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Improved Technical Specifications, Volume 12, Revision 0, ITS Section 3.7, Plant Systems.
	ML051990476
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Site:	Monticello
Issue date:	06/29/2005
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To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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{{#Wiki_filter:IMPROVED TECHNICAL SPECIFICATIONS MONTICELLO NUCLEAR GENERATING PLANT VOLUME 12 ITS Section 3.7, Plant Systems Comm'tted to Ncler Excellon _-

Attachment 1, Volume 12, Rev. 0, Page 1 of 161 ATTACHMENT 1 VOLUME 12 MONTICELLO IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.7 PLANT SYSTEMS Revision 0 Attachment 1, Volume 12, Rev. 0, Page 1 of 161

Attachment 1, Volume 12, Rev. 0, Page 2 of 161 LIST OF ATTACHMENTS

1. ITS 3.7.1

2. ITS 3.7.2

3. ITS 3.7.3

4. ITS 3.7.4

5. ITS 3.7.5

6. ITS 3.7.6

7. ITS 3.7.7

8. ITS 3.7.8 Attachment 1, Volume 12, Rev. 0, Page 2 of 161

Attachment 1, Volume 12, Rev. 0, Page 3 of 161 ATTACHMENT I ITS 3.7.1, Residual Heat Removal Service Water (RHRSW) System Attachment 1, Volume 12, Rev. 0, Page 3 of 161

Attachment 1, Volume 12, Rev. 0, Page 4 of 161 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 12, Rev. 0, Page 4 of 161

C C ITS 3.7.1 ITS 3.0 LIMING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.7.1 C. y/GoetlISystem System C. Residual Heat Removal Service Water (RHRSW) . Except as specitied In 3.5.C.2 below, both Demonstratl the operability of the drywelt spray I Im

1. See ITS 3.6.1.8 1 Subsystems shall be headers and nozzles with an air test during each J 0 0 opera b enever irradiated fuel is in the reacto 10 Year teriod.5 ApplicabiSity reaor water empens In oM C s A sists of the fwi et Add proposed Surveillance SR 3.7.1.1 M.2 C e TS3618ad ri ITS 3.6.2.3 0 ha r Se S6 CD{ See ITS 3.6.1.8 O CINA2. One -mra/:d d Subsystem may be<

< Inopealfo 7 days. 5

    °ACTION               C                   3. L"If the requirernents of 3.5.C.1 or 2 cannot oe met, an orderly shutdown of the reactor will be Initiated                                                                                              0 Dland                                          the reactor water temperature shall be reduced-and in MODE 3                                                                               t0t Ltto                                                                                                                                                                                      toe 0                                                      lessthan22121 withi                                 in 12 hours                                                                              CD-EJ I For aiiowed out of service times for tho RHR pumpi

-0 lsee Section 3.5.A. 1

                        .                        ~~t                                                             See ITS 3.6.1.8 and ITS 3.6.2.3 Ja 3.514.5                                                                                                        104        08001/01 Amendment No. 27, 77, 70, 06, t02 122 Page 1 of I

Attachment 1, Volume 12, Rev. 0, Page 6 of 161 DISCUSSION OF CHANGES ITS 3.7.1, RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM 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 3.5.C.1 is applicable whenever irradiated fuel is in the reactor vessel and reactor water temperature is greater than 212 0F. ITS LCO 3.7.1 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring two RHRSW subsystems to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.5.C.1 is to ensure the RHRSW subsystems are OPERABLE to mitigate the consequences of a design basis accident. The RHRSW subsystems are required to be OPERABLE during MODES 1, 2, and 3 when there is considerable energy in the reactor core and a DBA could cause a significant heat up of the suppression pool. In MODES 1 and 3, the reactor coolant temperature will always be above 212 0 F. 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. Therefore, it is necessary and acceptable to require the RHRSW subsystems to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 Currently, the CTS does not provide any specific Surveillances to verify OPERABILITY of the RHRSW subsystems. ITS SR 3.7.1.1 requires verification that each RHRSW subsystem manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position every 31 days. This changes the CTS by adding this Surveillance Requirement to the Technical Specifications. The purpose of ITS SR 3.7.1.1 is to provide assurance that the proper flow paths will exist for RHRSW System operation. This change is acceptable because it provides additional assurance that the RHRSW System will be capable of performing its function. This change is designated as more restrictive because it adds Surveillance Requirements to the CTS. RELOCATED SPECIFICATIONS None Monticello Page 1 of 3 Attachment 1, Volume 12, Rev. 0, Page 6 of 161

