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Improved Technical Specifications, Volume 9, Revision 0, ITS Section 3.4, Reactor Coolant System.
	ML051990501
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Site:	Monticello
Issue date:	06/29/2005
From:	; Nuclear Management Co
To:	; Office of Nuclear Reactor Regulation
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{{#Wiki_filter:IMPROVED TECHNICAL SPECIFICATIONS r4.. MONTICELLO NUCLEAR GENERATING PLANT,

         ;,VOLUME 9 ITS Section 3.4, Reactor Coolant System Commuttedto NudearExcellence

Attachment 1, Volume 9, Rev. 0, Page 1 of 255 ATTACHMENT I VOLUME 9 MONTICELLO IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.4 REACTOR COOLANT SYSTEM (RCS) Revision 0 Attachment 1,Volume 9, Rev. 0, Page 1 of 255

Attachment 1, Volume 9, Rev. 0, Page 2 of 255 LIST OF ATTACHMENTS

1. ITS 3.4.1

2. ITS 3.4.2

3. ITS 3.4.3

4. ITS 3.4.4

5. ITS 3.4.5

6. ITS 3.4.6

7. ITS 3.4.7

8. ITS 3.4.8

9. ITS 3.4.9

10. ITS 3.4.1 0

11. Relocated/Deleted Current Technical Specifications (CTS)

12. Improved Standard Technical Specifications (ISTS) not adopted in the Monticello ITS Attachment 1, Volume 9, Rev. 0, Page 2 of 255

Attachment 1, Volume 9, Rev. 0, Page 3 of 255 ATTACHMENT I ITS 3.4.1, Recirculation Loops Operating Attachment 1, Volume 9, Rev. 0, Page 3 of 255

Attachment 1, Volume 9, Rev. 0, Page 4 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 4 of 255

( ( ( ITS 3.4.1 ITS 3.0 LIMmNG CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 0 F. RecirculatIon System F. Recirculatlon System 0 LCO 3.4.1 1. Intional entry Into the stability exclusion region of I1. SeR Sqtificjation 4.6.GI 0 0 tho power-flow map dellned In tho Core Operating =r Umits Report (COIn) is prohlbilod.If cnlry Into the r Add proposed SR 3.4.1.2 M stabilily exclusion regldoes ocur, Immediately perform or reo of the in unl tho ACTION A stabirity exclusion region has been exitod: 0

a. Insert control ro CD b. Incroes speed of an oparaircutaton 0 pu/

0 LCO 3.4.1 buffer region of the power-ow map as defned In the COLR Is Ibited uness the power distributlon controls as -U LCO 3.4.1 ned 7n tihe COLR are In fe~ct II the power M ICo NOTE distribution controls are not In effect and entry into 0 the stabilty buffer reglon does occur, Immediately performlone or mrri ie to owin untit the ID ACTION A stability butfer region hias beenexiled: 0 0 01

-9                       a. Insert control f-n                                                                  g recirculation U'1                       b. Incro            speed of an ape                                                                                                 CIi 3514.5                                                                                                 107               9/17/90 Amendment No. 7-7 79 93 97 Page 1 of 2

C C ITS 3.4.1 ITS 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 sUnVEILLANCE REouinEMENTS 3.4.1 F. Recirculation System i n0 LCO 3.4.1 3 IThe rnadtor may be started and orated 0) nr operation may continue with only one recirculation 0 loop In operntion provided that: 3 LCO 3.4.1. to

a. The following changes to selpoints and safely a

ACTION B limit settings will he made within 24 hours after

-                                 initiating operation with only one recircunlioatl tD

3 _ oop in operation.

0 LCO 3.4.1.b I. The Operatint Unit MCI'R (MCPR) will be 0 0 dianged per Specificition 3.1 I.C. LCO 3.4.1.a

2. The Maximum Average Phnnnr Linear I lent Generation Rate (MAPLIIGR) will be 3

CD CD changed per Specification 3.11 .A. LCO 3.4.1.c 3. The APiM Neutron FluX 5 0 X 2 0 I ACO 3.4.1A l fspopnlns wilt be changed as noted in Tables 3.1.1 _fbhTThnIcal Specilleatlons 3.5.F.1 and 3.5.E.2 are

--=e 0 ACTION A I.-Mt.

0) to a) ACTION C - ih no reactor coolant system recirculation loops In M Co I operation: 4 M0 0) ACTION A rn p mply with TechnicaliS.edcatllons 3.5 F.1 a) 0 Land 3.5.F.2. I qu( m (comply with apecticntinnCland 3.5.E3 - KN ACTION B for operation with onl one rrdrarnlion lop in oaerallotrm i tn on n The reactor shall be placed In hot shutdown ACTION C wlthIn 12 hours. Two recirculation lops 4 I Add pnoposed SR 3.4.1.1 h -wILh matched flows shalt 3.51M.5 be in operation WS8 06/11/02 Amendment No. 7 7i 7 9 9 9 r 7, 120 4 l Add proposed ACToN B . Page 2 of 2

Attachment 1, Volume 9, Rev. 0, Page 7 of 255 DISCUSSION OF CHANGES ITS 3.4.1, RECIRCULATION LOOPS OPERATING ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.5.F.3.a.3 requires the Average Power Range Monitor (APRM) Rod Block setpoints to be adjusted if single recirculation loop operation is entered. ITS 3.4.1 does not include this requirement. This changes the CTS by deleting the requirement to adjust the APRM Rod Block setpoints when single recirculation loop operation is entered. The requirements for the APRM Rod Block have been relocated to the Technical Requirements Manual (TRM), as described in the Discussion of Changes in ITS 3.3.2.1. Therefore, a reference to this requirement in the ITS is not necessary. Any required changes to the APRM Rod Block setpoints will be controlled in accordance with changes to the TRM. As such, since the Specification has been relocated, the deletion of this specific requirement, which is simply a cross-reference to the affected Specification, is acceptable. This change is designated as administrative because it does not result in any technical changes to the CTS. A.3 CTS 3.5.F.4.a requires compliance with CTS 3.6.A.2 and CTS 3.5.F.3 for one recirculation loop in operation. ITS 3.4.1 does not include this cross reference to other requirements. This changes the CTS by deleting references to other TS requirements. The purpose of CTS 3.5.F.4.a is to ensure the temperature limitations of CTS 3.6.A.2 are met prior to starting an idle recirculation loop and that the requirements of CTS 3.5.F.3 are met when the single loop is in operation. This specification also implicitly provides permission to start a recirculation pump when both are out of service. ITS SR 3.4.9.3 includes the same requirements that are in CTS 3.6.A.2. ITS 3.4.1 does notinclude an explicit reference to this Surveillance Requirement since ITS SR 3.4.9.3 includes a Note that states that this Surveillance applies during a recirculation pump startup. CTS 3.5.F.3.a allows 24 hours to comply with the single loop requirements specified in CTS 3.5.F.3.a.1, 2, and 3 after initiating operation with only one recirculation loop in operation. This requirement is retained in ITS 3.4.1 ACTION B. Therefore, this cross reference is not needed. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 4.5.F.1 provides a cross reference to the Surveillance Requirements in CTS 4.6.G. However, these Surveillances are jet pump Surveillances and do not Monticello Page 1 of 3 Attachment 1, Volume 9, Rev. 0, Page 7 of 255

Attachment 1, Volume 9, Rev. 0, Page 8 of 255 DISCUSSION OF CHANGES ITS 3.4.1, RECIRCULATION LOOPS OPERATING cover stability monitoring issues. ITS SR 3.4.1.2 requires verification that either operation is in the Normal Region of the power to flow map every 24 hours or operation is in the Stability Buffer Region of the power to flow map and the power distribution controls specified in the COLR are in effect every 24 hours. This changes the CTS by deleting the cross references to the Surveillance requirements in CTS 4.6.G and adds a new Surveillance Requirement. The purpose of ITS SR 3.4.1.2 is to periodically ensure the unit is operating in an allowed region of the power to flow map. CTS 3.5.F.1 does not allow operation in the Exclusion Region of the power to flow map and CTS 3.5.F.2 allows entry into the Stability Buffer Region of the power to flow map as long as the power distribution controls are in effect. This change is acceptable because it is consistent with current requirements and provides additional assurance that these requirements are met. This change is designated as more restrictive because it adds a new SR to the CTS. M.2 ITS LCO 3.4.1 requires as one alternative, that two recirculation pumps "with matched flows" shall be in operation. If the requirements of this LCO are not met, ITS 3.4.1 ACTION B must be entered and the requirements of the LCO must be met within 24 hours (i.e., the unit is now operating in single loop and must meet the LCO 3.4.1 single loop requirements). ITS SR 3.4.1.1 requires the verification every 24 hours that jet pump loop flow mismatch with both recirculation loops in operation is: a. s 10% of rated core flow when operating at

      < 70% of rated core flow; and b. s 5% of rated core flow when operating at 2 70% of rated core flow. These requirements are not included in the CTS. This changes the CTS by adding the LCO requirement that two recirculation loops "with matched flows" shall be in operation, adds an ACTION to cover the condition when the flows are not matched, and adds a new Surveillance Requirement to verify every 24 hours that the mismatch criteria is met.

The purpose of the new LCO 3.4.1 matched flow requirement is to ensure that during a loss of coolant accident (LOCA) caused by a break of the piping of one recirculation loop, the assumptions of the LOCA analysis are satisfied. This change is acceptable because the proposed requirements help to ensure the unit is in a configuration consistent with the assumptions of the LOCA analysis. In addition, the proposed ACTION will ensure appropriate action is taken when the criteria is not met and the proposed Surveillance Requirement will periodically ensure the flows are matched. This change is designated as more restrictive because it adds the LCO requirement that two recirculation loops "with matched flows" shall be in operation, adds an ACTION to cover the condition when the flows are not matched, and adds a new Surveillance Requirement to verify every 24 hours the mismatch criteria is met. RELOCATED SPECIFICATIONS None Monticello Page 2 of 3 Attachment 1, Volume 9, Rev. 0, Page 8 of 255

Attachment 1, Volume 9, Rev. 0, Page 9 of 255 DISCUSSION OF CHANGES ITS 3.4.1, RECIRCULATION LOOPS OPERATING REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) If the Stability Exclusion Region of the power to flow map is entered, CTS 3.5.F.1 requires exiting the region by either inserting control rods or increasing the speed of an operating recirculation pump. If the Stability Buffer Region of the power to flow map is entered, CTS 3.5.F.2 requires exiting the region by the same two actions. If either region is entered as a result of both recirculation pumps tripping, CTS 3.5.F.4.a requires complying with CTS 3.5.F.1 and 2 by inserting control rods. ITS 3.4.1 Required Action A.1 requires action to be taken to restore operation to within the Normal Region of the power to flow map, but does not provide the specific details of how operation is to be restored. This changes the CTS by relocating the details that operation is restored to within the Normal Region of the power to flow map by either inserting control rods or increasing the speed of an operating pump (as applicable) to the ITS Bases. The removal of these details for performing Required Actions from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 3.4.1 Required Action A.1 still retains the requirement to restore operation to within the Normal Region of the power to flow map. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES None Monticello Page 3 of 3 Attachment 1, Volume 9, Rev. 0, Page 9 of 255

Attachment 1, Volume 9, Rev. 0, Page 10 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviation's (JFDs) Attachment 1, Volume 9, Rev. 0, Page 10 of 255

Attachment 1, Volume 9, Rev. 0, Page 11 of 255 Recirculation Loops Operating 3.4.1 K.> CTS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.5.F 3.4.1 Recirculation Loops Operating 3.5.F.1, LCO 3.4.1 Two recirculation loops with matched flows shall be in operation 3.5.F.2, DOC M.2 D (within [j the Normal Region of the power to flow map specified in the COLR I UK EX i 3.5.F.1, 3.5.F.2, MOne recirculation loop&2be in operatioprodethe following limits 3.5.F.3 w applied when the associated LCO is applicable: 3.5.F.3.a.2 a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits jspecified in the COLRNI 0 3.5.F.3.a.1 b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR) single loop operation limits Tspecified in the COLF~pnd A) 3.5.F.3.a.3 c. LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation," Function 2.b (Average Power Range Monitors Flowlji3ed ISimulate rmal PoweR - High), Allowable Value of FRenrecd 74 Table 3.3.1.1-1 is reset for single loop operation.] tNto Fux M 3.5.F.3 APPLICABILITY: MODES I and 2. ACTIONS _--------------------REVIEX'ER'S NOTE----_ _ Refer to the fo owing topical reports for th resolution for the Stability T chnical Specifications:

Enhance Option IANEDO-32339S pplement4
Option NEDO-31760 Supplemenf 1 and NEDO-32465
GE-Op ion III NEDC-32410 and N C-32410 Supplement 1 0
ABB ption III CENPD-400 Rev.

3.5.F.3.a, DOC M.2 BWR/4 STS 3.4.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 11 of 255

Attachment 1, Volume 9, Rev. 0, Page 12 of 255 3.4.1 I INSERT I CTS 3.5.F.2 ----- NOI to- --------- Operation within the Stability Buffer Region of the power to flow map specified in the COLR is allowed provided the power distribution controls specified in the COLR are in effect. I INSERT 2 3.5.F.1. A. Operation within the A.1 Initiate actions to Immediately 3.5.F.2 Stability Exclusion restore operation to Region of the power within the Normal to flow map specified Region of the in the COLR. power to flow map specified in the OR COLR. Operation within the Stability Buffer Region of the power to flow map specified in the COLR and the power distribution controls specified in the COLR not in effect. Insert Page 3.4.1-1 Attachment 1, Volume 9, Rev. 0, Page 12 of 255

Attachment 1, Volume 9, Rev. 0, Page 13 of 255 Recirculation Loops Operating 3.4.1 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.5.F.4.b 5 Required Action and associated Completion Time of Condition @3 not met. Be in MODE 3. emc 12 hours

                                                                                                               }D OR No recirculation loops in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOCM.2 SR 3.4.1.1 ---- ------------- OE--------- Not required to be performed until 24 hours after both recirculation loops are in operation. Verify eirailoorjet pump'flow mismatch with both recirculation loops in operation is: 24 hours 0 I

a. < [10% of rated core flow when operating at

                                    <[o1%of rated core flowande

b. < [;5% of rated core flow when operating at 0D 2_ 70O% of rated core flow.

              .4- -- - -- - -

INSERT 3 BWR/4 STS 3.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 13 of 255

Attachment 1, Volume 9, Rev. 0, Page 14 of 255 3.4.1 ,- CTS I INSERT 3 DOC M.1 SR 3.4.1.2 Verify either: 24 hours

a. Operation is in the Normal Region of the power to flow map specified in the COLR; or

b. Operation is in the Stability Buffer Region of the power to flow map specified in the COLR and the power distribution controls specified in the COLR are in effect.

Insert Page 3.4.1-2 Attachment 1, Volume 9, Rev. 0, Page 14 of 255

Attachment 1, Volume 9, Rev. 0, Page 15 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.1, RECIRCULATION LOOPS OPERATING

1. ITS 3.4.1 has been revised to be consistent with License Amendment 97, which resolved the core stability issues. Changes to the current requirements have been made in accordance with the changes in the Discussion of Changes. In addition, the ISTS 3.4.1 ACTIONS "REVIEWER'S NOTE" has been deleted since it was not intended to remain in the plant specific ITS. Subsequent Conditions and Required Actions have been renumbered, as required.

2. The brackets have been removed and the proper plant specific information/value has been provided.

3. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.

4. The Reactor Protection System (RPS) Instrumentation Function 2.b in ITS LCO 3.4.1 has been modified to reflect the plant specific name.

5. Typographical error corrected.

6. Changes have been made to reflect plant specific design requirements related to flow mismatch.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 15 of 255

Attachment 1, Volume 9, Rev. 0, Page 16 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 16 of 255

Attachment 1, Volume 9, Rev. 0, Page 17 of 255 Recirculation Loops Operating B 3.4.1 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.1 Recirculation Loops Operating I at afaster rate J BASES BACKGROUNID The Reactor Coan Recirculation System is designed to provide[A] forced coolant flow through the core to remove heat from the fuel. The 0 forced coolant flow removestmTrel heafifrom the fuel than would be possible with just natural circulation. The forced flow, therefore, allows 0 operation at significantly higher power than would otherwise be possible. The recirculation system also controls reactivity over a wide span of reactor power by varying the recirculation flow rate to control the void content of the moderator. The Reactor Co an Recirculation System consists of two recirculation pump loops external to the reactor vessel. 0 These loops provide the piping path for the driving flow of water to the reactor vessel jet pumps. Each external loop contains one variable speed by Jmotor driven recirculation pums a motor generator (MG) set to control r-,pump speeg and associated piping, jet pumps, valves, and 0 ins-trum a ion. The recirculation loops are part of the reactor coolant pressure boundary and are located inside the drywell structure. The jet pumps are reactor vessel internals. The recirculated coolant consists of saturated water from the steam separators and dryers that has been subcooled by incoming feedwater. This water passes down the annulus between the reactor vessel wall and the core shroud. A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high pressure flow into an external manifold, from which individual recirculation inlet lines are routed to the jet pump risers within the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps. This flow enters the jet pump at suction and result in a inlets and is accelerated by the driving flow. The drive flow and suction partial pressure flow are mixed in the jet pump throat secTio . The total flow then passes through the jet pump diffuser section into the area below the core (lower 0 plenum), gaining sufficient head in the process to drive the required flow upward through the core. The subcooled water enters the bottom of the fuel channels and contacts the fuel cladding, where heat is transferred to the coolant. As it rises, the coolant begins to boil, creating steam voids within the fuel channel that continue until the coolant exits the core. Because of reduced moderation, the steam voiding introduces negative reactivity that must be compensated for to maintain or to increase reactor power. The recirculation flow control allows operators to increase BWR/4 STS B 3.4.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 17 of 255

Attachment 1, Volume 9, Rev. 0, Page 18 of 255 Recirculation Loops Operating B 3.4.1 BASES BACKGROUND (continued) recirculation flow and sweep some of the voids from the fuel channel, overcoming the negative reactivity void effect. Thus, the reason for having variable recirculation flow is to compensate for reactivity effects of boiling over a wide range of power generation (i.e., Zoo 100% of RTP) (D without having to move control rods and disturb desirable flux patterns.> Each recirculation loop is manually started from the control room. The MG set provides regulation of individual recirculation loop drive flows. The flow in each loop is manually controlled. APPLICABLE SAFETY The operation of the Reactor1oa Recirculation System is an initial condition assumed in the design basis loss of coolant accident (LOCA) 0 ANALYSES (Ref. 1). During a LOCA caused by a recirculation loop pipe break, the intact loop is assumed to provide coolant flow during the first few seconds of the accident. The initial core flow decrease is rapid because the recirculation pump in the broken loop ceases to pump reactor coolant to the vessel almost immediately. The pump in the intact loop coasts down relatively slowly. This pump coastdown governs the core flow response for the next several seconds until the jet pump suction is uncovered (Ref. 1). The analyses assume that both loops are operating at the same flow prior to the accident. However, the LOCA analysis was reviewed for the case with a flow mismatch between the two loops, with the pipe break assumed to be in the loop with the higher flow. While the flow coastdown and core response are potentially more severe in this assumed case (since the intact loop starts at a lower flow rate and the core response is the same as if both loops were operating at a lower flow rate), a small mismatch has been determined to be acceptable based on engineering judgment. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during 0 abnormal operational transients (Ref. 2), which are analyzed in

              ~Chapte        of the~pSAR.                     E                                   0 A plant specific LOCA analysis has been performed assuming only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3).

The transient analyseof sChaptseI of the SAR have also been performed for single recirculation loop operation (Re.and demonstrate 0 sufficient flow coastdown characteristics to maintain fuel thermal margins \ during the abnormal operational transients analyzed provided the MCPR 0D requirements are modified. During single recirculation loop operation, modification to the Reactor Protection System (RPS) average power BWR/4 STS B 3.4.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 18 of 255

Attachment 1, Volume 9, Rev. 0, Page 19 of 255 B 3.4.1 4 INSERT I The recirculation flow also provides sufficient core flow to ensure thermal-hydraulic stability of the core is maintained. Insert Page B 3.4.1-2 Attachment 1, Volume 9, Rev. 0, Page 19 of 255

Attachment 1, Volume 9, Rev. 0, Page 20 of 255 Recirculation Loops Operating B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued) - 11AII.-hi. --- I

                                                                    -.-               I               Neutron Flux JK range monitor (APRM)linstrum                  etpointj is also required to account for the different relationships between recirculation drive flow and reactor core flow. The APLHGR and MCPR limits for single loop operation are Especified in the COLR. The APRM Flow[Biased Siu                               HERMAlI 0

I R High Allowable Valucrin LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation." Recirculation )tops perating satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). 0 LCO Two recirculation loops are requi to bk in operation with their flows matched within the limits specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loop the assumptions of the LOCA analysis are satisfied. With the limits specified in SR 3.4.1.1 not met, the recirculation loop with the lower flow must be considered not in operation. With only one recirculation loop in operation, modifications to the required APLHGR limits (LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)"), MCPR limits NeutroncFlux (LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), and APRM lo iBiased Simul MAL POWER- High Allowable Value mLCO 3.3.1.11riv be applied to allow continued operation consistent with 0 the assumptions of Reference 3. INSERT 3 APPLICABILITY In MODES 1 and 2, requirements for operation of the Reacto Co ant Recirculation System are necessary since there is considerable energy in 0 the reactor core.and the limiting design basis transients and accidents are assumed to ocorethermal-hydrauli instabilit rnaV occur l 0 In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important. ACTIONS L!LT4 E l 1( With the requirements of the LCO not me, the recirculation loops must be J restored to operation with matched flows within 24 hours. A recirculation loop is considered not in operation when the pump in that loop is idle or when the mismatch between total jet pump flows of the two loops is greater than required limits. The loop with the lower flow must be considered not in operation. Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited time is allowed to restore thel inmorablk loop to operating status. BWR/4 STS B 3.4.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 20 of 255

Attachment 1, Volume 9, Rev. 0, Page 21 of 255 B 3.4.1 4 INSERT 2 Operation of the Reactor Recirculation System also ensures adequate core flow at higher power levels such that conditions conducive to the onset of thermal-hydraulic instability are avoided. The USAR, Section 14.6 (Ref. 7) requires protection of fuel thermal safety limits from conditions caused by thermal-hydraulic instability. Thermal-hydraulic instabilities can result in power oscillations that could result in exceeding the MCPR Safety Limit. The MCPR Safety Limit is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2). Implementation of requirements for avoidance of, and protection from thermal-hydraulic instability, consistent with the BWR Owners' Group Long-Term Stability Solution Option I-D (Refs. 8 and 9), provides assurance that power oscillations are either prevented or can be readily detected and suppressed without exceeding the specified acceptable fuel design limits. To minimize the likelihood of thermal-hydraulic instability that results in power oscillations, a power to flow "Stability Exclusion" Region is calculated using the approved methodology specified in Specification 5.6.3. The resulting "Stability Exclusion" Region may change each fuel cycle and is therefore specified on the power to flow map in the COLR. Entries into the "Stability Exclusion" Region may occur as a result of an event, such as a recirculation pump trip or runback, an inadvertent control rod withdrawal, or a loss of feedwater heating during startup. Option l-D also has a "Stability Buffer" Region around the "Stability Exclusion" Region. The "Stability Buffer" Region is established to ensure the unit does not operate near the "Stability Exclusion" Region with power distributions more extreme than those used in the stability analysis. Entry into the "Stability Buffer" Region is allowed as long as the power distribution controls specified in the COLR are in effect. The Option I-D analysis demonstrates that in the unlikely event of a neutron flux oscillation, it will be detected and suppressed by the APRM Flow Referenced Neutron Flux - High scram before the MCPR SL is reached. INSERT 3 In addition, during two loop and single loop operation, the combination of core flow and THERMAL POWER must be within the "Normal" Region of the power to flow map specified in the COLR to ensure thermal-hydraulic instability does not occur. Alternately, as allowed in the Note to the LCO, operation in the "Stability Buffer" Region is allowed, provided the power distribution controls specified in the COLR are in effect. Insert Page B 3.4.1-3a Attachment 1, Volume 9, Rev. 0, Page 21 of 255

Attachment 1, Volume 9, Rev. 0, Page 22 of 255 B 3.4.1 4 INSERT 4 A.1 With operation within the "Stability Exclusion" Region of the power to flow map specified in the COLR, the core is in a condition where thermal-hydraulic instabilities are conservatively predicted to occur, and the reactor core must be brought to an operating state where such instabilities are not predicted to occur. With operation within the "Stability Buffer" Region of the power to flow map specified in the COLR and the power distribution controls as defined in the COLR are not in effect, the reactor core may be operating with power distributions more extreme than those used in the stability analysis. In either of these two conditions, the reactor core must be placed outside of these regions. To achieve this status, action must be taken immediately to exit the "Stability Exclusion" and "Stability Buffer" Regions of the power to flow map. This is accomplished by inserting control rods or increasing recirculation pump speed of an operating recirculation pump such that the combination of THERMAL POWER and core flow move to a point outside the "Stability Exclusion" and "Stability Buffer" Regions. The action is considered sufficient to preclude core thermal-hydraulic instabilities that could challenge the MCPR safety limit. The starting of a recirculation pump is not used as a means to exit the "Stability Exclusion" or "Stability Buffer" Region of the power to flow map. Starting an idle recirculation pump could result in a reduction in inlet core enthalpy and enhance conditions necessary for thermal-hydraulic instabilities. Alternately, operation may continue in the "Stability Buffer" Region, provided the requirements of the LCO Note are met. 4 INSERT 5 for reasons other than Condition A Insert Page B 3.4.1-3b Attachment 1, Volume 9, Rev. 0, Page 22 of 255

Attachment 1, Volume 9, Rev. 0, Page 23 of 255 Recirculation Loops Operating B 3.4.1 BASES ACTIONS (continued) Alternatively, if the single loop requirements of the LCO are applied to ooperation with only one recirculation e lfequirements of the LCO and the initial conditions of the accident sequence. The 24 hour Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. This Required Action does not require tripping the recirculation pump in the lowest flow loop when the mismatch between total jet pump flows of the two loops is greater than the required limits. However, in cases where large flow mismatches occur, low flow or reverse flow can occur in the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing pump speeds to re-establish forward flow or by tripping the pump. With no recirculation loops in operation or the Required Action and associated Completion Time of Condition fnEot met, 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. In this condition, the recirculation loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown characteristics. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loops are within the allowable limits for mismatch. At low core flow (i.e., < E70%/o of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety and the APLHGR Limi such that the potential adverse effect of early boiling transition 0 average planar bundle during a LOCA is reduced. A larger flow mismatch can therefore be power allowed when core flow is < [70Z% of rated core flow. The recir atio lljojet pump#flow, as used in this Surveillance, is the summation of the flows from all of the jet pumps associated with a single recirculation loop. BWR/4 STS B 3.4.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 23 of 255

Attachment 1, Volume 9, Rev. 0, Page 24 of 255 Recirculation Loops Operating B 3.4.1 BASES SURVEILLANCE REQUIREMENTS (continued) The mismatch is measured in terms of percent of rated core flow. If the flow mismatch exceeds the specified limits, the loop with the lower flow is

                                        . ThffSR is not required when both loops are not in (X (I operation since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours after both loops are in operation. The 24 hour Frequency is consistent with the Surveillance Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

REFERENCES E. TSAR, Section onl 6. 14..2 ' A' SAR, [SectiQP4.5.1.4.

3. Plant specific an ne 1oo operation. 17 INSERT 8

                                                                                            -- 0 BWR/4 STS                           B 3.4.1-5                               Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 24 of 255

Attachment 1, Volume 9, Rev. 0, Page 25 of 255 B 3.4.1 INSERT 6 SR 3.4.1.2 This SR ensures the combination of core flow and THERMAL POWER are within the "Normal" Region of the power to flow map specified in the COLR to prevent uncontrolled thermal-hydraulic oscillations. At low recirculation flow and high reactor power, the reactor exhibits increased susceptibility to thermal-hydraulic instability. Alternately, the SR ensures the combination of core flow and THERMAL POWER are within the "Stability Buffer" Region of the power to flow map specified in the COLR and the power distribution controls specified in the COLR are in effect. The "Stability Buffer" Region of the power to flow map is established to ensure the unit does not operate near the "Stability Exclusion" Region of the power to flow map specified in the COLR with power distributions more extreme than those used in the stability analysis. The 24 hour Frequency is based on operating experience and the operator's knowledge of the reactor status, including significant changes in THERMAL POWER and core flow. INSERT 7 NEDC-32514P, "Monticello SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis," October 1997. i INSERT 8

4. NEDO-24271, "Monticello Nuclear Generating Plant Single-Loop Operation,"

June 1980.

5. NEDC-30492, "Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvement (ARTS) Program for Monticello Nuclear Power Generating Plant," April 1984.

6. NEDC-32546P, "Power Rerate Safety Analysis Report for Monticello,"

Revision 1, July 1996.

7. USAR, Section 14.6.

8. NEDO-31960-A, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology" (revision specified in Specification 5.6.3).

9. NEDO-31960-A, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology" (revision specified in Specification 5.6.3).

Insert Page B 3.4.1-5 Attachment 1, Volume 9, Rev. 0, Page 25 of 255

Attachment 1, Volume 9, Rev. 0, Page 26 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.1 BASES, RECIRCULATION LOOPS OPERATING

1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

2. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

3. Typographical/grammatical error corrected.

4. Changes have been made to reflect those changes made to the Specification. The following requirements have been renumbered, where applicable, to reflect the changes.

