within the limits within 30 minutes. provides no further action. The intent of the action is more appropriately presented in ITS Required Action C.1 which provides for tuilt inuous Actions during all out of-limits conditions in Mode 4 and 5. As an enhanced presentation of the CTS intent, the change is deemed to be administrative. FERMI UNIT 2 1 REVISION 4 04/02/991

~ ..t, H'nf.., n E,' W, . w. - 4 . c. DISCUSSION OF CHANGES ITS: SECTION 3.4.1.0 -'RCS P/T LIMITS b A.ii (Not_used) lk ~ A.6 CTS 4.4.6.1.4.b requirement to verify the vessel flange and head flange temperature within 30 minutes prior to" tensioning of the head _ bolting studs has been deleted. This requirement is a duplication of CTS 4.0.4 and ITS SR 3.0.4. which' require the I Surveillances to be' current prior. to entering the condition in which the.SR.is ' applicable. Therefore, this change is an administrative change with no impact on safety. TECHNICAL CHANGES MORE RESTRICTIVE M.1 CTS 3.4.6.1 Action requires an evaluation to determine that the RCS remains acceptable for continued operation. -ITS imposes a specific limitation for the engineering evaluation with a specific completion time. The Completion Time of 72 hours is considered reasonable for operation in Modes 1. 2. and 3 (Required Action A.2) because the P/T. limits represent long term vessel fatigue and usage factors. In Modes 4 and 5. the Completion Time of "."rior.to entering MODE 2 or 3" (Required Action C.2) specifically gevents entry into operating Modes. This.is consistent with the CTS intent in applying LCO 3.0.4 The'open ended time limit on completing the evaluation when the unit is shutdown is a g reasonable interpretation of the CTS requirement, since it is not t expected that a P/T limit violation while shutdown would present an immediate threat to the RCS integrity. As a specific ( limitation consistent with the NUREG. and an enhanced presentation of the CTS intent.. the change will' not introduce any adverse impact on safety. Additionally, the CTS Actions are ' interpreted to imply that the . required evaluation must be completed regardless of the restoration. ITS Notes to Condition'A and Condition C provide explicit clarification of this' intent. As'an enhanced presentation of the CTS intent, the change is deemed to be administrative (and discussed here for completeness). 5 FERMI . UNIT 2-2 REVISION 4 04/02/99I

(.' ' ?. - bj ~ i g ~.... ',., ' 'N I DISCUSSION OF CHANGES f ITS: SECTION 3.4.10 - RCS P/T LIMITS l TECHNICAL CHANGES - LESS RESTRICTIVE " Generic" LA.1 CTS 4.4.6.1.3 and Table 4.4.6.1.3 1 details the schedule for removal of the reactor vessel material surveillance specimens, and states the function.of the examination results is to update P/T curves. -ITS does not retain these details. The requirement for an NRC approved withdrawal schedule and for use of the specimen results are contained in 10 CFR 50. Appendix H. and therefore, are not necessary to be repeated in the ITS to provide adequate protection of the public health and safety. LA.2 CTS 3.4.6.1 Agtions state the requirement to " perform an engineering evaluation to determine." ITS 3.4.10 Required Actions A.2 and C.2 state " determine." The understanding that the h determination is made by performance of an en41neering evaluation [ is relocated to the Bases, which requires changes to be controlled g in accordance with the ITS 5.5.10. Bases Control Program. This y relocation continues to provide adequate protection of the public health and safety since the requirement to " determine" continues to be required by the Technical Specifications, j LR.1 Not used. FERMI UNIT 2 3 REVISION 4 04/02/99 l

,.;.,,2 % u : ~~

2..

RCS P/T Limits 3.4.10 a INSERT 3.4.10 1 SURVEILLANCE REQUIREMENTS SURVEILLANCE' FREQUENCY SR 3.4.10.1 Verify: 5 e4 a. RCS pressure and RCS temperature are to the right of the limits specified in p Figure 3.4.10-1: and '9

b. RCS heatup and cooldown rates are limited to:

1. s 100 F i.n any 1 hour period: and

2. s 20*F in any 1 hour period during inservice hydrostatic and leak testing operations above the heatup and

' cooldown limit curves. FERMI - UNIT 2 Page 3.4-24 (INSERT) REVISION 4.04/02/99l

3 g.. p;. p. l l3 Specificafion 3.4.ll

i. 1[;

REACTOR COOLANT SYSTEM ,{ REACTOR STEAM DOME llMITING CONDITION FOR OPERATION ( (or ejad % -3r4:672 4g The pressure in the reactor steam dome shall be less than+1045 psig. l APPLICABILITY: OPERATIONAL CONDITIONS l b d 29 y,l ACTION: /(c.hort/ With the reactor steam dome pressure exceeding 1045 psig, reduce the pressure te less than 1045 psig within 15 minutes [6 rte in at least HOT SHUTDOWN within MC SR N /., $URVEftlANCE REOUTREMENTS ,p sg 3. 4.11.1 L.I -1.-'..S.2-The reactor steam dome pressure shall be verified to be less than / Qc 1045 psig at lea:t once per 12 hours. l l Notqpp11 cable during-antikated transhnts { FERMI - UNIT 2 3/4 4 23 Amendment No. 87 f" N PAGE I 0F 01

4 e" i e ~ l DISCUSSION OF CHANGES ITS: SECTION 3.4.11 - REACTOR STEAM DOME. PRESSURE 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 techn.ical 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. b A.2 Not used. ( IECHNICAL CHANGES - MORE RESTRICTIVE M.1 CTS 3.4.6.2 Applicability footnote identifying that the reactor steam dome pressure limit is not applicable during anticipated transients is deleted. ITS 3.4.11 is constructed with the reactor steam dome pressure limit as an operational limit. It is assumed that anticipated transients will result in exceeding operational l limits. It is further assumed that action (either automatic or operator, or both) is initiated to return the plant to acceptable steady state operation. No benefit or rationale is seen for the CTS exception. Therefore, the reactor steam dome pressure limit is applicable during such transients, and the Required Actions should be taken as part of the mitigation of the transient. This more restrictive change will have no negative impact on safety, since it elirinates a CTS exception which could be misinterpreted. TECHNICAL CHANGES - LESS RESTRICTIVE

• • Generic" None

~ )J FERMI UNIT 2 1 REVISION 4 04/02/99l

DISCUSSION OF CHANGES ITS: SECTION ?.4.11 REACTOR STEAM DOME PRESSURE TECHNICAL CHANGES - LESS RESTRICTIVE " Specific" L.1 CTS 3.4.6.2 and 4.4.6.2 requires t'he reactor steam dome pressure to be "less than 1045 psig," however, the associated Action applies with reactor steam c;ome pressure " exceeding 1045 psig." The requirement associated with pressure exactly equal to 1045 psig is not specifically addressed. ITS 3.4.11 limit on reactor steam dome pressure is specified as less than or equal to 1045 psig, which agrees with the Fermi-2 vessel overpressurization analyses. Since the difference is infinitesimal, and resolves an obvious discontinuity consistent with the CTS Action presentation. { the change will not result in any significant impact on ' safety. 1 RELOCATED SPECIFICATIONS None ) t TECHNICAL SPECIFICATION BASES The CTS Bases for this Specification have been replaced by Bases that reflect the format and applicable content of ITS 3.4.11 consistent with the BWR STS, NUREG 1433. Rev. 1. FERMI - UNIT 2 2 REVISIDN 4, 04/02/99l

NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.11 - REACTOR STEAM DOME PRESSURE TECHNICAL CHANGES - LESS RESTRICTIVE (Soecification 3.4.11 "L.1" Labeled Comments / Discussions) 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 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 evaluation are presented below. 1. Does the change involve a significant increase in the probability or' consequences of an accident previously evaluated? The proposed change resolves an obvious discontinuity between the LC0 statement and its surveillance, and the associated Action statement. The operating limit on reactor steam dome pressure is not considered as an initiator for any previously evaluated accident. Therefore, the-proposed change will not increase the probability of any accident previously evaluated. The difference is infinitesimal, and resolves an obvious discontinuity in o manner consistent with the Fermi-2 vessel overpressurization analyses and the CTS Action presentation. Therefore. the change will not significantly increase the consequences of any accident previously evaluated. 2. Does the change create the possibility of a new or different kind of l accident from any accident previously evaluated? 1 This proposed change will not involve any physical changes to plant systems, structures, or components (SSC), or measurable 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. FERMI UNIT 2 1 REVISION 4 04/02/99 i I

T I N0 SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.4.11 - REACTOR STEAM DOME PRESSURE j. p TECHNICAL CHANGES LESS RESTRICTIVE .LSpg.cification 3.4.11 "L.1" Labeled Comments / Discussions) i 3. Does this change involve a significant reduction in a margin of safety? The proposed change does not involve a significant reduction in a margin of safety because the change is consistent with the vessel overpressurization analysis assumptions. Furthermore, the difference is infinitesimal, and resolves an obvious discontinuity in a manner consistent with the CTS Action presentation as well as the Fermi-2 analyses. I FERMI - UNIT 2-2 REVISION 4. 04/02/99 3