Attachment 1, Volume 12, Rev. 0, Page 7 of 161 DISCUSSION OF CHANGES ITS 3.7.1, RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM REMOVED DETAIL CHANGES LA.1 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.5.C.1 states that an RHRSW subsystem consists of the following equipment powered from one division: 1 RHR Heat Exchanger, 1 RHR Service Water Pump, and valves and piping necessary for torus cooling. ITS 3.7.1 requires two RHRSW subsystems to be OPERABLE, but the details of what constitutes an OPERABLE subsystem are moved to the Bases. This changes the CTS by moving the details of what constitutes an OPERABLE subsystem to the Bases. The removal of these details, which are 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. The ITS still retains the requirement for two RHR service water subsystems to 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 of the ITS. This program provides for the evaluation of changes to ensure ihe 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. LESS RESTRICTIVE CHANGES L.1 (Category 4 - Relaxation of Required Action) When two RHRSW subsystems are inoperable, a unit shutdown is required by CTS 3.5.C.3; no time is provided to restore a subsystem prior to requiring the unit shutdown. With two RHRSW subsystems inoperable, ITS 3.7.1 ACTION B will allow 8 hours to restore one inoperable RHRSW subsystem prior to requiring a unit shutdown. This changes the CTS by allowing 8 hours to restore one of two inoperable RHRSW subsystems prior to requiring a unit shutdown. The purpose of CTS 3.5.C is to require sufficient containment cooling to ensure the primary containment conditions for the safety analyses are met. The proposed 8 hour Completion Time is acceptable since an immediate shutdown has the potential to result in a unit scram and discharge of steam to the suppression pool, when both RHRSW subsystems are inoperable and incapable of removing the generated heat. The 8 hours provides some time to restore one of the subsystems prior to requiring a shutdown (thus precluding the potential problem described above), yet is short enough that it does not significantly increase the probability of an accident to occur during this additional time. 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.2 (Category 3 - Relaxation of Completion Time) CTS 3.5.C.3 requires the unit to be shutdown and reactor water temperature reduced to less than 21 20 F within 24 hours if the requirements of CTS 3.5.C.1 or CTS 3.5.C.2 are not met. ITS 3.7.1 ACTION C requires the reactor be in MODE 3 in 12 hours and in MODE 4 in 36 hours. This changes the CTS by requiring the unit to be in Monticello Page 2 of 3 Attachment 1, Volume 12, Rev. 0, Page 7 of 161

Attachment 1, Volume 12, Rev. 0, Page 8 of 161 DISCUSSION OF CHANGES ITS 3.7.1, RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM MODE 3 in 12 hours and by extending the time to reduce reactor water temperature to < 212 0 F (i.e., MODE 4) from 24 hours to 36 hours. The purpose of CTS 3.5.C.3 is to place the unit outside the Applicability of the Specification 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. 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 also 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 amount of 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 3 of 3 Attachment 1, Volume 12, Rev. 0, Page 8 of 161

Attachment 1, Volume 12, Rev. 0, Page 9 of 161 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 9 of 161

Attachment 1, Volume 12, Rev. 0, Page 10 of 161 RHRSW System 3.7.1 CTS 3.7 PLANT SYSTEMS 3.5.C 3.7.1 Resid~ual Heat Removal Service Water (RHRSW) System 3.5.C.1 LCO 3.7.1 Two RHRSW subsystems shall be OPERABLE. 3.5.C.1 APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One RHRSW pu A.1 Restore R SW pump to 30 days inoperable. OPER E status. B. One RSW pump in B.1 Xestore one RHRSW pump 7 day 0D e subsystem to OPERABLE status. oprble.// 3.5.C.2 One RHRSW sub'system .1 ----- NOTE----- Enter applicable Conditions 0D inoperable otaitionnd. Re uired Actions of

                                             '-'LCO 3.4,        "Residual Heat Removal (RHR) Shutdown 0

Cooling System - Hot Shutdown," forjRHR shutdown cooling] made 0 inoperable by RHRSW System. Restore RHRSW 7 days subsystem to OPERABLE status. BWR/4 STS 3.7.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 10 of 161

Attachment 1, Volume 12, Rev. 0, Page 11 of 161 RHRSW System 3.7.1 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME DOC L.1 subsystems inoperable I-I .1 ------ NOTE-- Enter applicable Conditions 0D Ifor reasons C ~anI and Required Actions of C EJ. WL CO34?g for WRHR shutdown coolingM made 00O inoperable by RHRSW System. Restore one RHRSW subsystem to OPERABLE U8"Ehours 0 status. 3.5.C.3 Ra.1 Be in MODE 3. 12 hours 0D associated Completion Time not met. AND

                                             .2    Be in MODE 4.                   36 hours 0

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M.2 SR 3.7.1.1 Verify each RHRSW manual, power operated, and 31 days automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position or can be aligned to the correct-position. BWR/4 STS 3.7.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 11 of 161

Attachment 1, Volume 12, Rev. 0, Page 12 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.1, RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM

1. ISTS 3.7.1 ACTIONS A and B have been deleted because they are not applicable to Monticello. The Monticello USAR, Section 5.2.3.2.3 (for the long term primary containment response after a design basis loss of coolant accident analysis) only credits one RHRSW pump in one subsystem (i.e., a flow rate of 3500 gpm, which is the flow rate for one RHRSW pump). The following requirements have been renumbered, where applicable, to reflect this deletion.