5. Changes have been made to more closely match the LCO requirement.

6. The brackets have been removed and the proper plant specific information/value has been provided.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 26 of 255

Attachment 1, Volume 9, Rev. 0, Page 27 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 27 of 255

Attachment 1, Volume 9, Rev. 0, Page 28 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.1, RECIRCULATION LOOPS OPERATING There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 28 of 255

, Volume 9, Rev. 0, Page 29 of 255 ATTACHMENT 2 ITS 3.4.2, Jet Pumps , Volume 9, Rev. 0, Page 29 of 255

Attachment 1, Volume 9, Rev. 0, Page 30 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 30 of 255

( ( ( ITS 3.4.2 ITS ITS w 3.4.2 0 C) LCO 3.4.2 E3D C, SR 3.4.2.1.b 2 0 a ACTION A 03 0

                                                                  -C 0

to 2 to-0 M -To ID 0 VD0 C) -A) K) 0a 0 en) cn1 Ul Ln 3.6/4.6 128 3127/86 Amendment No. X, 42 I Page 1 of 1

Attachment 1, Volume 9, Rev. 0, Page 32 of 255 DISCUSSION OF CHANGES ITS 3.4.2, JET PUMPS ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 When the jet pump requirements of CTS 3.6.G are not met, CTS 3.6.G requires the unit to be in cold shutdown (MODE 4). ITS 3.4.2 ACTION A only requires a shutdown to MODE 3. This changes the CTS by stating the unit must be in shutdown to MODE 3 instead of to MODE 4. The purpose of CTS 3.6.G, in part, is to place the unit in a condition in which the jet pumps are not required to be OPERABLE. CTS 3.6.G requires the jet pumps to be OPERABLE in the Run mode (i.e., MODE 1). Thus, while the CTS Action requires a shutdown to MODE 4, in actuality, only a shutdown to MODE 2 is required. Once MODE 2 is achieved, continuation to MODE 4 is not required since the jet pumps are not required OPERABLE in MODES other then MODE 1. However, since the requirement that the jet pumps be OPERABLE in MODE 2 has been added (DOC M.1), ITS 3.4.2 ACTION A includes a shutdown to MODE 3. This change is acceptable because MODE 3 is outside the Applicability of the proposed Specification. Therefore, this change is considered a presentation preference change with the deletion of MODES 3 and 4 being made to be consistent with the actual CTS LCO statement and the inclusion of MODE 3 being made to be consistent with the change discussed in DOC M.1. As such, this change is considered an administrative change. A.3 CTS 4.6.G.1 states that the jet pump OPERABILITY Surveillance must be performed by "recording" jet pump loop flows, recirculation pump flows, recirculation pump speeds, and individual jet pump D/P. ITS SR 3.4.2.1 does not include this requirement to record the stated parameters. This changes the CTS by deleting the explicit requirements to record the unit parameters. The purpose of CTS 4.6.G.1 is to verify jet pump OPERABILITY. This change is acceptable because this requirement duplicates the requirements of 10 CFR 50 Appendix B, Section XVII (Quality Assurance Records): maintain records of activities affecting quality, including the results of tests (i.e., Technical Specification Surveillances). Compliance with 10 CFR 50 Appendix B is required by the Monticello Operating License. The details of the regulations within the Technical Specifications are repetitious and unnecessary. Therefore, retaining the requirement to perform the associated Surveillances and eliminating the details from Technical Specifications that are found in 10 CFR 50 Appendix B is considered a presentation preference. As such, this change is considered an administrative change. Monticello Page 1 of 4 Attachment 1, Volume 9, Rev. 0, Page 32 of 255

Attachment 1, Volume 9, Rev. 0, Page 33 of 255 DISCUSSION OF CHANGES ITS 3.4.2, JET PUMPS MORE RESTRICTIVE CHANGES M.1 CTS 3.6.G requires all jet pumps to be OPERABLE in the "Run" mode. ITS LCO 3.4.2 requires the jet pumps to be OPERABLE in MODES 1 and 2. This changes the CTS by requiring the jet pumps to be OPERABLE in MODE 2. The purpose of CTS 3.6.G is to ensure the jet pumps are OPERABLE so that the capability of reflooding the core to two-thirds core height after a design basis accident is possible. This change will require the jet pumps to be OPERABLE in MODE 2. In MODES 1 and 2, the jet pumps are required to be OPERABLE since there is a large amount of energy in the reactor core and since the limiting design basis accidents are assumed to occur in these MODES. This is consistent with the requirements for operation of the Reactor Recirculation System (LCO 3.4.1). This change has been designated as more restrictive because it requires the jet pumps to be in OPERABLE in MODE 2. M.2 CTS 3.6.G requires the unit to be in cold shutdown (MODE 4) within 24 hours if CTS 3.6.G is not met. ITS 3.4.2 ACTION A requires the unit to be in MODE 3 in 12 hours if ITS LCO 3.4.2 is not met. This changes the CTS by requiring the unit to be in a shutdown condition in 12 hours instead of 24 hours. The change to the unit condition required to be achieved (MODE 3 versus MODE 4) is discussed in DOC A.2. The purpose of CTS 3.6.G is to place the unit in a condition in which the jet pumps are not required to be OPERABLE. This change is acceptable because the allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. This change is designated as more restrictive because the shutdown condition (MODE 3) must be achieved sooner than is currently required. M.3 CTS 4.6.G.1 requires the jet pumps to be demonstrated OPERABLE by verifying that the recirculation pump flow/speed ratio deviation from normal expected operating range does not exceed 5% and the jet pump loop/speed ratio deviation from normal expected operating range to not exceed 5%. CTS 4.6;G.2 states if either of these conditions are not met with pump speed greater than or equal to 60%, determine individual jet pump D/P percent deviation from average loop D/P and compare to the Limiting Conditions for Operation. If the pump speed is less than 60% and the deviation of the jet pump D/P exceeds the Limiting Conditions for Operation criteria, the jet pump D/P. shall be monitored, and evaluated every 24 hours until such time as evaluation at the higher pump speed is made. CTS 3.6.G, the Limiting Conditions for Operation, in part, requires the individual jet pump diffuser to lower plenum differential pressure (D/P) percent deviation from average loop D/P to not differ by more than 20% deviation from its normal range of deviation. In addition, it states that if one or more jet pumps exceed the stated criteria, to evaluate the reason for the deviation, and in the circumstance that one or more of the jet pumps are determined to be inoperable, the unit is required to be placed in a cold shutdown condition. Thus the CTS allows, when pump speed is less than 60%, all the jet pump criteria of CTS 4.6.G.1 and CTS 3.6.G to not be met and operation to continue indefinitely. The CTS also allows, when pump speed is greater than or equal to 60%, all the jet pump Monticello Page 2 of 4 Attachment 1, Volume 9, Rev. 0, Page 33 of 255

Attachment 1, Volume 9, Rev. 0, Page 34 of 255 DISCUSSION OF CHANGES ITS 3.4.2, JET PUMPS criteria of CTS 4.6.G.1 and CTS 3.6.G to not be met and operation to continue provided an evaluation of the deviation is acceptable. ITS LCO 3.4.2 requires all jet pumps to be OPERABLE and all applicable criteria are stated in the ITS SR 3.4.2.1. The Surveillance allows either the criteria in CTS 4.6.G.1 or CTS 3.6.G to be met, consistent with the current allowances. However, the ITS does not allow continued operation with both criteria not met below 60% pump speed, and does not allow continued operation with both criteria not met when pump speed is greater than or equal to 60% provided an evaluation is performed. ITS 3.4.2 requires the jet pumps to be immediately declared inoperable and the unit shut down if both of the criteria are not met. This changes the CTS by not allowing continued operation under certain conditions when both of the jet pump criteria are not met. The purpose of CTS 3.6.G and CTS 4.6.G.1, in part, is to specify the jet pump OPERABILITY criteria and to ensure proper actions are taken when they are not met. This change is acceptable because it has been determined that the proposed Surveillance Requirement acceptance criteria are the criteria that determine whether or not the jet pumps are OPERABLE. If both jet pump OPERABILITY criteria are not met, then jet pump OPERABILITY is not assured, and the unit should be shut down until the problem is resolved. Time should not be provided to continue operation while performing evaluations since, with the jet pumps not OPERABLE, core coverage during an accident is not assured. This change is designated as more restrictive because more stringent OPERABILITY Requirements are being applied in the ITS than were applied in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES NONE LESS RESTRICTIVE CHANGES L.1 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.6.G states, in part, whenever there is recirculation flow with the reactor operating, jet pumps shall be demonstrated OPERABLE daily. ITS SR 3.4.2.1 Note 1 states the jet pump OPERABILITY Surveillance does not have to be performed until 4 hours after the associated recirculation loop is in operation. This changes the CTS by allowing a short time after the recirculation loop is placed in operation to evaluate whether or not the jet pumps are OPERABLE. The purpose of CTS 4.6.G, in part, is to require verification of jet pump OPERABILITY after the associated recirculation loop is placed in operation. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Monticello Page 3 of 4 Attachment 1, Volume 9, Rev. 0, Page 34 of 255

Attachment 1, Volume 9, Rev. 0, Page 35 of 255 DISCUSSION OF CHANGES ITS 3.4.2, JET PUMPS ITS SR 3.4.2.1 allows a 4 hour delay in performance of the jet pump OPERABILITY Surveillance after a recirculation loop is placed in service. The jet pump OPERABILITY criteria are only applicable when the associated recirculation loop is in service. The 4 hour time limit to perform the Surveillance Requirement is an acceptable time to establish conditions appropriate for data collection and evaluation. This change is designated less restrictive since a specific delay time has been added which will require the jet pump OPERABILITY Surveillance Requirement to be performed 4 hours after a loop is placed in service. L.2 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.6.G requires the jet pumps OPERABILITY Surveillance to be performed in the "Run" mode. ITS SR 3.4.2.1 requires the same verification to be performed however, its performance may be delayed until 24 hours after

      > 25% RTP, as described in Note 2. This changes the CTS by delaying the performance of the Surveillance until 24 hours after exceeding a THERMAL POWER of 25% RTP.

The purpose of CTS 4.6.G is to verify the jet pumps are OPERABLE. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. This change delays the performance of the jet pump Surveillance until 24 hours after exceeding a THERMAL POWER of 25% RTP. This change is acceptable because during low flow conditions, jet pump noise approaches the threshold response of the associated flow instrumentation and precludes the collection of repeatable and meaningful data. The proposed change is consistent with the ISTS. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.3 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.6.G.1, in part, requires the jet pump OPERABILITY Surveillance to be performed daily and following any unexplained change in core flow, jet pump loop flow, recirculation loop flow, or core plate differential pressure. ITS SR 3.4.2.1 only requires the Surveillance to be performed every 24 hours. This changes the CTS by deleting the requirement to perform the jet pump OPERABILITY Surveillance following the above changes in unit conditions. The purpose of CTS 4.6.G.1 is to verify each jet pump is OPERABLE. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. This change deletes the explicit requirement to perform the jet pump OPERABILITY Surveillance following certain changes in unit conditions. ITS SR 3.0.1 states, in part, that failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. If the unit undergoes an unexplained change in core flow, jet pump loop flow, recirculation loop flow, or core plate differential pressure and it is suspected that the jet pump OPERABILITY is affected, ITS SR 3.0.1 will, in effect, require an informed evaluation be performed to determine whether jet pump OPERABILITY has been impacted. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. Monticello Page 4 of 4 Attachment 1, Volume 9, Rev. 0, Page 35 of 255

Attachment 1, Volume 9, Rev. 0, Page 36 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 36 of 255

Attachment 1, Volume 9, Rev. 0, Page 37 of 255 Jet Pumps 3.4.2 CTS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 Jet Pumps 3.6.G LCO 3.4.2 All jet pumps shall be OPERABLE. 3.6.G APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.6.G A. One or more jet pumps A.1 Be in MODE 3. 12 hours inoperable. BWR/4 STS 3.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 37 of 255

Attachment 1, Volume 9, Rev. 0, Page 38 of 255 Jet Pumps 3.4.2 CTS SURVEILLANCE REQUIREMENTS I SURVEILLANCE FREQUENCY 3.6.G, SR 3.4.2.1 A----------------OTE-------- 4.6.G.1, 4.6.G.2 1. Not required to be performed until 4 hours after associated recirculation loop is in operation.

2. Not required to be performed until 24 hours after > 25% RTP.

Verify at least one of the following criteria (ajor) is satisfied for each operating recirculation loop: 24 hours 0D

a. Recirculation pump flow to speed ratio differs by

                                        < 5% from established patterns, and jet pump loop flow to recirculation pump speed ratio differs by < 5%from established pattemr4ns                            0

b. Each jet pump diffuser to lower plenum differential pressure differs by < 20% from established patterns.

c. Eachj ump flow differs by <

I established patterns. arom 0D

        ---------------------------I--            ------ REVIEWER'S NOTE-----           -----------

An acceptable optio STS, NUREG 4. these criteria for jet pump OPERABILITY can be d in the BWR/6 0 BWR/4 STS 3.4.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 38 of 255

Attachment 1, Volume 9, Rev. 0, Page 39 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.2, JET PUMPS

1. The specific criterion of ISTS SR 3.4.2.1.c, which is one method of verifying the jet pumps are OPERABLE, is not included in ITS SR 3.4.2.1. This change is consistent with the current Monticello licensing basis, which only allows the first two criteria listed in ISTS SR 3.4.2.1 to be used in determining jet pump OPERABILITY.

2. This Reviewer's Note has been deleted. This Note provides the location of an alternative set of criteria that is not used by Monticello. This is not meant to be retained in the final version of the plant specific submittal.

3. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 39 of 255

Attachment 1, Volume 9, Rev. 0, Page 40 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 40 of 255

Attachment 1, Volume 9, Rev. 0, Page 41 of 255 Jet Pumps B 3.4.2 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.2 Jet Pumps BASES BACKGROUND The Reactor Co an Recirculation System is described in the Background section of the Bases for LCO 3.4.1, "Recirculation Loops 0 Operating," which discusses the operating characteristics of the system and how these characteristics affect the Design Basis Accident (DBA) analyses. The jet pumps are part of the Reactor Co an Recirculation System and are designed to provide forced circulation through the core to remove 0 heat from the fuel. The jet pumps are located in the annular region between the core shroud and the vessel inner wall. Because the jet pump suction elevation is at two-thirds core height, the vessel can be reflooded and coolant level maintained at two-thirds core height even with the complete break of the recirculation loop pipe that is located below the jet pump suction elevation. Each reacto co ant recirculation loop contains ten jet pumps. Recirculated coolant passes down the annulus between the reactor vessel wall and the core shroud. A portion of the coolant flows from the vessel, through the two external recirculation loops, and becomes the driving flow for the jet pumps. Each of the two external recirculation loops discharges high pressure flow into an external manifold from which individual recirculation inlet lines are routed to the jet pump risers within 0D (3 the reactor vessel. The remaining portion of the coolant mixture in the annulus becomes the suction flow for the jet pumps. This flow enters the

       'and result in a  *et um at suction inlets and is accelerated by the drive flow. The drive recovery          flow and suction flow are mixed in the jet pump throat section. The total flow then passes through the jet pump diffuser section into the area below the core (lower plenum), gaining sufficient head in the process to drive the required flow upward through the core.

APPLICABLE Jet pump OPERABILITY is an explicit assumption in the design basis loss SAFETY of coolant accident (LOCA) analysis evaluated in Reference 1. ANALYSES The capability of reflooding the core to two-thirds core height is dependent upon the structural integrity of the jet pumps. If the structural system, including the beam holding a jet pump in place, fails, jet pump displacement and performance degradation could occur, resulting in an increased flow area through the jet pump and a lower core flooding elevation. This could adversely affect the water level in the core during the reflood phase of a LOCA as well as the assumed blowdown flow during a LOCA. Jet Pumps satisifCriterionrgof 10 CFR 50.36(c)(2)(ii). 0D BWR/4 STS B 3.4.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 41 of 255

Attachment 1, Volume 9, Rev. 0, Page 42 of 255 Jet Pumps B 3.4.2 BASES LCO The structural failure of any of the jet pumps could cause significant degradation in the ability of the jet pumps to allow reflooding to two-thirds core height during a LOCA. OPERABILITY of all jet pumps is required to ensure that operation of the Reactor o ant Recirculation System will be consistent with the assumptions used in the licensing basis analysis 0D (Ref. 1). APPLICABILITY In MODES 1 and 2, the jet pumps are required to be OPERABLE since there is a large amount of energy in the reactor core and since the limiting DBAs are assumed to occur in these MODES. This is consistent with the requirements for operation of the Reactor Co an Recirculation System (LCO 3.4.1). 0 In MODES 3, 4, and 5, the Reactor Co an Recirculation System is not required to be in operation, and when not in operation, sufficient flow is 0D not available to evaluate jet pump OPERABILITY. ACTIONS A.1 An inoperable jet pump can increase the blowdown area and reduce the Cog capabiity refloods during a design basis LOCA. If one or more of the jet pumps are inoperable, the plant must be brought to a MODE in which 0 the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. The Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.4.2.1 REQUIREMENTS This SR is designed to detect significant degradation in jet pump performance that precedes jet pump failure (Ref. 2). This SR is required to be performed only when the loop has forced recirculation flow since surveillance checks and measurements can only be performed during jet pump operation. The jet pump failure of concern is a complete mixer displacement due to jet pump beam failure. Jet pump plugging is also of concern since it adds flow resistance to the recirculation loop. Significant degradation is indicated if the specified criteria confirm unacceptable deviations from established patterns or relationships. The allowable deviations from the established patterns have been developed based on the variations experienced at plants during normal operation and with jet pump assembly failures (Refs. 2 and 3). Each recirculation loop must BWR/4 STS B 3.4.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 42 of 255

Attachment 1, Volume 9, Rev. 0, Page 43 of 255 Jet Pumps B 3.4.2 BASES SURVEILLANCE REQUIREMENTS (continued) satisfy one of the performance criteria provided. Since refueling activities (fuel assembly replacement or shuffle, as well as any modifications to fuel support orifice size or core plate bypass flow) can affect the relationship between core flow, jet pump flow, and recirculation loop flow, these relationships may need to be re-established each cycle. Similarly, initial entry into extended single loop operation may also require establishment of these relationships. During the initial weeks of operation under such conditions, while base-lining new "established patterminiengineering ( judgdment of the daily surveillance results is used to de ect significant 2,) abnormalities which could indicate a jet pump failure. The recirculation pump speed operating characteristics (pump flow and loop flow versus pump speed) are determined by the flow resistance from the loon suction through the jet pump nozzles. A change in the relation;hip indicates a plug, flow restriction, loss in pump hydraulic performance, leakage, or new flow path between the recirculation pump discharge and jet pump nozzle. For this criterion, the pump flow and loop flow versus pump speed relationship must be verified. Individual jet pumps in a recirculation loop normally do not have the same flow. The unequal flow is due to the drive flow manifold, which does not distribute flow equally to all risers. The flow (or jet pump diffuser to lower plenum differential pressure) pattern or relationship of one jet pump to the loop average is repeatable. An appreciable change in this relationship is an indication that increased (or reduced) resistance has occurred in one of the jet pumps. Thi may be indicated by an incr ase in the relative The deviations from normal are considered indicative of a potential problem in the recirculation drive flow or jet pump system (Ref. 2). Normal flow ranges and established jet pump flow and differential pressure patterns are established by plotting historical data as discussed in Reference 2. The 24 hour Frequency has been shown by operating experience to be timely for detecting jet pump degradation and is consistent with the Surveillance Frequency for recirculation loop OPERABILITY verification. This SR is modified by two Notes. Note 1 allows this Surveillance not to be performed until 4 hours after the associated recirculation loop is in operation, since these checks can only be performed during jet pump operation. The 4 hours is an acceptable time to establish conditions appropriate for data collection and evaluation. BWRI4 STS B 3.4.2-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 43 of 255

Attachment 1, Volume 9, Rev. 0, Page 44 of 255 Jet Pumps B 3.4.2 BASES SURVEILL ANCE REQUIREMENTS (continued) ti r e ote 2 allows this SR not to be performed THE RMAL POWERUE (3 g]25%M RTP. During low flow conditions, jet pump noise approaches the threshold response of the associated flow instrumentation and precludes the collection of repeatable and meaningful data. REFERENCS 1 SAR, Section 3 I dindudingSupplement 1, 'Jet Pump Beam Cracks,' I

D i\

2. GE Service Inform ation Letter No. 33tJ June 10. 'igI

3. NUREG/CR-3052, November 1984.

o fseE Bulletin 807: BWR Jet P AsebyFailure,' BWR/4 STS B 3.4.2-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 44 of 255

Attachment 1, Volume 9, Rev. 0, Page 45 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.2 BASES, JET PUMPS

1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

2. Typographical/grammatical error corrected.

3. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

4. The word "may" has been added since a change in the described relationship may be due to other factors.

5. This statement has been deleted since it is misleading; an increase in flow could be indicative of other problems.

6. Changes have been made to more closely match the LCO requirements.

7. The brackets have been removed and the proper plant-specific information/value has been provided.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 45 of 255

Attachment 1, Volume 9, Rev. 0, Page 46 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 46 of 255

Attachment 1, Volume 9, Rev. 0, Page 47 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.2, JET PUMPS There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 47 of 255

, Volume 9, Rev. 0, Page 48 of 255 ATTACHMENT 3 ITS 3.4.3, Safety/Relief Valves (S/RVs) , Volume 9, Rev. 0, Page 48 of 255

Attachment 1, Volume 9, Rev. 0, Page 49 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 49 of 255

c c c ITS 3.4.3 0 ITS ITS w 0e 0) C, 0 3 3 0

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 -9' LCO 3.4.3                                                                          CD co                                                                                    01 M                                                                                       -h 01 M                                                                                    N SR 3.4.3.1                                                                        An Ul                                                                                    01 ACTION B 127            0821/033 Amendment No. 20, 62, 786,02, 02, !I 4 122, 128y 137 Page 1 of 1

Attachment 1, Volume 9, Rev. 0, Page 51 of 255 DISCUSSION OF CHANGES ITS 3.4.3, SAFETY/RELIEF VALVES (S/RVs) ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, 'Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.6.E.1 includes a cross reference to other Specifications "(note: Low-Low Set and ADS requirements are located in Specification 3.2.H and 3.5.A, respectively)," that govern additional requirements associated with the S/RVs. CTS 4.6.E.2 includes a cross reference to Surveillance Requirements in CTS Table 4.2.1 associated with the Low-Low Set logic. These cross references to other Specifications or Surveillance Requirements are not included in ITS 3.4.3. This changes the CTS by deleting the cross reference to other Specification requirements. The purpose of CTS 3/4.6.E is to include the self actuation requirements for the S/RVs. ITS 3.4.3 does not include any cross references to other Specifications that govern other safety-related requirements for the S/RVs. This change is acceptable since the other Specifications prescribe the appropriate requirements for the other safety-related functions of the S/RVs and this cross reference is not necessary. This change is considered administrative because it does not result in technical changes to the CTS. A.3 CTS 3.6.E.1 states, in part, that "seven" S/RVs are required to be OPERABLE. However, CTS 3.6.E.1 also states that "8" valves shall be set within the prescribed limits. ITS LCO 3.4.3 requires "seven" valves to be OPERABLE and ITS SR 3.4.3.1 requires the verification that the safety function lift setpoints of the "required" S/RVs are within limits. This changes the CTS by only requiring the "required" valves to be set to the prescribed limits. The purpose of CTS 3.6.E.1 is to ensure that seven valves are OPERABLE and tested. Thus, in actuality, only "7" S/RVs need to be set at the proper setpoint. The "8" number is the total number of S/RVs installed at Monticello. Therefore, this change is acceptable since only "7" S/RVs are required to be OPERABLE (i.e., even if one of the "8" S/RVs is not set properly and declared inoperable, the LCO is still met). This change is considered administrative because it does not result in technical changes to the CTS. A.4 These changes to CTS 4.6.E.1.a and CTS 4.6.E.1.b are provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-04-036, from Thomas J. Palmisano (NMC) to USNRC, dated June 30, 2004. As such, these changes are administrative. Monticello Page 1 of 6 Attachment 1, Volume 9, Rev. 0, Page 51 of 255

Aftachment 1, Volume 9, Rev. 0, Page 52 of 255 DISCUSSION OF CHANGES ITS 3.4.3, SAFETY/RELIEF VALVES (SIRVs) MORE RESTRICTIVE CHANGES M.1 CTS 3.6.E.1 requires the S/RVs to be OPERABLE "During power operating conditions and whenever reactor coolant pressure is greater than 110 psig and temperature is greater than 3450F." CTS 3.6.E.2 states that if Specification 3.6.E.1 is not met, initiate an orderly shutdown and have reactor coolant pressure and temperature reduced to 110 psig or less and 3450 F or less. ITS LCO 3.4.3 is Applicable in MODES 1, 2, and 3 and ITS 3.4.3 ACTION B requires the unit to be in MODE 4. This changes the CTS by requiring the S/RVs to be OPERABLE in MODE 2 < 1% RATED THERMAL POWER (RTP) and in MODE 3 when reactor coolant pressure is less than 110 psig or temperature is less than 345 0F, and requires the unit to exit these new MODES when a shutdown is required. The purpose of CTS 3.6.E.1 is to ensure the appropriate number of S/RVs is OPERABLE whenever there is a potential to overpressurize the reactor coolant pressure boundary. This change expands the Applicability to require the S/RVs to be OPERABLE at all times when in MODE 2, instead of when > 1% RTP (the CTS 1.0.0 definition states that Power Operation is when reactor power is

      > 1% RTP) and in MODE 3 when the average reactor coolant system temperature is > 2120 F instead of when the reactor coolant pressure is
      > 110 psig and temperature is > 3450 F. The S/RVs must be OPERABLE in MODE 2 because the reactor is critical or control rods are withdrawn (thus a potential for pressurization exists), and in MODE 3 because considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in this MODE. Thus, this change is acceptable since the S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat. This change is designated as more restrictive because the LCO will be applicable under more reactor conditions.

M.2 CTS 3.6.E.1 requires the S/RVs to be set at < 1120 psig. ITS SR 3.4.3.1 states that the required S/RVs shall be set to 1109 +/- 33.2 psig. In addition, this Surveillance states that following testing, lift settings shall be within +/- 1%. This changes the CTS by providing a minimum setting for the S/RVs. The purpose of CTS 3.6.E.1 is to ensure the S/RVs will open to mitigate the consequences of an overpressurization event. This change is acceptable because the LCO requirements continue to ensure that the S/RVs are maintained consistent with the safety analyses. The new minimum values will ensure that the S/RVs do not open prematurely and cause a rapid depressurization of the Reactor Coolant System when not desired. This change is designated as more restrictive because it adds a minimum setting that does not appear in the CTS. M.3 ITS SR 3.4.3.2 requires the verification that each required S/RV opens when manually actuated every 24 months. A Note is included that allows this test to not be performed until 12 hours after reactor steam flow is adequate to perform the test. This Surveillance Requirement is not included in the CTS. This changes the CTS by adding a Surveillance Requirement to verify the required S/RVs can be manually actuated every 24 months. Monticello Page 2 of 6 Affachment 1, Volume 9, Rev. 0, Page 52 of 255

Attachment 1, Volume 9, Rev. 0, Page 53 of 255 DISCUSSION OF CHANGES ITS 3.4.3, SAFETY/RELIEF VALVES (S/RVs) This change is acceptable because it helps to ensure each required S/RV is functioning properly and that no blockage exists in the valve discharge line. This change is designated as more restrictive because it adds a Surveillance Requirement that does not appear in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA. 1 (Type 1- Removing Details of System Design and System Description, Including Design Limits) CTS 3.6.E.1 states that the safety valve function "(self actuation)" of seven S/RVs shall be OPERABLE. ITS LCO 3.4.3 requires the safety function of seven S/RVs to be OPERABLE. This changes the CTS by moving the detail that the safety function of S/RVs are by "self actuation" to the ITS Bases. The removal of this detail, which is related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS LCO 3.4.3 still retains the requirement that the safety function of seven S/RVs shall be OPERABLE. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.2 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.6.E.1.a states that the S/RVs shall be "tested or replaced each refueling interval" in accordance with the Inservice Testing Program. ITS SR 3.4.3.1 requires the verification that the safety function lift setpoints of the required S/RVs are 1109 + 33.2 psig in accordance with the Inservice Testing Program. This changes the CTS by relocating the procedural detail that the S/RVs "shall be tested or replaced each refueling interval" to the Inservice Testing Program. The removal of these details for performing Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains a requirement to perform the testing required by the Inservice Testing Program. Also, this change is acceptable because these types of procedural details will be adequately controlled in the Inservice Testing Program, which is controlled under 10 CFR 50.55a. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. Monticello Page 3 of 6 Attachment 1, Volume 9, Rev. 0, Page 53 of 255

Attachment 1, Volume 9, Rev. 0, Page 54 of 255 DISCUSSION OF CHANGES ITS 3.4.3, SAFETY/RELIEF VALVES (SIRVs) LA.3 (Type 6 - Removal of LCO, SR, or other TS requirement to the TRM, USAR, ODCM, OQAP, IST Program, or lIP) CTS 4.6.E.1 .b states that "At least two of the safety/relief valves shall be disassembled and inspected each refueling interval." This changes the CTS by relocating this Surveillances Requirement to the Inservice Testing Program. The removal of this Surveillance Requirement from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS LCO 3.4.3 still retains a requirement that seven S/RVs be OPERABLE and includes a Surveillance Requirement that ensures the setpoint of the S/RVs is within limit and another Surveillance Requirement that requires each S/RV to be manually opened. Also, this change is acceptable because this type of Surveillance Requirement will be adequately controlled in the Inservice Testing Program, which is controlled under 10 CFR 50.55a. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications. LA.4 (Type 6 - Removal of LCO, SR, or other TS requirement to the TRM, USAR, ODCM, OQAP, IST Program, or lIP) CTS 4.6.E.1 .c states that "The integrity of the safety/relief valve bellows shall be continuously monitored." CTS 4.6.E.1.d states that "The operability of the bellows monitoring system shall be demonstrated each operating cycle." This changes the CTS by relocating these Surveillances Requirements to the Technical Requirements Manual (TRM). The removal of these Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS LCO 3.4.3 still retains a requirement that seven S/RVs are OPERABLE and includes a Surveillance Requirement that ensures the setpoint of the S/RVs are within limit and another Surveillance Requirement that requires each S/RV to be manually opened. Also, this change is acceptable because the removed information will be adequately controlled in the Technical Requirements Manual (TRM). Any changes to the TRM are made under 10 CFR 50.59, which ensures changes to the TRM are properly evaluated. This change is designated as a less restrictive removal of detail change because requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 1- Relaxation of LCO Requirements) CTS 3.6.E.1 states, in part, that seven S/RVs shall be OPERABLE. It also states that 8 S/RVs shall be set at

      < 1120 psig. ITS SR 3.4.3.1 states that the required S/RVs shall be set to 1109 + 33.2 psig. In addition, this Surveillances states "Following testing, lift settings shall be within +/- 1%." This changes the CTS by allowing the S/RVs to be within + 3% of the nominal setpoint of 1109 during operation and only after testing are the S/RVs required to be set to + 1% of 1109 psig (i.e., 1120 psig).

The addition of the minimum allowed setpoint is discussed in DOC M.2. Monticello Page 4 of 6 Attachment 1, Volume 9, Rev. 0, Page 54 of 255

Attachment 1, Volume 9, Rev. 0, Page 55 of 255 DISCUSSION OF CHANGES ITS 3.4.3, SAFETY/RELIEF VALVES (S/RVs) The purpose of CTS 3.6.E.1 is to ensure the S/RVs will open to mitigate the consequences of an overpressurization event. This change is acceptable because the LCO requirements continue to ensure that the S/RVs are maintained consistent with the safety analyses. The safety analysis assumes that five S/RVs open at 1165 psig. The maximum value allowed in ITS SR 3.4.3.1 is 1142.2 psig, which is less than the safety analysis assumed value of 1165 psig. The Surveillance also requires the setpoint to be adjusted to 1109 psig + 1% (i.e., a maximum of 1120 psig) after testing. The upper limit of this value is consistent with the current requirements in CTS 3.6.E.1 that the valves should be set to < 1120 psig. The S/RV setpoint is + 3% for OPERABILITY, however, the values are reset to + 1% during the Surveillance. This is necessary since the actual setpoint tends to drift during the operating cycle. Test data show that the S/RV will not drift outside of the proposed OPERABILITY limits of 1109 +/- 3% psig. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS. L.2 (Category 4 - Relaxation of Required Action) CTS 3.6.E does not provide any specific time for restoring one or two required inoperable S/RVs; an immediate plant shutdown is required by CTS 3.6.E.2. ITS 3.4.3 ACTION A covers the condition when one or two required S/RVs are inoperable and allows 14 days to restore the S/RVs to OPERABLE status. This changes the CTS by adding an allowance to operate for up to 14 days when one or two required S/RVs are inoperable. The purpose of ITS 3.4.3 ACTION A is to allow time to restore one or two inoperable required S/RVs before requiring a reactor shutdown. The Monticello overprotection analysis only takes credit for five OPERABLE S/RVs, and seven S/RVs are required OPERABLE by LCO 3.4.3. Thus, when one or two required S/RVs are inoperable, a sufficient number of S/RVs remain OPERABLE to meet the analysis assumptions. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the remaining S/RVs. This includes the capacity and capability of remaining S/RVs, a reasonable time for repairs or replacement, and the low probability of a design basis accident occurring during the repair period. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.3 (Category 3 - Relaxation of Completion time) CTS 3.6.E.2 states that if Specification 3.6.E.1 is not met, initiate an orderly shutdown and have reactor coolant pressure and temperature reduced to 110 psig or less and 345°F or less. ITS 3.4.3 ACTION B requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours. This changes the CTS by requiring the unit to be in MODE 3 in 12 hours and extends the time to be outside of the Applicability of the Specification from 24 hours to 36 hours. The change to be in MODE 4 in lieu of the current requirement to reduce reactor coolant pressure to 110 psig or less and reactor coolant temperature to 345°F or less is discussed in DOC M.1. Monticello Page 5 of 6 Attachment 1, Volume 9, Rev. 0, Page 55 of 255

Attachment 1, Volume 9, Rev. 0, Page 56 of 255 DISCUSSION OF CHANGES ITS 3.4.3, SAFETY/RELIEF VALVES (S/RVs) The purpose of CTS 3.6.E.2 is to place the unit outside of the Applicability of the Specification within a reasonable amount of time. This change is acceptable because the Completion Time is-consistent with safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a transient occurring during the allowed Completion Time. The allowed Completion Times are reasonable, based on operating experience, to reach required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required (12 hours versus 24 hours). This portion of the change reduces the time the unit would be allowed to continue to operate in MODES 1 and 2 once the condition is identified. The consequences of a pressurization event are significantly reduced when the reactor is shutdown and a controlled cooldown is already in progress. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. Monticello Page 6 of 6 Attachment 1, Volume 9, Rev. 0, Page 56 of 255

Attachment 1, Volume 9, Rev. 0, Page 57 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 57 of 255

Attachment 1, Volume 9, Rev. 0, Page 58 of 255 S/RVs 3.4.3 CTS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety/Relief Valves (S/RVs) 3.6.E.1 LCO 3.4.3 The safety function of ] S/RVs shall be OPERABLE. 0 APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC L.2 A. [i One EortwoM.equirecd S/RV'si inoperable. A.1 Restore theErequirecd S/R\4sNto OPERABLE 14 days I 0 status. 3.6.E.2 B. E Required Action and associated Completion B.1 Be in MODE 3. 12 hours (0 Time of Condition A not AND met.E B.2 Be in MODE 4. 36 hours OR CThreef or more [requiredM S/RVs 0 inoperable. ____ ____ ____L .I BWR/4 STS 3.4.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 58 of 255

Attachment 1, Volume 9, Rev. 0, Page 59 of 255 S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

                                  ./                               1 3.6.E.1,   SR 3.4.3.1 4.6.E.1 .a               s [2] ir/quired] S/RVs4 ay be changed to a lower setp, t group.                                                              0 Verify the safety function lift setpoints of the     VIn accordance IrequirecU S/RVs ae                              s I  with the Inservice Testing Program Ior [1 WrronthS] I BWR/4 STS                                3.4.3-2                      Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 59 of 255

Attachment 1, Volume 9, Rev. 0, Page 60 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.3, SAFETY/RELIEF VALVES (SIRVs)

1. The brackets have been removed and the proper plant specific information/value has been provided.

2. The Note to ISTS SR 3.4.3.1 and the S/RV Table setpoint listing has been deleted since the Monticello S/RVs have the same setpoint range requirement. The S/RV Table setpoint listing has been replaced with the appropriate setpoint range for all the required S/RVs.