INSERT THIS PAGE IN FRONT OF VOLUME 10 y' jyndend10i SECTION 3.9/

• 1' Remove Replace B 3.9.2 ITS pg B 3.9.2-4 Rev 0 B 3.9.2 ITS pg B 3.9.2-4 Rev 4 B 3.9.2 NUREG M/U pg B 3.9-7 B 3.9.2 NUREG M/U pg B 3.9-7 Rev 4 B 3.9.2 ITS pg B 3.9-7 (Insert) Rev 4 3.9.2 JFDs pg i Rev 0 3.9.2 JFDs pg i Rev 4 3.9.4 DOCS pg 3 Rev 0 3.9.4 DOCS pg 3 Rev 4 3.9.4 NS11C pg i Rev 0 3.9.4 NSHC pg i Rev 4 B 3.9.5 ITS pg B 3.9.5-2 Rev 0 B 3.9.5 ITS pg B 3.9.5-2 Rev 4 B 3.9.5 NUREG M/U pg B 3.0-17 B 3.9.5 NUREG M/U pg B 3.9-17 Rev 4 3.9.5 JFDs pE 1 Rev0 3.9.5 JFDs pg i Rev 4 B 3.9.6 NUREG M/U pg B 3.9-19 B 3.9.6 NUREG M/U pg B 3.9-19 Rev 4 3.9.7 ITS pg 3.9-10 Rev 0 3.9.7 ITS pg 3.9-10 Rev 4 B 3.9.7 ITS pg B 3.9.7-2 Rev 0 B 3.9.7 ITS pg B 3.9.7-2 Rev 4 3.9.7 CTS M/U (3/4 9-16) pg 1 of 1 3.9.7 CTS M/U (3/4 9-16) pg 1 of 1 Rev 4 3.9.7 NUREG M/U pg 3.9-11 3.9.7 NUREG M/U pg 3.9-11 Rev 4 O

B 3.9.7 NUREG M/U pg 3.9-26 B 3.9.7 NUREG M/U pg 3.9-26 Rev 4 B 3.9.7 NUREG M/U pg B 3.9-26 (Insert) Rev 0 B 3.9.7 NUREG M/U pg B 3.9-26 (Insert) Rev 4 3.9.8 ITS pg 3.9-12 Rev 0 3.9.8 ITS pg 3.9-12 Rev 4 B 3.9.8 ITS pg B 3.9.8-2 Rev 0 B 3.9.8 ITS pg B 3.9.8-2 Rev 4 3.9.8 CTS M/U (3/4 9-17) pg 1 of 1 ' 3.9.8 CTS M/U (3/4 9-17) pg 1 of 1 Rev 4 3.9.8 DOCS pg 2 Rev 0 3.9.8 DOCS pg 2 Rev 4 3.9.8 NUREG M/U pg 3.9-14 3.9.8 NUREG M/U pg 3.9-14 Rev 4 B 3.9.8 NUREG M/U pg B 3.9-30 B 3.9.8 NUREG M/U pg B 3.9-30 Rev 4 B 3.9.8 NUREG M/U pg B 3.9-30 (Insert) Rev 0 B 3.9.8 NUREG M/U pg B 3.9-30 (Insert) Rev 4 O Rev 4 04/02/99 l

E 1 a Refuel Position One Rod Out Interlock L B 3.9.2 ). BASES: l' SURVEILLANCE REQUIREMENTS (continued) SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the L associated refuel position one-rod out interlock will function properly when a simulated or actual signal . indicative of a required condition is injected into the logic. A. successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This A . clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a WJ relay. This is acceptable because all of the other contacts of the. relay are verified by other Technical Specifications and non-Technical Specifications. tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST may be performed by any series cf sequential.' overlapping, or total channel steps so that the . entire channel is tested. The 7 day Frequency is considered . adequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the control room to alert. the operator to control rods not fully inserted. To j 9 perform the required testing the applicable condition must = be entered.(i.e., a control rod must be withdi awn from its . full-in position). Therefore. SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until 1 hour after any control rod is withdrawn. REFERENCES 1. 10 CFR 50. Appendix A. GDC 26. 2. UFSAR. Section 7.6.1.1. 3. UFSAR. Section 15.4.1.1. L l FERMI UNIT 2 B 3.9.2 -4 Revision 4. 04/02/99

.s t .a... R2 fuel Position One-Rid-Out Interlock B 3.9.2 BASES I l ACTIONS A.1 and A.2 (continued) fuel assemblies. Action must continue until all such control rods are fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted. SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Proper functioning of the refueling position one-rod-out interlock requires the reactor mode switch to be in Refuel. During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refueling position one-rod-out interlock will be OPERABLE when required. By " locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the cons ~ ole while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation. The Frequency of 12 hours is sufficient in view of other administrative controls utilized during refueling operations a to ensure safe operation. SR 3.9,2.2 fl. Perfomance of a CHANNEL FUNCTIONAL TEST en - d d.n.J demonstrates the associated refuel position one-rod-ogt interlock will function properly when a simulated or actual signal indicative of a required condition is injected into I INSEET the logicJ The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel g g'g._ l steps so that the entire channel is. tested. The 7 day 3 Frequency is consid red adequate because of demonstrated ~ circuit reliability, procedural controls on control rod g 6d withdrawals, and visual and audible indications available in the control room to alert the operator to control rods not fully inserted. To perform the required testing, the applicable condition must be entered (i.e...a control rod (continued) BWRf41TS' B 3.9-7 Rev-h--04/07/9!Ir-f. ResI 'l 3 wm a.w e ~

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t( sQ '1- .-3 g. .~ s- ,,,. c. :: y. .v.. Refueling Position One Rod Out Interlock B 3.9.2 INSERT B 3.9.2-1 A successful test of the required contact (s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an accratable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other contacts of the relay are verified by other Technical Specifications and non Technical Specifications tests at least once per refueling interval with applicable extensions. ,re \\ l FERMI UNIT 2 Page B 3.9 7 (Insert) REVISION 4 04/02/99 q l \\

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E: 4;: * ;{.Lu ..t. j t .t.2. %x5. : c. i. ; -s :./. ( - x l l JUSTIFICATION FOR DIFFERENCES FROM NUREG - 1433 ITS: SECTION 3.9.2 REFUEL POSITION ONE R00-0UT INTERLOCK 9 NON BRACKETED PLANT SPECIFIC CHANGES P.1 Not used. P.2 Bases changes are made to reflect plant specific design details. equipment terminology. and analyses. Specifically.' the Bases description of the one-rod-out' interlock as consisting of two channels is revised to reflect the Fermi-2 design. This interlock is not designed as a function with two redundant channels. P.3 Not used. P.4 The reference to the NRC Policy Statement has been replaced with a more appropriate reference to the Improved Technical Specification " split" criteria found in 10 CFR 50.36(c)(2)(ii). GENERIC CHNAGES C.1 TSTF 205: NRC approved change to NUREG-1433. 1 .aw, FERMI - UNIT 2 1 REVISION 4 04/02/99l I

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'q .it ..{ fk j } e DISCUSSION OF CHANGES ITS: SECTION 3.9.4 - CONTROL R0D POSITION INDICATION 4 TE_CHNICAL CHANGES - LESS RESTRICTIVE Speci fic" L.1 CTS 3.1.3.7 LC0 and Action b. an':' CTS 4.1.3.7. require the position indication system to indicate the current position of any withdrawn control rod in Mode 5. ITS 3.9.4 requires only ' full-in" position indication apability. This change is acceptable because the Operability of the control rod " full-in" position indication for each control rod (whether the control rod is inserted or withdrawn) is required to support Operability of the refueling interlocks (ITS 3.9.1) and Operability of the one rod-out interlock (ITS 3.9.2). Requiring only the " full-in" po:ition indication is acceptable because in Mode 5'the safety analysis is based on the assurance that interlocks ensure that either all or all but one control rod is fully inserted. There is no assumption lh in the safety analysis about rod position other than it is " full-in" or not " full in." Therefore, only the position indication channel associated with the " full in" indication is required to be Operable in Mode 5. ITS 3.9.4 modifies Surveillance Requirements to be consistent with the requirement that only the full-in indicator must be Operable. The new Surveillance (ITS SR 3.9.4.1) requires that each time a control rod is withdrawn from the full in position, the full-in indication is indicating correctly (i.e., it is not indicating full in when a control rod is withdrawn). (Note that failure to indicate " full in" when the control rod is fully inserted results in conservative actuation or the one rod out and refueling interlocks, and therefore, is not explicitly required to be verified.) Additionally, the current requirements to verify the position of the control rod every 24 hours (CTS 4.1.3.7.a). that the control rod position changes during exercise tests (CTS 4.1.3.7.b), and that the full out indicator functions during rod coupling checks (CTS 4.1.3.7.c). are deleted because these tests do not support Operability of the " full-in" position indication channel. These less restrictive changes have no impact on safety. RELOCATED SPECIFICATIONS None .q FERMI - UNIT 2 3 REVISION 4 04/02/99l

T NO SIGNIFICANT HAZARDS EVALUATION ITS: SECTION 3.9.4 CONTROL R00 POSITION INDICATION TECHNICAL CHANGES - LESS h 5fRICTIVE (Soecification 3.9.4 "L.1" 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. 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 50.97. The criteria and the conclusions of the evaluation are presented below. 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? This change replaces a Mode 5 requirement that the position indication system indicate the current position of any withdrawn control rod with a requirement that only the control rod ~" full in" position indication is Operable for all control rods including those not withdrawn. This change will not result in an increase in the probability or consequences of an accident previously evaluated because the Operability of the control rod " full in" position indication for each control rod (whether the control rod is inserted or withdrawn) is sufficient to support Operability of the refueling interlocks and Operability of the one-rod-out interlock. Requiring Operability of the " full-in" position indication only is acceptable because in Mode 5 the safety analysis is based on the assurance that interlocks ensure that either all or all but l@ one control rod is fully inserted. There is no assumption in the safety analysis about rod position other than it is " full-in" or not " full-in." Therefore, only the position indication channel associated with the " full in" indication is required to be Operable in Mode 5. 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) nr 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. FERMI - UNIT 2 1 REVISION 4 04/02/99l l 1

Control Rod OPERABILITY-Refueling B 3.9.5 BASES LC0 Each withdrawn control rod must be OPERABLE. The withdrawn control rod is considered OPERABLE if the scram accumulator pressure is a: 940 psig and the control rod is capable of being automatically inserted upon receipt of a scram signal: ) however, no specific scram time limit is imposed. Inserted-control rods have already completed their reactivity control function, and therefore are not required to be OPERABLE. APPLICABILITY During MODE 5, withdrawn control rods must be OPERABLE to ensure that in a scram the control rods will insert and provide the required negative reactivity to maintain the y reactor subcritical. i e-For MODES 1 and 2. control rod requirements are found in fl LC0 3.1.2. " Reactivity Anomalies." LC0 3.1.3. " Control Rod OPERABILITY." LC0 3.1.4. " Control Rod Scram Times." and LC0 3.1.5. " Control Rod Scram Accumulators." During MODES 3 l T and 4. control rods are not able to be withdrawn since the l reactor mode switch is in shutdown and a control rod block l is applied. This provides adequate requirements for control l