2. Changes have been made to reflect changes made to other Specifications.

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

Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 12 of 161

Attachment 1, Volume 12, Rev. 0, Page 13 of 161 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 13 of 161

Attachment 1, Volume 12, Rev. 0, Page 14 of 161 RHRSW System B 3.7.1 B 3.7 PLANT SYSTEMS B 3.7.1 Residual Heat Removal Service Water (RHRSW) System BASES BACKGROUND The RHRSW System is designed to provide cooling water for the Residual Heat Removal (RHR) System heat exchangers, required for a safe reactor shutdown following a Design Basis Accident (DBA) or transient. The RHRSW System is operated whenever the RHR heat exchangers are required to operate in the shutdown cooling mode r3flin I modesthe suppression pool coolingor spray mode of the RHR System. Q The RHRSW System consists of two independent and redundant%-f subsystems. Each subsystem is made up of a header, two m gpm pumps, a suction source, valves, piping, heat exchanger, and associated instrumentation. Either of the two subsystems is capable of providing the required cooling capacity with one pump operating to maintain safe manual 1 Inch shutdown conditions. The two subsystems are separated from each normally close cross tie valved, so that failure of This 1 inch cross tie lins one subsystem will not affect the OPERABILITY of the other subsystem. > (ID Installed between the two The RHRSW System is designed with sufficient redundancy so that no J subsystems to allow both subsystems to be single active component failure can prevent it from achieving its design pressurizedbyoneoperating function. The RHRSW System is described in the MSAR, Section9..7 RHRSW pump. Reference 1. e is to the reactor buildingsiip se=vw=aedischarge Cooling water is pumped by the RHRSW pumps from the I ne. which discharges Riveq through the tube side of the RHR heat exchangers, and discharges 7Icag iet h Tircltn waei A minimuMkA+fw line from the p

                                                                          /~]                                            !

discharge to tqa-mme structure pre s the pump fr erheating whe ping against a closeschar e valve. The system is initiated manually from the control room. If operating during a loss of coolant accident (LOCA), the system is automatically (i2 tripped to allow thi'diesel generators to automatically power only that equipment necessary to reflood the core. The system can be manually started mnuteath A, or manetad any time the LOCA signal is manually overridden or clears. APPLICABLE The RHRSW System removes heat from the suppression pool to limit the SAFETY suppression pool temperature and primary containment pressure ANALYSES following a LOCA. This ensures that the primary containment can perform its function of limiting the release of radioactive materials to the environment following a LOCA. The ability of the RHRSW System to support long term cooling of the reactor or primary containment is BWR/4 STS B 3.7.1-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 14 of 161

Attachment 1, Volume 12, Rev. 0, Page 15 of 161 RHRSW System B 3.7.1 BASES APPLICABLE SAFETY ANALYSES (continued 1 discussed in theffSR ~ae~ UIAR Uc~-era- ana (RS (ReM. 2 I5 J l rn m p Mtiv ). Thy analysu! explicitly assumethat the RHRSW System will provide adequate cooling support to the equipment required for safe LtJ shutdown. ThiZ analysts includekthe evaluation of the long term Lii primary containment response after a design basis LOCA. Ei ED' The safety analysts for long term cooling A performed for various 0 combinations of RHR System failures. The worst case single failure that would affect the performance of the RHRSW System is any failure that E-pwould disable one subsystem of the RHRSW System. As discussed in 00

           .3     thitMSAR, Saction                            tflsig analyses, manual Lii (Ref. fR4or 52    -   initiation of the OPERABLE RHRSW subsystem and the associated RHR System is assumed to occur 1El0 minutes after a DBA. The RHRSW flow assumed in the analyMs isf9             gpmpru        with                         (on operating in one loop. In this case, the maximum suppression chamber water temperaturela-nZd issur9                                 nlw I        J 9      s~

20.1t~ad[iS 14ll res ive ,well below the design temperature of 3 F lnd rna0(imum Knowable proe L{ofJs it The RHRSW System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO Two RHRSW subsystems are required to be OPERABLE to provide the required redundancy to ensure that the system functions to remove post accident heat loads, assuming the worst case single active failure occurs coincident with the loss of offsite power. An RHRSW subsystem is considered OPERABLE when: a pumps PERABLE and 0

b. An OPERABLE flow path is capable of taking suction from the intake structure and transferring the water to the RHR heat exchangers at the assumed flow rate. Additionally, the RHRSW cross tie valve (which allow'the two RHRSW loops to be connected e close so that failure of oKe subsystem will not affec hPRBLT of Ithe other s9ubfstems."_/g An adequate suction source is not addressed in this LCO since the 9 minimum net positive suction head ft mean sea level in thepr 0 3viie Water SW)] System andgUltimate Heat 0 W TUHSS) Sink eB3 R . /