3. The OPERABILITY of the S/RVs is not dependent upon electrical solenoids. The S/RV function is self actuated, as described in the Background section of the ITS Bases. Therefore, verifying all solenoids operated by the manual control switches (of which only four of the S/RVs actually have two solenoids) is not necessary for the OPERABILITY of the S/RVs. Therefore this requirement to test each valve solenoid, which is not part of the Monticello Licensing Basis, has not been included in the ITS.

4. Change made for consistency within the Surveillance Requirement.

5. A minimum reactor steam pressure is not necessary to prevent damage to the S/RVs. Only an adequate reactor steam flow is necessary to properly test the S/RVs (i.e., as long as the turbine bypass valves are controlling reactor steam pressure, the S/RVs can be tested safely).

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 60 of 255

Attachment 1, Volume 9, Rev. 0, Page 61 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 61 of 255

Attachment 1, Volume 9, Rev. 0, Page 62 of 255 S/RVs B 3.4.3 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.3 Safety/Relief Valves (S/RVs) BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB). The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The S/RVs can actuate by either of two modes: the safety mode or the relief mode. In the safety mode (or spring mode of operation), the spring loaded pilot (I.e.,Iis~ valve opens when steam pressure at the valve inlet overcomes the spring actuating holding the pilot vayve c osed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens 0 the main valve. This satisfies the Code requirement. Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool. The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves. The LLS requirements are E specified in LCO 3.6.f.0, "Low-Low Set (LLS) Valves," and the ADS requirements are specified in LCO 3.5.1, "ECCS - Operating." 0D APPLICABLE The overpressure protection system must accommodate the most severe SAFETY pressurization transient. Evaluations have determined that the most ANALYSES severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram assayed with MSIV position) (Ref. 1). For the purpose of the analyses, M[fJS/RVs are assumed to operate in the safety mode. The 3 0 analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110% x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event. From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above. Reference 2 discusses additional events that are expected to actuate the S/RVs. S/RVs satis Criterion 3 of 10 CFR 50.36(c)(2)(ii). 0 BWR/4 STS B 3.4.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 62 of 255

Attachment 1, Volume 9, Rev. 0, Page 63 of 255 S/RVs B 3.4.3 BASES 3 E LCO The safety u ior ofIR S/RVs are required to be OPERABLE to satisfy 0( the assumptions of the safety analysis (Refs. 1 and 2). The requirements M of this LCO are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function). However. two addibonal SIRVs are required to be LOPERABLE to provide addibional relief capacity. The SIRV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the [8 L2J (D ) rimLre based on these setpoints, but also include the additional uncertainties L~of+/-i <<poof the nominal setpoint drift to provide an added degree of conservatismm Operation with fewe valves OPERABLE than cified or with setpoints outside the ASME limits, could result in a more severe reactor response c9 to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded. APPLICABILITY In MODES 1, 2, and 3,[]S/RVs must be OPERABLE, since 0 considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat. In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents. In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions. ACTIONS EA.1 0 With the safety function of one [or twcIrequiredq S/R\Msj inoperable, the 0 remaining OPERABLE S/RVs are capable of providing the necessary overpressure protection. Because of additional design margin, the ASME Code limits for the RCPB can also be satisfied with two S/RVs inoperable. However, the overall reliability of the pressure relief system is reduced because additional failures in the remaining OPERABLE S/RVs could result in failure to adequately relieve pressure during a limiting event. For this reason, continued operation is permitted for a limited time only. BWRI4 STS B 3.4.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 63 of 255

Attachment 1, Volume 9, Rev. 0, Page 64 of 255 S/RVs B 3.4.3 BASES ACTIONS (continued)

                  *The 14 day Completion Time to restore the inoperable required S/RVs to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an event requiring S/RV actuation, and a reasonable time to complete the Required Action.E B.1 and B.2 With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure. If the safety function of the inoperable required S/RVs cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1, or if the safety function of Ithreej or more-Rrequiredq S/RVs is inoperable, 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 to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS This Surveillance requires that the [requiredc S/RVs will open at the E pressures assumed in the safety analysis of Reference 1. The demonstration of the S/RV safe lift settings must be performed during shutdown, since this is a bench test, Dto be done in accordance with the Inservice Testing Prograrri. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is +/- 3r/o for OPERABILITY; however the valves are reset to +/- 1%during the Surveillance to allow for drift [A Note is provided to a ow up to [two] of the required [11] S/R s to be physically replaced with /RVs with lower setpoints. This provid s operational flexibility whicfi maintains the assumptions in the ove pressure analysis.] The 18 morn requency was seTcted because this S reillance must be performe during shutdown co ditions and is based the time between refuelin g7s. BWR/4 STS B 3.4.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 64 of 255

Attachment 1, Volume 9, Rev. 0, Page 65 of 255 S/RVs B 3.4.3 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.4.3.2 A manual actuation of each 1requiredl SIRV 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 turbinelcontro5Fv~ves on bypass valves, by a change in the measured (3 steam flow, or bv anv other method suitable to veri steam flow. Adequate reactor ste dome pressure musXe available to pelofm this to avoid damacg the valve. Also Adequate steam flow must be I test passing through the r oines omain valves to continuep wturbine bypass control reactor pressure when the S divert steam flow upon open aRVs Sufficient time is therefore allowed after the required pressure antlo 2 A~prvchieved to perform this test. Adequate pressR. at which thio

             '- Ito be perfumed is [920t psig (the sofessure recoi mended     se       buo         alveing Imanufaclurer).Tdequ~ate steam flow is represented beyEat            leastl1T}E tubne u       bypass valvele openF o-r total steayWw 2:1 OUlb/hrlx Plant startup overprsre thesform        ing this test because valve OPERABILITY and the setpoints for overpressure protection are verified per ASME Code requirements, prior to valve installation.                      pis modified b l iseSRu a otethatstatestheSurveillance is          notredquireqtue        performed untihe
                                   \   12 hours                ~~~~~~~~after rator              sta ~es~r nllw~aeperform 1 hous{Te zthetes. alowe fo manal ctuaionafter the requi                          re Fr         is reached  is sufficient to achievtable f     aconditions   for testing Et  W,5,, an~d 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 S/RV is considered OPERABLE.

lThe [1 8] month on aSTAGGERED TEST BAOIS Frequency e ~st tat< Ieach solenoid forjeach S/RV is alternately dsted.1 The ftgS~ IOCodel Freqec waeeloped based on the S/RV tests required by the [D A Prssu>VsselCodeSectin Xl(Ref. OX pI rtn t j AME~soile~d experience has shown that these components usually pasth . Surveillance when performed at theLL8 month Frequency. Therefore, the (i Frequency was concluded to be acceptable from a reliability standpoint. REFERENCE<SAR, Section SAR, Section [= ASMET Band Press3 Co . ectivon XRVessel BWR/4 STS B 3.4.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 65 of 255

Attachment 1, Volume 9, Rev. 0, Page 66 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.3 BASES, SAFETY/RELIEF VALVES (S/RVs)

1. Changes have been made to reflect changes made to the Specification.

2. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

3. The brackets have been removed and the proper plant specific information/value has been provided.

4. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

5. Typographical/grammatical error corrected.

Monticello Pagel of 1 Attachment 1, Volume 9, Rev. 0, Page 66 of 255

Attachment 1, Volume 9, Rev. 0, Page 67 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 67 of 255

Attachment 1, Volume 9, Rev. 0, Page 68 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.3, SAFETY/RELIEF VALVES (S/RVs) There are no specific NSHC discussions for this Specification. Monticello Page 1of I Attachment 1, Volume 9, Rev. 0, Page 68 of 255

Attachment 1, Volume 9, Rev. 0, Page 69 of 255 ATTACHMENT 4 ITS 3.4.4, Reactor Coolant System (RCS) Operational LEAKAGE Attachment 1, Volume 9, Rev. 0, Page 69 of 255

Attachment 1, Volume 9, Rev. 0, Page 70 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 70 of 255

C C C ITS 3.4.4 ITS ITS 3.0 LIMmNG COND0ONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

0. Reactor Coolant System (RCS) D. Reactor Coolant System (RCS) a)

0rS 3.4A 1. Operational Leakage 'a ) 1. Operational Leakage 0 0 Applicability tie rrdated s by 1 Applicability Any time Irradiated fuel Is h the reactor vessel and F -- 1 pp a Y ocnttemperatureIs above 212F ery12hours reactor coolant system (RCS) leakage, shall be LCO 3.4.4 lUited to: 0

a. Unidentified Leakage Is within limits, tD CD LCO 3.4.4.b 1) s 5 gpm Unidentified Leakage SR 3.4.4.1 -

0 b. Unidentified Leakage Increase Is within limits,

2) < 2 gpm Increase In Unidentified Leakage within the previous 24 hour period while In and

." LCO 3.4.4.d ;U the run mode, a-"

c. Total Leakage Is within limits.

0) LCO 3.4.4.c 3) s 25 gpm Total Leakage averaged over the Co previous 24 hour period, and CD CD 0 X

0 LCO 3.4.4.a 4) no pressure boundary leakage 0 t3 tD 0 b. fWifh reactor coolant system leakage greater

-CD  ACTION A                               than 3.6.D.1.a.1) or3.6.D.1.a.3) above, reduce Wlbjeakage to within fimits within four hoursfr                                                                                             6 01                                        lbIn Hot Shutdown Within the next 12 hours ACTION C                             land In Cold Shutdown within the following                                                                                                   rul I2A hours.

C. Wih an Increase In Unidentified Leakage In excess of the rate specified In 3.6.D.1.a.2) reduce leakage to within limits within four hours, ACTION B -- or verity that the source of Increased leakage Is not service sensitive type 304 or type 316 Austenitic stainless steel within four hoursF5r be I-1 in Hot Shutdown within the next 12 hours and In ACTION C Cold Shutdown within the following 24 hours. 3.614.6 1264 08121/03 4 Amendment No. 1 , 17, 7, 4 r 137 Page 1 of 2

C C C 0 ITS 3.4.4 ITS 3.0 UMING CONDIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

I d. If any Pressure Boundary Leakage exists, be In

0) ACTION C s Hot Shutdown within the next 12 hours and in 0 Cold Shutdown within the following 24 hours. 3 CD 4

0 2. RCS Leakage Detection Instrumentation 2. RCS Leakage Detection Instrumentation

a. Any time Irradiated fuel Is In the reactor vessel and reactor water temperature Is above 21 2F the Dryweil Floor Drain Sump Monitoring RCS leakage detection Instrumentation shall be demonstrated OPERABLE by I a System shall be operable.* If the Drywall Floor a.. Primary containment atmosphere particulate 0S . Drain Sump Monitoring System Is not operable, monitoring system - perform a sensor check ;L then: once per 12 hours, a channel functional test at CD least monthly and a channel calibration at least once per cycle. CD

1) Restore the Drywall Floor Drain Sump 0

Co Monitoring System to operable status 0 within 30 days. 1 b. Required leakage detection Instrumentation - ED 0 2) Otherwise, be In Hot Shutdown within the next 12 hours and In Cold Shutdown within perform a sensor check once per 12 hours, a channel functional test" (flow instruments only) I- See ITS 3.4.5 } . at least monthly, and a channel calibration test 0) -o K) the following 24 hours. at least once per cycle. to b) 03 b. Any time Irradiated fuel Is in the reactor vessel CO and reactor water temperature Is above 21 2F 0) -4 0 the drywell particulate radioactivity monitoring C7n I. system shall be operable. If the drywell to

                                                                                                                                                                                 -9' 01 t1 particulate radioactity monitoring system is not                                                                                                   CD Kl                              operable, then:                                                                                                                                   01 U1 I   -.             1) Analyze grab samples of the primary containment atmosphere once per 12 hours.                                          ^^  A functional test of this Instrument means Injection of a I

simulated signal Into the Instrument (not primary sensor)

                    ^ A mode change Is allowed when this system is                         to verify the proper Instrument channel response alarm I         Inoperable.                                                          and/or Initlating action.

3.6/4.6 126a 08/21/03 Amendment No. 417 137 W Page 2 of 2

Attachment 1, Volume 9, Rev. 0, Page 73 of 255 DISCUSSION OF CHANGES ITS 3.4.4, REACTOR COOLANT SYSTEM (RCS) OPERATIONAL LEAKAGE ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.6.D.1 .a and CTS 4.6.D.1 are applicable any time irradiated fuel is in the reactor vessel and reactor water temperature Is above 212 0F. ITS LCO 3.4.4 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring the RCS Operational LEAKAGE to be within applicable limits in MODE 2 when reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.6.D.1.a and CTS 4.6.D.1 is to ensure the RCS Operational LEAKAGE is within the applicable limits when the potential for LEAKAGE exists. The potential for LEAKAGE exists only when the reactor has a potential for being pressurized. In the ITS, this occurs in MODES 1, 2, and 3 since the reactor pressure vessel head is attached and all reactor head bolts are fully tensioned. In MODES 1 and 3, the reactor coolant temperature will always be above 2120F. In MODE 2, the reactor coolant temperature may be less than or equal to 212 0 F when the reactor is subcritical but control rods are withdrawn (thus a potential for pressurization exists). Therefore, it is necessary and acceptable to require the RCS Leakage Detection Systems to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 73 of 255

Attachment 1, Volume 9, Rev. 0, Page 74 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 74 of 255

Attachment 1, Volume 9, Rev. 0, Page 75 of 255 RCS Operational LEAKAGE 3.4.4 CTS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.6.D.1 3.4.4 RCS Operational LEAKAGE 3.6.D.1.a LCO 3.4.4 RCS operational LEAKAGE shall be limited to: 3.6.D.1.a.4) Ia. No pressure boundary LEAKAGE = 3.6.D.1.e.1) b.

5 gpm unidentified LEAKAG a 3.6.D.I.a.3) c.
fJprtotal LEAKAGE averaged over the previous 24 hour (

periodK n 3.6.D.I.a.2) Md. < 2 gpm increase in unidentified LEAKAGE within the previous hour period in MODE 1. i 0 3.6.D.1.a. APPLICABILITY: MODES 1, 2, and 3. 4.6.D.1 ACTIONS CONDITION J REQUIRED ACTION COMPLETION TIME 3.6.D.1.b A. Unidentified LEAKAGE A.1 Reduce LEAKAGE to within 4 hours not within limit. limits. OR Total LEAKAGE not within limit. Reduce LEAKAGEIo within 4 B.1 3.6.D.1.c B. Unidentified LEAKAGE B.1 -Reduce LEAKAGEsto within 4 hours increase not within limit. limits. OR B.2 Verify source of unidentified 4 hours LEAKAGE increase is not service sensitive type 304 or type 316 austenitic stainless steel. BWRI4 STS 3.4.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 75 of 255

Attachment 1,Volume 9, Rev. 0, Page 76 of 255 RCS Operational LEAKAGE 3.4.4 ACTIONS (continued) CONDITION REQUIRED ACTION j COMPLETION TIME 3.6.D.1.b. C. Required Action and C.1 Be in MODE 3. 12 hours 3.6.D.l.c associated Completion Time of Condition A or B AND not met. C.2 Be in MODE 4. 36 hours OR 3.6.D.1.d Pressure boundary LEAKAGE exists. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY UJ 4.6.D.1, 4.6.D.1.a. SR 3.4.4.1 Verify RCS unidentified and total LEAKAGE and unidentified LEAKAGE increase are within limits. l > ` 0 4.6.D.1.b, 4.6.D.1.c BWR/4 STS 3.4.4:2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 76 of 255

Attachment 1, Volume 9, Rev. 0, Page 77 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.4, REACTOR COOLANT SYSTEM (RCS) OPERATIONAL LEAKAGE

1. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.

2. The brackets have been removed and the proper plant specific information/value has been provided.

3. The Surveillance Frequency has been extended from 8 hours to 12 hours consistent with the current licensing basis.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 77 of 255

Attachment 1, Volume 9, Rev. 0, Page 78 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 78 of 255

Attachment 1, Volume 9, Rev. 0, Page 79 of 255 RCS Operational LEAKAGE B 3.4.4 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4.4 RCS Operational LEAKAGE BASES BACKGROUND The RCS includes systems and components that contain or transport the coolant to or from the reactor core. The pressure containing components of the RCS and the portions of connecting systems out to and including the isolation valves define the reactor coolant pressure boundary (RCPB). The joints of the RCPB components are welded or bolted. During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. Limits on RCS operational LEAKAGE are required to ensure appropriate action is taken before the integrity of the RCPB is impaired. This LCO specifies the types and limits of LEAKAGE. This protects the RCS pressure boundary described in 10 CFR 50.2, 10 CFR 50.55a(c), and DC 55 of A, ppendix A. } (RefE 1L .ai ) The safety significance of RCS LEAKAGE from the RCPB varies widely depending on the source, rate, and duration. Therefore, detection of LEAKAGE in the primary containment is necessary. Methods for quickly separating the identified LEAKAGE from the unidentified LEAKAGE are necessary to provide the operators quantitative information to permit them to take corrective action should a leak occur that is detrimental to the safety of the facility or the public. A limited amount of leakage inside primary containment is expected from auxiliary systems that cannot be made 100% leaktight. Leakage from these systems should be detected and isolated from the primary containment atmosphere, if possible, so as not to mask RCS operational LEAKAGE detection. This LCO deals with protection of the RCPB from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident. BWR/4 STS B 3.4.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 79 of 255

Attachment 1, Volume 9, Rev. 0, Page 80 of 255 RCS Operational LEAKAGE B 3.4.4 BASES APPLICABLE The allowable RCS operational LEAKAGE limits are based on the SAFETY predicted and experimentally observed behavior of pipe cracks. The ANALYSES normally expected background LEAKAGE due to equipment design and the detection capability of the instrumentation for determining system LEAKAGE were also considered. The evidence from experiments suggests that, for LEAKAGE even greater than the specified unidentified LEAKAGE limits, the probability is small that the imperfection or crack associated with such LEAKAGE would grow rapidly. The unidentified LEAKAGE flow limit allows time for corrective action before the RCPB could be significantly compromised. The 5 gpm limit is a small fraction of the calculated flow from a critical crack in the primary system piping. Crack behavior from experimental programs (Refs. 2 and 3) shows that leakage rates of hundreds of gallons per minute will H precede crack instability 4 The low limit on increase in unidentified LEAKAGE assumes a failure mechanism of intergranular stress corrosion cracking (IGSCC) that produces tight cracks. This flow increase limit is capable of providing an early warning of such deterioration. No applicable safety analysis assumes the total LEAKAGE limit. The total LEAKAGE limit considers RCS inventory makeup capability and drywell floor sump capacity. RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO RCS operational LEAKAGE shall be limited to:

a. Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material degradation. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

b. Unidentified LEAKAGE The 5 gpm of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that thelcontaineamonitormn 1 ETcr I Drryw-elsump ENO monitoring and con aunair cooler I condensatere monitoring equipment can detect within a J reasonable time period. Violation of this LCO could result in continued degradation of the RCPB.

BWR/4 STS B 3.4.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 80 of 255

Attachment 1, Volume 9, Rev. 0, Page 81 of 255 RCS Operational LEAKAGE B 3.4.4 BASES LCO (continued)

c. Total LEAKAGE The total LEAKAGE limit is based on a reasonable minimum detectable amount. The limit also accounts for LEAKAGE from known sources (identified LEAKAGE). Violation of this LCO indicates an unexpected amount of LEAKAGE and, therefore, could indicate new or additional degradation in an RCPB component or system.

d. Unidentified LEAKAGE Increase An unidentified LEAKAGE increase of > 2 gpm within the previous F4- hour period indicates a potential flaw in the RCPB and must be quickly evaluated to determine the source and extent of the 0

LEAKAGE. The increase is measured relative to the steady state value; temporary changes in LEAKAGE rate as a result of transient conditions (e.g., startup) are not considered. As such, the 2 gpm increase limit is only applicable in MODE 1 when operating pressures and temperatures are established. Violation of this LCO could result in continued degradation of the RCPB. APPLICABILITY In MODES 1, 2, and 3, the RCS operational LEAKAGE LCO applies, because the potential for RCPB LEAKAGE is greatest when the reactor is pressurized. In MODES 4 and 5, RCS operational LEAKAGE limits are not required since the reactor is not pressurized and stresses in the RCPB materials and potential for LEAKAGE are reduced. ACTIONS A.1 With RCS unidentified or total LEAKAGE greater than the limits, actions must be taken to reduce the leak. Because the LEAKAGE limits are conservatively below the LEAKAGE that would constitute a critical crack size, 4 hours is allowed to reduce the LEAKAGE rates before the reactor must be shut down. If an unidentified LEAKAGE has been identified and quantified, it may be reclassified and considered as identified LEAKAGE; however, the total LEAKAGE limit would remain unchanged. BWR/4 STS B 3.4.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 81 of 255

Attachment 1, Volume 9, Rev. 0, Page 82 of 255 RCS Operational LEAKAGE B 3.4.4 BASES ACTIONS (continued) B.1 and B.2 An unidentified LEAKAGE increase of > 2 gpm within a4 hour period is an indication of a potential flaw in the RCPB and must be quickly evaluated. Although the increase does not necessarily violate the absolute unidentified LEAKAGE limit, certain susceptible components must be determined not to be the source of the LEAKAGE increase within the required Completion Time. For an unidentified LEAKAGE increase greater than required limits, an alternative to reducing LEAKAGE increase to within limits (i.e., reducing the LEAKAGE rate such that the current rate is less than the "2 gpm increase in the previous hours" limit; either by (i) isolating the source or other possible methods) is to evaluate service sensitive type 304 and type 316 austenitic stainless steel piping that is subject to high stress or that contains relatively stagnant or intermittent flow fluids and determine it is not the source of the increased LEAKAGE. This type piping is very susceptible to IGSCC. The 4 hour Completion Time is reasonable to properly reduce the LEAKAGE increase or verify the source before the reactor must be shut down without unduly jeopardizing plant safety. C.1 and C.2 If any Required Action and associated Completion Time of Condition A or B is not met or if pressure boundary LEAKAGE exists, 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 to MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power condition's in an orderly manner and without challenging plant safety systems. SURVEILLANCE SR 3.4.4.1 REQUIREMENTS The RCS LEAKAGE is monitored by a variety of instruments designed to provide alarms when LEAKAGE is indicated and to quantify the various types of LEAKAGE. Leakage detection instrumentation is discussed in more detail in the Bases for LCO 3.4., '"RCS Leakage Detection WJ Instrumentation." Sump level and flow rate are typically monitored to BWR/4 STS B 3.4.4-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 82 of 255

Attachment 1, Volume 9, Rev. 0, Page 83 of 255 RCS Operational LEAKAGE B 3.4.4 an altemate method that may be used to Identify LEAKAGE Is via the drywell equipment drain sump monitoring system i.e., wvhen the drywell floor drain sump Is overflowing to the drywell equipment drain sump and the alarm settings are reset to detect the low limit for unidentifed LEAKAGE) BASES 11_ _ SURVEILLANCE REQUIREMENTS (continued)

                                                                                . -A determine actual LEAKAGE rates; however,lany rn,0thod may be used t I quantify LEAKAGE within the Guidelines of Referefice 5'. In conjunction with alarms and other administrative controls, aF Qjogu4r Frequency for                                            3 this Surveillance is ap opriate for identifying LEAKAGE and for tracking                                          {id required trends (Ref. A.                       -f3l REFERENCES       1. I 10CFR 5Q..Appi-ix A. GDC 3 .                                                                                   0D

2. GEAP-5620 April 1968. MA106BPipes I

                     ,                                                      Cntan ng Axal Through-Wall Flows,-        l

3. NUREG-067,ctober 1975.

Generic'Investigation and Evaluation of

4. 52/] InE~cin Austenitic Stainless Steel Ppn f A Gu/.5 I.4gltr - oln aerRatrPa

 -(      13-     . Generic Letter 88-ol,,sSupplement 1'                                    192    'rar r NRC Position on tntergranular Stress Corrosion ICracking (IGSCC) In BWR Austenitic Stainless LSteel Piping,'

BWR/4 STS B 3.4.4-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 83 of 255

Attachment 1, Volume 9, Rev. 0, Page 84 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.4 BASES, REACTOR COOLANT SYSTEM (RCS) OPERATIONAL LEAKAGE

1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

2. The brackets have been removed and the proper plant specific information/value has been provided.

3. Changes have been made to reflect those changes made to the Specification or other Specifications.

4. Typographical/grammatical error corrected.

Monticello Page 1of 1 Attachment 1, Volume 9, Rev. 0, Page 84 of 255

Attachment 1, Volume 9, Rev. 0, Page 85 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 85 of 255

Attachment 1, Volume 9, Rev. 0, Page 86 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.4, REACTOR COOLANT SYSTEM (RCS) OPERATIONAL LEAKAGE There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 86 of 255

Attachment 1, Volume 9, Rev. 0, Page 87 of 255 ATTACHMENT 5 ITS 3.4.5, Reactor Coolant System (RCS) Leakage Detection Instrumentation Attachment 1, Volume 9, Rev. 0, Page 87 of 255

Attachment 1, Volume 9, Rev. 0, Page 88 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 88 of 255

C C C ITS 3.4.5 ITS ITS 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS See ITS 3.4.4 } w d. if any Pressure Boundary Leakage exists, be in t d

                                                                                                          ,o,             t            s Ihot Sbnutlow witnin tri next i flours an in j                                                                                                                S 0

s tD 3.4.5 2 Cold Shutdown within the following 24 hours. RCS Leakage Detection Instrumentation R. dran p ovml oy reqft draki stmoP 2. ltemith5.d s.3OIo to the drytwU RCS Leakage Detection Instrumentation 7-0 0

   -   Applicability    t               n and readtor w
                                                                 ,                        a..

I Any time Irradiated fuel Is In the reactor vessel I temerature is above 212' the Drywell FloorlDralnSum gonioring g/

FlCS leakage detection Instrumentation shall be demonstrated OPERABLE by, I C' 0 2 0 0 LCO 3.4.5.a Systertshall be operable. IlltheJ rwl lo SR 3.4.5.1. a. Primary containment etmosphere particulate jurm~onlong amp aslel~snoloperable.\ SR 3.4.5.2, monitoring system - perform a sensor check then: \SR 3.4.5.3 once per 12 hours, a channel functional test at tD least monthly end a channel calibration at least ACTION A -- tD 0 1) ain ume once per 2 n A3 CD 0 M LI MofluotLrO _ within 30 days. tsemno operable status

b. Required leakage detection InstrumentatIon -

M (0

   -U                                                                                                    SR 3.4.5.1.

X ACTION D 2) Otherwise, be in Hot Shutdown within the SR 3.4.5.2. perform a sensor check once per 12 hours, a 0 0 channel functional test" (fow instruments only) next 12 hours and In Cold Shutdown within SR 3.4.5.3 at least monthly, and a channel calibration test .l CD the following 24 hours. 0 at least once pe ila co Applicability b. Any tine irradiated tuel Is in the reactor vessel 24' moth CD 0 tD and readorwalertemveratureisabove 212OF LCO 3.4.5.b _ the drywell particulate radioactivity monitoring I - co sysiem shall be operable.^/i the drywell U3 CD I 01 paticulale radioacivily monitoring system is not 01 operable, then: U1 tn ACTION B 01 tLt I 1) Analyze grab samples of the primary containment atmosphere once per 12 hours. " A functional test of this instrunmotIfmeans Injection of a Required simulated signal Into the ~l1ment (not primary sensor) A A mode change Is allowed when this system IsH X to verify the proper Inspinent channol response alarm Action A.1 I I.o4. and/or initiating actlin. Note 3.614.8 126a 08/21/03 Amendment No. 17 7, 137 Page 1 of 2

C C c ITS 3.4.5 ITS 3.0 LIMmNG CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 4

0) ACTION D 2) Otherwise, be In Hot Shutdown within the next 12 hours and In Cold Shutdown within W 0E the following 24 hours.

3 CD

      -Applicability 0

r ACTION C 0 particulate radioactIvity monitoring system) are 0

ad (0 Inoperable. restore at least one channel of the See ITS 3.4.3 and 3 CD required leakage detection Instrumentation to ITS 3.6.1.5 operable status within 1 houror be In Hot 0 0 Shutdown within the next 2 hours and In Cold ACTION D (1) Shuldown within the following 24 hours. (a E. Safety/Relief Valves 0 Safety/F L ED 0 E. CD

1. a. Safety/relief valves shall be tested or replaced 0)

                               . During power operating conditions and whenever                               each refueling outage In accordance with the to                               reactor coolant pressure Is greater than 10 psi                              Inservice Testing Program.

0

 .4 01) to valve function (self actuation) of seven safey/relie                      b. At least two of the safety/relief valves shag be                        CD CD                                                                                                                                                             See ITS 3.4.3}

a) valves shall be operable (note: Low-Low Set and disassembled and Inspected each refueling D ADS requirements are located In Specification I 'I outage. 0W UP 3.2.H. and 3.5.A, respectively). 0

c. The Integrity of the safety/relief valve bellows 0TnD Valves shall be set as follows: shall be continuously monitored.

Ln 8 valves at s 1120 psig

d. The operability of the bellows monitoring system

2. If Specification 3.6.E.1 is not met, initiate an ordert) shall be demonstrated each operating cycle.

shutdown and have reactor coolant pressure and temperature reduced to 110 psig or less and 345F 2. Low-Low Set Logic surveIllance shall be performed or less within 24 hours. In accordance with Table 4.2.1. 3.614.8 127 08/21/03 See ITS 3.4.3)

Amendment No. 30, 62, 76, 02, 0, 14442.2 137 Page 2 of 2

Attachment 1, Volume 9, Rev. 0, Page 91 of 255 DISCUSSION OF CHANGES ITS 3.4.5, REACTOR COOLANT SYSTEM (RCS) LEAKAGE DETECTION INSTRUMENTATION ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 4.6.D.2.b requires a CHANNEL FUNCTIONAL TEST of the required leakage detection instrumentation (flow instruments only) and footnote ** states "A functional test of this instrument means injection of a simulated signal into the instrument (not primary sensor) to verify the proper instrument channel response alarm and/or initiating action." ITS SR 3.4.5.2 requires the performance of a CHANNEL FUNCTIONAL TEST, but the footnote words are not included. This changes the CTS by deleting the modifying words of the footnote. The purpose of CTS 4.6.D.2.b is to require the performance of a CHANNEL FUNCTIONAL TEST, as modified by the allowance of footnote **. This change is acceptable because the allowance of the footnote are included in the ITS definition of CHANNEL FUNCTIONAL TEST. The allowance was added as described in ITS Chapter 1.0, DOC L.3. Therefore, this change is considered administrative because it does not result in technical changes to the CTS. A.3 CTS 4.6.D.2.a requires a CHANNEL CALIBRATION of the primary containment atmosphere particulate monitoring system at least once per operating cycle. CTS 4.6.D.2.b requires a CHANNEL CALIBRATION of the required leakage detection instrumentation at least once per operating cycle. ITS SR 3.4.5.3 requires similar tests every "24 months." This changes the CTS by changing the Frequencies from once per "operating cycle" to "24 months." This change is acceptable because the current "operating cycle" is "24 months." In letter L-MT-04-036, from Thomas J. Palmisano (NMC) to the USNRC, dated June 30, 2004, NMC has proposed to extend the fuel cycle from 18 months to 24 months and the same time has performed an evaluation in accordance with Generic Letter 91-04 to extend the unit Surveillance Requirements from 18 months to 24 months. CTS 4.6.D.2.a and 4.6.D.2.b were included in this evaluation. This change is designated as administrative because it does not result in any technical changes to the CTS. A.4 CTS 3.6.D.2.a allows 30 days to restore the inoperable drywell floor drain sump monitoring system to OPERABLE status. CTS 3.6.D.2.b allows the plant to operate continuously when the drywell particulate radioactivity monitoring system is inoperable as long as grab samples of the primary containment atmosphere are analyzed every 12 hours. CTS 3.6.D.2.a and CTS 3.6.D.2.b both include a footnote

that states, "A mode change is allowed when this system is inoperable." ITS 3.4.5 ACTION A covers the condition when LCO 3.4.5.a is not met (i.e., both the drywell floor drain sump monitoring system and the drywell Monticello Page 1 of 4 Attachment 1, Volume 9, Rev. 0, Page 91 of 255

Attachment 1, Volume 9, Rev. 0, Page 92 of 255 DISCUSSION OF CHANGES ITS 3.4.5, REACTOR COOLANT SYSTEM (RCS) LEAKAGE DETECTION INSTRUMENTATION equipment drain sump monitoring system are inoperable), and requires LCO 3.4.5.a to be met in 30 days. A Note is also included that states LCO 3.0.4.c is applicable. ITS 3.4.5 ACTION B covers the condition for when the drywell particulate radioactivity monitoring system is inoperable, and requires a grab sample to be analyzed every 12 hours. However, it does not include a Note similar to the ACTION A Note. This changes the CTS by deleting the modifying words of the footnote for the drywell particulate radioactivity monitoring system. Other changes to CTS 3.4.6.D.2.a are discussed in DOC A.5. The purpose of the CTS 3.6.D.2.a and b footnote (footnote *) is to allow entry into the Applicability of the Specification with an inoperable required RCS leakage detection monitoring system. Since footnote

is not applicable to CTS 3.6.D.2.c, it cannot be used when both systems are inoperable. ITS LCO 3.0.4 has been added in accordance with the Discussion of Changes for ITS Section 3.0, DOC L.1. This LCO allows entry into a MODE or other specified condition in the Applicability under certain conditions when a Technical Specification required component is inoperable. ITS LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability of a Specification when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. ITS LCO 3.0.4.c allows entry into a MODE or other specified condition in the Applicability of a Specification when an allowance is stated in the individual value, parameter, or other Specification. ITS 3.4.5 ACTION A requires LCO 3.4.5.a to be met within 30 days, therefore, the allowance in ITS LCO 3.0.4.c is needed to be consistent with the CTS allowance. On the other hand, the explicit allowance in LCO 3.0.4.c is not needed for the inoperable drywell particulate radioactivity monitoring system action (ITS 3.4.5 ACTION B) because this action allows continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This action only requires an analysis of primary containment atmosphere grab samples every 12 hours and does not require the inoperable drywell particulate radioactivity monitoring system to be restored to OPERABLE status. The allowances in ITS LCO 3.0.4.a apply to this ACTION and entry into the MODE or other specified condition in the Applicability will be allowed in the ITS. This change is acceptable because the allowances of the CTS footnote will apply in the ITS. This change is considered administrative because it does not result in technical changes to the CTS.