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rod OPERABILITY during these conditions. l l l ACTIONS M i With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert'the inoperable control rod (s). Inserting the control rod (s) ensures the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod (s) is fully ir, rted. i 1 FERMI UNIT 2 B 3.9.5-2 Revision 4 04/02/99 9 1p*4 e v 1

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BASES (continued) imrt eq @posec/ _ p-LC0 Each withdrawn control rod must be OPERABLE. The withdrawn control rod is con ered OPERABLE if the scram accumulator pressure is 2 $ 4 psig and the control rod is capable of being automatical y inserted upon receipt of a scram signalE Inserted control rods have already completed their reactivity control function, and. therefore are not required to be OPERABLE. APPLICABILITY During MODE 5, withdrawn control rods must be OPERABLE to ensure that in a scram the control rods will insert and provide the required negative reactivity to maintain the reactor subcritical. d i For MODES I and 2, control rod requirements are found in-LCO 3.1.2, " Reactivity Anomalies," LCO 3.1.3, " Control Rod l% OPERABILITY," LC0 3.1.4, " Control Rod Scram Times," and LCD 3.1.5, " Control Rod Scram Accumulators." During MODES 3 and 4, coatrol rods are not able to be withdrawn since the reactor mode switch is it. shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during these conditions. w /. ACTIONS Al .1 With one or more withdrawn control rods inoperable, action. must be immediately initiated to fully insert the inoperable control rod (s). Inserting the control rod (s) ensures the shutdown and scram capabilities are not adversely affected. Actions must continue until the increrable control rod (s) is fully inserted. . SURVEILLANCE SR 3.9.5.1 and SR 3.9.5.2 REQUIREMENTS During MODE 5, the OPERABILITY of control rods is primarily required to ensure a withdrawn control rod will j automatically insert if a signal requiring a reactor shutdown occurs. Because no explicit analysis exists for automatic shutdown during refueling, the shutdcwn function is' satisfied if t.t: withdrawn control rod is capable of (continued)- BWRf4-STS B 3.9-17 Rev L J/4/.02/SE 5

r; .y.< p : >,.. ' y.. t JUSTIFICATION FOR DIFFERENCES FROM NUREG 1433 ITS: SECTION 3.9.5 CONTROL R00 OPERABILITY - REFUELING l 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. P.2 'Not used. f P.3 Not used. 6 P.4 Not used. lg P.5 Editorial change made for consistency with other ITS Bases that discuss the same function. P.6 The reference to the NRC Policy Statement has been replaced with a j more appropriate reference to the Improved Technical Specification "solit" criteria found in 10 CFR 50.36(c)(2)(ii). j l ' FERMI - UNIT 2 1 REVISION 4 04/02/99l

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A ' f y r. 3 {.'. N .c RPV Water Leve1[-Iri di:t:d Sc1]- B 3.9.6 i B 3.9 REFUELING OPERATIONS B 3.9.6 Reactor Pressure Vessel (RPV) Water Leve1[-deradi= tad Fuell M BASES s BACKGROUND IIemov nt of [irra.ated] fuel asse 'es'[orhandli Y i I /N6@ [ contr rods] withi the RPV require a minimum wate e f I ft above e inn nf the RP flanneJ During ugd g 3,9,Q -l ' refueling, this maintains e uffici=te ter level in thF e wa I j - reactor vessel cavity and spent fuel pool -- L TT M ;..t m te @ i gp4;c;, ps necessary to retainfiodine fission product activity +in D the water in the event of a fuel handling accident (Refs."1 i and 7). Sufficient iodine,rtivity would be retained to k "I"d limit offsite doses from the accident to s 25% of 10 CFR 100 i limits, as provided by the guidance of Reference 3. ov r o&v irrod;ahdassemWes seated wh 1(4t Rfv e APPLICABLE During movemen o -ef ;er.tr:1 r:d:f, the water lev}el mi tt.e ;""Qs _an initial 44te6 fuel. asse SAFETY ANALYSE 3 condition design parameter in the analysis of a rues gg handling accident in containment postulated by Regulator j""- i _ _'_~ D Guide 1.25 (Ref.1).y minimum water level of 23 ft C 6 b "000f d# (Regulatory Position 1.c of Ref.1) allows a j%}uidt, V75 decontamination factor of 100 (Regulatory Position C.I.g of Q45tv/ -- Ref. 1) to be used in the accident analysis for iodine. This relates to the assumption that 99% of the total iodine released from the pellet-to-cladding gap of all the 4eop damaged ? fuel assembly rods is retained by the water. The fuel pellet-to-cladding gap is assumed to contain 10% of the total fuel rod iodine inventory (Ref.1). soumdse Analysis of the fuel handling accident inside ontainmeh is described in Reference 2. -With-.a-ainimum_ water level 4 '(. 't -d-e e.inimur; A;n U-.T M h;u-- -i r40-fuelJ b":adling, Me analysis \\and test ;r;;r=:En;stratdthat the iodine Yelease due to a postulated fuel handling accident is adequately captured by'the water and that offsite doses are maintained within allowab limits ~ (Ref. 4). p g 41 p; While the worst case assumptions nelude the dropping of the irradiated fuel assembly being handled onto the reactor core, the possibility. exists of the dropped assembly striking the RPV flange and releasing fission products. Therefore, the minimum depth for water coverage to ensure abo (continued) BWRf4-STS B 3.9-19 Rev-17-04/07/95 fest Y

k; h. ?b ;$ Thll. {ll _3;f.hk;3( 1 RHR-High Water Level 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Residual Heat Removal (RHR)-High Water Level LCO 3.9.7 One RHR shutdown cooling subsystem shall be OPERABLE. . APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel b (RPV). the water level = 20 ft 6 inches above the top of s the RPV flange, and heat losses to ambient not greater 96 than or equal to heat input to reactor coolant. ) -4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required RHR shutdown A.1 Verify an alternate 1 hour cooling subsystem method of decay heat inoperable. removal is available. E e===47 ~~ Once per 24 hours thereafter B. Required Action and B.1 Suspend loading Immediately associated Completion irradiated fuel Time of Condition A assemblies into the not met. RPV. E l (continued) e .? I FERMI - UNIT 2 3.9 10 Revision 4 04/02/99 1 I

'[b h h h $k 0.. ' R. 'Y RHR-High Water Level B 3.9.7 BASES LC0 (continued) line may be used to ' allow pumps in one loop to discharge into the op>osite loop's recirculation line to make a complete suasystem. Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous'or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. APPLICABILITY One RHR shutdown cooling subsystem must be OPERABLE in MODE 5, with irradiated fuel in the reactor pressure vessel. with the water level a 20 ft 6 inches above the top of the RPV flange. and heat losses to ambient not greater than or equal to heat input to the reactor coolant to provide decay heat removal. RHR System requirements in'other MODES are covered by LCOs in Section 3.4. Reactor Coolant System. (RCS): Section 3.5. Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System: and. Section 3.6. Containment Systems. RHR Shutdown Cooling System requirements in MODE 5 with irradiated fuel in the h reactor )ressure vessel and with the water level < 20 ft 6 ( inches a)ove the RPV flange are given in LC0 3.9.8. C6 ACTIONS fu,1 With no RHR shutdown cooling subsystem OPERABLE. the availability of an alternate method of decay heat removal must be established within 1 hour. In this condition. the volume of water 'above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the overall reliability is reduced because loss of water level could result in reduced decay heat removal capability. The 1 hour Com)letion Time is based on decay heat removal function and t1e probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these l FERMI - UNIT 2 B 3.9.7 - 2 Revision 4 04/02/99

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n,' {r , Jr 4 .q f Spea6ca:Lon 3,9 7 I i [ l REFUELING OPERATIONS l 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CTREULATION HIGH WATER LEVEL [fMITING CONDITION FOR OPERATION l -2.0.!!.; At least one shutdown _ cooling mode loop of the residual heat removal NO 3 A*) (RHR) system shall be OPERABLEJw hatleast:} a. e OPERABLE RHR ump, and h b. One OPERABLE heat exchan r. 4 APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel is in the td reactor vessel and the water level is greater than or eaual to 20 feet p g,.,y,p g 6 inches above the top of the reactor pressure vessel flange and heat losses to ambient

• are not sufficient to maintaf n OPERATIONAL CONDITION 5.

l ACTION: With r ' RHR shutdown cooling mode loop OPERABLE, within I hour and at leas M bon.A once - 24 hours thereafter, verify the OPERABILITY of at least one/ alternate Otherwise, syspend a MA tg p .apable of decay heat removal. b m t" rede derav neat loafand establishtSECOSUK(F to art iner-- Mod .iM mt_INTEGRIT g thin 4 hours 1 __ s SURVElllANEE REOUTREMENTS ".0.M,4-At least once per 12 hours verify at least one RHR shutdown _ cooling mode loop is ceDable_o.fftakino suction from he reactor vesses ano als a ta' t excnanger with avalla argi]n sR 3 9 7.1 (,c*ooling wat,pdtor vessel through an RHR y s, e \\o C' /1f f[',d,hy

• Ambient losses must be such that no increase in reactor vessel w A

d j temperature will occur (even though COLD SHUTDOWN conditions are being maintained). j l FERMI - UNIT 2 3/4 9-16 l l PAGE l OF 01 Res 4 i

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'd v. i.- u i7 ~~ i RHR-High Water Level 3.9.K 7 3.9 REFUELING OPERATIONS 3.9. Residual Heat Removal (RHR)-High Water Level C T5 7 LCO 3.9. One RHR shutdown cooling subsystem shall be OPERABLE dEE3h 3.9,ll,l {

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-NOTE------ The r tred shutdo cooling su ystem ma ' remove ./ fr operatio for up o 2 hours pe 8 hour riod. s r "No$ IflCD 20 SA.lf.I APPLICABILITY: MODE S with irradiated fuel in the eactor pressure vessel 4ptaba# (RVP) and the water level 2 }.44-above the top of th t** JRPV ; flange ~ y GM hea+ lossef 6 ainbl*11 do w f,l 9 m4w% we a% heatinpul f. \\ ACTIONS 40 M M $*fCC*l2n _ e a k CONDITION REQUIRED ACTION COMPLETION TIME A. Required RHR shutdown A.I Verify an alternate I hour cooling subsystem method of decay heat / 3* 9

• ll*I inoperable, removal is available.