BWR/4 STS B 3.7.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 15 of 161

Attachment 1, Volume 12, Rev. 0, Page 16 of 161 B 3.7.1 Q INSERT I may be opened since the cross tie valve is only 1 inch in size and the RHRSW pump flow requirements (tested per the requirements of the Inservice Testing Program) account for the flow through the open cross tie valve. Insert Page B 3.7.1-2 Attachment 1, Volume 12, Rev. 0, Page 16 of 161

Attachment 1, Volume 12, Rev. 0, Page 17 of 161 RHRSW System B 3.7.1 BASES APPLICABILITY In MODES 1, 2, and 3, the RHRSW System is required to be OPERABLE to support the OPERABILITY of the RHR System for primary containment cooling (LCO 3.6.2.3, "Residual Heat Removal (RHR) Suppression Pool and therefore, the requirements are not the same for all facets of Coolingm _ "-kni _ _ _ __ (R_____ edo gpr,3 ) and decay heat removal (LCO 3.4."'Residual Heat __oval C nI 0 operatoninMODES4 Removal (RHR) Shutdown Cooling System - Hot Shutdown"). The and s. The LCOs of the system supported by the Applicability is therefore consistent with the requirements rqieet of these RHRSW System wilt systems. govern RHRSW System OPERABILITY In MODES 4 and 5, the OPERABILITY requirementsof the RHRSW requirements In MODES System are determined by the systems it supports. 0 ACTIONS A.1 With one RHRSW ump inoperable, the in perable pump must be restored to OPE LE status within 30 ys. With the unit in this condition, the re ining OPERABLE RH SW pumps are adequate to perform the RH W heat removal funct n. However, the overall reliability is redu ed because a single f ilure in the OPERABLE subsystem cou result in reduced RHISW capability. The 30 day Completion Ti e is based on the rem ining RHRSW heat removal 0 capability, inc ding enhanced reliabipty afforded by manual cross connect cap ility, and the low prob bility of a DBA with concurre worst case single ilure. B.1 With on RHRSW pump inoper ble in each subsystem, if no dditional failures ccur in the RHRSW stem, and the two OPERAB pumps are ali ned by opening the no mally closed cross tie valves, hen the remali ng QPERABLE pump and flow paths provide adeq ate heat remo al capacity following a/design basis LOCA. Howeve capability for this lignment is not assumed in long term containment re ponse analysis and an additional single fai ure in the RHRSW System co Id reduce the sy em capacity below th assumed in the safety analy is. Therefore, c tinued operation is pe mitted only for a limited time. One inoperable ppmp is required to be r stored to OPERABLE status ithin 7 days. The day Completion Time/or restoring one inoperable R RSW pump to PERABLE status is b sed on engineering judgment considering the level of redundancy provided. BWR/4 STS B 3.7.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 17 of 161

Attachment 1, Volume 12, Rev. 0, Page 18 of 161 RHRSW System B 3.7.1 BASES ACTIONS (continued) Required Action 1 is intended to handle the inoperability of one RHRSW subsystem e .an The Completion Time of 7 days is allowed to restore the RHRSW subsystem to OPERABLE status. With the unit in this condition, the remaining OPERABLE RHRSW subsystem is adequate to perform the RHRSW heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE RHRSW subsystem could result in loss of RHRSW function. The Completion Time is based on the redundant RHRSW capabilities afforded by the OPERABLE subsystem and the low probability of an event occurring requiring RHRSW during this period. The Re~quiread Action isrmodified by 'a Note'indicating that the applicable - Conditions of LCO 3.4J, be entered and Required Actions taken if the YU inoperable RHRSW subsystem results in inoperableTRHR shutdown cooling. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. With both RHRSW subsystems inoperablelfor reasons other that Condition B (e.g. subsystemsw noperable flow , or one subsysterg~wtnan ino eralump and one subsiha ino e flow pathF,the RHRSW System is not capable of performing its intended function. At least one subsystem must be restored to OPERABLE status within 8 hours. The 8 hour Completion Time for restoring one RHRSW subsystem to OPERABLE status, is based on the Com letion Times provided for the RHR suppression pool cooling !MdII M~functions. The Required Action is modified by a Note indicating that the applicable Conditions of LCO 3.4, be entered and Required Actions taken if the inoperable RHRSW subsystem results in inoperable MRHR shutdown coolingl. This is an exception to LCO 3.0.6 and ensures the proper (D actions are taken for these components. BWRI4 STS B 3.7.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 18 of 161

Attachment 1, Volume 12, Rev. 0, Page 19 of 161 RHRSW System B 3.7.1 BASES ACTIONS (continued)