A.5 CTS 3.6.D.2.a requires the drywell floor drain sump monitoring system to be OPERABLE. CTS 3.6.D.2.a.1) covers the condition for an inoperable drywell floor drain sump monitoring system and it allows 30 days to restore the inoperable drywell floor drain sump monitoring system to OPERABLE status. CTS 3.6.D.2.c covers the condition for all channels of both systems (drywell floor drain sump monitoring system and drywell particulate radioactivity monitoring system) inoperable. ITS LCO 3.4.5.a requires either the drywell floor drain sump monitoring system or the drywell equipment drain sump monitoring system with the drywell floor drain sump overflowing into the drywell equipment drain sump system. ITS 3.4.5 ACTION A covers the condition when LCO 3.4.5.a is not met, and requires LCO 3.4.5.a to be satisfied. ITS 3.4.5 ACTION C covers the condition when all "required" leakage detection systems are inoperable. This Monticello Page 2 of 4 Attachment 1, Volume 9, Rev. 0, Page 92 of 255

Attachment 1, Volume 9, Rev. 0, Page 93 of 255 DISCUSSION OF CHANGES ITS 3.4.5, REACTOR COOLANT SYSTEM (RCS) LEAKAGE DETECTION INSTRUMENTATION changes the CTS by providing the option to allow the drywell equipment drain sump monitoring system with the drywell floor drain sump overflowing into the drywell equipment drain sump system to be used instead of the drywell floor drain monitoring system and adjusts the Actions, as required. The purpose of CTS 3.6.D.2.a is to ensure an approved method of monitoring unidentified LEAKAGE is available. The proposed option of allowing the drywell equipment drain sump monitoring system with the drywell floor drain sump overflowing into the drywell equipment drain sump to be used instead of the drywell floor drain sump monitoring system is currently allowed by the CTS Bases. This change is acceptable because this allowance was requested by NMC in a License Amendment Request dated January 29, 2003 and approved by the NRC in a Safety Evaluation to License Amendment 137, dated August 21, 2003. Therefore, this change is considered administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 CTS 3.6.D.2.a, CTS 3.6.D.2.b, and CTS 3.6.D.2.c are applicable any time irradiated fuel is in the reactor vessel and reactorwater temperature is above 2120 F. ITS LCO 3.4.5 is Applicable in MODES 1, 2, and 3. This changes the CTS by requiring the RCS Leakage Detection Instrumentation to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 212 0 F. The purpose of CTS 3.6.D.2.a, CTS 3.6.D.2.b, and CTS 3.6.D.2.c is to ensure the RCS Leakage Detection Instrumentation is OPERABLE when the potential for LEAKAGE exists. The potential for LEAKAGE exists only when the reactor has a potential for being pressurized. In the ITS, this occurs in MODES 1, 2, and 3 since the reactor pressure vessel head is attached and all reactor head bolts are fully tensioned. In MODES 1 and 3, the reactor coolant temperature will always be above 212 0F. In MODE 2, the reactor coolant temperature may be less than or equal to 2120 F when the reactor is subcritical but control rods are withdrawn (thus a potential for pressurization exists). Therefore, it is necessary and acceptable to require the RCS Leakage Detection Systems to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None Monticello Page 3 of 4 Attachment 1, Volume 9, Rev. 0, Page 93 of 255

Attachment 1, Volume 9, Rev. 0, Page 94 of 255 DISCUSSION OF CHANGES ITS 3.4.5, REACTOR COOLANT SYSTEM (RCS) LEAKAGE DETECTION INSTRUMENTATION LESS RESTRICTIVE CHANGES None e Monticello Page 4 of 4 Attachment 1, Volume 9, Rev. 0, Page 94 of 255

Attachment 1, Volume 9, Rev. 0, Page 95 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 95 of 255

Attachment 1, Volume 9, Rev. 0, Page 96 of 255 RCS Leakage Detection Instrumentation 3.4 REACTOR COOLANT SYSTEM (RCS) 3.6.D.2 3.4 RCS Leakage Detection Instrumentation 0 3.6.D.2.a, LCO The following RCS leakage detection instr 3.6.D.2.b OPERABLE: Orthe dweleu Ipmentdrain qstom wi the dreI lmponitoring umPp loordrain drain *urni: to toe eqtuipment

                                                                     'sump overflowing 3.6.D.2.a t arywell floor drain sump monil P

3.6.D2.b

                                                 . lune Cha              e            rimary cona monitoring systerrm atmosp enq particulate EZ
                                                                                                                                             } 0 J 0 3

[c. Primary contain cooler condensate flitoring I syste 3.6.D.2.a. APPLICABILITY: MODES 1, 2, and 3. 3.6.D2.b NOTE LCO 3.0.4.c Isapplicable. I k

                                                                                                                                               -0 ACTIONS                      3.4.5 not met CONDITION                                      REQUIRED ACTION                              COMPLETION TIME 3.6.D.2.a.

3.6.D.2.a.1) A. D elI floor am sump rronitoring sstem o eP A.1 Restor drywel floor drain] surnmonitor, g system Wt BLE tatuSatisfythe 30 days Lf - require entsL

                                                                                                                                             }05 3.6.D.2.b, 3.6.0.2.b.1)

B. Re uired pr a Iccqhtainme tF B.1 Analyze grab samples of primary containment Once per 12 hours 0 9 monitoring atmosphere. system inoperable.

                              \                      .       ~AND                           -

Drywell particulate radioactMty B.2 [Rest e required primary co sment atmospheric XZ0 days] 0 rti oring system to OPERABLE status./ BWR/4 STS 3.4.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 96 of 255

Attachment 1, Volume 9, Rev. 0, Page 97 of 255 RCS Leakage Detection Instrumentation CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. [ Primary containment C.1 -NOTE---- air cooler condensate Not applicable when flow rate monitoring required primary system inoperable. containment atmospheric monitoring system is inoperable. Perform SR 3.4.6.1. nce per 8 hours] 0 D. [Required primary D.1 Restore required primary 30 days containment containment atmospheric atmospheric monitor ng monitoring system to system inoperable. OPERABLE status. AND OR Primary conta ment air D.2 Restore primary 30 days] cooler conde sate flow containment air co ler rate monitor g system condensate flow rote inoperable monitoring syst to OPERABLE st us. 3.6.D.2.a.2), 3.6.D.2.b.2). 7 Required Action and associated ComDletion el Be in MODE 3. 12 hours 3.6.D.2.c 1 AND Time of Condion A,B [6or D] not met. 2 Be in MODE 4. 36 hours

908 3.6.D.2.c All required leakage detection systems 0 I I BWR/4 STS 3.4.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 97 of 255

Attachment 1, Volume 9, Rev. 0, Page 98 of 255 3.4.5 INSERT I C.1 Satisfy the requirements of 1 hour LCO 3.4.5.a. OR C.2 Restore drywell particulate radioactivity monitoring system to 1 hour OPERABLE status. Insert Page 3.4.6-2 Attachment 1, Volume 9, Rev. 0, Page 98 of 255

Attachment 1, Volume 9, Rev. 0, Page 99 of 255 RCS Leakage Detection Instrumentation /S) 3.4 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE leakage detection InstrumentationI FREQUENCY

                              -         r--l                 ---                  11 A_

4.6.D.2.a, 4.6.D2.b SR 3.4.9. Perform a CHANNEL CHECK of required:12 hours - U7 I containment atmosiaherremonitoring systen-l-4.6.D2.a SR 3.4 Perform a CHANNEL FUNCTIONAL TEST of 31 days 0 I required leak i . on instrumentation I

                                                                                            *1-4 M24 4.6.D.2.a, 4.6.D.2.b SR 3.43         Perform a CHANNEL CALIBRATION of required leakage detection instrumentation.

UT-[nfhs-' 00O the drywell particulate radioactivity monitoring system and the flow instrumentation of the required drywell drain sump monitoring system (D BWR/4 STS 3.4.6-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 99 of 255

Attachment 1, Volume 9, Rev. 0, Page 100 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.5, REACTOR COOLANT SYSTEM (RCS) LEAKAGE DETECTION INSTRUMENTATION

1. ISTS 3.4.6 is renumbered as ITS 3.4.5 since ISTS 3.4.5, "Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage," is not included in the Monticello ITS.

2. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.

3. The brackets have been removed and the proper plant specific information/value has been provided.

4. Changes have been made to reflect plant specific nomenclature and current licensing basis requirements.

5. The bracketed requirement/information (i.e., the primary containment air cooler condensate flow rate monitoring system) has been deleted since it is not applicable to Monticello. The following requirements have been renumbered, where applicable, to reflect this deletion.

6. A Note stating that "LCO 3.0.4.c is applicable" has been added to ITS 3.4.5 Required Action A.1. This allowance is consistent with footnote (*) to CTS 3.6.D.2.a, which was approved by the NRC in a recent Technical Specification amendment as documented in the Safety Evaluation Report for the amendment (Amendment 137, dated August 21, 2003).

7. The Surveillance Requirements have been revised to reflect the current requirements in CTS 4.6.D.2.a and b.

8. ISTS 3.4.6 ACTION F (ITS 3.4.5 ACTION C) has been modified to cover the condition when all required leakage detection systems are inoperable. The ACTION requires either to satisfy the requirements of LCO 3.4.5.a or to restore the drywell particulate radioactivity monitoring system within 1 hour. This is consistent with a recently approved Technical Specification Amendment (Amendment 137, dated August 21, 2003). In addition, the ACTION has been placed in the proper order.

ISTS 3.4.6 ACTION E (ITS 3.4.5 ACTION D) has been revised to require entry when proposed ITS 3.4.5 ACTION C is not met (by deleting the words "of Condition A, B, [C, or D]," which make ITS 3.4.5 ACTION D applicable to all ITS 3.4.5 ACTIONS).

9. The bracketed requirement to restore the required atmospheric monitoring system to OPERABLE status in 30 days (ISTS 3.4.6 Required Action B.2) has been deleted.

This is consistent with current licensing basis, and was recently approved by the NRC in a recent Technical Specification amendment (Amendment 137, dated August 21, 2003).

10. ISTS LCO 3.4.6.a (ITS LCO 3.4.5.a) has been mnodified to allow the option of allowing the drywell equipment drain sump monitoring system with the drywell floor drain sump overflowing into the drywell equipment drain sump to be used instead of the drywell floor drain sump monitoring system as is currently allowed by the CTS Bases. This change is consistent with current licensing basis, and was recently approved by the NRC in a recent Technical Specification amendment (Amendment 137, dated August 21, 2003). ITS 3.4.5 ACTION A has been modified to reflect this change to the LCO.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 100 of 255

Attachment 1, Volume 9, Rev. 0, Page 101 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 101 of 255

Attachment 1, Volume 9, Rev. 0, Page 102 of 255 RCS Leakage Detection Instrumentation/ B 3.4 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4. S Leakage Detection Instrumentation 0 BASES jUA eto4.. BACKGROUND OlDC 30 of 1Q~fAppend~ix A](Ref. 1), requires means for detecting and, to the extent practical, identifying the location of the source of RCS LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems. Limits on LEAKAGE from the reactor coolant pressure boundary (RCPB) are required so that appropriate action can be taken before the integrity of the RCPB is impaired (Ref. 2). Leakage detection systems for the RCS are provided to alert the operators when leakage rates above normal background levels are detected and also to supply quantitative measurement of leakage rates. The Bases for LCO 3.4.4, "RCS Operational LEAKAGE," discuss the limits on RCS LEAKAGE rates. Systems for separating the LEAKAGE of an identified source from an unidentified source are necessary to provide prompt and quantitative information to the operators to permit them to take immediate corrective action. drain sump fill rate. drainsumpflow,drain) LEAKAGE rom the RCPB inside the drywell is detected by at least one of IJ F independently monitored variables, such as'sump level 1f-our- changesaind drywell ase s an particulate radioactivity levels. The primary means of quantifying LEAKAGE in the drywell is the drywell floor drain sump monitoring system. l or pumped out of l id is ciassfied as) The drywell floor drain sump monitorin gsystem monitors the LEAKAGE collected irrthe floor drain sum ~i nidentified LEA AGE pns sts ofl LEAKAGE from contro rod drives, valve flanges or packing , floor drains the Closed Cooling rater System,aand drywell air cooling nit (i condensate drains nd anv LEAKAGE not collected in e dmell elrI equipment drain pii!md. Thelprimaryvo-iainmentlfloor drain sump has l I transmitters that supple in the main control room ndras I FINSERT 11 The floor drain sump level indtors have switches that start a stop the sump pumps when requ . A timer starts each time the p is pumped down to the w level setpoint. If the sump fio the high level CD setpoint before timer ends, an alarm sounds ie control room, indicating a AKAGE rate into the sump in ess of a preset limit. A flow in o in the discharge line of the drywell floor drain sump

                        ;pumps    provides       flow indicatio*nthe control room. She purcan alsoI                      (i
                          /estarted rcncontrol                roomr                                                :lo BW34      S346-                                                                    ev.03/10 BWR/4 STS                                           B 3.4.6-1                                      Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 102 of 255

Attachment 1, Volume 9, Rev. 0, Page 103 of 255 B 3.4.5 K) INSERT I a transmitter that supplies a level recorder in the control room and one drywell floor sump fill rate computer point (rate of change) which alarms in the control room. Q INSERT2 The drywell floor drain sump monitoring system also has a pump flow transmitter on the discharge line of the drywell floor drain sump pumps. Two drywell floor drain sump pumps take suction from the drywell floor drain sump and discharge to the floor drain collector tank. Three level switches are provided on the drywell floor drain sump. One level switch starts the auto pump, the second level switch starts the standby pump. The third level switch actuates a high level alarm in the control room. Hand switches for manual operation of the pumps and indicating lights are provided in the control room. The flow integrator monitors the flow when the pump(s) are operating. Q INSERT3 The drywell floor drain sump monitoring system measures the time interval between actuation of two level switches located in the drywell floor drain sump as it fills. At any time that the interval decreases over the previous minimum interval (indicating an increased leak rate) an alarm operates in the control room (computer point). Each method of LEAKAGE determination of the drywell floor drain sump monitoring system described above (level recorder in the control room, drywell floor sump fill rate, or flow integrator) is sensitive enough to detect leak rate changes better than one gallon per minute in a one hour period and therefore may be used to satisfy the LCO requirement. Insert Page B 3.4.6-1a Attachment 1, Volume 9, Rev. 0, Page 103 of 255

Attachment 1, Volume 9, Rev. 0, Page 104 of 255 B 3.4.5 Qj) INSERT 3A The plant also includes a drywell equipment drain sump monitoring system. This monitoring system normally monitors identified LEAKAGE collected in or pumped out of the drywell equipment drain sump. This identified LEAKAGE consists of LEAKAGE piped from recirculation pump seals, valve stem leakoffs, reactor well bulkhead and bellow drains, and the reactor vessel flange leakoff. The drywell equipment drain sump monitoring system includes the same types of instruments described for the drywell floor drain sump. An alternate to the drywell floor drain sump monitoring system is the drywell equipment drain sump system. Because of the physical size of the sumps, it is possible through detection or calculation to verify the operational unidentified LEAKAGE limit (! 5 gpm) and unidentified LEAKAGE rate limit (! 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1) during the period of time it takes to actually overflow from one sump to the other. Once the drywell floor drain sump is overflowing to the drywell equipment drain sump, the drywell equipment drain sump monitoring system can be used to quantify unidentified LEAKAGE. However, the alarm settings for the equipment drain sump instruments must be reset to detect the lower limit for unidentified leakage. In this condition, all additional LEAKAGE measured by the drywell equipment drain sump system is assumed to be unidentified LEAKAGE unless the leakage has been identified and quantified. Each method of LEAKAGE determination of the drywell equipment drain sump monitoring system (level recorder in the control room, drywell equipment sump fill rate, or flow integrator) as described for the drywell floor drain sump monitoring system is sensitive enough to detect leak rate changes better than one gallon per minute in a one hour period and therefore may be used to satisfy the LCO requirement. Insert Page B 3.4.6-1b Attachment 1, Volume 9, Rev. 0, Page 104 of 255

Attachment 1, Volume 9, Rev. 0, Page 105 of 255 RCS Leakage Detection Instrumentation B 3.4- (D BASES BACKGROUND (continued) I drywell particulate radioactivity I Thelprmary inmentaaid monitoring systerrn- continuously mcntohe primary containment atmosphere for airborne particulate and radioactivity. A sudden increase of radioactivity, which may be attributed eous 0 to RCPB steam or reactor water LEAKAGE, is annunciated in the control room. rina contaimint atmos herd particulate land eous radioactivity monitoring syste not capable of quantifying LEAKAGE rates, bu esensitive enough tolindicate increased-E rates of

                                  , gpm n 1 ou . Larger changes in                               rates are detected in imonnor leakage at

[low as I E-9 pCVcc as eas8 poortionally-hrtertimes (Ref.3. 0 Condensate from four of the primary containment coolers i outed to the primary containment or drain sump and is monitored a flow transmitter that provid indication and alarms in the coXrol room. This primary containm air cooler condensate flow rateqronitoring system serves as an a ed indicator, but not quantifier, RCS unidentified \ LEAKAGE APPLICABLE A threat of significant compromise to the RCPB exists if the barrier SAFETY contains a crack that is large enough to propagate rapidly. LEAKAGE ANALYSES rate limits are set low enoughto detect the LEAKAGE emitted from a (dryweln foor drain sump [monitoring system ma single crack in the RCPB (Refs~ndM NRd~ leava inside the drywell is designed with the capability of detecting ecionl 0D LEAKAGE less than the established LEAKAGE rate limits and providing appropriate alarm of excess LEAKAGE in the control room.LZZD 0D A control room alarm allows the operators to evaluate the significance of the indicated LEAKAGE and, if necessary, shut down the reactor for further investigation and corrective action. The allowed LEAKAGE rates are well below the rates predicted for critical crack sizes (Ref. M*ffJ Therefore, these actions provide adequate response before a significant 0D break in the RCPB can occur. RCS leakage detection instrumentation satisfies Criterion 1 of 10 CFR 50.36(c)(2)(ii). LCO ih/he drywell floor drain sump monitoring system s required to quantify the unidentified LEAKAGE from the RCS. Thus, for the system to be I1NSERT 4A considered OPERABLE,1flither the flow moaenno-r the sump ievel I )~ foneof thdtrain sump l I mdniloring pgfi TT ntte systeml must be OPERABL~D sThe other I, monitoring system monitoring systems provid(vearly alarms to the operatorksso3J-er examination of other detection systems will be made to determine the extent of any corrective action that may be required. With the leakage ) detection systems inoperable, monitoring for LEAKAGE in the RCPB is degraded. BWR14 STS B 3.4.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 105 of 255

Attachment 1, Volume 9, Rev. 0, Page 106 of 255 B 3.4.5 0 INSERT 4 The drywell particulate radioactivity monitoring system provides a means to detect changes in LEAKAGE rates (Ref. 1). 0 INSERT4A or the drywell equipment drain sump monitoring system with the drywell floor drain sump overflowing into the drywell equipment drain sump Q INSERT 4B or one of the three drywell equipment drain sump monitoring system methods must be OPERABLE with the drywell floor drain sump overflowing into the drywell equipment drain sump. Insert Page B 3.4.6-2 Attachment 1, Volume 9, Rev. 0, Page 106 of 255

Attachment 1, Volume 9, Rev. 0, Page 107 of 255 RCS Leakage Detection Instrumentation B 3.4. (F BASES APPLICABILITY In MODES 1, 2, and 3, leakage detection systems are required to be OPERABLE to support LCO 3.4.4. This Applicability is consistent with that for LCO 3.4.4. ___k3 [&TAIOT-0N7S 4-* /~ LO 3..5.anot met 1 0 With the drywell floor drai n ni orinq sstem inoperabl, no other de.l form of sampling can provide the equivalent information to quantify _=1oC~patry1cuoanit'F eaae However, th~rmr contai m eric activity monitory E g and the primary containm mt e ndensate flow rate moni or will provide indication of changes in leak 3.4.5.anotmet Withithe drywell floor drain ring system inoperabl, but with own RCS unidentified and total LEAKAGE being determined every A0 (SR 3.4.4.1), operation may continue for 30 days. The 30 day a eCompletion Time of Required Action A.1 is acceptable, based on operating experience, considering th mL fornril of leakage detection thallJlstill availab T _ C ~ 3 IINSNST5RT 6 B.11aS.1(A Withlbothgaj5Ebs and particulatenimary conlamem atmospheric monitoring channelE inoperable, grab samples of the primary containmentJd atmosphere must be taken and analyzed to provide periodic leakage informaltion. l [P-rovided a sample is obtained and finalyzed once every lA j12 hours, the plan~tmay be operated for up to 30/days; to allow restortv lof at least one of She required monitors.11P-rovided a sample is obtaine(E and analyzed every 12 hours, the plant may continue operation since at nornaliv least one other form of drywell leakage detection (i.e., air o er Icondensate rate monito ) i availablej drywell floor drain sump monitoring system or the drywell The 12 hour interval provides periodic information that is adequate to equipment drain detect LEAKAGE. [The70 day Completion Time or restoration sump monitoring system with the recognizes that at lea( one other form of leakae detection is available. drywell floor drain sump overflowing into 0 the drywell equipment drain sump BWR/4 STS B 3.4.6-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 107 of 255

Attachment 1, Volume 9, Rev. 0, Page 108 of 255 B 3.4.5 0 INSERT 5 and the fact that the LEAKAGE is still being determined every 12 hours. 0 INSERT 6 Required Action A.1 is modified by a Note that states that the provisions of LCO 3.0.4.c are applicable. As a result, a MODE change is allowed when LCO 3.4.5.a is not met. This allowance is provided because other instrumentation is normally available to monitor RCS leakage. Insert Page B 3.4.6-3 Attachment 1, Volume 9, Rev. 0, Page 108 of 255

Attachment 1, Volume 9, Rev. 0, Page 109 of 255 RCS Leakage Detection Instrumentation 5 B 3.4 0 BASES ACTIONS (continued) [-i With the required primary co ainment air cooler condensate fi w rate monitoring system inoperabi SR 3.4.6.1 must be performed very 8 hours to provide periodic i formation of activity in the prima containment at a more freq ent interval than the routine Fre uency of SR 3.4.7.1. The 8 hour in erval provides periodic informati that is adequate to detect LEA GE and recognizes that other fo ms of leakage detection are available. ,flowever, this Required Action is odified by a Note that allows this acton to be not applicable if the req ired primary containment atmospheic monitoring system is inoperabf . Consistent with SR 3.0.1, Surveil nces are not required to be perf rmed on 0D inoperable equipme .1 [D.1 and D.2 With both the pri ary containment gaseous and p ticulate atmospheric monitor channej and the primary containment ai cooler condensate flom rate monitor in perable, the only means of detec ing LEAKAGE is the drywell floor dain sump monitor. This conditio does not provide the required dive se means of leakage detection. he Required Action is to restore eith of the inoperable monitors to 0 RABLE status within 30 days to gain the intended leakage detecion diversity. The 30 day Completio Time ensures that the plant will ot be operated in a degraded onfiguration for a lengthy time pnod.] I0-11--COW and F[.2 I and associated Completion Time I If any Required Action of Condit. [C. or D11cannot be met w :n Ithe associatedr; etion Time, 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 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. With all required monitors inoperableJi required automatic means of monitoring LEAKAGE are availabl [accordance wji .3 is required immea iania rsr urdown In I INER I 0 BWR/4 STS B 3.4.64 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 109 of 255

Attachment 1, Volume 9, Rev. 0, Page 110 of 255 B 3.4.5 0 INSERT 7 Therefore, either LCO 3.4.5.a must be satisfied or the drywell particulate radioactivity monitoring system must be restored to OPERABLE status within 1 hour. The 1 hour Completion Time is acceptable because the likelihood of an increase in LEAKAGE in this 1 hour period is small and ensures the plant will not be operated in a degraded configuration for a lengthy time period. Insert Page B 3.4.64 Attachment 1, Volume 9, Rev. 0, Page 110 of 255

Attachment 1, Volume 9, Rev. 0, Page 111 of 255 RCS Leakage Detection Instrumentation/ B 3.4.t (9 BASES P SURVEILLANCE lieakage detection in LCO 3.4.5.a andInstrumentation channels the drywell particulate (both the required radioactivity equipment monitoring system)zspecified I ) REQUIREMENTS This SR is for the performance of a CHANNEL CHECK of the requred A I primary contaInmgniieq%7bnc monitorngq systerff. 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. JC 3.4 .2 rdrywell particulate radioactivity monitoring system and the flow Instrumentation of the required drain sump monitoring system I(drywehl floor or drywell equipment) J This SR is for the performance of a CHANNEL FUNCTIONAL TEST of[ thelrequired RCS lea eteon instrumentation: I he test ensures that the monitors can perform their function in the desired manner Th 0 test alserifies the alasetpoint and relative accura the ins ment string. A successful test of the required contact(s) of a 0D 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 acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the other required 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 Frequency of 31 days considers instrument reliability, and operating experience has shown it proper for detecting degradation. SR 3.4 .3 0 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 of months is a typical refueling cycle and considers channel reliability. perating experience has roven this Frequency is acceptable. 24 0D REFERENCES 1. 110 CFR 50, p IXA, USAR. Section 4.3.3.3 0

2. Regulatory Guidel1.45, May`1973. A FSARl.1. -lure5Behavior a3. InASTM A106B Pipes Containing C) 0-iir AaGEAP-5620Flaws,i Arough-Wall (ZFT NUREG-75/067, October 1975.

                                                                     'Investigation and Evaluation of Cracking In Austenitic Stainless Steel Piping of Boiling Water Reactor Plants.'

0 lo. "'AR, S-ectionF[5.2i (5.2 . 0 6.3.1 BWRI4 STS B 3.4.6-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 111 of 255

Attachment 1, Volume 9, Rev. 0, Page 112 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.5 BASES, REACTOR COOLANT SYSTEM (RCS) LEAKAGE DETECTION INSTRUMENTATION

1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

2. The brackets have been removed and the proper plant specific information/value has been provided.

3. Changes have been made to reflect those changes made to the Specification. The following requirements have been renumbered, where applicable, to reflect the changes.

4. Typographical/grammatical error corrected.

5. This sentence has been deleted since it is not consistent with the definition of CHANNEL FUNCTIONAL TEST (CFT) in ITS Section 1.1. A CFT does not require the setpoint to be checked or the relative accuracy of the instrument string to be verified.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 112 of 255

Attachment 1, Volume 9, Rev. 0, Page 113 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 113 of 255

Attachment 1, Volume 9, Rev. 0, Page 114 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.5, REACTOR COOLANT SYSTEM (RCS) LEAKAGE DETECTION INSTRUMENTATION There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 114 of 255

Attachment 1, Volume 9, Rev. 0, Page 115 of 255 ATTACHMENT 6 ITS 3.4.6, Reactor Coolant System (RCS) Specific Activity Attachment 1, Volume 9, Rev. 0, Page 115 of 255

Attachment 1, Volume 9, Rev. 0, Page 116 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 116 of 255

C C C ITS 3.4.6 0 ITS ITS 3.0 LIMIING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

4. The reactor vessel head boling studs shall not be 4. en the reactor vessel head studs areunder under tensIon unless the temperature of the vessel S ITS 3.4.9 tension and the reactor Is in the Cold Shuldown

0) head flange and the head are Z700 F. Condition, the reactor vessel shell flange temverature shall be permanently rrlrnrrtai S

Dl 0 _ . _. . -t 3.4.6 C. Coolant Chemistry R r nt System Specific Atvy

1. (a) The steady state radioodine concentration In

(;iA)C. Coolant Chemistry 0 LCO 3.4.6 0 the reactor coolant shall not exceed 2.0 1. (a) A sample of reactor colant shall be taken at microcrires of 1-131 dose equivalent per gram least ever yh r adena ed or of waler. SR 3.4.6.1 radloactive lodines of 1-131 through 1.135 Idu o roMODE1 CD (b) The steady state radiolodine concentration In Co Ihe reactor coolant shall not exceed 0.02 ,_ (b) A sample of reactor coolant shall be taken and Dl microcuries of 1-131 dose equivalent per gram analyzed for radioactive lodines of 1-13 1 00 0 of waler when the reactor coolant temperature through 1.135 wfthin 24 hours prior to raising (D is > 21 2F, the reactor is not critical, and primary containment integriy has not been the reactor coolant temperature >212F. with the reactor not critical, and with primary See ITS 3.10.1 } Dl establlshed. 0 containment Integrity not established. rD 716 X L. _- 01 tD -4 Co _P U' Add proposed APPLICABIIflY 6 rX f-la 3.6/4.6 123 03107/01 Amendment No. Or10114107r 117 Page 1 of 3

C C C ITS 3.4.6 0 3.0 LIMmNG CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS I (c) When the ma condenser offgas sysm .I pretreatmep monitors Indicate an irease in a)

2) 0 L3 radioacf gaseous ealluents of 0

C) ste state percent or 5000 1/sec, whichever Is gr er, during iy reactor operati a reactor coolant . a 0 0 a mpie shall be taken an nalyzod for

                                                           /rndiective lodines.                              I                   0 2                                                      l(d) Isoto Ianalysis of r ctor coolant sa       es s     be made at leest once per                       L3 271 (e) Wheneve e steady stater diolodina concen lion of prir opefain is greater th                             0 3                                                                                                                                 CD 1 perg nt but less than 0 percent of CD 0                                                           Sp ficatton 3.6.C.1 A) a sample of re or cant shall be la n within 24 hour f any 0

1 L3-E) eador startup a analyzed for oactive 0 0 /odines of 1-13 hrough 1-135.

CD M 6 (I) Whenever the sfidy state radhioldine 0 concentration prlor operalton i greate han IV 10 percent Section 3.6.C.1.(a). a sq ple of reacor 1ant shall be taken daily d prior to any reador startup and analyz or radioactive L3e a) todi s of l-t31 through l-135 awell as the 10 c olant sample and analyserequired by co

                                                           /pacifcation 4.6.C.1.(e) aove.