8!(Q 2 Oc.$ son. Once per 24 hours thereafter i B. Required Action and B.1 Suspend loading Immediately associated Completion irradiated fuel Time of Condition A assemblies into the 3, C).11. I, not met. RPV. O.Cd!o' (1 8!!D (continued) i j - -BWR/4-STS--- 3.9-11 -Rev-17-04-/nv ion m ReA

1 e l RHR-High Water Level B 3.9,Y 7 BASES LCO. An OPERABLE RHR shutdown cooling. subsystem consists of an (continued) RHR pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. In-HODE4; -the-TAe RH RHR-cross tie valve is-not-required-to-be-closed;-thus' the /m, e halve-may4e-opened to allow pumps in one loop to discharge c ross - f,.e

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/HQ y />c uccol fD complete subsystem. het.igt.uja Additionally,eachRHRshutdowncoolin$lyaligned(remoteo subsystem is i Ib considered OPERABLE if it can be manua local) in the shutdown cooling mode for removal of decay j heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However.-te-ensur4-adequate cere 1 flow-to-aMow-for ascuf5tYMYBJe~Yrsutor cochmt l ,f -temper 4ture-mon 4 tor 4n9r-near}y-continuousaper_ation ts. sh..q%.. f....Ngte 1. n uv iped y.How.,2pur-except. ion-to. ~ M_ .,,... u ny u u ay....,....., APPLICABIL One RHR shutdown cooling subsystem must be OPERABLE and-tn l P. -oper-ation in MO with irradiated fuel in the reactor 209[cp pressure vess ith the water level h P3}-feet ove 6,ne the top of th lange, to provide decay heat removal. RHR Syste rements other MODES are covered by LCOs (in Section 3.4, Reac or Coolant System (RCS); Section 3.5, a Emergency Core Cooling Systems (ECCS) and Reactor Core d f Isolation Cooling (RCIC) System; and Section 3.6, C Containment Systems. RHR Shutdown Cooling System 4 1 requirements in MODE 5 with irradiated fuel in the reactor .7ns e d k) w pressure vessel and with the water level < e the RPV flange are given in LCO 3.9 co/1.g,h cAc.s, L

3. 9.7 - j ACTIONS L1 With no RHR shutdown cooling subsystem OPERABLE, Van alternate method of decay heat removal must be established within I hour. In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the overall reliability is reduced because loss of water level could (continued)

BWR/4-STS-B 3.9-26 Rev-1~,--04/0E 8 1 ~ g, L

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RHR - High Water Level B 3.9.7 l 1 I 9: - Insert B 3.9.7-1 h I l s b 1 , and heat-losses to ambient not greater than or equal to 6 ~ heat input to the reactor core s Q S ] i i i 9 FERMI - UNIT 2 Page.B 3.9 26 (Insert) REVISION 4 04/02/99l l

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t.. RHR-Low Water Level 3.9.8 3.9 REFUELING OPERATIONS i 1 l 3.9.8 Residual Heat Removal (RHR)-Low Water Level LC0 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and with no recirculation pump in operation, one RHR shutdown l cooling subsystem.shall be in operation. ............................N0TES----- 1. The required operating RHR shutdown cooling subsyst' m e may be removed from operation for up to 2 hours per 8 hour period. l 2. One RHR shutdown cooling subsystem may be ine erable for up to 2 hours for surveillance testing. fAPPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel n (RPV), the water level < 20 ft 6 inches above the top of i the RPV flange, and heat losses to ambient not greater ~ Q than or equal to heat input to reactor coolant. m. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Verify an alternate 1 hour RHR shutdown cooling method of decay heat subsystems inoperable. removal is available AND for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours thereafter I (continued) ? { FERMI - UNIT 2 3.9 12 Revision 4 04/02/99

RHR-Low Water Level B 3.9.8 BASES LCO (continued) opposite loop's recirculation line to make a complete subsystem. Additionally. each RHR shutdown cooling subsystem'is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactc" coolant temperature es required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation of either an RHR pump or a recirculation pump is required. Note 1 is provided to allow a 2 hour. exception to shut down the operating subsystem every 8 hours. Note 2 is provided to allow a 2 hour exception for a single subsystem inoperability due to surveillance testing, l APPLICABILITY Two RHR shutdown cooling subsystems are required to be ~ OPERABLE, and one RHR pump or recirculation pump must be in pl operation in MODE 5, with irradiated fuel in the RPV with the water level < 20 ft 6 inches above the top of the RPV 4, flange, and heat losses to ambient not greater than or equal 0-to heat input to the reactor coolant to provide decay heat M removal. RHR System requirements in other MODES are covered by LC')s in Section 3.4. Reactor Coolant System (RCS): s k Section 3.5. Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System: and Section 3.6. Containment Systems. RHR Shutdown Cooling System requirements in MODE 5 with irradiated fuel in the RPV and with the water level a 20 ft 6 inches above the RPV flange are given in LC0 3.9.7. " Residual Heat Removal (RHR)-High Water Level." 3 6 g l FERMI - UNIT 2 B 3.9.8-2 Revision 4 04/02/99 4,.--. ~

9 y 3,_ 7, g <9+. .c -, 7 _,,,j,p.m ' O. a specificalien. 3 9 8 .f REFUELING OPERATIONS i LOW WATER LEVEL LIMITING CONDITION FOR OPERATION '. mytntained cubater than/6r equ d to k_R) Q. 4.0.11.LjReactAr water 7evel stra i l 1 Li*p d two' shutdown cooling mode loops of the residual heat removal (Kn in.es 3,9 a sy s all be OPERABLE and, at least one recirculation pump shall be in U o eration, or at least one shutdown cooling mode loop shall be in operation *88 wt ea cop consi ting f at least: Ye h /a OP LE Rp ne QPERABL RHR eat xch ger. l EfLICABILITY: OPERATIONAL CONDITION 5 when ir. radiated fuel is in the reactor vessel and the water level is less than 20 feet 6 inches above the top of the td reactor pressure vessel flange and heat losses to ambient" are not sufficient to 8Pf :nb!M/ maintain OPERATIONAL CONDITION 5. 'g { l U ' M/ ACTION: a. With less than the above required RHR shutdown cooling mode loops i OPERABLE, within I hour and at least once per 24 hours thereafter, verify bf the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling mode loop. b. With neither a recirculation pump nor an RHR shutdown cooling mode loop in operation immediately initiate corrective action to return either at least one recirculation pump or at least one RHR shutdown cooling mode loop to hcf,,ioV\\ b operation as soon as possible. Within I hour establish reactor coolant circulati by an_ alternate method and monitor reactor coolant temperature ~ (- a least once per hour. c. W In reactor water level less Inan 21 inches, witnin i nour resto e eactor wat r level to e recuired vel or place t a recircula on pumps ~ in operati n or place o RHR shutd n cooling mode loops in op ation. //73 SURVElltANCE pE0VIREMENTS g 3 9,8.\\ 9-ttt At least one nutdown cooling mode loop of the residual heat removal system or at least one re':irculation pumo shall be determined to be in operation and circulating reactor coolant at least once per 12 hours. f 4'.V.ll.2.2 Aerify refa tor water 34 vel to be gr4ater than or eoM1 to 214 incheJ - (af least orfte per 12 poursj j SA'5 9 3/2 4.9.11.2.3 At least once per 12_ hours verify the recuired RHR shutdown cooling i fDac5ftu tne repctor vessel thrgugn the RHR hegt exchanger (sl wgth their asso[ro' mode loop (s) are caoalltofjtakylig suction tropf ine reactor vesytt and discha g*g ated (cooying water /available. j CD 3 M 80ne RHR shutdown cooling mode loop may be inoperable for up to 2 hours for A ggt surveillance testing. d 3 A-

• The shutdown cooling pump may be removed from operation for up to 2 hours per

'L(Ndt6 I _-hour period. h.1 pTh own cooliffg pumo may/De removed frfm operation dup (ng hydrosta c R /)glioblDfy "Ainbient losses must be such that no increase in reactor vessel water temperature will occur (even inaugh COLD SHUTDOWN conditions are being e i maintained)., Q FEPJil - LIN11 2 3/4 9-17 M PAGE \\ 0F 0-