                              .                                                                    0 If the RHRSW subsystems cannot be not restored to OPERABLE status within the associated Completion Times, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.1.1 REQUIREMENTS Verifying the correct alignment for each manual, power operated, and automatic valve in each RHRSW subsystem flow path provides assurance that the proper flow paths will exist for RHRSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. A valve is also allowed to be in the nonaccident position, and yet considered in the correct position, provided it can be realigned to its accident position. This is acceptable because the RHRSW System is a manually initiated system. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. REFERENCES 1 AR, Section E 1c .4ID 0 0 AR,2.31 (i) C) ate. (0 1 R o4AR-ecBne[6.cr (E0 BWR/4 STS B 3.7.1-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 19 of 161

Attachment 1, Volume 12, Rev. 0, Page 20 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.1 BASES, RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM

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

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. Changes are made to reflect those changes made to the Specifications.

4. These punctuation corrections have been made consistent with the Writers Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 5.1.3.

5. Typographical error corrected.

Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 20 of 161

Attachment 1, Volume 12, Rev. 0, Page 21 of 161 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 12, Rev. 0, Page 21 of 161

Attachment 1, Volume 12, Rev. 0, Page 22 of 161 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.1, RESIDUAL HEAT REMOVAL SERVICE WATER (RHRSW) SYSTEM There are no specific NSHC discussions for this Specification. I Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 22 of 161

Attachment 1, Volume 12, Rev. 0, Page 23 of 161 ATTACHMENT 2 ITS 3.7.2, Emergency Service Water (ESW) System and Ultimate Heat Sink (UHS) Attachment 1, Volume 12, Rev. 0, Page 23 of 161

Attachment 1,Volume 12, Rev. 0, Page 24 of 161 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 12, Rev. 0, Page 24 of 161

, Volume 12, Rev. 0, Page 25 of 161 ITS 3.7.2 4          Add propose-d IT-S 3.77.2 ]

Page 1 of 1 , Volume 12, Rev. 0, Page 25 of 161

Attachment 1, Volume 12, Rev. 0, Page 26 of 161 DISCUSSION OF CHANGES ITS 3.7.2, EMERGENCY SERVICE WATER (ESW) AND ULTIMATE HEAT SINK (UHS) ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any specific requirements for the Emergency Service Water (ESW) System or ultimate heat sink (UHS). The ESW System and UHS requirements are governed by the systems they support. ITS LCO 3.7.2 requires two ESW subsystems and the UHS to be OPERABLE. Appropriate ACTIONS and Surveillance Requirements are also provided. This changes the CTS by incorporating the requirements of ITS 3.7.2. The ESW System and UHS are necessary to support the equipment required for long term cooling of the reactor containment following a Design Basis Accident (DBA). The requirement to maintain two ESW subsystems and the UHS OPERABLE assures adequate cooling capacity is available for the removal of heat from equipment such as residual heat removal and core spray pump coolers, and ECCS room coolers required for safe shutdown following a DBA. The ESW System, together with the UHS satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). This change is acceptable because the ability of the ESW System to provide adequate cooling to the identified safety equipment is an implicit assumption for the safety analysis evaluated for a DBA. The ITS restoration actions for when an ESW subsystem or the UHS is inoperable is consistent with or more restrictive than the Technical Specification Systems they support. In addition, specific temperature and level requirements are now specified, as well as Surveillance Requirements. 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 12, Rev. 0, Page 26 of 161

Attachment 1, Volume 12, Rev. 0, Page 27 of 161 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 27 of 161

Attachment 1, Volume 12, Rev. 0, Page 28 of 161 [3A JBSWESystem and MUHSJ 3.7.2 0D CTS 3.7 PLANT SYSTEMS 3.7.2 jPan Service Water (13SWflSystem andgUltimate Heat Sink (UHS)l 0D DOCM.1 LCO 3.7.2 TwoiSWjjsubsystems andQUHSj shall be OPERABLE. 0 APPLICABILITY: MODES 1, 2, and 3. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One [PSW] pump A.1 Restore [P pump to 30 days inoperable. OPE E status.. B. One e W] pump in subsystem B.1 estore one [PSW] pump to OPERABLE status. 7 day 0 noperable.// C. [ One or more c g C.1 Restore ing tower 7 days] towers w ne cooling an inoperable. fa status. o OPERABLE 0

        ---REVIEWER'S NOTE               D.1    Verify water t perature of     Once per hour]

The [ ]OF is the maximm the UHS is 90]OF allowed UHS temper ure averaged the previous value and is based/n 24 hour eriod. temperature limit ions of the equipment at is relied upon for acci ent mitigation 0D and safe s tdown of the unit. D. AWater temperature of

             /the UHS> [90]OF and
         / S ]0F.