N 01 tn 3.614.6 124 07/25101 Amendment No. 1-4O7.121 Page 2 of 3

C C C ITS 3.4.6 0 ITS 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

2. During stnirup and at stoaming rates below 100,000

2. (a) The reactor coolant water shalt not exceed the a) following limits with steaming rates less than pounds per hour, a sample of reactor coolant shalt n) a 0 0 100.000 pounds per hour except nasspedined In be taken every four hours and analyzed for 3.6.C.2.b. conductivity and chloride content.

a) Conductivity 5 pmhoScm 2r= Chloride ion 0.1 ppm 0 (b) For rractor stArtulipS the maxImurm value for conductivity shall not exceed 10 jumhotcm and { (0 thte maximum value for clorido Ion concentration shaltnot exceed 0.1 ppm for the liMr 24 hours after placing thie reactor In the power operatIng condition. l See CTS Y4.6.C.2. CTS 314.6.C.3. rand CTS 3/4.6.C.4

                                                                                                                                                                                              }  a CD 0rs" 3,

3. Except as specified In 3.G.C.Zb above, the reaclor 3.(n) With steaming rates greater than or equal to 0 coalant wnter shal not exceed Itie folowing limits 100.000 lis. per hour, a reactor coolant sample wth steaming rates greater than or equal to shall be taken at least every Pa hours and wheni the 100.000 lbs. per hour. continuous conductivity monitors Indicate abnormal CD 3

-U conducttvity (other than short-term spikes) and Conductivty 5 jimholcm analyzed for conducivity and chloride Ion content. a a, a) rhkmrto Ion n!, nwnI 1-(h) When the contintious eonductivily monitor is

4. If Specifictions 3.5.C.1 Iiousgt!3.0.C.3 re nol~ Inoperabte, during power operation, a reactor (n ACTION B met.%n orderly shutdown shal be initated eia d Ihe e coolant samplt shou ld be tnken once per 12 hours I 0 reactor shal bo in~ho cold shutdown conilition arid nnalyzed for conductivIty and chlorkn Ion 0 withinlv hours. eonnirt Ur' I MODE 3 In Add proposed Add proposed Required Add pnvposed Required Ut I2hoursand ACTION A O Action B.1 II 1 Ut Add proposed Reurseird L2 \ L Ain B.2.1 L5 3.0/4.6 - 125 12/24193 Amendment No. 104 Page 3 of 3

Attachment 1, Volume 9, Rev. 0, Page 120 of 255 DISCUSSION OF CHANGES ITS 3.4.6, REACTOR COOLANT SYSTEM (RCS) SPECIFIC ACTIVITY ADMINISTRATIVE CHANGES A.1 Inthe conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 If the iodine concentration limits are not met, CTS 3.6.C.4 requires the unit to be shutdown. ITS 3.4.6 ACTION B also requires a unit shutdown (see DOC L.4 for the change related to when the unit shutdown commences), but also requires a periodic determination of DOSE EQUIVALENT 1-131 every 4 hours while the unit is being shut down (ITS 3.4.6 Required Action B.1). This changes the CTS by requiring a periodic determination of DOSE EQUIVALENT 1-131 to monitor any changes in specific activity during the unit shutdown. The purpose of ITS 3.4.6 Required Action B.1 is to ensure changes in the specific activity concentration are adequately monitored. This is acceptable because it ensures a proper response to an increasing specific activity level during the shutdown can be taken. This change is designated as more restrictive because more stringent sampling requirements are being applied in the ITS than in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 7 - Relaxation of Surveillance Frequency, Non-24 Month Type Change) CTS 4.6.C.1.(a) requires a reactor coolant sample be taken and analyzed for radioactive iodines of 1-131 through 1-135 every 96 hours during power operation. ITS SR 3.4.6.1 requires verifying DOSE EQUIVALENT 1-131 specific activity every 7 days when in MODE 1. This changes the CTS by extending the Surveillance Frequency from 96 hours to 7 days and eliminates the requirement to perform the Surveillance in MODE 2 above 1% Rated Thermal Power (RTP). Monticello Page 1 of 4 Attachment 1, Volume 9, Rev. 0, Page 120 of 255

Attachment 1, Volume 9, Rev. 0, Page 121 of 255 DISCUSSION OF CHANGES ITS 3.4.6, REACTOR COOLANT SYSTEM (RCS) SPECIFIC ACTIVITY The purpose of CTS 4.6.C.l.(a) is to ensure Reactor Coolant System iodine is within limits so that any release of radioactivity to the environment during a main steam line break (MSLB) outside primary containment is less than a small fraction of the 10 CFR 100 and 10 CFR 50, Appendix B, GDC 19 limits. This change extends the Surveillance Frequency from 96 hours to 7 days and eliminates the requirement to perform the Surveillance in MODES 2 above 1% RTP. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of reliability. The change in Surveillance Frequency from 96 hours to 7 days maintains a Frequency that is adequate to trend changes in the iodine activity level in MODE I because the level of fission products generated in other MODES is much less. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.2 (Category 2 - Relaxation of Applicability) CTS 3.6.C does not specify Applicability requirements. The Applicability of MODES 1, 2, and 3 is inferred in the CTS 3.6.C.4 Action to place the plant in cold shutdown when the Specification is not met. ITS 3.4.6 Applicability is MODE 1, and MODES 2 and 3 with any main steam line not isolated. In addition, as a result of this change, ITS 3.4.6 Required Action B.2.1 allows all main steam lines to be isolated in lieu of a shutdown to MODE 4 (i.e., cold shutdown). This changes the CTS by specifying that MODES 2 and 3 are applicable only when the main steam lines are not isolated. The purpose of the CTS 3.6.C Applicability is to specify the MODES and conditions in which a potential for iodine leakage to the environment exists due to a main steam line break (MSLB) outside primary containment. This change is acceptable because the requirements continue to ensure that the process variable is maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. This change allows, in MODES 2 and 3 with the main steam lines isolated, the limits on primary coolant radioactivity to not apply because an escape path does not exist for release of radioactive material from the primary coolant to the environment in the event of an MSLB outside of primary containment. This change is designated as less restrictive because LCO requirements in ITS are applicable in fewer operating conditions than in CTS. L.3 (Category 5 - Deletion of Surveillance Requirement) CTS 4.6.C.1 .(c) requires a reactor coolant sample be taken and analyzed for radioactive iodines when the main condenser offgas system pretreatment monitors indicate an increase in radioactive gaseous effluents of 25 percent or 5000 pCi/sec, whichever is greater, during steady state reactor operation. CTS 4.6.C.1 .(d) requires an isotopic analysis of reactor coolant samples be made at least once per month. CTS 4.6.C.1.(e) requires a reactor coolant sample be taken within 24 hours of any reactor startup and analyzed for radioactive iodines 1-131 through 1-135 whenever the steady state radioiodine concentration of prior operation is greater than 1 percent but less than 10 percent of Specification 3.6.C.1.(a). CTS 4.6.C.1 .(f) requires a reactor coolant sample be taken daily and prior to any reactor startup and analyzed for radioactive iodines 1-131 through 1-135 whenever the steady state radioiodine concentration of prior operation is greater than 10 percent of Specification 3.6.C.1.(a). These Surveillance Requirements Monticello Page 2 of 4 Attachment 1, Volume 9, Rev. 0, Page 121 of 255

Attachment 1, Volume 9, Rev. 0, Page 122 of 255 DISCUSSION OF CHANGES ITS 3.4.6, REACTOR COOLANT SYSTEM (RCS) SPECIFIC ACTIVITY are not retained in the ITS. This changes the CTS by deleting Surveillance Requirements to sample and analyze for radioactive iodines that are event based and a Surveillance that requires a complete isotopic analysis monthly. The purpose of CTS 4.6.C.1.(c), CTS 4.6.C.l.(d), CTS 4.6.C.1.(e), and CTS 4.6.C.1.(f is to ensure Reactor Coolant System iodine is within limits so that any release of radioactivity to the environment due to a MSLB outside primary containment is less than a small fraction of the 10 CFR 100 and 10 CFR 50, Appendix B, GDC 19 limits. This change is acceptable because the deleted Surveillance Requirements are not necessary to verify that the reactor coolant specific activity used to meet the LCO is consistent with the safety analysis. Thus, appropriate values continue to be verified in a manner and at a Frequency necessary to give confidence that the assumptions in the safety analysis are protected. This change is acceptable because proposed ITS SR 3.4.6.1 provides the required Surveillance to ensure the radiological dose at the exclusion area boundary is within the requirements of 10 CFR 100 and 10 CFR 50, Appendix B, GDC 19. This change is designated as less restrictive because Surveillances required in the CTS will not be required in the ITS. L.4 (Category 4 - Relaxation of Required Action) CTS 3.6.C.4 requires an orderly shutdown be initiated and the reactor be in cold shutdown condition within 24 hours if the iodine concentration limit is not met. ITS 3.4.6 ACTION A allows 48 hours to restore the iodine concentration to within limits (ITS 3.4.6 Required Action A.2) prior to requiring a unit shutdown, provided the iodine concentration is < 4.0 pCi/gm. However, during this 48 hour period, a determination of DOSE EQUIVALENT 1-131 is required every 4 hours (ITS 3.4.6 Required Action A.1). In addition, during this 48-hour period, changes in MODES or other specified conditions in the Applicability are allowed (ITS 3.4.6 Required Action A Note). This changes the CTS by allowing 48 hours, under certain conditions, to restore iodine concentration to within limits prior to requiring a unit shutdown. The change also allows changes in MODES or other specified condition in the Applicability during this 48 hour time. The purpose of CTS 3.6.C.4 is to ensure the plant reactor coolant specific iodine activity is placed in a condition that will reduce the amount of activity released to the environment due to a MSLB outside primary containment to a value less than a small fraction of the 10 CFR 100 and 10 CFR 50, Appendix B, GDC 19 limits. The proposed ACTION A allows 48 hours, under certain conditions, to restore iodine concentration to within limits prior to requiring a unit shutdown. The change also allows changes in MODES or other specified condition in the Applicability during this 48-hour time. This change is acceptable because the Required Actions are used to establish remedial measures taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to restore inoperable features. The Required Actions are consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the restoration period. This change ensures the reactor coolant specific activity level is maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. The change requires a Monticello Page 3 of 4 Attachment 1, Volume 9, Rev. 0, Page 122 of 255

Attachment 1, Volume 9, Rev. 0, Page 123 of 255 DISCUSSION OF CHANGES ITS 3.4.6, REACTOR COOLANT SYSTEM (RCS) SPECIFIC ACTIVITY periodic determination of DOSE EQUIVALENT 1-131 to monitor any changes in specific activity during the 48 hour period. If the reactor coolant specific activity exceeds 4.0 pCi/gm DOSE EQUIVALENT 1-131, a unit shutdown will be required. In addition, the requirement to restore within 48 hours provides a reasonable time for temporary coolant activity increases to be cleaned up with normal processing systems. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in CTS. L.5 (Category 3 - Relaxation of Completion Time) CTS 3.6.C.4 requires an orderly shutdown be initiated and the reactor be in cold shutdown condition within 24 hours if the iodine concentration limit is not met. ITS 3.4.6 Required Actions B.2.2.1 and B.2.2.2 require the reactor be in MODE 3 in 12 hours and in MODE 4 in 36 hours, respectively. This changes the CTS by adding a requirement to be in MODE 3 in 12 hours and by extending the time allowed to be in cold shutdown (i.e., MODE 4) from 24 hours to 36 hours. The purpose of CTS 3.6.C.4 is to place the unit in a MODE in which the LCO is no longer applicable within a reasonable amount of time. This change is acceptable because the Completion Time is consistent with the safe operation under the specified Condition, considering the OPERABLE status of the redundant systems or features. This includes the capacity and capability of remaining systems or features, a reasonable time for repairs or replacement, and the low probability of a DBA occurring during the allowed Completion Time. This change is also acceptable because it requires the unit to be in an intermediate condition (MODE 3) sooner than is currently required. This portion of the change reduces the amount of time the unit would be allowed to continue to operate in MODES 1 and 2 once the condition is identified. In addition, the time allowed for being in MODE 4 is extended from 24 hours to 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. This change is designated as less restrictive because additional time is allowed to place the unit outside the LCO Applicability than is allowed in the CTS. Monticello Page 4 of 4 Attachment 1, Volume 9, Rev. 0, Page 123 of 255

Attachment 1, Volume 9, Rev. 0, Page 124 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 124 of 255

Attachment 1, Volume 9, Rev. 0, Page 125 of 255 RCS Specific Activity 3.41 ,J CTS 0 3.4 REACTOR COOLANT SYSTEM (RCS) 3.41 3.6.C RCS Specific Activity 0 3.6.C.1.(a) LCO 3.4 The specific activity of the reactor coolant shall be limited to DOSE EQUIVALENT 1-131 specific activity< S* Ci/gm.

E 0D APPLICABILITY: MODE 1, MODES 2 and 3 with any main steam line not isolated.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC L.4 A. Reactor coolant specific activity > iCi/gm

                                                -------       NOTE--

LCO 3.0.4.c is applicable. 0D and 5 4.0 pCi/gm DOSE EQUIVALENT 1-131. A.1 Determine DOSE Once per 4 hours EQUIVALENT 1-131. AND A.2 Restore DOSE 48 hours EQUIVALENT 1-131 to within limits. B. Required Action and B.1 Determine DOSE Once per 4 hours 3.6.C.4 associated Completion EQUIVALENT 1-131. Time of Condition A not met. AND OR B.2.1 Isolate all main steam lines. 12 hours Reactor Coolant specific OR activity >P.0g pCi/gm 0 DOSE EQUIVALENT B.2.2.1 Be in MODE 3. 12 hours 1-131. AND B.2.2.2 Be in MODE 4. 36 hours BWR/4 STS 3.4.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 125 of 255

Attachment 1, Volume 9, Rev. 0, Page 126 of 255 RCS Specific Activity CTS 3.4% 0 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.6.C.1 .(a) SR 3.4A NOTES---- 0D Only required to be performed in MODE 1. Verify reactor coolant DOSE EQUIVALENT 1-131 7 days specific activity is

0 BWR/4 STS 3.4.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 126 of 255

Attachment 1, Volume 9, Rev. 0, Page 127 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.6, REACTOR COOLANT SYSTEM (RCS) SPECIFIC ACTIVITY

1. ISTS 3.4.7 is renumbered as ITS 3.4.6 since ISTS 3.4.5, "Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage," is not included in the Monticello ITS.

2. The brackets have been removed and the proper plant specific information/value has been provided.

Monticello Page 1 of I Attachment 1, Volume 9, Rev. 0, Page 127 of 255

Attachment 1, Volume 9, Rev. 0, Page 128 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 128 of 255

Attachment 1, Volume 9, Rev. 0, Page 129 of 255 RCS Specific Activity B 3.4 B 3.4 REACTOR COOLANT SYSTEM (RCS) 0 B 3.4 gCS Specific Activity BASES BACKGROUND During circulation, the reactor coolant acquires radioactive materials due to release of fission products from fuel leaks into the reactor coolant and activation of corrosion products in the reactor coolant. These radioactive materials in the reactor coolant can plate out in the RCS, and, at times, an accumulation will break away to spike the normal level of radioactivity. The release of coolant during a Design Basis Accident (DBA) could send radioactive materials into the environment. Limits on the maximum allowable level of radioactivity in the reactor coolant are established to ensure that in the event of a release of any radioactive material to the environment during a DBA, radiation doses are maintained within the limits of 10 CFR 100 (Ref. 1). This LCO contains iodine specific activity limits. The iodine isotopic activities per gram of reactor coolant are expressed in terms of a DOSE EQUIVALENT 1-131. The allowable levels are intended to limit the 2 hour radiation dose'to an individual at the site boundary to a small fraction of the 10 CFR 100 limit. APPLICABLE SAFETY Analytical methods and assumptioninvolving radioactive material in the primary coolant are presented in thelAR (Ref. 2). The specific activity 0D ANALYSES in the reactor coolant (the source term) is an initial condition for evaluation of the consequences of an accident due to a main steam line break (MSLB) outside containment. No fuel damage is postulated in the MSLB accident, and the release of radioactive material to the environment is assumed to end when the main steam isolation valves (MSIVs) close completely. and control roam This MSLB release forms the basis for determining offsite~doses (Ref. 2). The limits on the specific activity of the primary coolant ensure that the 0D 2 hour thyroid and whole body doses at the site boundary, resulting from Te limits on the specific an MSLB outside containment during steady state operation, will not activity of the primary coolant also ensure the thyroid dose to control room operators. exceed 10% of the dose guidelines of 10 CFR 100Jr 0D resulting from a MSLB outside containment during steady state operation will The limits on specifi adivityl typical site locations. T valuo m a parametric evaluation of limits onservative because the 0 not exceed the limits of 10 evaluation considered more restrictive parameters than for a specific site, CFR 50. Appendix A GDC such as the location of the site boundary and the meteorological 19 (Ref. 3). conditions of the site. RCS specific activity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). BWR/4 STS B 3.4.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 129 of 255

Attachment 1, Volume 9, Rev. 0, Page 130 of 255 RCS Specific Activity B 3.4X 0 ED BASES ,-F2.10 LCO The specific iodine activity is limited to pCi/gm DOSE EQUIVALENT 1-131. This limit ensures the source term assumed in the 0 safety analysis for the MSLB is not exceeded, so any release of radioactivity to the environment during an MSLB is less than a small fraction of the 10 CFR 100 limits I and 10 CFR 50, Appendix A,GDC 19 (Ref. 3)limits 0 APPLICABILITY In MODE 1, and MODES 2 and 3 with any main steam line not isolated, limits on the primary coolant radioactivity are applicable since there is an escape path for release of radioactive material from the primary coolant to the environment in the event of an MSLB outside of primary containment. In MODES 2 and 3 with the main steam lines isolated, such limits do not apply since an escape path does not exist. In MODES 4 and 5, no limits are required since the reactor is not pressurized and the potential for leakage is reduced. ACTIONS A.1 and A.2 When the reactor coolant specific activity exceeds the LCO DOSE EQUIVALENT 1-131 limit, but is s 4.0 pCi/gm, samples must be analyzed for DOSE EQUIVALENT 1-131 at least once every 4 hours. In addition, the specific activity must be restored to the LCO limit within 48 hours. The Completion Time of once every 4 hours is based on the time needed to take and analyze a sample. The 48 hour Completion Time to restore the activity level provides a reasonable time for temporary coolant activity increases (iodine spikes or crud bursts) to be cleaned up with the normal processing systems. A Note permits the use of the provisions of LCO 3.0.4.c. This allowance permits entry into the applicable MODE(S) while relying on the ACTIONS. This allowance is acceptable 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. B.1. B.2.1, B.2.2.1. and B.2.2.2 If the DOSE EQUIVALENT 1-131 cannot be restored to < pCi/gm 0 within 48 hours, or if at any time it is > 4.0 pCi/gm, it must be determined at least once every 4 hours and all the main steam lines must be isolated within 12 hours. Isolating the main steam lines precludes the possibility of releasing radioactive material to the environment in an amount that is more than a small fraction of the requirements of 10 CFR 101during a postulated MSLB accident. I and 10 CFR 50, Appendix A, GDC 19 (Ref. 3) 0 BWRI4 STS B 3.4.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 130 of 255

Attachment 1,Volume 9, Rev. 0, Page 131 of 255 RCS Specific Activity B 3.4i 0 BASES ACTIONS (continued) Alternatively, the plant can be placed in MODE 3 within 12 hours and in MODE 4 within 36 hours. This option is provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads). In MODE 4, the requirements of the LCO are no longer applicable. The Completion Time of once every 4 hours is the time needed to take and analyze a sample. The 12 hour Completion Time is reasonable, based on operating experience, to isolate the main steam lines in an orderly manner and without challenging plant systems. Also, the allowed Completion Times for Required Actions B.2.2.1 and B.2.2.2 for placing the unit in MODES 3 and 4 are reasonable, based on operating experience, to achieve the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE REQUIREMENTS SR 3.4yi.1 0 This Surveillance is performed to ensure iodine remains within limit during normal operation. The 7 day Frequency is adequate to trend changes in the iodine activity level. This SR is modified by a Note that requires this Surveillance to be performed only in MODE 1 because the level of fission products generated in other MODES is much less. REFERENCES 1. 10SCFR 100.11 0 2~ -SAR, Section i a ( DO0

13. 10 CFR 50, Appendix A, GDC 19.-

0 BWR/4 STS B 3.4.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 131 of 255

Attachment 1, Volume 9, Rev. 0, Page 132 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.6 BASES, REACTOR COOLANT SYSTEM (RCS) SPECIFIC ACTIVITY

1. Changes have been made to reflect those changes made to the Specification.

2. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

3. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

4. Changes have been made to more closely match the LCO requirement.

5. The brackets have been removed and the proper plant specific information/value has been provided.

Monticello Page 1 of I Attachment 1, Volume 9, Rev. 0, Page 132 of 255

Attachment 1, Volume 9, Rev. 0, Page 133 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 133 of 255

Attachment 1, Volume 9, Rev. 0, Page 134 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.6, REACTOR COOLANT SYSTEM (RCS) SPECIFIC ACTIVITY There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 134 of 255

Attachment 1, Volume 9, Rev. 0, Page 135 of 255 ATTACHMENT 7 ITS 3.4.7, Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown Attachment 1, Volume 9, Rev. 0, Page 135 of 255

Attachment 1,Volume 9, Rev. 0, Page 136 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 136 of 255

, Volume 9, Rev. 0, Page 137 of 255 ITS 3.4.7 Page 1 of 1 , Volume 9, Rev. 0, Page 137 of 255

Attachment 1, Volume 9, Rev. 0, Page 138 of 255 DISCUSSION OF CHANGES ITS 3.4.7, RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM - HOT SHUTDOWN ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any requirements for the Residual Heat Removal (RHR) Shutdown Cooling System during hot shutdown operations. ITS LCO 3.4.7 requires two RHR shutdown cooling subsystems to be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation. Appropriate ACTIONS and a Surveillance Requirement are also provided. This changes the CTS by incorporating the requirements of ITS 3.4.7. Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. The change is acceptable since the RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). This change is designated as more restrictive because it adds new requirements to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 138 of 255

Attachment 1, Volume 9, Rev. 0, Page 139 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 139 of 255

Attachment 1, Volume 9, Rev. 0, Page 140 of 255 RHR Shutdown Cooling System - Hot Shutdown 3.4k( CTS 3.4 REACTOR COOLANT SYSTEM (RCS) DOC ML 3.4 & Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown 0D

Doc LCO 34 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with MA no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

                                --                               &IfyTUe j I _ A___________

1. Both RHR shutdown cooling subsystems and recirculation pumps 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 the performance of Surveillances.

APPLICABILITY: MODE 3e with reactor steam dome pressu the RHRut in missiv 0 Iprewure]h o ACTIONS

              --------        ----        ------ NOTE--------

DOC Separate Condition entry is allowed for each RHR shutdown cooling subsystem. M.A CONDITION REQUIRED ACTION COMPLETION TIME DOC A. One or two RHR A.1 Initiate action to restore Immediately M.A shutdown cooling RHR shutdown cooling subsystems inoperable. subsystem(s) to OPERABLE status. AND A.2 Verify an alternate method 1 hour of decay heat removal is available for each inoperable RHR shutdown cooling subsystem. AND J n BWR/4 STS 3.4.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 140 of 255

Attachment 1, Volume 9, Rev. 0, Page 141 of 255 RHR Shutdown Cooling System - Hot Shutdown CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME A.3 Be in MODE 4. 24 hours DOC B. No RHR shutdown B.1 Initiate action to restore one Immediately M.1 cooling subsystem in RHR shutdown cooling operation. subsystem or one recirculation pump to AND operation. No recirculation pump in AND operation. B.2 Verify reactor coolant 1 hour from discovery circulation by an alternate of no reactor coolant method. circulation AND Once per 12 hours thereafter AND B.3 Monitor reactor coolant Once per hour temperature and pressure. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M.1 SR 3.4. ~~~NOTE------- 0 Not required to be met until 2 hours after reactor steam dome pressure is [9he RHR I-shutdown cooling supplyl I permisie-pressure]; I;Q^O; QFI#4IF I Uussollao I-W"_ lenc __ _ __ _ _ _I Verify one RHR shutdown cooling subsystem or 12 hours recirculation pump is operating. BWRI4 STS 3.4.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 141 of 255

Attachment 1, Volume 9, Rev. 0, Page 142 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.7, RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM - HOT SHUTDOWN

1. ISTS 3.4.8 is renumbered as ITS 3.4.7 since ISTS 3.4.5, 'Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage," is not included in the Monticello ITS.

2. The brackets have been removed and the proper plant specific information/value has been provided.

3. Editorial change made for enhanced clarity.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 142 of 255

Attachment 1, Volume 9, Rev. 0, Page 143 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 143 of 255

Attachment 1, Volume 9, Rev. 0, Page 144 of 255 RHR Shutdown Cooling System - Hot Shutdown B 3.4 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.41 Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown aResidual 0 BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the 2 decay of fission products and increases the temperature of the reactor 2coolant. This deca heat must be removed to reduce the temperature of the reactor coolantto s ay remoal in preparation fn for performingi/efueling o; maintenance operations, or for keeping the J reactor in the Hot Shutdown condition. The two redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump discharges the reactor coolant, after circulation through the respective heat exchanger, to the reactor vialthe ass6ciatedi recirculation loop. The RHR heat exchangers transfer heat to the RHR Service Water 0. System (LCO 3.7.1, "Residual Heat Removal Service Water (RHRSW) System"). APPLICABLE Decay heat removal by operation of the RHR System in the shutdown SAFETY cooling mode is not required for mitigation of any event or accident ANALYSES evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. The RHR shutdown tooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). 0 LCO Two RHR shutdown cooling subsystems are required to be OPERABLE, and when no recirculation pump is in operation, one shutdown cooling subsystem must be in operation. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associated piping and valves.. The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping. Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE. Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems. Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the BWR14 STS B 3.4.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 144 of 255

Attachment 1, Volume 9, Rev. 0, Page 145 of 255 RHR Shutdown Cooling System - Hot Shutdown B 34 BASES LCO (continued) shutdown cooling mode for removal of decay heat. In MODE 3; one RHR shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required. I v o ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. land reIrlation pumpsJ.. Note 1 permits both RHR shutdown cooling subsystems to be removed 0 from operation for a period of 2 hours in an 8 hour period. Note 2 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours for the performance of Surveillance tests. These tests may be on the-affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy. APPLICABILITY In MODE 3 with reactor steam dome pressure belowathe RHRici0in [shutdowncoolin supply permissivressurell (i.e., the actual pressure at which the interlock isolaboninterlock f resets) the RH SSysten may be operated in the shutdown cooling mode to remove decay heat to reduce or maintain coolant temperature. Shutdown-Coaling l Otherwise, a recirculation pump is required to be in operation. must be OPERABLE and one RHR shutdown cooling subsystem I In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure supply isolation

        ,interlock greater than or equal to the RHR cut in per ke pressure], this LCO is applicabe. Operation of the R                   ystem in the shutdown cooling snot 0

mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHRliIl 0 permiss iv ressur is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 21below t s-ressur<, the (i OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, "ECCS - Operating") do not allow placing the RHR shutdown cooling subsystem into operation. The requirements fo ecay heat removal in MODES 4 and 5 are 7 discussed in LCO 3.4 "Residual Heat Removal (RHR) Shutdown '-- Cooling System - Cold Shutdown," LCO 3.9.l,"Residual Heat Removal (RHR) - High Water Level," and LCO 3.9.0, "Residual Heat Removal J (RHR) - Low Water Level." BWRI4 STS B 3.4.8-2 Rev. 3.0, 03131/04 Attachment 1, Volume 9, Rev. 0, Page 145 of 255

Attachment 1, Volume 9, Rev. 0, Page 146 of 255

                                                                      *;                                T7 RHR Shutdown Cooling System - Hot Shutdown               s B 3 .4.1f (U BASES ACTIONS         A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem.

A.1. A.2. and A.3 With one required RHR shutdown cooling subsystem inoperable for decay heat removal, except as permitted by LCO Note 2, the inoperable subsystem must be restored to OPERABLE status without delay. In this condition, the remaining OPERABLE subsystem can provide the necessary decay heat removal. The overall reliability is reduced, however, because a single failure in the OPERABLE subsystem could result in reduced RHR shutdown cooling capability. Therefore, an alternate method of decay heat removal must be provided. With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The 1 hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. The required cooling capacity of the alternate method should be ensured Condensate/Feed by verifying (by calculation or demonstration) its capability to maintain or Systems reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability. Iternate methods that can be used include (but are not limited to) the ent Fuel oo in se and the Reactor Water Cleanup System. However, due to the potentially reduced reliability of the alternate methods of decay heat removal, it is also required to reduce the reactor coolant temperature to the point where MODE 4 is entered. BWR/4 STS B 3.4.8-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 146 of 255

Attachment 1, Volume 9, Rev. 0, Page 147 of 255 RHR Shutdown Cooling System - Hot Shutdown B 3.4( BASES ACTIONS (continued) B.1. B.2, and B.3 With no RHR shutdown cooling subsystem and no recirculation pump in operation, except as permitted by LCO Note 1, reactor coolant circulation by the RHR shutdown cooling subsystem or recirculation pump must be restored without delay. Until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The 1 hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours thereafter. This will provide assurance of continued temperature monitoring capability. During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR shutdown cooling subsystem or recirculation pump), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate. SURVEILLANCE SR (I) REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room. This Surveillance is modified by a Note allowing sufficient time to align the RHR System for shutdown cooling operation after clearing the pressure interlock that isolates the system, or for placing a recirculation pump in operation. The Note takes exception to the requirements of the Surveillance being met (i.e., forced coolant circulation is not required for this initial 2 hour period), which also allows entry into the Applicability of this Specification in accordance with SR 3.0.4 since the Surveillance will not be "not met" at the time of entry into the Applicability. REFERENCES None. BWR/4 STS B 3.4.84 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 147 of 255

Attachment 1, Volume 9, Rev. 0, Page 148 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.7 BASES, RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM - HOT SHUTDOWN

1. Changes have been made to reflect those changes made to the Specification.

2. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

3. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

4. Changes have been made to more closely match the LCO requirement.

5. The brackets have been removed and the proper plant specific information/value has been provided.

6. Typographical/grammatical error corrected.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 148 of 255

Attachment 1, Volume 9, Rev. 0, Page 149 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 149 of 255

Attachment 1, Volume 9, Rev. 0, Page 150 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.7, RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM - HOT SHUTDOWN There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 150 of 255

Attachment 1, Volume 9, Rev. 0, Page 151 of 255 ATTACHMENT 8 ITS 3.4.8, Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown Attachment 1, Volume 9, Rev. 0, Page 151 of 255

Attachment 1, Volume 9, Rev. 0, Page 152 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 152 of 255

, Volume 9, Rev. 0, Page 153 of 255 ITS 3.4.8 Add proposed ITS 3.4.8 a

Page 1 of 1 , Volume 9, Rev. 0, Page 153 of 255

Attachment 1, Volume 9, Rev. 0, Page 154 of 255 DISCUSSION OF CHANGES ITS 3.4.8, RESIDUAL HEAT REMOVEAL (RHR) SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any requirements for the Residual Heat Removal (RHR) Shutdown Cooling System during cold shutdown operations. ITS LCO 3.4.8 requires two RHR shutdown cooling subsystems to be OPERABLE, and, with no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation. Appropriate ACTIONS and a Surveillance Requirement are also provided. This changes the CTS by incorporating the requirements of ITS 3.4.8. Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. This change is acceptable since the RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). This change is designated as more restrictive because it adds new requirements to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of I Attachment 1, Volume 9, Rev. 0, Page 154 of 255

Attachment 1, Volume 9, Rev. 0, Page 155 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 155 of 255

Attachment 1, Volume 9, Rev. 0, Page 156 of 255 RHR Shutdown Cooling System - Cold Shutdown 3.4 j/D CTS 3.4 REACTOR COOLANT SYSTEM (RCS) DOC MA1 3.4 1; Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown 0D D00 LCO 3 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with M.l no recirculation pump in operation, at least one RHR shutdown cooling subsystem shall be in operation.

1. Both RHR shutdown cooling subsystems and recirculation pumps 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 the performance of Surveillances.

APPLICABILITY: MODE 4. ACTIONS Kl PTYr

                                           ------    NL      I k rc Separate Condition entry is allowed for each shutdown cooling subsystem.

A _ _ __M _ __ _ _ _ _ _ __ _ CONDITION REQUIRED ACTION COMPLETION TIME DOC A. One or two RHR A.1 Verify an alternate method 1 hour M.1 shutdown cooling of decay heat removal is subsystems inoperable. available for each AND inoperable RHR shutdown cooling subsystem. Once per 24 hours thereafer DOC M.1 B. No RHR shutdown B.1 Verify reactor coolant 1 hour from discovery cooling subsystem in circulating by an alternate of no reactor coolant operation. method. circulation AND AND No recirculation pump in Once per 12 hours operation. thereafter AND BWR/4 STS 3.4.9-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 156 of 255

Attachment 1, Volume 9, Rev. 0, Page 157 of 255 RHR Shutdown Cooling System - Cold Shutdown CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B.2 Monitor reactor coolant Once per hour temperatureupwressj 0D SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 0 DOC MA1 S R 3.4 .r/ Verify one RHR shutdown cooling subsystem or recirculation pump is operating. 12 hours 0 BWR/4 STS 3.4.9-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 157 of 255

Attachment 1, Volume 9, Rev. 0, Page 158 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.8, RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN

1. ISTS 3.4.9 is renumbered as ITS 3.4.8 since ISTS 3.4.5, "Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage," is not included in the Monticello ITS.