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[ ~ L ~ i, .. a,. DISCUSSION OF CHANGES ITS: SECTION 3.9.8 - RHR - LOW WATER LEVEL TECHNICAL CHANGES - LESS RESTRICTIVE Generic" LA.1 CTS 3.9.11.2' includes details relating to system design, function. and Operability for the Residual Heat Removal System'in Shutdown Cooling Mode. ITS 3.9.8 includes only a requirement for Operability and moves details of system' design and specific Operability requirements to the Bases.' This is acceptable because the requirement to maintain the function Operable is not impacted by the relocation of the design information. Therefore, these details can be' adequately defined and controlled in the Bases which require change control-in accordance with ITS 5.5.10. Bases Control Program. LA.2 CTS 4.9.11.2.3 requires verification of RHR shutdown cooling capability. and includes details of the required capability. ITS SR 3.9.8.2 requires this same surveillance, but relocates the details _ of RHR shutdown cooling capability to the Bases. This is acceptable because these details do not impact the ITS requirement to maintain the system Operable. These details can be adequately defined and controlled in the Bases which require change control c in. accordance with ITS 5.5.10, Bases Control Program,- The system configuration and capability requirements are also described in the UFSAR. which are controlled in' accordance with 10 CFR 50.59. LA.3 CTS 3.9.11.2 during Mode 5 operation with the reactor cavity not completely flooded requires reactor water level to be maintained 2 214 inches in addition to a means of forced circulation. CTS 3.9.11.2 Action c requires two means of forced circulation from either two recirculation pumps or two RHR-SDC loops in the ~ event 'evel is not maintained 2 214 inches. ITS 3.9.8 does not l retain these restrictions -- only requiring one means of forced circulation regardless of reactor water level. This is consistent with the NUREG 1433. This water level requirement can be adequately defined and controlled in the Technical Rec;uirements Manual (TRM), which requires revisions to be controlled by 10 CFR '50.59. This relocation continues to provide adequate protection > of the public health and safety 'since the requirement for forced circulation, and alternate means of reactor coolant circulation on loss of forced circulation, continues' to be required by the Technical Specifications. _j FERMI - UNIT 2 2 REVISION 4 04/02/99l

I RHR-Low Water Level'

3. 9.J 3.9 REFUELING OPERATIONS 3.9p Residual Heat Removal (

-Low Water Level 3 W4, no recireglahon pump. . in opmaken, LCO 3.9.g Two RHR shutdown cooling subsystemi snail De UPLMABLE, and (7.9 S one RHR shutdown conling subsystem shall be in operation. -- - NOT @ The required operating khutdown cooling subsystem saay be [5.9.fl.2,h removed from operation for up to 2 hours per 8 hour period. OnL Rne s ho+dov>n coolir,g saMYs4e n itny be iosopeak f,,. 4f 9.Ilhg / c Op*; \\

2..

r L hoors he sursglia nct le sHn3.y APPLICABILITY: MODE 5 with irradiated fuel in the reactor pressure vessel if"*,.*4 g.9 and-the water level <L{Mhf4-above the top of the $# (RPV)j (2 O H.,6 i rt ckM {RPVflang heel losses do an1bs cHT int BreaW &an ACTIONS f>tol do hea f onsat H reathe uo/arod.,. , s 3* CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Verify an alternate 1 hour RHR shutdown cooling method of decay heat 3 9.11. 2 3 subsystems inoperable, removal is available A!Q Aclien a. for each inoperable required RHR shutdown Once per cooling subsystem. 74 hours thereafter B. Required Action and B.1 Initiate action to. Immediately restorettecondaryjf associated Completion (Doc At) j Time of Condition A containment to not met. OPERABLE status. (continued) BWR/4 STS 3.9-14 Rev 1, 04/07/95 q.

f y,, ih fl C' U'e ~ RHR-Loc Water Level B 3.g!9: 5' BASES j; ~' allow pumps in one loop to discharge @A a}.. the opposite 1o b0 LCO' (con nue loop's gat-exchangar to make a complete subsystem. / /o c Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay.

i. eat.

Operation (either continuous or intermittent) of ene subsystem can maintain and reduce the reactor coolant temperature as required. ' However, to ensure adequate core flow to allow for accurate average reactor coolant @ p(4gt.T temperature mont ing, nearly continuous operation 85.9.S-5 required. t('No provided to allow a 2 hour excee.an to shut down the op ting subsystem every 8 hours. 4 einiu tw AIM a w pamp ora retkeulah% =- un twpor onMMlahMeed p APPLICABILITY Two RHR shutdo cooling subsystems are required to be OPERABLE, and one must be ration in MODE 5, with irradiated fuel in the R and-ith the water level g [gC,e SPft-above the top RPV flar.ge,pto provide decay }}q 4 1 at_ removal. RHR System requirements in/other MODES are covered by LCOs in Section 3.4, Reactor Coolant System 3 w-(RCS); Section 3.5, Emergency Core Cooling Systems and Reactor Core Isolation Cooling (RCIC) System; an(dECCS) d-Section 3.6, Containment Systems. 4 RHR Shutdown Cooling System requirements in MODE 5 with irradiated ful in the 0 r IMg7 RPV and with the water lesel 2 [ above the RPV flange are given in LCO 3.9 " Residual Heat Removal (RHR)-High 8 38.2-lj Water Levei.- g ACTIONS L,1 With one of the two required RHR shutdown cooling subsystems g e7 inoperable, the remaining subsystem is capable of providing the required decay heat removal. However, the overall made reliability is reduced. Therefore an alternate method of decay heat removai must b.,provided. g y,j/a g ~ shutdown-tV5Tfn-g subsystems inoperable, an alterna~te methodWith bcts requir of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establisbes backup decay heat removal capabilities, similar to the requirements Of the LCO. The I hour Completion Time is based on the decay heat removal function and the probability of a loss of the (continued)

• BWRf4W B 3.9-3ts R u 1, 0410F/95

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. :((. Jl; } RHR - Low Water Level B 3.9.8 ~ IDsert B 3.9.8 1 G \\- . and heat losses to ambient not greatec than or equal to' k heat input to the reactor core g Insert B 3.9.8-3 Note 2 is provided to allow a 2 hour e::.c ption for a single subsystem inoperability due to surveillance testi ). i i 1 -m f FERMI - UNIT 2 Page B 3.9 30 (Insert) REVISION 4 04/02/95l

INSERT THIS PAGE IN FRONT OF VOLUME 11 'W (Volunse linCTS MARKUP COMPil%TIONo I s Remove Replace l 3.4.I CTS M/U (3/4 4-2) pg 3 of 6 3.4.1 CTS M/U (3/4 4-2) pg 3 of 6 Rev 4 3.4.4 CTS M/U (3/4 4-10) pg 1 of 2 3.4.4 CTS M/U (3/4 4..)0) pg 1 of 2 Rev 4 3.4.5 CTS M/U (3/4 412) pg 3 of 3 3.4.5 CTS M/U (3/4 4-12) pg 3 of 3 Rev 4 3.4.7 CTS M/U (3/4 4-17) pg 2 of 3 3.4.7 CTS M/U (3/4 4-17) pg 2 of1 Rev 4 3.4.7 CTS M/U (3/4 4-18) pg 3 of 3 3.4.7 CTS M/U (3/4 4-18) pg 3 of 3 Rev 4 3.4.10 CTS M/U (3/4 4-19) pg 5 of 8 3.4.10 CTS M/U (3/4 4-19) pg 5 of 8 Rev 4 3.4.11 CTS M/U (3/4 4-23) pg 1 of 1 3.4.11 CTS M/U (3/4 4-23) pg i of 1 Rev 4 3.9.7 CTS M/U (3/4 9-16) pg 1 of 1 3.9.7 CTS M/U (3/4 9-16) pg i of 1 Rev 4 j 3.9.8 CTS M/U (3/4 9-17) pg I of 1 3.9.8 CTS M/U (3/4 9-17) pg i of 1 Rev 4 i l v Rev 4 04/02/99

5eav r=t CAmoN 3 Q.l Ouosmaken ssw> ( Aiso see Sfeelfica Hon 3 *T' I) REACrOR COOLANT S'YSTEM SURVEILLANCE REQUIREMFNTS Sh* 5[4.4.1.1.1 Each pump discharge valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full' travel dirring SP 3 g,, each STA.RTUP* prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER. 4.4.1.1.2 uREiZu y .4.1.1. With one tactor cools system recirc atton loop noV in l operati, at least nce per.12 h rs verify that-a. ERMAL POWER s less than or equal to 67. of RATED THERMAL POWE, and b. The individ 1 recireviati pump flow co roller for t operating ecirculation p p is in the H ual mode, a e The speed f the operati recirculatic pump is les tha J or equal o 75% of rate pump speed. - j I.4.1.1.4 With one reactor coolant system locp not in operation with THERMAL 4 POWER less than or equal to 30% of RATED THERMAL POWER or with recirculation loop flow in the operating loop less than or equal to 50% of rated loop flow, verify the 'ollowing differential temocrature reautrements are met within no (more than 10 minutes prior to either THERMAL POWER increase or recirculation ow increase: a. Less~ than or equal to 145'T between reactor vessel sten N space coolant and bottom head drain line coolant, and gi b'er,firh b. Less than or equal to 50'F between the reactor ~ coolant within the loop not it. cperation and the coolant in the p,go j r9 actor pressure vessel", and c. Less than or equal to 50'F between the reactor coolant within the Icop not in operation and the operating loop.** i s tL Stail;eak

• 1f not performed within the previous 31 da'ys.

3 5i t I"

• • Requirement does not apply when the recirculation ' loop not in operation is

%'f,[ isolated from the reacter pressure vessel. FERMI UNIT 2 3/4 4 2 Amendment No. JJ. ff.37,130 f a 'l PAGE 3 0F 06

8C ht:a o fl S* 9 See $ c"Ct hWIl0Il b' k' i REACTOR COOLANT SYSTEM Al OPERATIONAL LEAKAGE LIMITING CONDTTION FOR OPERATION j LC0 g g 3.t..E Reactor coolant system leakage shall be limited to: 3/3.cL-ar No PRESSURE BOUNDARY LEAKAGE. 3.dd.6 -tri 5 gpm UNIDENTIFIED LEAKAGE. 3,d A.C c. 25 gpm total leakage averaged over any 24-hour period. Cec d. Leakage specified in Table 3.4.3.2-1 at a reactor coolant system speri4'd8 R pressure of 1045 s 10 psig from any reactor coolant system pressure 345 isolation valve specified in Table 3.4.3.2-1. g g g y,f/,d p 2 gpm increase in UNIDENTIFIED LEAKAGE within 24 hour period during OPERATIONAL CONDITION 1. N 2 APPLICABILITY: OPERATIONAL CONDITION 5 1. 2, and 3. ) EU.DIi: ~ a. With any PRESSURE BOUNDARY !EAKAGE, be in at least HOT SHUTDOWN within ] Ach.on C 12 hours and in COLD SHUTDOWN slt' tin the next 24 hours. i b. With any reactor coolant system 1eakage greater than the limits in b MCYloh A and/or c._above, reduce the ler age rate to within the limits within Lhouts/or be in at least HOT SriUTDOWN within the next 12 hours and in gppg C COLD SHUTDOWN within the following 24 hours. c. With any reactor coolant system pressure isolation valve leakage greater GCC than the above limit, isolate the high pressure portion of the affected system from the be pressure portion within 4 hours by use of at least SpeCig;on one other closed manual, activated automatic, or check

• valve, or be SAS in at least HOT SHUTDOWN. thin the next 12 hours and in COLD SHUTDOWN within the following 24 N

.t. i d. With one or more of the high/ low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2 2 inoperable, restore the inoperable monitor (s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm setpoint at least once per 12 hours; restore the inoperable monitor (s) to OPERABLE status within 30 days or be in at Itast HOT SHUTDOWN within tt.e next 12 hours and in COLD HUTDOWN within the following 24 hours. (;cp h *Which has been verified not to exceed the allowable leakage limit at the .S[3 'g last refueling outage or after the last time the valve was disturbed, whichever is more recent. FERMI. UNIT 2 3/4 4-10 Amendment No. E7, D, 98 i PAGE 0F 02. ber 4 ..m.