BWR/4 STS 3.7.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 28 of 161

Attachment 1, Volume 12, Rev. 0, Page 29 of 161

                                                                                    . D-"

1nSW1System and EJHSE 3.7.2 0 K'-) CT5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME DOCM.1 I OnehSsubsyster gnprblloreasop- [.1

1. Ent r applicle 00 other tCondjtn[s] A Co iditions nd I[agdC R quired Actions of/

yCO 3.8. ,",AC-/

                                                                /Sources/- Operatin ,"                             0 for die I generat made noperable y

[PS . m1 Enter applicable Conditions and 0 Reuired Actions of

                                                           ~LCO 3.4, "Residual Heat Removal (RHR)                                   0 Shutdown Cooling System - Hot Shutdown," for RHR shutdown co ling made inopera~bleby SWJ                                    0D Restore thenSVdV subsystem to OPERABLE                                  (00 status.

______________________ .1. BWR/4 STS 3.7.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 29 of 161

Attachment 1, Volume 12, Rev. 0, Page 30 of 161 Sstm and TUHSM 0jWSSWElSystem 3.7.2 e L< CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME DOC M.1 W Rquired Action and M.1 Be in MODE 3. 12 hours 0 associated Completion Time of Condition A:% AND 0 IEoiI not met. f-f J OR[.2 Be in MODE 4. 36 hours 0 OR Both~fSV44subsystems

                          .rreason other tbVCondijky[sl 13l

[adC . MOR UHS9 inoperableEfoi reasons WwthenanHT)( CondikKC [or DIDG SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Vaeri the water lev -e [PSW] cooling tower 24 ho DOC M.1 SR 3.7.2 0 Verify the water levelin PS4m well o 4 hoursO1 2each E} the intake structure] is t mean sea level. DOC M.1 SR 3.7 Verify the average water temperature of JHSI is 24 hours i SR 3.7.2.4 [Operate each [PS ingtower fan for 31 d 0

                              / [15] minute BWR/4 STS                                     3.7.2-3                           Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 30 of 161

Attachment 1, Volume 12, Rev. 0, Page 31 of 161 ESWESystem and jUHSf mi3.7.2 0D SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY DOC M. SR 3.7.2 --- NOTE-Isolation of flow to individual components does not 0 System inoperable. 0D Ver~ify -eac~h jSWlj subsystem manual bo~ 31 days 0 oeted] and automatic valve in the flow paths servicing safety related systems or components, 0 that is not locked, sealed, or otherwise secured in position, is in the correct position. DOC M.1 SR 3.7.2.- Verify each?SV1 subsystem actuates on an actual or simulated initiation signal. n months 0(E(! BWR/4 STS 3.7.2-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 31 of 161

Attachment 1, Volume 12, Rev. 0, Page 32 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.2, EMERGENCY SERVICE WATER (ESW) AND ULTIMATE HEAT SINK (UHS)

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

2. ISTS 3.7.2 ACTIONS A and B have been deleted because they are not applicable to Monticello. The Monticello ESW System design contains only one ESW pump per subsystem.

3. The bracketed ISTS 3.7.2 ACTION C has been deleted because it is not applicable to Monticello. Monticello does not include cooling towers for OPERABILITY of the ESW System or UHS.

4. The bracketed ISTS ACTION D has been deleted as it is not part of the plant specific ITS. The 900 F limit in ITS SR 3.7.2.2 is the maximum water temperature assumed in the accident analysis. Therefore, when 900 F is exceeded, the UHS is inoperable and ISTS 3.7.2 ACTION F (ITS 3.7.2 ACTION B) must be taken. The following requirements have been renumbered to reflect this and other ACTION deletions.

5. ISTS 3.7.2 Required Action E.1 Note 1 has been deleted because it is not applicable to Monticello. The Emergency Diesel Generator-Emergency Service Water System provides the Monticello emergency diesel generator cooling water. The following Note has been renumbered to reflect this deletion.

6. Changes have been made to reflect changes made to other Specifications.

7. The ISTS 3.7.2 Required Action E.1 Completion Time has been changed from 72 hours to 7 days. The requirement for 72 hours was based on the more limiting Completion Time associated with restoring an inoperable DG. The Monticello EDGs are supported by the EDG-ESW system. The Completion Time has been revised to be consistent with the time provided to restore an inoperable RHR and core spray subsystem, which are the subsystems supported by the ESW System.

8. Each ESW subsystem includes both manual and automatic valves. There are no power operated valves other than automatic valves: therefore, reference to "power operated" valves in ISTS SR 3.7.2.5 (ITS SR 3.7.2.3) has been deleted.

Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 32 of 161

Attachment 1, Volume 12, Rev. 0, Page 33 of 161 Improved Standard Technical Specifications (ISTS) Bases Markup

  -     and Justification for Deviations (JFDs)

Attachment 1, Volume 12, Rev. 0, Page 33 of 161

Attachment 1, Volume 12, Rev. 0, Page 34 of 161

                                                                                             -p VSWSystem and [UHSE B3.7.2      0 B 3.7 PLANT SYSTEMS B 3.7.2 P)Anl Service W                           SW)] System and WUltimate Heat SinkiUHS                                     0 BASES                                         l BACKGROUND                           The [SVV4 System is designed to provide cooling water for the removal of heat from equipment such as th                                       residual heat removal (RHR) pump coolers, and room coolers for Emergency Core Cooling System equipment, required for a safe reactor shutdown the condensers of the                  following a Design Basis Accident (DBA) or transient. Thd`MSWI System                     0 contiuornlmom ardspn loss of offsite power-generatr (EncDGbere closureora4.16 kV also provides cooling tolunf-epo-nents, as required~kig-no-malI ao. receipt ofp la l Upon accident (LOCA) sign

_essentialjI otsite power or loss of Errsam essential loads are aut oa~dsutorratically divided E lly isolated, the PSWM Divisions 1 andlone !SW1pump is automatically started in each division.

                                                                                                                             }  0 essential bus transfer to                                      c           o the alternate offsite powesour~e,,      S         he SV4System consists of theJUHtg nd two independent and                          }

reaundant susysems. tacn OT the two.5WE5su0systems is maoe up J oneffl gpm pumps, a suction source, valves, piping and (E C assrmenta tion. Either of the two subsystems is capable of suedin Eanlysts providing the required one o cooling capacity to support the re re systems peratin. Th t are separated from each C2 0 other so failure of one subsystem will not affect the OPERABILITY of the other system. Cooling water is pumped from the [Altah River] by thepSWq pumps to the essential components through the two main headers. After 0D removing heat f om the components, the water is discharged to the f discharge Ilint circulating water Ulaito revp.ace-Mporation losses fr4 [water system, or tthe river via a valv`. circulating 0D APPLICABLE SAFETY Sufficient water inventory is available for all ESWESystem post LOCA cooling requirements for a 30 day period with no additional makeup 0 ANALYSES water source available. The ability of th0IRSWJ System to support long term coolinc of the reactor containment is assumed in evaluations of the 0 ipment required for safe reactor shutdown presented in theTESAR,

                 -                                              (Refs. I land 2,5     0ective ). TI inclluud the evaluation of the long term primary containment response na          l 0
                                     ~afer a design basis LOCA.

The ability of thejSWj System to provide adequate cooling to the identified safety equipment is an implicit assumption for the safety 0 D- analysis evaluated in ReferenceM 1 1 . IThe ability to pro si te I emergenc e is dependent on the abili W] System to 0 BWR/4 STS B 3.7.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 34 of 161

Attachment 1, Volume 12, Rev. 0, Page 35 of 161 MSWESystem and UHS4 0 SB 3.7.2 BASES APPLICABLE SAFETY ANALYSES (continued) The long term cooling capability of the RHRjcore spray C50H . by thRE3SWV System. seiseFa dependent on the cooling provided 00 The SV System, together with theJUH¶, satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO TeWV subsystems are independent of each other to the degree that each has separate controls, power supplies, and the operation of one does not depend on the other. Inthe event of a DBA, one subsystem of Er SWEis required to provide the minimum heat removal capability 0

                 \assumed in the safety analysis for the system to which it supplies cooling water. To ensure this requirement is met, two subsystems ofIYSWj must be OPERABLE. At least one subsystem will operate, if the worst single active failure occurs coincident with the loss of offsite power.

stem is considered OPERABLE when it has an OPERABLE H OPERABLE pum, and an OPERABLE flow path capable ofo taking suction from the intake structure an tRAnsferring the water to the appropriate equipment. The OPERABILITY of the [UH$] is based on having a minimum water 0 level in the pump well of the intake structure of [P7] ft mean sea level and a maximum water temperature of 9C°F. t

                 ,The isolation of th [9g]O ,the design basis assumption associated with in al UHS temperature ar bounded provided the temperature of th UHS averaged over t e previous 24 hour period is s [90]0 F. With t e water temperature o the UHS > [90]0 F, long term cooling capabiyty of the ECCS loads d DGs may be affected.