2. This change is made to be consistent with NUREG-1434, ISTS 3.4.10, as well as NUREG-1433, ISTS 3.4.8.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 158 of 255

Attachment 1, Volume 9, Rev. 0, Page 159 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 159 of 255

Attachment 1, Volume 9, Rev. 0, Page 160 of 255 RHR Shutdown Cooling System - Cold Shutdown B34 B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4*Residual Heat Removal (RHR) Shutdown Cooling System - Cold Shutdown 0 BASES BACKGROUND Irradiated fuel in the shutdown reactor core generates heat during the decay of fission products and increases the temperature of the reactor coolant. This decay heat must be removed to maintain the temperature I} of the reactor coolant 5Fs de in preparation for performingfefuelingW maintenance operations, or for keeping the J reactor in the Cold Shutdown condition. The two redundant, manually controlled shutdown cooling subsystems of the RHR System provide decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Ees the reactor coolant, after circulation through the respective loop. The heat RHR exchhe heat exchangers transfer heat to assciated recirculation the RHR Service Water 0D System. APPLICABLE Decay heat removal by operation of the RHR System in the shutdown SAFETY cooling mode is not required for mitigation of any event or accident ANALYSES evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. The RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). LCO Two RHR shutdown cooling subsystems are required to be OPERABLE, and when no recirculation pump is in operation, one RHR shutdown cooling subsystem must be in operation. An OPERABLE RHR shutdown and the necessary cooling subsystem consists of one OPERABLE RHR pump, one heat RHRSW System capable of providing xsohe r,and the associated piping and valves. The two subsystems have a common suction source and are allowed to have a common heat 0 coolengwater tothe exchanger and common discharge piping. Thus, to meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE. Since the piping and heat exchangers are passive components that are assumed not to fail, they are allowed to be common to both subsystems. [Fn MODE 4, the RHR crsste valve (2E1 1-F01 0)/ [may be openeisoalow pumps in one loop toDshrethr~oLuh the/ opposite recircmation loop to make a complet subs stem. Additionally, each 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. In MODE 4, one RHR shutdown cooling BWR/4 STS B 3.4.9-1 . Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 160 of 255

Attachment 1, Volume 9, Rev. 0, Page 161 of 255 RHR Shutdown Cooling System - Cold Shutdown B 3.4.-Ji BASES LCO (continued) subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required. o r ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. [and recirculation pumpsil? Note 1 permits both RHR shutdown cooling subsystems'to be removed ) from operation for a period of 2 hours in an 8 hour period. Note 2 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours for the performance of Surveillance tests. These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy. I must be OPERABLE and one RHR shutdown cooling subsystemr I0 APPLICABILITY In MODE 4, the RHR Shutdown Cooling System'may be operated in the 212l shutdown cooling mode to remove decay heat to maintain coolant temperature belowL2NF. Otherwise, a recirculation pump is required to be in operation. [shutdown cooling supply Isolation intemock] In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure greater than or equal to the RHR cu in perjre ressure, this LCO is / not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at reactor pressures greater than or equal to the RHR ermis ressurq is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 21below t ressuro, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) '(LCO 3.5.1, "ECCS - Operating") do not allow placing the RHR shutdown cooling subsystem into operation. m The requirements for decay heat removal in MODE 3 below the cut in permissive pressure and in MODE 5 are discussed in LCO 3.4.jV7'IJ "Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown," LCO 3.A, "Residual Heat Removal (RHR) - High Water Level," and LCO 3.9.8, "Residual Heat Removal (RHR) - Low Water

Level." BWR/4 STS B 3.4.9-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 161 of 255

Attachment 1, Volume 9, Rev. 0, Page 162 of 255 RHR Shutdown Cooling System - Cold Shutdow j B 3.4 i BASES ACTIONS A Note has been provided to modify the ACTIONS related to RHR shutdown cooling subsystems. Section 1.3, Completion Times, specifies once a Condition has been entered, subsequent divisions, subsystems, components or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable shutdown cooling subsystems provide appropriate compensatory measures for separate inoperable shutdown cooling subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable RHR shutdown cooling subsystem. A.1 With one of the two required RHR shutdown cooling subsystems inoperable, except as permitted by LCO Note 2, the remaining subsystem is capable of providing the required decay heat removal. However, the overall reliability is reduced. Therefore, an alternate method of decay heat removal must be provided. With both RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The 1 hour Completion Time is based on the 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 alternate method(s) must be reconfirmed every 24 hours thereafter. This will provide assurance of continued heat removal capability. The required cooling capacity of the alternate method should be ensured by verifying (by calculation or demonstration) its capability to maintain or reduce temperature. Decay heat removal by ambient losses can be considered as, or contributing to, the alternate method capability. Alternate methods that can be used include (but are not limited to)t lt Go Spent Fuel P ing System an the Reactor Water Cleanup Systerr, {by Itself or using feed and bleed In combination with Control Rod Drive System or Condensate/Feed Systems BWR/4 STS B 3.4.9-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 162 of 255

Attachment 1, Volume 9, Rev. 0, Page 163 of 255 RHR Shutdown Cooling System - Cold Shutdown B 3.4gQ BASES ACTIONS (continued) B.1 and B.2 With no RHR shutdown cooling subsystem and no recirculation pump in operation, except as permitted by LCO Note 1, and until RHR or recirculation pump operation is re-established, an alternate method of reactor coolant circulation must be placed into service. This will provide the necessary circulation for monitoring coolant temperature. The 1 hour Completion Time is based on the coolant circulation function and is modified such that the 1 hour is applicable separately for each occurrence involving a loss of coolant circulation. Furthermore, verification of the functioning of the alternate method must be reconfirmed every 12 hours thereafter. This will provide assurance of continued temperature monitoring capability. During the period when the reactor coolant is being circulated by an alternate method (other than by the required RHR Shutdown Cooling System or recirculation pump), the reactor coolant temperature and pressure must be periodically monitored to ensure proper function of the alternate method. The once per hour Completion Time is deemed appropriate. SURVEILLANCE SR 3.4.**I- (0 REQUIREMENTS This Surveillance verifies that one RHR shutdown cooling subsystem or recirculation pump is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The Frequency of 12 hours is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystem in the control room. REFERENCES None.. BWR/4 STS B 3.4.9-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 163 of 255

Attachment 1, Volume 9, Rev. 0, Page 164 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.8 BASES, RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN

1. Changes have been made to reflect those changes made to the Specification.

2. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

3. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

4. Changes have been made to more closely match the LCO requirement.

5. This allowance has been deleted since it is not necessary. The RHR crosstie valves are normally open since Monticello utilizes Low Pressure Coolant Injection Loop Select logic.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 164 of 255

Attachment 1, Volume 9, Rev. 0, Page 165 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 165 of 255

Attachment 1, Volume 9, Rev. 0, Page 166 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.8, RESIDUAL HEAT REMOVAL (RHR) SHUTDOWN COOLING SYSTEM - COLD SHUTDOWN There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 166 of 255

Attachment 1, Volume 9, Rev. 0, Page 167 of 255 ATTACHMENT 9 ITS 3.4.9, Reactor Coolant System (RCS) Pressure and Temperature (PIT) Limits Attachment 1,Volume 9, Rev. 0, Page 167 of 255

Attachment 1, Volume 9, Rev. 0, Page 168 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 168 of 255

C C C ITS 3.4.9 ITS ITS 3.0 UMING COIDmONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.6 PRIMARY SYSTEM BOUNDARY 4.6 PRIMARY SYSTEM BOUNDARY

                                                                                                                                                                                               %D C,

0 0 3 a Applies to the operating tatus of the reactor coolant system. Applies to the perIodIc examination d testing requirements Ca, for the reactor coolant system. U.t 0 0 OblectQe _bletive 3 0 To determine the condition of e reactor coolant system and To assure the Int rty and sale operation of the reactor coolant system. the operation of the safety d ces related to it. 0 CD

-o                                                                                                                                                                                             0 0

i)Spedciathin e) M 5 3.4.9 A. Reactor Coolant Heatup and Cooldown A. Reactor Coolant Heatup and Cooldown 0 ;n

1. The average rate of reactor coolant temperature SR 3.4.9.1 During heatups and cooldowns the foflol@ng3 L (0)

LCO 3.4.9. 0 SR 3.4.9.1 change durIng normal heatup or cooldown shall not temperatures shall ber ee ly to

-9                            exceed 100°F/hr. when averaged over a one-hour                           __i[nutesluU 3 consecutvrednsaaclotil_

to period. 0 I _ n_ _.- - 1 . . - - I ............ 2

                                                                                                                                         . _-_ _^_n------__nn_

CD 2Z The pump In an Idle redrculation loop shag not b a. Reacto vessel snenl a818a to snenl flange. Co K0 LCO 3.4.9, o (D SR 3.4.9.3 started unless the temperature of the coolant withn 00, the Idle recirculation loop Is within 50°F of the b. React r vessel bottom dr. 0 (0 reactor coolant temperature.

c. RePCflation loops A and G M

                                                                                                                                                                                                -9 Cn

(. tn

d. Reaptor vessel bottom h ad.

                                                                                                ,-1Add           proposed Note to SR 3.4.9.3    --

3.614.6 121 1/9181 Amendment No. 0 Page 1 of 7

C C C ITS 3.4.9 0 ITS ITS 3.0 LIMITING CONDIONS FOR OPERATION 4.0 SURVEILLANCE REOUIREMENTS B. Reactor Vessel Temperature and Pressure B. Reactor Vessel Temperature and Pressure C, 0

1. During in-service hydrostatic or leak testing, the SR 3.4.9.1 1. During In-service hydrostatic or leak LCO 3.4.9, reactor vessel shell temperatures specified In SR 3.4.9.1 I mne vessel Dr 2(D 0

4.6.0.1, except for the reactor vessel bottom head, temperatures t CD shall be at or above the temperatures shown on the minutes. two curves of Figure 3.6.2, where the dashed curve, RPV Core Beltlfine, Is Increased by the core - bettline temperature adjustment from Figure 3.6.1. 0 The reactor vessel bottom head temperature shall 0 be at or above the temperatures shown on the soiid curve of Figure 3.6.2, RPV Remote from Core 0 3 ,Beltrine, with no adjustment from Figure 3.6.1. 3 (D 0(D I LCO 3.4.9.1, 2. During heatup by non-nudear mea SR 3.4.9.1 sel yen cooldown following Co ~ lysieSthe 0 nuclear shutdown, reactor vessel shell and fluid temperatures specified DX In 4.6-A shall be at or above the higher of the 0 ltemperatures of Figure 3.6.3 where the dashed curve, RPV Core Befttlne, Is increased by the _X expected shift.in KTNOT from Figure 3.6.1.

 -4                                                                                                                                                          ;U is to      LCO 3.4.9,        3      During al operation with a critical reacF                                                                                    6 SR 3.4.9.2             Ifor low leyray                                                                                                                la co CD Ma'                           Ireactor vrs      s yes     the reaor vessel shel and fluid temperatures specified In 4.6.A shall be at or above the higher of the temperatures of Figure 3.6.4 where the dashed curve, 'RPV Core Belttine, Un                             Is Increased by the expected shift In TNOT from Figure 3.6.1.

4 3.614.6 122 04122/03 Amendment No. 3S20f, 135 , Page 2 of 7

C C C ITS 3.4.9 ITS ITS 3.0 LIMING CONDITIONS FOR OPERATION a 4.0 SURVEILLANCE REQUIREMENTS a) LCO 3.4.9. SR 3.49.4. 4. The reactor vessel head boating studs shall not be LIl

4. When the reactor vessel head studs are under 0

SR 3.4.9.5. SR 3.4.9.5 under tension unless the temperature of the vessel _ head flange and the head are Z70 0 F. SR 3.4.9.4, SR 3.4.9.5. tensIon and the reactor Is In the Cold Shutdown' Condition, the reactor vessel shall flange

                                                                                                                                                          -1l 3                        FC.Coolt es          JL SR 3.4.9.8 C.

temperature shall be permanently recorded. Coolant Chemistry 03

1. (a) The steady state radfoiodine concentration In r4' 0

0 I I the reactor coolant shall not exceed 2.0 microcurfes of 1-131 dose equivalent per gram of water. (b) The steady state radiotodine concentration In

                                                                                 -I-      m I

1. (a) A sample of reactor coolant shall be taken at I--

least every 96 hours and analyzed for radioactive lodines of 1-131 through 1-135 during power operation, I See ITS 3.4.61 0 0 3 Ihe reactor coolant shall not exceed 0.02 (b) A sample of reactor coolant shall be taken and CD (0 microcurfes of 1-131 dose equivalent per gram analyzed for radioactive lodfnes of 1-131 of water when the reactor coolant temperature -oW through 1-135 within 24 hours prior to raising See ITS 3.10.1 -u is > 212F. the reactor is not critical, and the reactor coolant temperature >212F. with ;U primary containment Integrity has not been ee the reactor not critical, and with primary established. containment Integrity not establlshed. 0 0 C -J N) (0) (1) 101 CiD to

n -J

-4 0 0 K) CA (1` CM tM M 3.6/4.6 123 03107/01 Amendment No. 0 1 01 4 0 7 r 1r 117 Page 3 of 7

C C C. ITS 3.4.9 ITS Figure 3.4.9-1 MONTICELLO UMITING BELTUNE SHIFT 1"W 1X 0

2) 0 f1U 0 100 Cl 90, 2 2 CD a) 3 5U :E. go.

0 I 70 0 CD 0 8 60

0) -U a: so.

CD CD 0 40 -o n 30 ax 0 20 01 10 ID N, 0 0 to

                                                                                                                                  -4 N,                         0 0.5           i            tis            2             2.5            3           3.5                01 (1/4 Vessel Wall Thickness Fluence (1018 n/cr 2))                                     01 Figure 3.6.1 Core Beltline Operating Umits Curve Adjustment vs. Fluence 3.614.6                                                                        133          10/12/99 Amendment No. 72, 1Os Page 4 of 7

c c C ITS 3.4.9 0 ITS Figure 3.4.9-2. w 0 0) 0 0 2 c R ReIv%Im _ 0 =o 0 0 9 2 a 0 9 3 t 0 tD t t: 1.. - I 1 1 - - L11- . . - -U 15 2 (D 0P 0) 1 2 IN t~~o 11111+ 11111 a

                                                                                                           ' RS) 0 04 CD

01

-IV 01 to N 01 cn 0 2C 400 600 800 IOM 1200 14W IE600 18 2D00 tn Prewurelhiit In Vegsd Top Heed Figure 3.6.2 Minimum Temperature vs. Pressure for Pressure Tests 3.614.6 134 0V24/03 Amendment No. 026,8S 133 Page 5 of 7

C C C 0 ITS 3.4.9 ITS Figure 3.4.9-3 s W a) W 0 0 0 0 0 3 0 o M CD 0 E) DX 0 0 (D

                                                                                                                                  -J CD

.4 n 0 2 r OcO m cO 1C0X M0 I 19D1 n, Irkeexe Unit In Vesed Top HNed (PSIG) n' Ln U' Figure 3.6.3 Minimum Temperature vs. Pressure Mechanical Heatup or Cooldown without the Core Critical 3.614.6 135 0224/03 Amendment No. 72,104 133 Page 6 of 7

c c C ITS 3.4.9 ITS Figure 3.4.9-4 is w C) 3 ED 0 X-1 0 M 0 P 3 CD o D CD 5 CD en lD m) CD D -4 to

                                                                                                                                    -.1 01

-4' 0 01 01 0 200 400 Ow WW 1200 1400 160) 1800 2a0) tn PnuUrr Iness TopHmd(PSlG) (n Figure 3.6.4 Minimum Temperature Vs. Pressure for Core Critical Operation 3.6/4.6 136 02/24103 Amendment No. 7-2,06,133 Page 7 of 7

Aftachment 1, Volume 9, Rev. 0, Page 176 of 255 DISCUSSION OF CHANGES ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (PIT) LIMITS ADMINISTRATIVE CHANGES A.1 In the conversion of the Monticello Current Technical Specifications (CTS) to the plant specific Improved Technical Specifications (ITS), certain changes (wording preferences, editorial changes, reformatting, revised numbering, etc.) are made to obtain consistency with NUREG-1433, Rev. 3, "Standard Technical Specifications General Electric Plants, BWR/4" (ISTS). These changes are designated as administrative changes and are acceptable because they do not result in technical changes to the CTS. A.2 CTS 3.6.A.1 includes a limit for average rate of reactor coolant temperature change during normal heatup and cooldown. CTS 3.6.A.2 includes a limit for the differential temperature between an idle recirculation loop and the reactor coolant temperature prior to an idle recirculation loop startup. CTS 3.6.B includes limitations on the reactor vessel temperature and pressure during various plant conditions. ITS LCO 3.4.9 states "RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained within limits" and includes an Applicability of "At all times." The applicable limits are included in the Surveillance Requirements associated with ITS 3.4.9. This changes the CTS by combining the requirements of CTS 3.6.A.1, CTS 3.6.A.2, and CTS 3.6.B in one LCO, including the limits in the Surveillance Requirements, and providing an Applicability. The purpose of ITS LCO 3.4.9 is to include all the requirements associated with limitations for RCS heatup and cooldown rates, limitations for the differential temperature between an idle recirculation loop and the reactor coolant temperature prior to an idle recirculation loop startup, and the limitations for the reactor vessel temperature and pressure. ITS LCO 3.4.9 states the RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be maintained with limits. The actual limits have been included in the Surveillance Requirements and an Applicability is included. This change is acceptable because it provides a clear statement concerning the requirements and when they must be met. The addition of the Applicability statement is consistent with the current requirements since there is no such statement for when they apply. This change is considered a presentation preference change only and, as such, is considered an administrative change. A.3 CTS 3.6.B.2 states that P/T limits of Figure 3.6.3 are applicable during a heatup by non-nuclear means "(except with the reactor vessel vented)" and CTS 3.6.B.3 states that PIT limits of Figure 3.6.4 are applicable during all operation with a critical core "other than...at times when the reactor vessel is vented." ITS LCO 3.4.9 and SRs 3.4.9.1 and 3.4.9.2 are applicable even when the reactor vessel is vented. This changes the CTS by requiring the applicable PIT limits to be met when the reactor vessel is vented. The purpose of CTS 3.6.B.2 and CTS 3.6.B.3 is to ensure applicable PIT limits are not exceeded. The minimum temperature allowed by CTS Figures 3.6.3 and Monticello Page 1 of 7 Aftachment 1, Volume 9, Rev. 0, Page 176 of 255

Attachment 1, Volume 9, Rev. 0, Page 177 of 255 DISCUSSION OF CHANGES ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (P/T) LIMITS 3.6.4 at 0 psig (i.e., reactor vessel vented) is 70 0F. CTS 3.6.B.4 requires a minimum temperature of 700 F when the reactor vessel head bolting studs are under tension. Therefore, since the two temperatures are the same, requiring the PIT limits of CTS Figures 3.6.3 and 3.6.4 (ITS Figures 3.4.9-3 and 3.4.9-4) to be met when the reactor vessel is vented is a presentation preference only. Therefore, this change is considered an administrative change. A.4 CTS 4.6.A requires various RCS temperatures to be "recorded" during heatup and cooldowns. CTS 4.6.B.1 requires various RCS temperatures to be "recorded" during the inservice hydrostatic or leak testing. ITS SR 3.4.9.1 requires a verification that the RCS pressure and temperature and heatup and cooldown rates are within the applicable limits. This changes the CTS by deleting the specific requirement to "record" the temperatures. The purpose of CTS 4.6.A is to ensure sufficient temperature data is taken to calculate the RCS heatup or cooldown rate. The purpose of CTS 4.6.B.1 is to ensure sufficient data is taken in order to determine whether the other P/T limits are met. This change is acceptable because this requirement duplicates the requirements of 10 CFR 50 Appendix B, Section XVII (Quality Assurance Records) to maintain records of activities affecting quality, including the results of tests (i.e., Technical Specification Surveillances). Compliance with 10 CFR 50 Appendix B is required by the Monticello Operating License. The details of the regulations within the Technical Specifications are repetitious and unnecessary. Therefore, retaining the requirement to perform the associated Surveillances and eliminating the details from Technical Specifications that are found in 10 CFR 50 Appendix B is considered a presentation preference. As such, this change is considered an administrative change. MORE RESTRICTIVE CHANGES M.1 CTS 3.6.A.2 states that the pump in an idle recirculation loop shall not be started unless the temperature of the coolant within the idle recirculation loop is within 500 F of the reactor coolant temperature. However, no specific Surveillance Requirement exists to verify the limit is met prior to starting a recirculation pump. ITS SR 3.4.9.3 includes a requirement to verify the limit specified in CTS 3.6.A.2 is met "Once within 15 minutes prior to each startup of a recirculation pump." This changes the CTS by adding a specific Surveillance Requirement. The purpose of CTS 3.6.A.2 is to ensure thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. ITS SR 3.4.9.3 includes a requirement to verify the same temperature differential limitation specified in CTS 3.6.A.2 "Once within 15 minutes prior to each startup of a recirculation pump." The proposed Surveillance Requirement is considered acceptable because performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limit will not be exceeded between the time of the Surveillance and the time of the idle pump start. This change is designated as more restrictive because it adds a Surveillance Requirement that does not appear in the CTS. Monticello Page 2 of 7 Attachment 1, Volume 9, Rev. 0, Page 177 of 255

Attachment 1, Volume 9, Rev. 0, Page 178 of 255 DISCUSSION OF CHANGES ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (P/T) LIMITS M.2 CTS 3.6.A.1 includes a limit for average rate of reactor coolant temperature change during normal heatup and cooldown. CTS 3.6.A.2 includes a limit for the differential temperature between an idle recirculation loop and the reactor coolant temperature prior to an idle recirculation loop startup. CTS 3.6.B includes limitations on the reactor vessel temperature and pressure during various plant conditions. There are no specified actions to take when the limitations are not met. Therefore, 10 CFR 50.36(c)(2) requires the unit to be shut down until the LCO is met. Thus, if any of the limitations of CTS 3.6.A.1, 3.6.A.2, or 3.6.B are not met, the unit is required to be shut down until the limitation not being met is back within limits. In addition, no time limit in which to complete the unit' shutdown is specified in 10 CFR 50.36(c)(2). ITS 3.4.9 ACTION A covers the condition when the requirements of the LCO are not met in MODE 1, 2, or 3, and requires the restoration of the parameters to within limit(s) within 30 minutes and a determination that the RCS is acceptable for continued operation within 72 hours. The action to determine whether the RCS is acceptable for continued operation must be completed even if the requirements of the LCO are restored to within limits. If these actions are not met, ITS 3.4.9 ACTION B requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours. ITS 3.4.9 ACTION C covers the condition when the requirements of the LCO are not met in conditions other than MODES 1, 2, and 3, and requires the immediate initiation of action to restore the parameters to within limit(s) and the determination that the RCS is acceptable for continued operation prior to entering MODE 2 or 3. The action to determine whether the RCS is acceptable for continued operation must be completed even if the requirements of the LCO are restored to within limits. This changes the CTS by adding finite times to shut down the unit when it is operating, and provides specific actions to take when the unit is already in the shutdown condition. The purpose of ITS 3.4.9 ACTIONS A, B, and C is to provide specific compensatory actions for when RCS pressure, RCS temperature, RCS heatup and cooldown rates, and the recirculation pump starting temperature requirements are not maintained within limits. This change is acceptable because it provides the necessary and specific actions to take when the requirements are not met and provides appropriate times to complete the actions. This change is designated as more restrictive because it adds specific compensatory actions and times for when PIT limits are not met in all conditions. M.3 CTS 3.6.B.3 includes P/T limits during all operation with a critical reactor. However, there is no specific Surveillance Requirement for verification that the RCS pressure and temperature are within the P/T limits prior to criticality. ITS SR 3.4.9.2 requires a verification that the RCS pressure and temperature are within the criticality limits once (ITS Figure 3.4.9-4) within 15 minutes prior to control rod withdrawal for the purpose of achieving criticality. This changes the CTS by adding a specific Surveillance Requirement to verify the criticality limits are met. The purpose of CTS 3.6.B.3 is to specify the P/T limits during all operation with a critical reactor. The CTS does not include a Surveillance Requirement for verification of the specified P/T limits. This change adds a specific Surveillance Requirement. This change is acceptable because the proposed Surveillance Monticello Page 3 of 7 Attachment 1, Volume 9, Rev. 0, Page 178 of 255

Attachment 1, Volume 9, Rev. 0, Page 179 of 255 DISCUSSION OF CHANGES ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (PIT) LIMITS Requirement (ITS SR 3.4.9.2) will verify the criticality limits are met within 15 minutes before control rod withdrawal for the purpose of achieving criticality. This change is designated as more restrictive because it adds a specific Surveillance Requirement not currently required by the CTS. M.4 CTS 3.6.B.2 states that the P/T limits in Figure 3.6.3 apply during low power physics tests and CTS 3.6.B.3 states that the P/T limits of Figure 3.6.4, which provides the criticality PIT limits, do not apply during low power physics tests. ITS SR 3.4.9.2 will require the criticality P/T limits provided in ITS Figure 3.4.9-4 to apply during low power physics test. This changes the CTS by applying the criticality PIT limits in lieu of the non-criticality P/T limits during low power physics tests. The purpose of CTS 3.6.B 2 and CTS 3.6.B.3 is to ensure applicable P/T limits are not exceeded. The criticality P/T limits of CTS Figure 3.6.4 (ITS Figure 3.4.9-4) are more limiting (i.e., require a higher temperature for a given pressure) than the non-criticality P/T limits of CTS Figure 3.6.3 (ITS Figure 3.4.9-3). Therefore, this change is acceptable since it helps to ensure that any time the reactor is critical, the criticality PIT limits are met, thus ensuring vessel thermal limits are not exceeded. This change is designated as a more restrictive change since more stringent requirements are being applied in the ITS than in the CTS. M.5 CTS 4.6.B.1 requires recording various temperatures during inservice hydrostatic testing or leak testing only "when the vessel pressure is above 312 psig." ITS SR 3.4.9.1.a will require verifying the temperatures "at all times" during inservice hydrostatic testing or leak testing. This changes the CTS by requiring the temperature verification "at all times" during inservice hydrostatic testing and leak testing, which includes when the reactor vessel pressure is less than or equal to 312 psig. The change to require a verification in lieu of recording the temperatures is discussed in DOC L.2. The purpose of CTS 4.6.B.1 is to periodically verify that the applicable PIT limits are being met. Therefore, this change is acceptable since verifying the limits are met any time during an inservice hydrostatic test or leak test will help to preclude an unrecognized'P/T limit violation. This change is designated as a more restrictive change since more stringent requirements are being applied in the ITS than in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.6.A requires the reactor coolant temperatures at the following locations to be monitored: a. reactor vessel shell adjacent to shell Monticello Page 4 of 7 Attachment 1, Volume 9, Rev. 0, Page 179 of 255

Attachment 1, Volume 9, Rev. 0, Page 180 of 255 DISCUSSION OF CHANGES ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (PIT) LIMITS flange; b. reactor vessel bottom drain; c. recirculation loops A and B; and d. reactor vessel bottom head temperatures. The temperatures are to be monitored during heatups and cooldowns "until 3 consecutive readings at each location are within 50 F." CTS 4.6.B.1 requires the RCS temperatures at the following locations to be monitored during inservice hydrostatic or leak testing: a. reactor vessel shell adjacent to shell flange; b. reactor vessel bottom head; and c. reactor vessel shell or coolant temperature representative of the minimum temperature of the beltline region. ITS SR 3.4.9.1 requires the RCS pressure, RCS temperature, and RCS heatup and cooldown rates to be within the applicable limits. This changes the CTS by relocating the details of the specific reactor coolant temperature locations that must be monitored and the criteria for ending the RCS heatup and cooldown rates verification to the ITS Bases. The removal of these details for evaluating Surveillance Requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS SR 3.4.9.1 still retains the requirement to verify that the RCS pressure, RCS temperature, and RCS heatup and cooldown rates are within the applicable limits. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 2 - Relaxation Of Applicability) CTS 3.6.A.2 states that the pump in an idle recirculation loop shall not be started unless the temperature of the coolant within the idle recirculation loop is within 500 F of the reactor coolant temperature. ITS LCO 3.4.9 and ITS SR 3.4.9.3 includes the same requirement, however a Note has been included in ITS SR 3.4.9.3 that states the Surveillance is only required to be met in MODES 1, 2, 3, and 4 during recirculation pump startup. This changes the CTS by excluding the idle loop temperature requirement during MODE 5 and when fuel is not in the reactor. The purpose of CTS 3.6.A.2 is to ensure thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances. This change is acceptable because it has been determined that the LCO requirements are not necessary for verification that the equipment can perform its required functions. This change allows the temperature differential between the idle recirculation loop and the reactor coolant system to be greater than the specified limit during an idle recirculation pump startup. This change is acceptable because in MODE 5 and when fuel is not in the reactor, the overall stress on limiting components is lower. Therefore, the AT limit is not required. This Monticello Page 5 of 7 Attachment 1, Volume 9, Rev. 0, Page 180 of 255

Attachment 1, Volume 9, Rev. 0, Page 181 of 255 DISCUSSION OF CHANGES ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (P/T) LIMITS change is designated as less restrictive because less stringent Applicability requirements are being applied in the ITS than were applied in the CTS. L.2 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.6.A requires a verification of the heatup and cooldown rate every 15 minutes. CTS 4.6.B.1 requires a verification that the P/T limits are within limits every 15 minutes during inservice hydrostatic or leak testing. ITS SR 3.4.9.1 requires the RCS pressure, RCS temperature, and RCS heatup and cooldown rates to be within the applicable limits every 30 minutes. This changes the CTS by extending the Surveillance Frequency from every 15 minutes to every 30 minutes. The purpose of CTS 4.6.A is to ensure the RCS heatup and cooldown rates are within appropriate limits while the purpose of CTS 4.6.B.1 is to ensure the P/T limits during inservice hydrostatic or leak testing are within appropriate limits. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. This change increases the RCS heatup and cooldown rate and P/T limits verification from every 15 minutes to every 30 minutes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.3 (Category 7 - Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 3.6.B.4 states that the reactor vessel head bolting studs shall not be under tension unless the temperature of the vessel head flange and the head are > 70 0F. CTS 4.6.B.4 states that when the reactor vessel head studs are under tension and the reactor is in the Cold Shutdown Condition, the reactor vessel shell flange temperature shall be permanently recorded. ITS SR 3.4.9.4, SR 3.4.9.5, and SR 3.4.9.6 include the requirement to verify the same limit specified in CTS 3.6.B.4 (reactor vessel flange and head flange temperatures are

      > 70 0F). ITS SR 3.4.9.4 requires a verification that the reactor vessel flange and head flange temperatures are within limits every 30 minutes when tensioning the reactor vessel head bolting studs. ITS SR 3.4.9.5 requires the same verification every 30 minutes, however the verification is not required to be performed until 30 minutes after RCS temperature is < 800 F in MODE 4. ITS SR 3.4.9.6 requires the same verification every 12 hours, however the verification is not required to be performed until 12 hours after RCS temperature is < 100F in MODE 4. This changes the CTS by deleting the requirement to permanently record the reactor vessel shell flange temperature at all times when the reactor vessel head studs are under tension and the reactor is in MODE 4, and includes a requirement to verify the limits are met at periodic Frequencies.

The purpose of CTS 4.6.B.4 is to ensure the specified limit in CTS 3.6.B.4 is met when the reactor vessel head bolting studs are under tension. This change is acceptable because the Surveillance Frequencies have been evaluated to Monticello Page 6 of 7 Attachment 1, Volume 9, Rev. 0, Page 181 of 255

Attachment 1, Volume 9, Rev. 0, Page 182 of 255 DISCUSSION OF CHANGES ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (PIT) LIMITS ensure that they provide an acceptable level of equipment reliability. This change deletes the requirement to permanently record the reactor vessel shell flange temperature when the reactor vessel head studs are under tension and the reactor is in MODE 4 and includes a requirement to verify the limits are met at certain periodic Frequencies. The flange temperatures must be verified to be above the limits 30 minutes before the start of and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature s 800F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature s 100 0F, monitoring of the flange temperature is required every 12 hours to ensure the temperature is within the limits. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. Monticello Page 7 of 7 Attachment 1, Volume 9, Rev. 0, Page 182 of 255

Attachment 1, Volume 9, Rev. 0, Page 183 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 183 of 255

Attachment 1, Volume 9, Rev. 0, Page 184 of 255 RCS P/T Limits

34. (0 3.4 REACTOR COOLANT SYSTEM (RCS) 3/4.6A 3.4 RCS Pressure and Temperature (P/T) Limits 314.6.B 3 3.&A.i.