\\ j-7:. Spec &cdwn 3.M TABLE 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES /_d,\\ / / MAXIMUM VALVE UMBER VALVE DESCRIPTION LEAKAGE foom) t 1 RHR stem 3f*y,g, . 4 "' E 1-F015A .LPCI Loop A Injection I ation Valve .4"' 11-F0158 LPCI L40p B Injectio solation Valve E11-F050 LP ' Loop A Inject n Line Testable S

1. N l

Check Valve Ell-50B PCI Loop B In etion Line Testable siASAC-Q0 l Check Val,vi / 11-F008 ShutdownCoolingRPVSuctionOu[ bard Isol'ation Valve / \\ E11-F0 9 Shutdown Cooling RPV Suction inboard ) / / Isolation Valve ./ l l E17-F608 , Shutdown Cooling Suct' ion / 'l / Isolation Valve' J ./ /, '), . Core Spray ystem Loop A,Iriboard Isolatio,lalve / / ,/r' E21-F005A E214c.iB B' Inboard Isolatitn Valve Loop,A Containmentjcheck Valve / E21-h.J6A Loop E Ltiop B Contains nt Check Valve f SR /21-F006B3. High Pressur/ Coolant 3.4.MA Injection, System j E41-F007 Pump D'ischarge Outboa,rd [ j / , Isolation Valve f l Ed-F006 ,Pemp Discharge I tioard [ j { / j isolation ve 4.RectorCo[eIsolation i Cooling' System \\ /' / l' E51 F012 Pu Discharge lation Valve ( Q F013 P p Dischar to feedwater Header Isolation valve y (a) External Leakage from this valve shall be lirited to 5 ml/ min. g REACTOR COOL T YSTE kNTRFACEVALV[ LEAKAGE PRESStfRE MONITORS / ALARM / SETPOINT - VALVE NUM M f ona) $( Ell-F035A & B. El 050A & RHR I 449 Ellf008 F009/F608 R Shutdown ling s 135 E2T-T005A 1,8. E21-F00 &B ore Spray s4 F (,.E41 F006 / 007 HPCI '2. F013 RC!C s 71 ^ FERMI - UNIT 2 3/4 4-12 Amendment No. JA, SJ, SJ, 98 PAGE 3 OF 03-

a 'J.,' 'I 0,n : s.g; So CCr'fCQ OR REACTCR COOLANT SYSTEM LIMITING CONDITION FOR OPtWATION (Continued) (, ACTION: (Continued) [ c. In OPERATIO f. CONDITION or 2, with: I 1. T L POWER nged by more an 15% of RA THERMAL P WER n I hour *, ~ l l The o gas level, e delay pipe nereased b re than 10 microcurie [seratesles' r second in our during ady-state eration at r e an 75,000 m ocuries per second, or / 3. The of level, at) e delay pipe, i reased by more an 15% ne hour dur (g' steady-state o ration at rele rates at r than 75,0 microcuries per econd,- the sampi and analysis r irements of em 4b of Table 4.4.5-1 until' thi specific activit' of the pri p coolant is restored to within its limit. SURVEILLANCE REOUTREMENTS M. -414:& The specific activity of the reactor coolant shall be demonstrated to L A.7.l be within the limits by performance of the sampling and analyris program of i Table 4.4.5-1. b LM NBtappliNbTe'Wric9 D, .i FERMI - UNIT 2 3/4 4-17 Amendment No. 6 PAGE S OF 03 e

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F I' Specmcoma 3 4.10 REACTOR COOLANT SYSTEM 3/4.4.6 PRESSURE / TEMPERATURE LIMITS i REACTOR C0OLANT SYSTEM LIMITING CONDITION FOR OPERATION L.co 3.tJ.lO The' reactor coolant system temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4.6.1 1 (1) Curve A for 6E I'UO'I A j hydrostatic or leak testing; (2) Curve B for heatup by non nuclear means, cooldown following a nuclear shutdown and low power PHYSICS TESTS: and (3) 'scurve C for operations with a critical core other than low power PHYSICS SR3.84 lo 'LitTESTS, with: q a. A maximum heatup of 100*F in any 1-hour period, 9 3 410.1.b.1 hb. A maximum cooldown of 100*F in any 1-hour period, c. A maximum temperature change of less than or equal to 20*F in any 1-hour period during inservice. hydrostatic and leak testing "R 3.WO.M'1 operations above the heatup and cooldown limit curves, and ln c4 d. The reactor vessel flange and head flange temperature greater than 4 SR 3.t/.iD. 7 or equal to 71'F when reactor vessel head bolting studs are under. tension. APPLICABILIT,1: At all times. ) ...( R.A. C l) gp. E *'#'0. U With ar.y of the above limits e'xceeded. ore the temperature and/or pressure b'I i to withJn the limits within $ minutes; cerrorm an -no n.--, ino evaiuhion havim / knee 4 determint the ettects of Ine oui.-oT-ilmn conomon on Ine nrvuura L ff,; g,.g, 4 c,t integritl' of the reactu coolant system: determine that the reactor coolant a j system rammine accentable for continued operations me in at least HOT i Acpp) 8 SHUTDOWN within 12 hours and in COLD SHUTDOWN within the following 24 hours. g dDO.AcaoN A NOTE \\ N \\ +4hodC MM [ Y SURVElllANCE REl'UIREMENTS 4 4.0.C.f.1-During system heatu'p, cooldown and inservice leak and hydrostatic $iA 3.54.101 testing operations, the reactor coclant system temperature and. pressure shall be determined to be within the above required heatup and cooldown limits and to the right of the limit lines of Figure 3.4.6.1 1 Curves A, B, or C, as applicable, at least once per 30 minutes. ~ 1 4-19 Amendment No. 77 FERMI - UNIT 2 PAGE 5 or os 8" 4 L

}.. .(, u. p g-Specificabtr 3.4.ll \\ ~ i REACTOR C0OLANT SYSTEM REACTOR STEAM DOME g LIMITING CONDITION FOR OpERATTON eci a& ' 9 uo -324:6 d 4g The pressure in the reactor steam dome shall be less than41045 psig. j APPLICABILITY: OPERATIONAL CONDITIONS l b d 20 (Ql ACTION: /c. hon. A With the reactor steam dome pressure exceeding 1045 psig, reduce the pressure ) __to less than 1045 psig within 15 minutes [bdi Fe in at least HOT SHUTOOWN within _JZ hours. i i b SURVElltANCE REOUIREMENTS~ (v7 pad..'U) [- sa s.4.n.1 k -4v4,Sr2 -The reactor steam dome pressure shall be verified to be less than / 1045 psig at least once per 12 hours. M.I NNppl-icable curt 119 antiDated.dransihts. FERMI - UNIT 2 3/4 4 23 Amendment No. 87 0 PAGE OF 01

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Q. Spt cificcedion 8 9 7 REFUELING OPERATIONS 3/4.9.11 RES100At HEAT REMOVAL AND COOLANT CIRCUL ATION HIGH WATER LEVEL [fMITING CONOTTTON FOR OPERATION -2.0.;;.; - At least one shutdown _ cooling mode loop of the residual heat removal [(O 3Sh (RHR) system.shall be OPERABLEfw hatleast:} j 'l a. e OPERABLE RHR ump, and b 4 b. One OPERABLE heat exchan r. APPLICABILITY: OPERATIONAL CONDITION 5, when irradiated fuel 't in the f5 reactor vessel and the water level is greater than or eoval to 20 feet ppppg 6 inches above the top of the reactor pressure vessel flange and heat losses D; i to ambient

• are not aufficient to maintain OPERATIONAL CONDITION 5.