3.6A2, LCO 3.4Z.It RCS pressure, RCS temperature, IRCS heatup and cooldown rates, and the recirculation pump starting temperature requirements shall be 0D 3.6.8.1, 3.6.82, 3.6.B.3. maintained within@fIlimits specified e PT . 0 3.6.B.4 DOCA-2 APPLICABILITY: At all times. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC M.2 A. ---- NOTE--- A.1 Restore parameter(s) to 30 minutes Required Action A.2 within limits. shall be completed if this Condition is entered. AND A.2 Determine RCS is 72 hours Requirements of the acceptable for continued LCO not met in operation. MODEJE 1, 2, ii 3. I t 0 DOCM.2 B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met. B.2 Be in MODE 4. 36 hours

4. 4 DOC M.2 C. -----NOTE- --- C.1 Initiate action to restore Immediately Required Action C.2 parameter(s) to within shall be completed if this limits.

Condition is entered. AND Requirements of the C.2 Determine RCS is Prior to entering LCO not met in other acceptable for operation. MODE 2 or 3 than MODES 1, 2, and 3. BWR/4 STS 3.4.10-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 184 of 255

Attachment 1, Volume 9, Rev. 0, Page 185 of 255 RCS P/T Limits./ 3.4. ( SURVEILLANCE F 3.6A1, SR 3.4 E NOTEDLY 3.6.B.1E 3.6..2, iOnly required to be performed during RCS heatup 4.6A and cooldown operations and RCS inservice leak 4.6.3.1 and hydrostatic testina. fII T1F-- 3.6.B.3. DOC M.3 SR 3.4.iN Verify RCS pressure and RCS temperature are within the criticality limits specified in the,?TLF<. Once within 15 minutes prior 0 to control rod [Figue34= withdrawal for the purpose of achieving criticality SR 3.4.10.3 Only required to be met in ODES 1, 2, 3, and 4 during recirculation pump tartup [with reactor steam dome pressure 2 2 psig]. 0 Verify the difference be een the bottom head On within coolant temperature a the reactor pressure vessel 1 minutes prior (RPV) coolant temper ture is within the limits t each startup of specified in the PTLR recirculation ump BWRI4 STS 3.4.10-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 185 of 255

Attachment 1, Volume 9, Rev. 0, Page 186 of 255 3.4.9 0 INSERT I 3.6.2.1 2. Figures 3.4.9-2 and 3.4.9-3 shall be adusted as required by Figure 3.4.9-1 for the reactor vessel shell and fluid temperature limits. Q INSERT 2 3.6A.1 b. RCS heatup and cooldown rates are < 100F average over a 1 hour period. Q/ INSERT 3 laAM i.w.w. - 2 ----- NOTE Figure 3.4.9-4 shall be adjusted as required by Figure 3.4.9-1 for the reactor vessel shell fluid temperature limits. Insert Page 3.4.10-2 Attachment 1, Volume 9, Rev. 0, Page 186 of 255

Attachment 1, Volume 9, Rev. 0, Page 187 of 255 RCS P/T Limits 3.4 (i) SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY 3.6A-2 DOCI. SR SR .4M1,C

                                              &I MR TU Itr---a_

Only required to be met in MODES 1, 2, 3, and 4 00 during recirculation pump startup. Verify the difference between the reactor coolant Once within temperature in the recirculation loop to be started 15 minutes prior and the RPV coolant temperature islwithine limits1 [specifiegfe PTL'R.h!1 : 5O0-F to each startup of a recirculation 0 pump SR 3.4.113.5 [Only required to be met du ing a THERMAL POWER increase or recirc ation flow increase in MODES 1 and 2 with one i le recirculation loop when [THERMAL POWE is

30% RTP or when operating loop flow is
5 % rated loop flow].

0D Verify the difference be een the bottom head On within coolant temperature an the RPV coolant 15 minutes prior temperature is [ < 1450 to/a THERMAL P WER increase recirculation ow increase ] SR 3.4.1 0.6 [Only required to be met du ing a THERMAL POWER increase or recircu ation flow increase in MODES 1 and 2 with one n-isolated idle recirculation loop when [T ERMAL POWER is

30% RTP or when oper ng loop flow is
50%

rated loop flow]. 0 Verify the difference be een the reactor coolant On within temperature in the idle circulation loop and the 15 inutes prior RPV coolant temperat re is [

500QF. to THERMAL PpWER increase ot recirculation fQbw increase ]

BWR14 STS 3.4.10-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 187 of 255

Attachment 1, Volume 9, Rev. 0, Page 188 of 255 RCS P/T Limits 3.4A ( CTS SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY 3.6.8.4, 4.6.B.4 SR 3.4.?IA" . kIATC

                                                  . A NC I I-Only required to be performed when tensioning the 0 0 reactor vessel head bolting studs.

Verify reactor vessel flange and head flange 30 minutes temperatures arelwithin the liaec 0 LI AT-3.6.B.4, 4.6.8.4 SR 3.4-W t . a_ I----NC I t---

                                                                                                      ~D00 Not required to be performed until 30 minutes after RCS temperature

800F in MODE 4.

Verify reactor vessel flange and head flange 30 minutes temnperatures are Within the IiriLesa;&R fied in the I 0 tern70 . 3.6.8.4, 4.6.B.4 SR 3.4.8 NOTE-Not required to be performed until 12 hours after 0 0 RCS temperature S 100OF in MODE 4. Verify reactor vessel flange and head flange 12 hours temperatures are Within hp Iinflsecifed in ea Fp-T7M 0 3~D-0 BWR/4 STS 3.4.10-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 188 of 255

Attachment 1, Volume 9, Rev. 0, Page 189 of 255 3.4.9 INSERT 4 Figure 3.6.1 140 130 120 110 100 ofi X 80 un 70 60 50 40 30 20 10 0. o 0.5 1.5 2.5 3 3.5 1/4 Vessel Wall Thickness Fluence (10' 8n/cm 2 ) Figure 3.4.9-1 Core Beltline Operating Limits Curve Adjustment Versus Fluence Insert Page 3.4.10-4a Attachment 1, Volume 9, Rev. 0, Page 189 of 255

Attachment 1, Volume 9, Rev. 0, Page 190 of 255 3.4.9 INSERT 5 Figure 3.6.2 [L C (D a)

         -R a)

CL E W

          -a a)

E E C-0 200 400 (00 . 600 1(00 1200 1400 1600 1860 2000 Pressure Limit in Vessel Top Head (psig) Figure 3.4.9-2 RCS Pressure Versus Temperature Limits Inservice Leak and Hydrostatic Testing Insert Page 3.4.10-4b Attachment 1, Volume 9, Rev. 0, Page 190 of 255

Attachment 1, Volume 9, Rev. 0, Page 191 of 255 3.4.9 INSERT 6 Figure 3.6.3 0 a) L. a) E 100

      .c I-E o 20   a                  e           WW

1X0 in SM Pressure Limit in Vessel Top Head (psig)

Figure 3.4.9-3 RCS Pressure Versus Temperature Limits Non-Nuclear Heatup and Cooldown Insert Page 3.4.10-4c Attachment 1, Volume 9, Rev. 0, Page 191 of 255

Attachment 1, Volume 9, Rev. 0, Page 192 of 255 3.4.9 Figure 3.6.4 Q INSERT 7 L-C a) E a) I-a) E

     ,3 E

0 200 400 68 1 X1X 12:0 1400 1600 1800 200 Pressure Limit in Vessel Top Head (psig) Figure 3.4.9-4 RCS Pressure Versus Temperature Limits Critical Operation Insert Page 3.4.10-4d Attachment 1, Volume 9, Rev. 0, Page 192 of 255

Attachment 1, Volume 9, Rev. 0, Page 193 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (PIT) LIMITS

1. ISTS 3.4.10 is renumbered as ITS 3.4.9 since ISTS 3.4.5, "Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage," is not included in the Monticello ITS.

2. The utilization of a Pressure and Temperature Limits Report (PTLR) requires the development, and NRC approval, of detailed methodologies for future revisions to P/T limits. At this time, Monticello does not have the necessary methodologies submitted to the NRC for review and approval. Therefore, the proposed presentation removes references to the PTLR and proposes that the specific limits and curves be included in the RCS P/T Limits Specification (ITS 3.4.9).

3. Editorial changes have been made to achieve consistency with the wording for similar types of Conditions in other Specifications (e.g., ISTS 3.1.1 Condition A, ISTS 3.6.1.3 Condition F, ISTS 3.6.4.1 Condition A, ISTS 3.6.4.2 Condition C, ISTS 3.6.4.3 Conditions B and D, ISTS 3.7.4 Conditions B, C, and E, and ISTS 3.7.5 Conditions B and D) and with the three PWR ISTS (NUREGs -1431, -1432, and

  -1433, Rev 3). This change has also previously been approved by the NRC for Quad Cities, Dresden, and LaSalle ITS conversions.

4. ISTS SR 3.4.10.3, SR 3.4.10.5, and SR 3.4.10.6 have been deleted because they are not required in the CTS. Subsequent SR's have been renumbered, as applicable.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 193 of 255

Attachment 1, Volume 9, Rev. 0, Page 194 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 194 of 255

Attachment 1, Volume 9, Rev. 0, Page 195 of 255 RCS P/T Limits B 3.4. ( B 3.4 REACTOR COOLANT SYSTEM (RCS) B 3.4. RCS Pressure and Temperature (PIT) Limits 0 BASES BACKGROUND All components of the RCS are designed to withstand effects of cyclic loads due to system pressure and temperature changes. These loads are introduced by startup (heatup) and shutdown (cooldown) operations, power transients, and reactor trips. This LCO limits the pressure and temperature changes during RCS heatup and cooldown, within the design f assumptions and the stress limits for cyclic operation. I ton rchitcality, and also limits he P contains P/T limit curve r heatup, cooldown, and inservice leaky and hydrostatic testing, an at for the maximum rate of change of reactor coolant temperature. The heatuD cu - eslimits for both 1(i)e heatup riticalitI Each PIT limit curve defines an acceptable region for normal operation. The usual use of the curves is operational guidance during heatup or cooldown maneuvering, when pressure and temperature indications are monitored and compared to the applicable curve to determine that operation is within the allowable region. The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the reactor coolant pressure boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel. 10 CFR 50, Appendix G (Ref. 1), requires the establishment of P/T limits for material fracture toughness requirements of the RCPB materials. Reference I requires an adequate margin to brittle failure during normal operation, anticipated operational occurrences, and system hydrostatic tests. It mandates the use of the ASME Code, Section III, Appendix G (Ref. 2). The actual shift in the RTNDT of the vessel material will be established periodically by removing and evaluating the irradiated reactor vessel material specimens, in accordance with ASTM E 185 (Ref. 3) and Appendix H of 10 CFR 50 (Ref. 4). The operating P/T limit curves will be adjusted, as necessary, based on the evaluation findings and the recommendations of Reference 5. BWRI4 STS B 3.4.1 0-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 195 of 255

Attachment 1, Volume 9, Rev. 0, Page 196 of 255 RCS P/T Limits B 3.4.L1 (i) BASES BACKGROUND (continued) The P/T limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the reactor vessel and head that are the most restrictive. At any specific pressure, temperature, and temperature rate of change, one location within the reactor vessel will dictate the most restrictive limit. Across the span of the P/T limit curves, different locations are more restrictive, and, thus, the curves are composites of the most restrictive regions. The heatup Orve represents a differ t set of restrictions than th cooldown rve because the directi ns of the thermal gradients rough the vess1wall are reversed. The ermal gradient reversal al rs the 0 locatiorv6f the tensile stress beteen the outer and inner wa '3 The criticality limits include the Reference 1 requirement that they be at llea-sft4F above thheatup curve or the cooldown curve and not lower than the minimum permissible temperature for the inservice leal4f and J 0 hydrostatic testing.t 0 The consequence of violating the LCO limits is that the RCS has been operated under conditions that can result in brittle failure of the RCPB, possibly leading to a nonisolable leak or loss of coolant accident. In the event these limits are exceeded, an evaluation must be performed to determine the effect on the structural integrity of the RCPB components. ASME Code, Section XI, Appendix E (Ref. 6), provides a recommended methodology for evaluating an operating event that causes an excursion outside the limits. APPLICABLE The P/T limits are not derived from Design Basis Accident (DBA) SAFETY analyses. They are prescribed during normal operation to avoid ANALYSES encountering pressure, temperature, and temperature rate of change conditions that might cause undetected flaws to propagate and cause nonductile failure of the RCPB, a condition that is unanalyzed.

            'prvrhRefe              rR       eslishes the meftho ec                      -ordre rminin& the P/T limit.               0-Rp Tre         Since the P/T limits are not derived from any DBA, there are no rpecuiredbcaho acceptance limits related to the Prr limits. Rather, the P/T limits are acceptance limits themselves since they preclude operation in an unanalyzed condition.

RCS P/T limits satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO The elements of this LCO are: Fiurs .492 a. RCS pressur temperatureia heatuDWcooldown rat are within the limits specified iafthe PIRj E duringRCS heatup, cooldown, andl 0D inservice leak and hydrostatic tes in 00F average overa 1 hourperiod;` BWR/4 STS B 3.4.10-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 196 of 255

Attachment 1, Volume 9, Rev. 0, Page 197 of 255 B 3.4.9 0 INSERT I During inservice leak and hydrostatic testing, the reactor vessel shell temperatures (reactor vessel shell adjacent to shell flange, reactor vessel shell or coolant temperature representative of the minimum temperature of the beltline region) shall be at or above the temperatures shown on the two curves of Figure 3.4.9-2, where the dashed curve, "RPV Core Beltline (Full Power Years)," is increased by the core beltline temperature adjustment from Figure 3.4.9-1. The reactor vessel bottom head temperature shall be at or above the temperatures shown on the solid curve of Figure 3.4.9-2, "RPV Remote from Core Beltline," with no adjustment from Figure 3.4.9-1. During heatup by non-nuclear means and cooldown following nuclear shutdown the RCS temperatures (reactor vessel shell adjacent to shell flange, reactor vessel bottom drain, recirculation loops A and B, reactor vessel bottom head) shall be at or above the higher of the temperatures of Figure 3.4.9-3 where the dashed curve, "RPV Core Beltline (Zero Full Power Years)," is increased by the expected shift in RTNDT from Figure 3.4.9-1. 0 INSERT 2 During all operation with a critical reactor, the RCS temperatures (reactor vessel shell adjacent to shell flange, reactor vessel bottom head, and reactor vessel shell or coolant temperature representative of the minimum temperature of the beltline region) shall be at or above the higher of the temperatures of Figure 3.4.9-4 where the dashed curve, "RPV Core Beltline (Zero Full Power Years)," is increased by the expected shift in RTNDT from Figure 3.4.9-1. Insert Page B 3.4.10-2 Attachment 1, Volume 9, Rev. 0, Page 197 of 255

Attachment 1, Volume 9, Rev. 0, Page 198 of 255 RCS PIT Limits B 3.4. F(i) BASES LCO (continued)

b. Thytemperature difference bet een the reactor vessel bottom ead c olant and the reactor press e vessel (RPV) coolant is with' the it of the PTLR during recir ulation pump startup, and durig increases in THERMAL P0 ER or loop flow while operati g at low

                      / THERMAL POWER or 1p flow, The temperature difference between the reactor coolant in the respective recirculation loop and in the reactor vessel mee                       imi
               *IHt         t      rIKduring recirculation pump startuq and f-gring increases-I W in TH FRMA               WowRi or loop flow While operating'bt low THERMALIl RCS pressure and temperature are within the criticality limits seiidi         innrt AI;[      RE 1Digcitc A)lFigure 3.4.94   C--7 L<.      The reactor vessel flange and the head flange temperan es are within thee                  ILKlwhen tensioning the reactor vessel head th bolting studs.                              ___

and when the reactor head is tensioned These limits define allowable operating regions and permit a large number of operating cycles while also providing a wide margin to nonductile failure. The rate of change of temperature limits control the thermal gradient through the vessel wall and are used as inputs for calculating the heatup, cooldown, and inservice leakW and hydrostatic testing P/T limit curves. Thus, the LCO for the rate of change of temperature restricts stresses caused by thermal gradients and also ensures the validity of the P/T limit curves. Violation of the limits places the reactor vessel outside of the bounds-of the stress analyses and can increase stresses in other RCS components. The consequences depend on several factors, as follows:

a. The severity of the departure from the allowable operating pressure temperature regime or the severity of the rate of change of temperature&ij.

b. The length of time the limits were violated (longer violations allow the temperature gradient in the thick vessel walls to become more pronounced and BWR/4 STS B 3.4.10-3 Rev. 3.0, 03/31/04 Attachment 1,Volume 9, Rev. 0, Page 198 of 255

Attachment 1, Volume 9, Rev. 0, Page 199 of 255 RCS P/T Limits B3.4. F3 BASES LCO (continued)

c. The existences, sizes, and orientations of flaws in the vessel material. 0 APPLICABILITY The potential for violating a P/T limit exists at all times. For example, PIT limit violations could result from ambient temperature conditions that result in the reactor vessel metal temperature being less than the minimum allowed temperature for boltup. Therefore, this LCO is applicable even when fuel is not loaded in the core.

ACTIONS A.1 and A.2 Operation outside the P/T limits while in MODl 1, 2, 3 must be corrected so that the RCPB is returned to a condition that has been 0 verified by stress analyses. The 30 minute Completion Time reflects the urgency of restoring the parameters to within the analyzed range. Most violations will not be severe, and the activity can be accomplished in this time in a controlled manner. Besides restoring operation within limits, an evaluation is required to determine if RCS operation can continue. The evaluation must verify the RCPB integrity remains acceptable and must be completed if continued operation is desired. Several methods may be used, including comparison with pre-analyzed transients in the stress analyses, new analyses, or inspection of the components. ASME Code, Section Xl, Appendix E (Ref. 6), may be used to support the evaluation. However, its use is restricted to evaluation of the vessel beltline. 0 The 72 hour Completion Time is reasonable to accomplish the evaluation of a mild violation. More severe violations may require special, event specific stress analyses or inspections. A favorable evaluation must be completed if continued operation is desired. Condition A is modified by a Note requiring Required Action A.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action A.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. BWR/4 STS B 3.4.10-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 199 of 255

Attachment 1, Volume 9, Rev. 0, Page 200 of 255 RCS P/T Limits, B 3.4.tlO(i) BASES ACTIONS (continued) B.1 and B.2 If a Required Action and associated Completion Time of Condition A are not met, the plant must be placed in a lower MODE because either the RCS remained in an unacceptable P/T region for an extended period of increased stress, or a sufficiently severe event caused entry into an unacceptable region. Either possibility indicates a need for more careful examination of the event, best accomplished with the RCS at reduced pressure and temperature. With the reduced pressure and temperature conditions, the possibility of propagation of undetected flaws is decreased. Pressure and temperature are reduced by placing the plant in at least MODE 3 within 12 hours and in MODE 4 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. C.1 and C.2 Operation outside the P/T limits in other than MODES 1, 2, and 3 (including defueled conditions) must be corrected so that the RCPB is returned to a condition that has been verified by stress analyses. The Required Action must be initiated without delay and continued until the limits are restored. Besides restoring the P/T limit parameters to within limits, an evaluation is required to determine if RCS operation is allowed. This evaluation must verify that the RCPB integrity is acceptable and must be completed before approaching criticality or heating up to > OM. Several methods (E) may be used, including comparison with pre-analyzed transients, new E analyses, or inspection of the components. ASME Code, Section Xl, Appendix E (Ref. 6), may be used to support the evaluation; however, its Barn '.- -_#AtintA to -4: lto - I ..h ltlin o I M7,1% Ubt: l IZ)1VZl1tt:U UW VCHUCItLIII VI tilt- UtVlIL1lU.4 IN tK L.tj SURVEILLANCE REQUIREMENTS SR 3.4.1 0 Verification that operation is within l limits is required every 30 minutes when RCS pressure and temperature conditions are undergoing planned changes. This Frequency is considered reasonable in view of the control room indication available to monitor RCS status. Also, since temperature rate of change limits are specified in hourly increments, 30 minutes permits a reasonable time for assessment and correction of minor deviations. [ NET 4l 0 BWR/4 STS B 3.4.10-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 200 of 255

Attachment 1, Volume 9, Rev. 0, Page 201 of 255 B3.4.9 Cii INSERT 3 Condition C is modified by a Note requiring Required Action C.2 be completed whenever the Condition is entered. The Note emphasizes the need to perform the evaluation of the effects of the excursion outside the allowable limits. Restoration alone per Required Action C.1 is insufficient because higher than analyzed stresses may have occurred and may have affected the RCPB integrity. Q INSERT 4 The following locations must be monitored during RCS heatup and cooldown operations: a) reactor vessel shell adjacent to shell flange; b) reactor vessel bottom drain; c) recirculation loops A and B; and d) reactor vessel bottom head. The following locations must be monitored during inservice leak and hydrostatic testing: a) reactor vessel shell adjacent to shell flange; b) reactor vessel bottom head; and c) reactor vessel shell or coolant temperature representative of the minimum temperature of the beltline region. Insert Page B 3.4.10-5 Attachment 1, Volume 9, Rev. 0, Page 201 of 255

Attachment 1, Volume 9, Rev. 0, Page 202 of 255 RCS P/T Limits / B 3.4.f 0 BASES may bedscontinued when thre SURVEILLANCE REQUIREMENTS (continued) oecutve measurements at e and3 t~ocbtion l are within 5F. Surveillancefor Surveillance for heatu cooldow inservice lea a dtatic testing may be discontinued when the criteria given in the relevant plant procedure for ending the activity are satisfied. This SR has been modifiedwl Not 5requires this Surveillance to E 1i) be performed only during system heatup and cooldown operations and inservice leako§ and hydrostatic testing. 0 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical. Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal. [SR 3 J10.MSR 3.4..AIISR 3.4.10. 0. 3.4.10.6W CD t th~e Differential temperatures within theapli atPIL limitiJensur hat (i nozzles and thermal stresses resultin from the startup of an idle recirculation pump botom will not exceed design allowance .on addition, compliance with theEj4EEl head region limits ensures that the assumptions of the analysis for the startup of an Owl idle recirculation loop (Ref. 8) 7saed<< WPerforming the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limit will not be exceeded between the time of the Surveillance and the time of the idle CD07 pump starts [Limiting differential temperatures wit in the applicable limits durig a THERMAL OWER increase or re(r culation flow increase in si gle loop operation, hile THERMAL POW.,iR s 30% RTP or operating op flow 0 s 50% of ated loop flow, ensure tat resulting thermal stress s will not exceed esign allowances. BWRI4 STS B 3.4.10-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 202 of 255

Attachment 1, Volume 9, Rev. 0, Page 203 of 255 B 3.4.9 0 INSERT 4A Note 2 requires Figures 3.4.9-2 and 3.4.9-3 to be adjusted as required by Figure 3.4.9-1 (core beltline temperature adjustment) for the reactor vessel shell and fluid temperatures. This ensures the proper RCS pressure and temperature limits are met based on reactor fluence. INSERT 5 The following locations must be monitored to verify compliance with the P/T criticality curve limits: a) reactor vessel shell adjacent to shell flange; b) reactor vessel bottom drain; c) recirculation loops A and B; and d) reactor vessel bottom head. 0 0 This SR has been modified by a Note that requires Figure 3.4.9-4 to be adjusted as required by Figure 3.4.9-1 (core beltline temperature adjustment) for the reactor vessel shell and fluid temperatures. This ensures the proper RCS pressure and temperature limits are met based on reactor fluence. Insert Page B 3.4.10-6 Attachment 1, Volume 9, Rev. 0, Page 203 of 255

Attachment 1, Volume 9, Rev. 0, Page 204 of 255 RCS P/T Limits B 3.4 (i) BASES SURVEILLANCE REQUIREMENTS (continued) Performing t e Surveillance within 15minutes before starting the die recirculatio pump, THERMAL P0 R increase during single I op operation, r recirculation flow incr ase during single loop operDtion,0 provides dequate assurance that/he limits will not be excee d between the time hehe Surveillance and t e time of the idle pump sta , power increas, or flow increase.] An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.t[and .4.10.6is to compare Y1 the temperatures of the operating recirculation loop and the idle loop. RSR 3.4.0M3[Thes fCs have lbeen modified byla NoteMJ N es 1 that requires the Surveillance to be performed only injMODES 1, 2, 3, G 7. an ith ractor steam dompsig] [Certain MODES] In J E 5, the overall stress on limiting components is lower. Therefore, EATli not requiredIfor SKs 3.4.1Ua4d0.4 in MU E MODES 3, 4,an5, THERML POWER increases re not possibe and recirculation increases will not result in additinast resses. tlyd () Therefore, AT/Iimits are only required for SRs 3,X.10.5 and 3.4.10.6. The Notes also gfate that the SR is only required tg/be met during the event ofl concern (e>/g., pump startup, power increase/or flow increase) since this is when tje stresses occurill.1 The Note also states the SR is only required to be met during a recirculation pump startup, since this is when the stresses occur. SR3.4.9.4. SR 3.4.9.5, and SR 3.4.9.6~ ISR 3.4.10.7. SR-4H8and SR 3.4.10.91 i Limits on the reactor vessel flange and head flange temperatures are generally bounded by the other P/T limits during system heatup and cooldown. However, operations approaching MODE 4 from MODE 5 and in MODE 4 with RCS temperature less than or equal to certain specified values require assurance that these temperatures meet the LCO limits. The flange temperatures must be verified to be above the limits 30 minutes before and while tensioning the vessel head bolting studs to ensure that once the head is tensioned the limits are satisfied. When in MODE 4 with RCS temperature s 800F, 30 minute checks of the flange temperatures are required because of the reduced margin to the limits. When in MODE 4 with RCS temperature < 100 0F, monitoring of the flange temperature is required every 12 hours to ensure the temperature is within thellimit( BWR/4 STS B 3.4.10-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 204 of 255

Attachment 1, Volume 9, Rev. 0, Page 205 of 255 RCS P/T Limits B 3.4. ( BASES SURVEILLANCE REQUIREMENTS (continued) The 30 minute Frequency reflects the urgency of maintaining the temperatures within limits, and also limits the time that the temperature limits could be exceeded. The 12 hour Frequency is reasonable based on the rate of temperature change possible at these temperatures. REFERENCES 1. 10 CFR 50, Appendix G. 0

2. ASME, Boiler and Pressure Vessel Code, Section III, Appendix G.

3. ASTM E 185, ju2 98 0

4. 10 CFR 50, Appendix H.

5. Regulatory Guide 1.99, Revision 2, May 1988.

6. ASME, Boiler and Pressure Vessel Code, Section Xl, Appendix E.

7. lNEDO-2177acembeD6 1978b\ 0 Mt8. [FSAR, Septkm5.1.26].]l 0D

{Letter from Dart Hood (NRC) to Jeffrey S. Forbes (NMC), -Montireo Nuclear Generating Plant - Issuance of Amendment 133, Reviint Pressure-Temperature Curves,' dated February 24, 2003 GE Service Information Letter No. 517, Supplement 1, 'Analysis Basis for Idle Recirculation Loop Startup,' dated August 26, 1998. BWR/4 STS B 3.4.10-8 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 205 of 255

Attachment 1, Volume 9, Rev. 0, Page 206 of 255 B 3.4.9 0 INSERT 6 SR 3.4.9.4 is modified by a Note that requires the Surveillance to be performed only when tensioning the reactor vessel head bolting studs. SR 3.4.9.5 is modified by a Note that requires the Surveillance to be initiated 30 minutes after RCS temperature < 800 F in MODE 4. SR 3.4.9.5 is modified by a Note that requires the Surveillance to be initiated 12 hours after RCS temperature < 1000F in MODE 4. The Notes contained in these SRs are necessary to specify when the reactor vessel flange and head flange temperatures are required to be verified to be within the specified limits. Insert Page B 3.4.10-8 Attachment 1, Volume 9, Rev. 0, Page 206 of 255

Attachment 1, Volume 9, Rev. 0, Page 207 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.9 BASES, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (PIT) LIMITS

1. Changes have been made to reflect those changes made to the Specification.

2. Changes have been made to more closely match the Specifications.

3. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

4. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

5. These punctuation corrections have been made consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.

6. Typographical/grammatical error corrected.

7. The brackets have been removed and the proper plant-specific information/value has been provided.

Monticello Page I of 1 Attachment 1, Volume 9, Rev. 0, Page 207 of 255

Attachment 1, Volume 9, Rev. 0, Page 208 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 208 of 255

Attachment 1, Volume 9, Rev. 0, Page 209 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.9, REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE (PIT) LIMITS There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 209 of 255

Attachment 1, Volume 9, Rev. 0, Page 210 of 255 ATTACHMENT 10 ITS 3.4.10, Reactor Steam Dome Pressure . Attachment 1, Volume 9, Rev. 0, Page 210 of 255

Attachment 1, Volume 9, Rev. 0, Page 211 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 211 of 255

, Volume 9, Rev. 0, Page 212 of 255 ITS 3.4.10 Page 1 of 1 , Volume 9, Rev. 0, Page 212 of 255

Attachment 1, Volume 9, Rev. 0, Page 213 of 255 DISCUSSION OF CHANGES ITS 3.4.10, REACTOR STEAM DOME PRESSURE ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any requirements for Reactor Steam Dome Pressure. ITS LCO 3.4.10 requires reactor steam dome pressure to be < 1025.3 psig. This changes the CTS by incorporating the requirements of ITS 3.4.10. Appropriate ACTIONS and a Surveillance Requirement are also provided. The vessel overpressure protection analysis assumes that the reactor steam dome pressure is 1025.3 psig prior to event initiation. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. This change is acceptable since the determination of compliance with the overpressure protection criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved. This change is designated as more restrictive because it adds new requirements to the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 213 of 255

Attachment 1, Volume 9, Rev. 0, Page 214 of 255 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 214 of 255

Attachment 1, Volume 9, Rev. 0, Page 215 of 255 Reactor Steam Dome PressurwiT 3.4i 0 CTS 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4-.4 Reactor Steam Dome Pressure 0 DOC M.1 LCO 3.4i0 The reactor steam dome pressure shall be < 00O APPLICABILITY: MODES I and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC A. Reactor steam dome A.1 Restore reactor steam 15 minutes M.1 pressure not within limit. dome pressure to within limit. DOC B. Required Action and B.1 Be in MODE 3. 12 hours M.1 associated Completion Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE AFREQUENCY SR 3.4.V DOC Verify reactor steam dome pressure is

IO20] psig. 12 hours (00 BWR/4 STS 3.4.11-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 215 of 255

Attachment 1, Volume 9, Rev. 0, Page 216 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.10, REACTOR STEAM DOME PRESSURE

1. ISTS 3.4.11 is renumbered as ITS 3.4.10 since ISTS 3.4.5, "Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) Leakage," is not included in the Monticello ITS.

2. The brackets have been removed and the proper plant specific information/value has been provided.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 216 of 255

Attachment 1, Volume 9, Rev. 0, Page 217 of 255 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 217 of 255

Attachment 1, Volume 9, Rev. 0, Page 218 of 255 Reactor Steam Dome Pressurwe" B 3.4.W ( B 3.4 RETOR COOLANT SYSTEM (RCS) B 3.4.[ Reactor Steam Dome Pressure 0 BASES BACKGROUND The reactor steam dome pressure is A assumed initial condition of desi n basis accidents and transientE Ia1an assume value in I the determination o compliance with reactor pressure vessel overpressure protection criteria fej j APPLICABLE The reactor steam dome pressure of FLf-1501fpsig is an initial condition of SAFETY the vessel overpressure protection analysis of Reference 1. This ANALYSES analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety/relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection }e-analysis are conserved. Reference 2 also assumes an initial reactor steam dome pressure for the analyis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see and LCO 3.2.3, Bases for LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)")

       'UNEAR HEAT GENERATION       and 1% cladding plastic strain (see Bases for LCO 3.2.1. "AVERAGE RATE (LHGR)-     PLANAR LINEAR HEAT GENERATION RATE (APLHGR4".                                                           0 Reactor steam dome pressure satisfies the requirements of Criterion 2 of 10 CFR 50.36(c)(2)(ii).

L 3 LCO The specified reactor steam dome pressure limit of 5[1001psig ensures the plant is operated within the assumptions of theltransien-gna se Operation above the limit may result in a transient response more severe\ (3 I than analyzed. reactor overpressure protection analysisI APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these MODES, the reactor may be generating significant steam and the desiqn las-scidents and5I Itransieontsaoundin (events that maychallenge the overpressure limits are possible 0 In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits. BWR/4 STS B 3.4.1 1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 218 of 255

Attachment 1, Volume 9, Rev. 0, Page 219 of 255 Reactor Steam Dome Pressurwei B 3.4A (0 BASES ACTIONS A.1 With the reactor steam dome pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the bounds of the analyses. The 15 minute Completion Time is reasonable considering the importance of maintaining the pressure within limits. This Completion Time also ensures that the probability of an accident occurring while pressure is greater than the limit is minimized, l If ~the operator is unable to re tre the reactortemoe m Bpressure tobelow the limit, then the react hould be pla Ito be operating within the assumptions of te transient anaye. B.1 If the reactor steam dome pressure cannot be restored to within the limit within the associated Completion Time, 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 at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE S 40 REQUIREMENTS Verification that reactor steam dome pressure is <- psig ensures 0 that the initial conditions of the design basis accidents and transients are met. Operating experience has shown the 12 hour Frequency to be sufficient for identifying trends and verifying operation within safety analyses assumptions. REFERENCE AR, Section 5 . AR, Sectionj 1.l BWR/4 STS B 3.4.11-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 219 of 255

Attachment 1, Volume 9, Rev. 0, Page 220 of 255 JUSTIFICATION FOR DEVIATIONS ITS 3.4.10 BASES, REACTOR STEAM DOME PRESSURE

1. Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature, number, reference, system description, analysis description, or licensing basis description.

2. Typographical/grammatical error corrected.

3. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

4. Changes have been made to reflect those changes made to the Specification.

5. Changes have been made to more closely match the LCO requirements.

6. The brackets have been removed and the proper plant-specific information/value has been provided.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 220 of 255

Attachment 1, Volume 9, Rev. 0, Page 221 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 221 of 255

Attachment 1, Volume 9, Rev. 0, Page 222 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.4.10, REACTOR STEAM DOME PRESSURE There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 222 of 255

Attachment 1, Volume 9, Rev. 0, Page 223 of 255 ATTACHMENT 11 Relocated/Deleted Current Technical Specifications Attachment 1, Volume 9, Rev. 0, Page 223 of 255

Attachment 1, Volume 9, Rev. 0, Page 224 of 255 CTS 3/4.6.C.2, CTS 3/4.6.C.3, and CTS 3/4.6.C.4, Reactor Coolant Chemistry Attachment 1, Volume 9, Rev. 0, Page 224 of 255

Attachment 1, Volume 9, Rev. 0, Page 225 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 225 of 255

C C C, CTS 3/4.6.C.2, CTS 314.6.C.3, and CTS 3/4.6.C.4 3.0 LIMITIHIG CONDmINS I' Font OPERATION I 4.0 SURVEILLANCE nEmUIREMENTS I 2- (a) The r itor conlint water shAll not exceed the 2. During startup and at stea ing rates below 100,000 folio eq limiNs with steaming rates less than pounds per hour. n sampt of rendor cnoonnt shall 100, 0 poundn per hour except as rspociflMd hi be tnken every four hours and analyzed for 3.6. .2k.b conductivity and chlorldd ontent. CD Con udivily 5 I mholcm Chic ide ion 0.1 ppm CD 0 (b) For reacor slartups the maximum value for cr ucality shall not exceed 10 a mholcmnnd 0 l maximum vahle for chlorklt lol 3~ co entrotlice shall not exceed 0.1 ppm for th R.1e (01 fir il 24 hours after placing tle reactor In the wer operating condition. (0

3. Excel as specified In 3.6.C.2.b above, the react 3.(a) With steaming rates g eater than or equal to CD 0

coola it water shall not exceed tila following limit 100,000Wbs. per hour n reactor coolant sample witl. iecamitiu rates greater than or equal to shell be taken at lees every 96 hours and when the t00W thq. por hour. continuous contduct lty monitors indicate abnormal CD N, conductiity (other lhhn short-term spikes) and N, onductrvity 5 limhofern analyzed for conduc ly and chloride ton content. { See ITS 3.4.6 hlride Ion 0.5 ppr (b) When the contin'ot conduetivily monitor Is

4. I S1eclicalienss 3.6.0.1 through 3.6.C.3 ere ni ID inoporable, during For operation, a reactor
U me an orderly shutown shall be initiated andthe coolant sample sh id be taken once per 12 hours N,

to re or shal be in the cold shutdmon condition end analyzed for ducivity and chloride ion i4.24 hours. content. 0 IN, CA U1 3.6t4.ti 1 125 12t24/98 Am ndmenl No. 104 Page 1 of 2

CTS 3/4.6.C.2, CTS 314.6.C.3, and CTS 3/4.6.C.4 oooI I _ - 0 . FAILURE l 0 0 Cone2ftress pn CD 22 (0 . 0o 0. -4 O~N FAIl ' - 0 0 3.6/4.6 -low1118 u Figre4..2Chor4.Sres oroso3Tess eslt @00 Page 2 of 2

Attachment 1, Volume 9, Rev. 0, Page 228 of 255 DISCUSSION OF CHANGES CTS 3/4.6.C.2, CTS 314.6.C.3, and CTS 3/4.6.C.4, REACTOR COOLANT CHEMISTRY ADMINISTRATIVE CHANGES A.1 CTS Figure 4.6.2 provides an illustration of the chloride stress corrosion test results of stressed 304 stainless steel specimens. This figure is not included in the ITS. This changes the CTS by deleting Figure 4.6.2. CTS Figure 4.6.2 was included in the CTS as part of the Coolant Chemistry Bases. It is not used in the evaluation of any Coolant Chemistry Surveillance Requirement or LCO. The text in the Bases associated with Figure 4.6.2 was deleted in a previous Bases revision and Figure 4.6.2 should have been deleted with that revision. This change is designated as an administrative change and is acceptable because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS R.1 CTS 3/4.6.C.2 and CTS 3/4.6.C.3 provide the requirements on the conductivity and chloride ion content in the Reactor Coolant System (RCS) and CTS 3/4.6.C.4 provides the Actions if CTS 3/4.6.C.2 or CTS 3/4.6.C.3 is not met. Poor coolant water chemistry contributes to the long term degradation of system materials of construction, and thus is not of immediate importance to the unit operator. Reactor coolant water chemistry is monitored for a variety of reasons. One reason is to reduce the possibility of failures in the RCS System pressure boundary caused by corrosion. However, the chemistry monitoring activity is of a long term preventative purpose rather than mitigative. This Specification does not meet the criteria for retention in the ITS; therefore, it will be retained in the Technical Requirements Manual (TRM). This change is acceptable because CTS 3/4.6.C.2 and CTS 3/4.6.C.3 do not meet the10 CFR 50.92(c)(2)(ii) criteria for inclusion into the ITS. 10 CFR 50.36(c)(2)(ii) Criteria Evaluation:

1. The RCS chemistry limits are not installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary prior to a DBA. The RCS Chemistry Specification does not satisfy criterion 1.

2. The RCS chemistry limits are not a process variable that is an initial condition of a DBA or Transient Analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The RCS Chemistry Specification does not satisfy criterion 2.

3. The RCS chemistry limits are not a structure, system, or component that is part of the primary.success path and which functions or actuates to mitigate a DBA or Transient that either assumes the failure of or presents a challenge to Monticello Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 228 of 255

Attachment 1, Volume 9, Rev. 0, Page 229 of 255 DISCUSSION OF CHANGES CTS 314.6.C.2, CTS 3/4.6.C.3, and CTS 314.6.C.4, REACTOR COOLANT CHEMISTRY the integrity of a fission product barrier. The RCS Chemistry Specification does not satisfy criterion 3.

4. As discussed in Sections 3.5 and 6, and summarized in Table 4-1 (item 21 1) of NEDO-31466, the reactor coolant water chemistry was found to be a non-significant risk contributor to core damage frequency and offsite releases.

Nuclear Management Company, LLC has reviewed this evaluation, considers it applicable to Monticello, and concurs with the assessment. Since the 10 CFR 50.36(c)(2)(ii) criteria have not been met, the RCS Chemistry LCO and associated Surveillances may be relocated out of the Technical Specifications. The RCS Chemistry Specification will be relocated to the TRM. Changes to the TRM will be controlled by the provisions of 10 CFR 50.59. This change is designated as relocation because the LCO did not meet the criteria in 10 CFR 50.36(c)(2)(ii) and has been relocated to the TRM. REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 229 of 255

Attachment 1, Volume 9, Rev. 0, Page 230 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 230 of 255

Attachment 1, Volume 9, Rev. 0, Page 231 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 314.6.C.2, CTS 3/4.6.C.3 and CTS 3/4.6.C.4, REACTOR COOLANT CHEMISTRY There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 231 of 255

, Volume 9, Rev. 0, Page 232 of 255 CTS 314.6.H, Snubbers , Volume 9, Rev. 0, Page 232 of 255

Attachment 1, Volume 9, Rev. 0, Page 233 of 255 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 9, Rev. 0, Page 233 of 255

C C C CTS 3/4.6.H H.

3.0 LIMITINlCONDmONS FOR OPERATION l I l 4.0 SURVEILLANCE REQUIREMENT 1 I I H. snubbers

0) The following surveillance r quirements apply to oil 0) 0 1. Except as permitted below, all safety related 0 snubbers shall be operable whenever the sup orted safety related snubbers.

3 0 system Isrequired to be Operable.

1. Visual inspections:

3 0

2. With one or more snubbers made or found to e Snubbers are calegori ad as Inaccessible or

-L                  Inoperable for any reason when Operablily I                        accessible during rca or operation. Each of these required, within 72 hours:                                         categories (Inaccessi le or accessible) may be inspected Independe Ily according to the schedule 0                  a. Replace or restore the inoperable snub rs to                    determined by Table .6-1. The visual inspection           LAAe Operable status and perform an engine ring                    interval for each typ of snubber shall be                       0 3

0 evaluation or Inspection of the support determined based u on the crieria provided in (0 (0 components, or Table 4.6-1. The In at Inspection interval for new types of snubbers s al be established at 1 months 0

b. Determine through engineering evalu i[on that +25%. 0

.M the as-found condition of the snubber ad no adverse affect on the supported com nents -o CD and that they would retain their struct ral 101 Integrity in the event of design basis clrsmic event, or

c. Declare the supported system Mope able and take the action required by the Tech Ical 0) to Specifications for Inoperability of th system. -9 CD N 0W K' U'

U' U' 3.61 129 08/01/01 Amendment. .9,43A54,42, 122 Page 1 of 5

c C C CTS 3/4.6.H l I 3.0 iUMING CONqMONS FOR OPERATION

3. Al sat ty-related snubbers Installed or planned for _

I.. 4.0 SURVEILLANCE REOUIREMENTS I

2. Visual inspections shall verify tht (1)the snubber Do use a Monticello are hydraulic snubbers. No has no visible Indications of darage or Impaired mechnical snubbers are used on safety-related operability, (2) attachments to toe foundation or 0 0 syste s at Monticello. IfInstalled Inthe future, supporting strucure are funclt al, and (3) 3 to app priate Technical Specifications changes will be fasteners for the attachment o the snubber to the prop sad within 60 days of Installation. component and to the snubbe anchorage are 0 0r functional.

3 3 0 Snubbers which appear to Inoperable as a result =1 of visual Inspection shall be assified as 0 unacceptable, but may be re lassified as acceptable for the purpose f establishing the next LM-( 0 visual Inspection Interval, pr vided that (1)the CD cause of the rejection Is claely established and 0 remedied for that particular nubber and for other snubbers, irrespective of ,that may be a) generically susceptible; an (2)the affected

                                                                                                                                                    .P CD snubber Is functionally te ad In the as-found CD                                                                                  condition and determined       arable per Specification 4.6.H.4.                                            CD IDw 0)

A review and evaluation hall be perfornned and to documented to justify co linued operation with an unacceptable snubber. continued operation K) cannot be justified, the ubber shall be declared to Inoperable and the act n requirements shall be CD 0 met.

-9 01 N,

C-n cn 3.614.6 130 7/15/92 Amen ent No.3,9, 39, 82 Page 2 of 5

C C C CTS 3/4.6.H I 3.0 LIMING CONDI IONS FOR OPERATION I I Il 4.0 SURVEILLANCE REOUIREMENTS I I/l 3 Functional testin of snubbers s all be conducted at least once per Tonths +25% during cold 0) shutdown. Ten percent o I e Ia number of ea

0) brand of snubber shal be fund nally tested either CD in place or In a bench lest. For each snubber that does not meet the functional t I acceptance criteria 0 0

In Specification 4.6.H.4 below, n additional ten a percent of that brand shall be cunilonally tested until no more failures are foun or all snubbers of CD that brand have been tested. 0.1 The representative sample s ected for functional testing shall Indude the as configurations, operating environments, an the range of size and L_(9 0 0 capacity of the snubbers. 5, -o 0 In addition to the regular sa plo and specified -o CD re-samples, snubbers whic faRed the previous functional test shall be relte ed during the next test 0) 0 Co period If they were reinstall d as a safety-related 0 snubber. If a spare snubb r has been Installed In place of a failed safety rel ated snubber, It shall be N tested during the next pe

0) If any snubber selected fIr functional testing either

-to Cai falls to lockup or fails to .e. frozen In place) (Ive N the cause shall be evalu led and If caused by (71 U' manufacturer or design eficlency, all snubbers of the same design subje o the same defect shall be 01 functionally tested. (TI 3.614.6 131 08/01/01 Amendment No. 122 Page 3 of 5

C C C CTS 3/4.6.H I I.N I 3.0 LIMITNG COHDITONS FOR OPERATION Im 4.0 SURVEILIANCE REQUIREMENTS I I 4. Hydraulic snubber functional ests shall verify that:

0) 0)

a. Activation (restraining a ion) Is achieved within the specified range of v locity or acceleration In both tension and comp esslon. CD

.0 b. Snubber bleed, or rel aserate, where required, Is within the specified ange In compression or 2 tension. CD 0

5. For any snubbers found 1operable, an engineering evaluation or Inspection all be performed on the components which are pported by the snubbe rs.

-A                                                      The purpose of this eng ieerlng evaluation or Inspection shall be to d ermine l the components                   Lkle  CD 03                                                      supported by the snubb were adversely affected CO                                                       by the Inoperability of e snubbers In order to ensure that the sup          component remains                           CD capable of meeting th designed service.

CD 6.. The Installatlon and intenance records for each 0) safety related snubb shall be reviewed once every ionI to veri that the Indlcated A1 0 CA) service rife will not exc ed prior to the next to scheduled snubber rvice life review. If the CD Indicated service lif will be exceeded, the snubber 0 U' service life shall be a-evaluated or the snubber shall be replaced o reconditioned to extend Is service life beyond he date of the next scheduled 01 U' service fle review. This reevaluation, replacement, or reconditioning s all be indicated In the records. 3.64 132A 3/13/86 Am nment No. a, 39 Page 4 of 5

C C C CTS 3/4.6.H ble 4.6.11 SNUBBER VISU L INSPECTION INTERVAL

0) B Number of nubbers

                                                                                .M Q~acpal                                                                      02 0

Population Colun A Column B Column C or Category Extend nterval Repeat Interval Reduce Interval l(Notes t and 2L ¢ IotesM4 and 6n (Notes 5 and 61 1 0 1 80 0 2 0 100 0 1 4 150 0 3 200 2 (0 5 13 Ca 300 15 12 25 0, 0 ; Note 1: e next visual Inspection Interval for a snubber opulatlon or category size shall be determined ased upon the previous Ispeclion interval and the number of unaccept le snubbers found during that interval. Snubbe s may be categorized, based pon their accessibility during power operatlon, s accessible or Inaccessible. These categories ay be examined separately jointly. However, that decision must be mad and documented before any Inspection and the decision shall be used as the

                                                                                                                                                  ,3_9           0
                                                                                                                                                                 -o asis upon which to determine the next inspecon interval for Ihat category.

Note 2: Interpolation between population or category ses and the number of unacceptable snubbers i permissible. Use next lower integer for the value of the limit for Columns A.B or C If that Integer includes a fractional value f unacceptable snubbers as A) 101 determined by interpolation. co Note 3: If the number of unacceptable snubbers is eq at to or iess than the number In Column A, the xt inspection Interval may be twice the previous Interval but not greater the 48 months. Note 4: tIthe number of unacceptable snubbers Is e ual to or less than the number in Column B but eater than the number In Column 00 0 A, the next Inspection Interval shall be the s e as the previous Interval. 0' CA' Note 5: lIthe number of unacceptable snubbers is e uIt to or greater than the number In Column C, e next Inspection Interval shall K' CM be two-thirds of the previous Inlerval. Howeler, if tha number of unacceptable snubbers is le s than the number in Column C U' but greater than the number In Column B. t e next Interval shall be reduced proportionally b Interpolation, that Is, the previous Interval shall be reduced by a factor that Is ne-third of the ratio of the difference between tIhi number ol unacceptable snubbers found during the previous Interval and the number In Column B to the difference I the numbers In Columns B and C. Note : Alt Inspection iniervals up to and Including 8 months may be adjusted a maximum of plus 5%. I 3.6146 132a 08/01/01 Amendment o. 82r 122 Page 5 of 5

Attachment 1, Volume 9, Rev. 0, Page 239 of 255 DISCUSSION OF CHANGES CTS 3/4.6.H, SNUBBERS ADMINISTRATIVE CHANGES A.1 These changes to CTS 4.6.H.1, CTS 4.6.H.3, and CTS 4.6.H.6 are provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-04-036, from Thomas J. Palmisano (NMC) to USNRC, dated June 30, 2004. As such, these changes are administrative. MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LAA (Type 6 - Removal of LCO, SR, or other TS requirement to the TRM, UFSAR, ODCM, QAPD, or I1P) CTS 3/4.6.H provides the requirements for all safety related snubbers. This Specification is not included in the ITS. This changes the CTS by moving the explicit snubber requirements from the Technical Specifications to the Technical Requirements Manual (TRM). The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of CTS 3/4.6.H is to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses and licensing basis. The requirement to perform snubber inspections is specified in 10 CFR 50.55a and the requirement to perform snubber inspections and testing is specified in ASME Section XI. Therefore, both Monticello commitments and NRC Regulations or generic guidance will contain the necessary programmatic requirements for the inspection and testing of safety related snubbers without repeating them in the ITS. With the removal of OPERABILITY requirements from the Technical Specification, snubber OPERABILITY requirements will be determined in accordance with Technical Specification system OPERABILITY requirements. Also, this change is acceptable because the removed information will be adequately controlled in the TRM. The TRM is incorporated by reference into the USAR and any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications. Monticello Page 1 of 2 Attachment 1, Volume 9, Rev. 0, Page 239 of 255

Attachment 1, Volume 9, Rev. 0, Page 240 of 255 DISCUSSION OF CHANGES CTS 314.6.H, SNUBBERS LESS RESTRICTIVE CHANGES None Monticello Page 2 of 2 Attachment 1, Volume 9, Rev. 0, Page 240 of 255

Attachment 1, Volume 9, Rev. 0, Page 241 of 255 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 9, Rev. 0, Page 241 of 255

Attachment 1, Volume 9, Rev. 0, Page 242 of 255 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 314.6.H, SNUBBERS There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 242 of 255

Attachment 1, Volume 9, Rev. 0, Page 243 of 255 ATTACHMENT 12 Improved Standard Technical Specifications (ISTS) not adopted in the Monticello ITS Attachment 1, Volume 9, Rev. 0, Page 243 of 255

Attachment 1, Volume 9, Rev. 0,. Page 244 of 255 ISTS 3.4.5, RCS Pressure Isolation Valve (PIV) Leakage Attachment 1, Volume 9, Rev. 0, Page 244 of 255

Attachment 1, Volume 9, Rev. 0, Page 245 of 255 ISTS 3.4.5 Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 245 of 255

Attachment 1, Volume 9, Rev. 0, Page 246 of 255 RCS PIV Leakage 3.4.5 3.4 REACTO COOLANT SYSTEM (RCS 3.4.5 RC Pressure Isolation Valve ( IV) Leakage LCO 3.4.5 The leakage from ea h RCS PIV shall be within Ii. it. APPLICABILI MODES 1 and 2, MODE 3, except val es in the residual heat remo al (RHR) shutdown cooling flow paih when in, or during the tran tion to or from, the shutdown cooli g mode of operation. ACTIONS

                                                -NOTES-

1. Separat> Condition entry is allowed r each flow path.

2. Enter a plicable Conditions and Required Actions for systems mape inoperable by PlVs.

                                                                                                     -o C NDITION                         REQUIRED ACTION                 COMPLETION TIME A. One r more flow paths           --NOTE--a--

with I akage from one or Each alve used to satisfy Required more RCS PlVs not Action A.1 and Required Action A.2 withi limit, must ave been verified to meet SR 3. .5.1 and be in the reactor coolat pressure boundary for the high ressure portion of the l ~syste] A.1 Isolate the high pressure 4 hours portion of the affected system from the low pressure portion by use o one closed manual, de-activated automatic, or check valve. BWR/4 STS 3.4.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 246 of 255

Attachment 1, Volume 9, Rev. 0, Page 247 of 255 I RCS PIV Leakage 3.4.5 ACTIONS (co inued) CON ITION RE UIRED ACTION MPLETION TIME A.2 Is ate the high pressure 7 hours po ion of the affected sy tem from the low pr ssure portion by use of a s cond closed manual, de-a tivated automatic, or c eck valve. B. Require Action and B.1 e in MODE 3. 2 hours associa ed Completion Time n t met. AND B.2 e in MODE 4. 36 hours SURVEILL NCE REQUIREMENTS -o0 SURVEILL CE l FREQUENCY SR 3.4.51 -- NOTE -- - Not required to be erformed in MODE 3. Verify equivalent le kage of each RCS PIV is [In accordance l 0.5 gpm per nom nal inch of valve size up to a with the Inservice maximum of 5 gp , at an RCS pressure 2 [] an Testing Program

                  <   I psig.                                               or [18] months]

BWRI4 STS 3.4.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 247 of 255

Attachment 1, Volume 9, Rev. 0, Page 248 of 255 JUSTIFICATION FOR DEVIATIONS ISTS 3.4.5, REACTOR COOLANT SYSTEM (RCS) PRESSURE ISOLATION VALVE (PIV) LEAKAGE

1. NUREG-1433, ISTS 3.4.5, sets forth Limiting Conditions for Operation and Surveillance Requirements for Reactor Coolant System (RCS) pressure isolation valve (PIV) leakage. PIVs are defined as any two valves in series within the reactor coolant pressure boundary (RCPB) which separate the high pressure RCS from an attached low pressure system. These valves are normally closed during power operation.

The Reactor Safety Study (WASH-1400) identified the potential intersystem loss of coolant accident (Event V) in a PWR as a significant contributor to the risk of core melt. In this scenario, check valves fail in the injection lines of the RHR or low pressure injection systems, allowing high pressure reactor coolant to enter low pressure piping outside containment. Subsequent failure of this low pressure piping would result in loss of reactor coolant outside containment and subsequent core meltdown. Similar scenarios were also determined to be possible in BWRs. All plants licensed since 1979 have PlVs listed in their Technical Specifications, along with testing intervals, acceptance criteria, and limiting conditions for operation. Certain older plants were required to periodically leak test, on an individual basis, only those PIVs which were listed in an Order dated April 20, 1981 (Event V Order). That Order was sent to 32 operating PWRs and 2 operating BWRs. Other older plants have had no specific requirements imposed to individually leak test any of their PIVs. A number of events have occurred involving leakage past PlVs, failures of the valves, inadvertent actuation of the valves, or mispositioning of the valves. As a result, the NRC issued Generic Letter 87-06, "Periodic Verification of Leak Tight Integrity of Pressure Isolation Valves," which requested that licensees submit (1) a list of all PIVs in their plant; and (2) a description of the periodic tests or other measures performed to assure the integrity of the valve as an independent barrier of the RCPB, along with the acceptance criteria for leakage, operational limits, and frequency of test performance. NMC, LLC responded to Generic Letter 87-06 by letter dated June 1, 1987. All PlVs are tested in accordance with the requirements of ASME Section Xl, paragraph IWV-3420. In addition, the motor-operated PlVs are also tested in accordance with 10 CFR 50, Appendix J at a Frequency consistent with the testing required by the ASME OM Code tests. Monticello was licensed prior to 1979, and was not a recipient of the Event V Order to perform periodic leak tests of PlVs. Therefore, the requirements of NUREG-1433 Specification 3.4.5 do not currently apply to Monticello and are not incorporated in the ITS. Subsequent Specifications are renumbered accordingly. Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 248 of 255

Attachment 1, Volume 9, Rev. 0, Page 249 of 255 ISTS 3.4.5 Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 9, Rev. 0, Page 249 of 255

Attachment 1, Volume 9, Rev. 0, Page 250 of 255 RCS PIV Leakage B 3.4.5 B 3.4 REACT R COOLANT SYSTEM (RC ) B 3.4.5 RCS ressure Isolation Valve (PIV Leakage BASES BACKGROU D The function of RCS PIVs is to separate the high ressure RCS from an attached low pressu system. This protects the CS pressure boundary described in 10 CF 50.2, 10 CFR 50.55a(c), and GDC 55 of 10 CFR 50, Appendix A (Refs. 1 2, and 3). RCS PIVs are de ned as any two normally closed vales in series within the reacto coolant pressure boundary (RCPB). lVs are designed to meet th requirements of Reference 4. Durin their lives, these valves can produce varying amounts of reactor oolant leakage through eithe normal operational wear or mechanica deterioration. The RCS PIV LCO allows RCS high pressure op ration when leakage through these valv s exists in amounts that do n t compromise safety. The PIV leakage Iiit applies to each individual alve. Leakage through these valves is not included in any allowable LE KAGE specified in LCO 3.4.4, "RCS perational LEAKAGE."

-{

Although this spe ification provides a limit on al wable PIV leakage rate, its main purpose i to prevent overpressure fail re of the low pressure portions of conne ting systems. The leakage Ii it is an indication that the PIVs between the RCS and the connecting sys ems are degraded or degrading. PIV I akage could lead to overpres ure of the low pressure piping or compon nts. Failure consequences ould be a loss of coolant accident (LOCA) utside of containment, an u analyzed event that could degrade the abili for low pressure injection. A study (Ref. 5) valuated various PIV configu ations to determine the probability of int rsystem LOCAs. This study oncluded that periodic leakage testing f the PIVs can'substantially r duce intersystem LOCA probability. PIVs are providd to isolate the RCS from th following typically connected syst is:

a. Residual H at Removal (RHR) System,

b. Core Spra System,

c. High Pres ure Coolant Injection System and

d. Reactor C re Isolation Cooling System.

The PIVs are I sted in Reference 6. BWRI41STS B 3.4.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 250 of 255

Attachment 1, Volume 9, Rev. 0, Page 251 of 255 I I RCS PIV Leakage B 3.4.5 BASES APPLICABLE Reference 5 evaluated arious PIV configurations, le kage testing of the SAFETY valves, and operational hanges to determine the eff ct on the ANALYSES probability of intersyste LOCAs. This study conclu d that periodic leakage testing of the lVs can substantially reduce e probability of an intersystem LOCA. PIV leakage is not con idered in any Design Basis cident analyses. This Specification prov des for monitoring the conditi n of the RCPB to detect PIV degradatio that has the potential to cau a LOCA outside of containment. RCS PI leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) -CO RCS PIV leakage is I akage into closed systems c nected to the RCS. Isolation valve leakag is usually on the order of dr ps per minute. Leakage that increas significantly suggests that omething is operationally wrong a d corrective action must be taken. Violation of this LCO could result in c ntinued degradation of a PIV which could lead to overpressurization of a low pressure system and the loss of the integrity of a fission product b rrier. The LCO PIV leakag limit is 0.5 gpm per nominal inch of valve size with a maximum limit of 5 gpm (Ref. 4). -o_ Reference 7 permits leakage testing at a lower pr ssure differential than between the specifi maximum RCS pressure a d the normal pressure of the connected sy tem during RCS operation (t e maximum pressure differential). The o erved rate may be adjusted o the maximum pressure differentia by assuming leakage is dire ly proportional to the pressure differentia to the one-half power. APPLICABI TY In MODES 1,2, an 3, this LCO applies becaus the PIV leakage potential is greates when the RCS is pressurize . In MODE 3, valves in the RHR shutdown cooling flow path are not req ired to meet the requirements of thi LCO when in; or during tran ition to or from, the RHR shutdown cooling ode of operation. In MODES 4 and ,leakage limits are not provi ed because the lower reactor coolant pr ssure results in a reduced po ential for leakage and for a LOCA outside t e containment. Accordingly, he potential for the consequences of eactor coolant leakage is far ower during these MODES. 3WR/4 S S B 3.4.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume u, Rev. U, Page 251 of 255

Attachment 1, Volume 9, Rev. 0, Page 252 of 255 I I l RCS PIV Leakage

                                                                   /                B 3.4.5 BASES ACTIONS      The ACTIONS are m dified by two Notes. Note 1 as been provided to modify the ACTIONS elated to RCS PIV flow path . Section 1.3, Completion Times, s ecifies once a Condition has een entered, subsequent divisions subsystems, components, o variables expressed in the Condition disc vered to be inoperable or not within limits will not result in separate en ry into the Condition. Sectio 1.3 also specifies Required Actions of e Condition continue to appl for each additional failure, with Complet on Times based on initial ent into the Condition.

However, the Requi d Actions for the Condition RCS PIV leakage limits exceeded pro e appropriate compensato measures for separate affected RCS PIV fi w paths. As such, a Note ha been provided that allows separate Co dition entry for each affected CS PIV flow path. Note 2 requires an valuation of affected system if a PIV is inoperable. The leakage may h ye affected system OPERA LITY, or isolation of a leaking flow path wi h an alternate valve may hay degraded the ability of the interconnected ystem to perform its safety f nction. As a result, the applicable Conditio s and Required Actions for s stems made inoperable by PIVs must be e tered. This ensures appropri te remedial actions are taken, if necessary for the affected systems. A.1 and A.2 If leakage from on or more RCS PIVs is not wi hin limit, the flow path

                                                                                            -o0 must be isolated y at least one closed manual, deactivated automatic, or check valve withi 4 hours.

Required Action .1 and Required Action A.2 ate modified by a Note. stating that the v Ives used for isolation must eet the same leakage requirements as e PIVs and must be on the CPB [or the high pressure portion of the sy em]. Four hours provi es time to reduce leakage in excess of the allowable limit and to isola the flow path if leakage can ot be reduced while corrective action to reseat the leaking PIVs a e taken. The 4 hours allows time for t ese actions and restricts the ime of operation with leaking valves. Required Actio A.2 specifies that the double solation barrier of two valves be resto d by closing another valve q alified for isolation or restoring one le king PIV. The 72 hour Com letion Time considers the time required t complete the action, the low robability of a second valve failing during th s time period, and the low probability of a pressure boundary ruptu eof the low pressure ECCS iping when overpressurized to reactor pres ure (Ref. 7). BWRI4 TS B 3.4.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 9, Rev. 0, Page 252 of 255

Attachment 1, Volume 9, Rev. 0, Page 253 of 255 I I I RCS PIV Leakage B 3.4.5 BASES ACTIONS (con Inued B.1 and B.2 !the If leakage cannot be r duced or the system isolat the plant must be brought to a MODE in which the LCO does not appl . To achieve this status, the plant must be brought to MODE 3 within 12 hours and MODE 4 within 36 ho rs. This action may reduce t e leakage and also reduces the potential or a LOCA outside the conta nment. The Completion Times arreasonable, based on opera ng experience, to achieve the required lant conditions from full p rcditions in an orderly manner and ithout challenging plant syst is. SURVEILLAN E SR 3.4.5.1 REQUIREM TS Performance of leak ge testing on each RCS PIV s required to verify that leakage is below th specified limit and to identify ach leaking valve. The leakage limit of .5 gpm per inch of nominal alve diameter up to. 5 gpm maximum ap lies to each valve. Leakage esting requires a stable pressure condition. For the two PlVs in series, th leakage requirement applies to each val individually and not to the c mbined leakage across -o both valves. If the WVs are not individually leaka e tested, one valve may have failed co pletely and not be detected the other valve in series meets the le kage requirement. In this sit ation, the protection provided by redun ant valves would be lost. The 18 month Fre uency required by the Inservi e Testing Program is within the ASME de, Section Xl, Frequency r quirement and is based on the need to pe orm this Surveillance during n outage and the potential for an un lanned transient if the Surve lance were performed with the reactor at power. This SR is modifi d by a Note that states the le kage Surveillance is not required to be pe ormed in MODE 3. Entry int MODE 3 is permitted for leakage testing a high differential pressures wi h stable conditions not possible in the lo er MODES. REFERE CES 1. 10 CFR50.

2. 10 CFR 50. 5a(c).

3. 10 CFR 50, ppendix A, GDC 55.

BWR/4 TS B 3.4.5-4 Rev. 3.0, 03/31104 Attachment 1, Volume 9, Rev. 0, Page 253 of 255

, Volume 9, Rev. 0, Page 254 of 255 , Volume 9, Rev. 0, Page 254 of 255

Attachment 1, Volume 9, Rev. 0, Page 255 of 255 JUSTIFICATION FOR DEVIATIONS ISTS 3.4.5 BASES, REACTOR COOLANT SYSTEM (RCS) PRESSURE ISOLATION VALVE (PIV) LEAKAGE

1. Changes have been made to reflect those changes made to the Specification.

Monticello Page 1 of 1 Attachment 1, Volume 9, Rev. 0, Page 255 of 255}}

ML051990501

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