I% ACTION: With no RHR shutdown cooling mode loop OPERABLE, within I hour and at least fe,{ son,A once per 24 hours thereaf ter, verify the OPERABILITY of at least nne/ alternate meIhod capable of decay heat removal. Otherwise. syseend alifgergia@Eg w - - e r'ne o rav reat loaSanc establish (SECON 6nvolvina an ince- - t M atth 'f wn t AINMLN I INTEGRIT)(Fithin4 hours 1 q SURVElllANCE REOUTREMENTS '.0.!!.'- At least once per 12 hours verifv at least one RHR shutdown cooling mode loop is cacable of ft& Kino suction f rom /ne reactor vessei ano ois. argingj sR3 91'1 gcooling wate7 tor vessel through an RHRgeat excnanger with avalla 'a tae e y s u, e 'A ff l water Id / if f[jd,, y *Ambtent losses must be such that no increase in reactor vesse { temperature will occur (even though COLD SHUT 00WN conditions are being maintained). l FERMI UNIT 2 3/4 9-16 /* U-PAGE l OF 01 Res 4

r*'"~ ', l-1' **^ A- ... A ,R ,7 -) ly ...43 +^ ~ heCIhca. lc TL 3 h b f REFUELING OPERATIONS J, Lod WATER LEVEL i Ll 41 TING CONDITION FOR OPERATION -0.0.!!.2 r water /evel spall De m4Antained ontater thanj6r equt( to h. thutoown coonng moot loops of the residual heat removal ( LM a has i syste. sha 1 be.PERA8LE and, at least one rc.;irculation pump shall be in 39u ope _ ration, or at least one shutdown cooling enode loop shall be in operation *88 wit ens _ivo @ nsi ting f at least: a OP LE Rp e RA8 RHR eat ch ger. APPLICABILITY: OPERATIONAL CONDITION 5 when irradiated fuel is in the reactor tb i vessel and the water level is less than 20 feet 6 inches above the top of the td l J reactor pressure vessel flange and heat losses to ambient ** are not sufficient to I kpl:nkf# - maintain OPERATIONAL CON 0!T10N 5. f OOD: AwonB) . pp, y, With less than the above reoutred RHR shutdown cooling mod:,10005 l a. OPERABLE, within I hour and at least once per 24 hours thereafter, verify i i the operability of at least one alternate method capable of deca / heat ggkf removal for each inoperable RHR shutdown cooling mode loop. b. With neither a recirculation pump tor an RV.R shutocwn cooling mode loop in operation immediately initiate corrective action to return either it least one recirculation pump or at least one RHR shutdown ecoling mode loop to hCf,.loV't b operation as soon as possible. Within I hour establish reactor coolant circulatiott by an alternate methsd anc W9itor reactor coolant tempertture aT least once per cour. (---- c. In reactor ater level IEn Insn z1 inches, witnin no7EsTct. ~ ector wat r level te e recuired vel or place t - recircul on ptaip in operati n or place o RHR shute n cooling mode cops in op stion. [43 [URVE1LLANCEDE0VIREMENTS j g 3,9,0 ! 'Ar9:tt-in At least one shutoown cooling mode leon of the residual hest removal system or at least one recirculation pu'no shall te determined to be in operation and circulating reacter coolant at least once per 12 hours. ~ 4erify reafter water 34 vel to be gp(ater than or easici to 214 trsdiej C4.V.11.2.2 [ce per 12 fours.j a/' least or SA 3 q.$.7. - 4.9.11.2.3 - At least once per 12 hours _ verify t'ie reouired RHR shutdown cooling i mode loop (s) are capath g suction f rTpf the reactor vespel and discha M9] g'g oncr. o tne r tar vesses th ugn the RHR hegt exchanger (s.),w(th their asso teo con ng water vallable. f" k.o 3 4 %

1. 0ne RHR shutdown cooling moce loco may be inoperible for up to 2 hours for C

g g,, t LGO N 0,y

• surveillance testing.The shutdown cooling pump may be removed frcm operat Nelf,6 4 hour period.

4.1 T HR toown coot y'g pumo mayAe remove f r,dm operation dup (ng hydrosta e t tino, r g gl; cab,'lWy ** Ambient losses must be such that no increase 19 reactor vessel water g temperature will occur (evon tnough COLD SHUTDOWN conditions are being e,- i tq maintained). FERMI - UNIT 2 3/4 9-17 f W PAGE \\ OF 0L

4. _

INSERT THIS PAGE IN FRONT OF VOLUME 12 ) W 'E 01 /Volunne 12: IMPROVED 12CHNICAL SPECIFICATIONS Remove Replace l 3.4.1 ITS pg 3.4-1 Rev 2 3.4.I ITS pg 3.4-1 Rev 4 3.4.5 ITS pg 3.4-11 Rev 0 3.4.5 ITS pg 3.4-11 Rev 4 3.4.3 ITS pg 3.4-12 Rev 0 3.4.5 ITS pg 3.412 Rev 4 3.4.6 ITS pg 3.4-13 Rev 0 3.4.6 ITS pg 3.4-13 Rev 4 3.4.6 ITS pg 3.4-14 Rev 0 3.4.6 ITS pg 3.4-14 Rev 4 3.4.10 ITS pg 3.4-24 Rev 0 3.4.10 ITS pg 3.4-24 Rev 4 3.9.7 ITS pg 3.9-10 Rev 0 3.9.7 ITS pg 3.910 Rev 4 3.9.8 ITS pg 3.9-12 Rev 0 3.9.8 ITS pg 3.912 Rev 4 O I () \\.J Rev 4 04/02/99

y, 71 jl' l.;j iZ p n Recirculation Loops Operating 3.4.1 ^ 3.4 REACTOR C00l>NT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 The. reactor core shall not exhibit core thermal-hydraulic instability or operate in the " Scram" or " Exit' Regions. I 8E j a. Two recirculation loops with matched recirculation loop jet pump flows shcll be in operation: E b. One recirculation loop may be in operation provided: 1. LCO 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 3.3.1.11 is reset for single loop i operation, when in MODE 1. i I --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 1 and 2. ACTIONS l 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." i i (continued) 1 i j FERMI UNIT 2 '3.4 1 Revision 4 04/02/99 l 1 I

g ,LM '[j ! [.' '. f . [} ?h,..* i RCS PIV Leakage 3.4.5 ACTIONS ..................................... NOTES - - - 1. Separate Condition entry is allowed for each flow path. 2. Enter applicable Conditions and Required Actions for systems made inoperable by PIVs. CONDITION REQUIRED ACTION COMPLETION TIME A. One or more flow paths NOTE----- -- ps _g j with leakage from one Each check valve used to' or more RCS PIVs not satisfy Required Action A.1 <(( within limit. must have been verified to meet SR 3.4.5.1 at the last refueling-outage or after the last time the valve was disturbed, whichever is more recent. A.1 Isolate the high 4 hours pressure portion of the affected system from the low pressure portion by use of one other closed manual. de activated automatic. or check valve. B. Required Action and B.1 Be in MODE 3. 12 hours i associated Completion Time not met. AND B.2 Be in MODE 4. 36 hours 4 l FERMI UNIT 2 3.4 11 Revision 4 04/02/99

.e. ,e Y l, < y; .~!:. V. I RCS PIV Leakage 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 -.--NOTE--- Not required to be performed in H00E 3. . Verify eauivalent leakage of each RCS PIV. In accordance 44 at an RCS pressure a 1035 and s 1055 psig: with the Inservice 1 a. For PIVs other than LPCI loop A and B Testing Program injection isolation valves is s 0.5 gpm per nominal inch of valve bl size up to a maximum of 5 gpm-i b. For LPCI loop A and B outboard i injection isolation valves is s 0.4 gpm through-seat, and s 5 ml/ min external leakage: and 1 C. For LPCI loop A and B inboard injection isolation testable check valves is s 10 gpm. i I i 1 r l FERMI UNIT 2 3.4 12 Revi sic., 4 04/02/99

+ RCS Leakage Detection Instrumentation 3.4.6 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Leakage Detection Instrumentation LC0 3.4.6 The following RCS leakage detection instrumentation shall be OPERABLE: a. Drywell floor drain sump flow. monitoring system: -4l b. The primary containment atmosphere gaseous radioactivity monitoring system; and c. Drywell floor drain sump level monitoring system. APPLICABILITY: MODES 1. 4, and 3. ACTIONS .................................. NOTE --- - -- LC0 3.0.4 is not applicable. CONDITION REQUIRED ACTION COMPLETION TIME A. Drywell floor drain A.1 Rest. ore drywell floor 30 days ' sump flow monitoring drain sump flow system inoperable. monitoring system to OPERABLE status. ^ B. Required primary B.1 Analyze grab samples Once per containment atmosphere of-primary 24 hours gaseous radioactivity-containment monitoring system atmosphere. inoperable. (continued) 'l l rERMI! : UNIT 2 - 3.4 13 Revision 4 04/02/99 7

g. n p ~ RCS Leakage Detection Instrumentation 3.4.6 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Drywell floor drain C.1

---

NOTE----

4 sump level monitoring Not applicable when {l system inoperable. primary containment atmosphere gaseous -Q radioactivity monitoring system is inoperable. Perform SR 3.4.6.1. Once per 8 hours D. Primary containment D1 Restore primary 30 days atmosphere gaseous containment radioactivity atmosphere gaseous monitoring system radioactivity inoperable. monitoring system to OPERABLE status. AND 08 Drywell floor drain sump level monitoring D.2 Restore drywell floor 30 days system inoperable, drain sump level monitoring system to OPERABLE status. E. Required Action and E.1' Be in MODE 3. 12 hours associated Completion Time of Condition A. MQ B. C. or D not met. E.2 Be in MODE 4 36 hours F. All required leakage F.1 Enter LC0 3.0.3. Immediately detection systems inoperable. ,f l FERMI - UNIT 2 3.4 14 Revision 4 04/02/99 t

V .h, p , ~ c.i n RCS P/T Limits 3.4.10 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. - -NOTE------ C.1 Initiate action to Immediately Required Action C.2 restore parameter (s) shall be completed if to within limits. this Condition is entered. E C.2 Determine RCS is Prior to Requirements of the acceptable for entering MODE 2 LC0 not met in other operation. or 3 than MODES 1. 2 and 3. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 NOTE- - -- Only required to be performed as applicable during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing. Verify: 30 minutes a. RCS pressure and RCS temperature are gl to the right of the limits specified in Figure 3.4.10-1: and -E b. RCS heatup and cooldown rates are { limited to: 1. s 100*F in any 1 hour period: and 2. s 20*F in any 1 hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. (continued) m l FERMI UNIT 2 3.4 24 Revision 4 04/02/99 b

k;: ll$.j5gh;;$.; {i { i.5 *;l ', ' RlR-High Water Level ) 3.9.7 ) lj 3.9 REFUELING OPERATIONS l 3.9.7 Residual Heat Removal (RHR)-High Water Level -LC0 3.9.7 One RHR shutdown cooling subsystem shall be OPERABLE. f MODE 5 with irradiated fuel in the reactor pressure vessel Y. APPLICABILITY: (RPV). the water level = 20 ft 6 inches above the top of o-the RPV flange. and heat losses to ambient not greater 4 than or equal to heat input to reactor coolant. -4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required RHR shutdown A.1 Verify an alternate 1 hour cooling subsystem method of decay heat inoperable. removal is available. 8NQ Once per 24 hours thereafter B. Required Action and B.1 Suspend loading Immediately associated Completion irradiated fuel i Time of Condition A assemblies into the not. met. ~ RPV. htLQ (continued) 7. a 1 FERMI - UNIT 2 3.9 10 Revision 4 04/02/99

.i- -. 7 ,,. ~ g ~ RHR-Low Water Level 3.9.8 f 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal.(RHR)-Low Water Level LCO 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and with no recirculation pump in operation, one RHR shutdown cooling subsystem shall be in operation. ............................N0TES 1. The required operating RHR shutdown cooling subsystem may be removed from operation for up to 2 hours per 8 hour period. 2. One RHR shutdown cooling subsystem may be inoperable for up to 2 hours for surveillance testing. fAPPLICABILITY:MODE 5 with irradiated fuel in the reactor pressure vessel M (RPV). the water level < 20 ft 6 inches above the top of the RPV flange, and heat losses to ambient not greater ~ Q than or equal to heat input to reactor coolant. c. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Verify an alternate 1 hour RHR shutdown cooling method of decay heat subsystems inoperable. removal is available AND for each inoperable required RHR shutdown Once per cooling subsystem. 24 hours thereafter (continued) n 1 FERMI UNIT 2 3.9 12 Revision 4, 04/02/99

i~ i l i INSERT THIS PAGE IN FRONT OF VOLUME 13 l + ' ' Volume 13: IMPROVED TECHNICAllSPECIFICATIONS BASES : Remove Replace B 3.4.3 ITS pg B 3.4.3-4 Rev 0 B 3.4.3 IT.S pg B 3.4.3 4 Rev 4 B 3.4.6 ITS pg B 3.4.6-5 Rev 0 B 3.4.6 ITS pg B 3.4.6-5 Rev 4 B 3.9.2 ITS pg B 3.9.2-4 Rev 0 B 3.9.2 ITS pg B 3.9.2-4 Rev 4 ) B 3.9.5 ITS pg B 3.9.5-2 Rc y 0 B 3.9.5 ITS pg B 3.9.5-2 Rev 4 B 3.9.7 ITS pg B 3.9.7-2 Rev 0 B 3.9.7 ITS pg B 3.9.7-2 Rev 4 B 3.9.8 ITS pg B 3.9.8-2 Rev 0 B 3.9.8 ITS pg B 3.9.8-2 Rev 4 i l O Rev 4 04/02/99

SRVs B 3.4.3 1 BASES SURVEILLANCE REQUIREMENTS (continued) The SR gives set pressures for all 15 SRVs installed. However, since only 11 SRVs are required, the SR is met if 11 SRVs are set properly. The Frequency is required by the Inservice Testing Program and is consistent with the fact that Surveillance must be performed during shutdown conditions. SR 3.4.3.2 A manual actuation of each required SRV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the SRVs divert steam flow upon opening. i Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is = 850 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by l turbine bypass valves open at least 20%. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours after reactor steam pressure and flow are adequate to perform the test. The 12 hours allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR. If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the SRV is considered OPERABLE. l FERMI UNIT 2 B 3.4 3-4 Revision 4 04/02/99

'I RCS Leakage Detection Instrumentation B 3.4.6 BASES ACTIONS (continued) i Action is to restore either of the inoperable monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a degraded configuration for a lengthy time period. E.1 and E.2 If any Required Action of Condition A, B, C, or D cannot be met within the associated Completion Time, 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 at least i MODE 3 within 12 hours and MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience. to perform the actions in an orderly manner and without challenging plant systems. M With all required monitors inoperable, no required automatic means of monitoring LEAKAGE are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required. SURVEILLANCE SR 3.4.6.1 REQlIIREMENTS This SR is for the performance of a CHANNEL CHECK of the required primary containment atmosphere gaseous radioactivity monitoring system. The check gives reasonable confidence that the channel is operating properly. The Frequency of 12. hours is based on instrument reliability and is reasonable for detecting off normal conditions. SR 3.4.6.2 This SR is for the po formance of a CHANNEL FUNCTIONAL TEST of the required RCS leakage detection instrumentation. The test ensures that the monitors can perform their function in l the desired manner. The Frequency of 31 days considers instrument reliability, and operating experience has shown it proper for detecting degradation. I FERMI UNIT 2 B 3.4.6 - 5 Revision 4, 04/02/99

Refuel Position One-Rod Out Interlock B 3.9.2 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST demonstrates the associated refuel position one rod out interlock will 1 ' function properly when a simulated or actual signal indicative of a. required condition is injected into the logic. A successful test of the required contact (s) of a i channel relay may be performed by the verification of the change.of state of a single contact of the relay. This A clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a WJ relay. This is acceptable because all of the other contacts of the relay are verified by other Technical Specifications and non Technical Specifications. tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. The 7 day Frequency is considered adequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the control room to alert the operator to control rods not fully inserted. To 1 perform the required testing, the applicable condition must 1 be entered'(i.e.. a control rod must be withdrawn from its full in position). Therefore. SR 3.9.2.2 has been modified by-a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until I hour after any control rod is withdrawn. I REFERENCES 1. 10 CFR 50. Appendix A. GDC 26. 2. UFSAR. Section 7.6.1.1. 3. UFSAR. Section 15.4.1.1. i 1 .j.-FERMI.- UNIT 2 B 3.9.2 - 4 Revision 4 04/02/99 e.

Control Rod OPERABILITY-Refueling B 3.9.5 BASES LCO Each withdrawn control rod must be OPERABLE. The withdrawn control ' od is considered OPERABLE if the scram accumulator i pressure is = 940 psig and the control rod is capable of being automatically inserted upon receipt of a scram signal: however, no s)ecific scram time limit is imposed. Inserted control rods lave already completed their reactivity control function, and therefore are not required to be OPERABLE. l APPLICABILITY During MODE 5 withdrawn control rods must be OPERABLE to ensure that in a scram the control rods will insert and provide the required negative reactivity to maintain the d reactor subcritical. i o-For MODES 1 and 2. control rod requirements are found in fI LC0 3.1.2. " Reactivity Anomalies." LCO 3.1.3. " Control Rod OPERABILITY." LCO 3.1.4. " Control Rod Scram Times." and LC0 3.1.5 " Control Rod Scram Accumulators." During MODES 3 T and 4. control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during these conditions. ACTIONS .A_J With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert the inoperable control rod (s). Inserting the control rod (s) ensures the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod (s) is fully inserted. .~ l FERMI UNIT 2 83.9.5-2 ' Revision 4 04/02/99 ,r-7 7 L

j h, ' . hk['lQ,l.,. ' g. . % g;g Q.j., ; ; .) RIR-High Water Level B 3.9.7 BASES LC0 (continued) line may be used to allow pumps in one loop to discharge into the opposite loup's recirculation line to make a complete subsystem. Additionally. each RHR shutdown cooling subsystem is 1 considered OPERABLE if it can be manually aligned (remote or local).in the shutdown cooling mode for removal of decay heat. Operation (either continuous'or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. j g-APPLICABILITY One RHR shutdown cooling subsystem must be OPERABLE in MODE 5, with irradiated fuel in the reactor pressure vessel. i with the water level = 20 ft 6 inches above the top of the RPV flange. and heat losses to ambient not greater than or g equal to heat input to the reactor coolant to provide decay oc heat removal. RHR System requirements in other MODES are covered by LCOs in Section 3.4. Reactor Coolant System (RCS): Section 3.5. Emergency Core Cooling Systems (ECCS) and Reactor Core Isolation Cooling (RCIC) System: and Section 3.6. Containment Systems. RHR Shutdown Cooling System requirements in MODE 5 with irradiated fuel in the h reactor 3ressure vessel and with the water level < 20 ft 6 inches a)ove the RPV flange are given in LC0 3.9.8. 4 ACTIONS A.,.1 With no RHR shutdown cooling subsystem OPERABLE. the availability of an alternate method of decay heat removal must be established within 1 hour. In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the overall reliability is reduced because loss of water level could result in reduced decay heat removal capability. The 1 hour Completion Time is based on decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore. . verification of the functional availability of.these i 1 1 FERMI UNIT 2 B 3.9.7 - 2 Revision 4 04/02/99 L s

g RHR-Low Water Level B 3.9.8 BASES LCO (continued) op>osite loop's recirculation line to make a complete su) system. Additionally. each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or i local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. _However, to ensure adequate core flow to allow for accurate average reactor coolant tem)erature monitoring, nearly continuous operaticn of i eitler an RHR pump or a recirculation pump is required. Note 1 is provided to allow a 2 hour exception to shut down the operating subsystem every 8 hours. Note 2 is provided to allow a 2 hour exception for a single subsystem inoperability due to surveillance testing. APPLICABILITY Two RHR shutdown cooling subsystems are required to be OPERABLE, and one RHR pump or recirculation pump must be in pj 0-)eration in MODE 5, with irradiated fuel in the RPV. with g t1e water level < 20 ft 6 inches above the top of the RPV flange and heat losses to ambient not greater than or equal to heat input to the reactor coolant to provide decay heat M removal. RHR System requirements in other MODES are covered i by LCOs in Section 3.4. Reactor Coolant System (RCS): I w k Section 3.5, Emergency Core Cooling Systems (ECCS) and I Reactor Core Isolation Cooling (RCIC) System; and Section 3.6. Containment Systems. RHR Shutdown Cooling System requirements in MODE 5 with irradiated fuel in the l RPV and with the water level = 20 ft 6 inches above the RPV flange are given in LC0 3.9.7. " Residual Heat Removal (RHR)-High Water Level." T j r l FERMI - UNIT 2 B 3.9.8-2 Revision 4 04/02/99 I 1 .}}