Therefore, to nsure long term cooliig capability is provided to the CCS loads when/ ater temperature oft UHS is > [90]0 F, Required lion D.1 is pro ded to more frequentl monitor the water temperatur of the UHS an verify the temperature 's [90] 0 F when averaged ov the previou 24 hour period. The ce per hour Completion Tim takes into (3 consi ration UHS temperat e variations and the increase monitoring freqency needed to ensure design basis assumptions an equipment lim ations are not exceed in this condition. If the wate temperature of thy UHS exceeds [90]F hen averaged over the previ s 24 hour period rthe water temperatur of the UHS exceeds [ ]0 F, C dition F must be
~~/entered immediately.]~~
~~WS subsystem inoperable for reasons othr iton Al :3~~
- n Cniion C1 e k flowaherbhpumps inoperable i Wit th aSWE subsystem must be restored to OPERABLE status With the unit in this condition, the remaining OPERABLE 0,
~~subsystem is adequate to perform the heat removal function. [ I However the overall reliability is reduced because a single failure in the I~~
\OPERABL w3SWIsubsystem could result in loss ofSWjfunction. ,,,S-ucin (ID In 7 day The` 72(ou Completion Time is based on the redundant[SWESystemr 1I capabilities afforded by the OPERABLE subsystem, the low probability of an accident occurring during this time period, and is consistent with the allowed Completion Time for restoring an inoperable Fg. 0!
Required Actionx[3.l is modified by IM Notes indicating that the afppliable Conditions of LCO 3.8.1, "A Se Operating," CO e, "Residual Heat Removal (RHR) Shutdown Cooling System - Hot Shutdown," be entered and Required Actions taken if the inoperable L ]SWJsubsystem results in an inoperable 11RHR shutdown cooling 3 subsystem res ye. This is in accordance with LCO 3.0.6 and (i ensures the proper actions are taken for these components. BWR/4 STS B 3.7.2-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 37 of 161
Attachment 1, Volume 12, Rev. 0, Page 38 of 161 pmSWISystem andQUHSM ( T0 B 3.7.2 BASES ACTIONS (continued)
~~. a 2 0D cccZi If the fSWsubsystem cannot be restored to OPERABLE status within the associated Completion Time, or bothvSWs subsystems are (I~~
inoperable fIasonsroeR ion B and iti], Tor the lJUHSJ is determined inoperablelfor reasons othgtharrGondition C or D the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems. SURVEILLANCE [ SR 3.7.2.1 REQUIREMENTS This SR ensm s adequate long ter 30 days) cooling can be maintained /With the [UHS] wat source below the minimum I el, the affecte SW] subsystem m be declared inoperable. T 24 hour 0 Freg ency is based on ope ting experience related to tr ding of the p ameter variations dug the applicable MODES.] I R37..M ( }D Thi SRverifies the water level n well of the intake structure to be sufficient for the proper operation of theSV pumps (net positive suction head and pump vortexing are considered in determining this limit). The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES. 1@ ISR 3.7.21B1 (CI Verification of the IHUHSjtemperature ensures that the heat removal capability of thdSSWI System is within the assumptions of the DBA analysis. The 24 hour Frequency is based on operating experience related to trending of the parameter variations during the applicable 0 MODES. a BWR/4 STS B 3.7.2-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 38 of 161
Attachment 1, Volume 12, Rev. 0, Page 39 of 161 m3SWqSystem andRUHSM r B3.7.2 0 BASES SURVEILLANCE REQUIREMENTS (continued) (SR 3.7.2.4 Operating ea cooling tower fan for 15 minutes ensures that all ns are OPE LE and that all asso ted controls are functioning operly. Italso e ures that fan or moto ilure, or excessive vibration an be 0 detec d for corrective action/he 31 day Frequency is ba d on ope ting experience, the own reliability of the fan unit the I r undancy available, a the low probability of signifi nt degradation of
~~/thecooling tower fan ccurring between surveillan cs.]~~
~~S R 3.7.21g 0D~~
, Verifying the correct alignment for each manual, powei: eraTed3 and automatic valve in eachMSWVjsubsystem flow path provides assurance 0D that the roper flow paths will exist fS~qoperation. This SR does to valves that are locked, sealed, or otherwise secured in 0D position, since these valves were verified to be in the correct Position pnor to locking, sealing, or securing. lA valve is also allowed to, b~nte nonaccidentpppgs~n, and yet considered in the correct-oiion, provided 0D it can bea~fratically realigned to its accidenion within the manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
l-This SR is modified by a Note indicating that isolation of theMSWE System to components or systems may render those components or E s/sytems inoperable but does not affect the OPERABILITY of the-MSWilj System. As such, when alTESWM pumps, valves, and piping are 0 OPERABLE, but a branch connection off the main header is isolated, the SWJJ System is still OPERABLE. The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. BWR/4 STS B 3.7.2-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 39 of 161
Attachment 1, Volume 12, Rev. 0, Page 40 of 161 j3SWIjSSystem andRUHSq ] I B 3.7.2 BASES ~D SURVEILLANCE REQUIREMENTS (continued) SR 3.7.21 ( This SR verifies will autom the automatic isolaa ly switch to the saf cool ater exclusive! to lves of the [PSW] r emergency position t safetv related equipment ring an em ovide 10 linitiationsignal.This SRnd verifiestheautomaticstartcapabilityOf.\ ffjthe MSWpum in each subsystem.f : 2 Operating experience has shown that these components usually pass the SR when performed at thejf[month Frequency. Therefore, this 0 Frequency is concluded to be acceptable from a reliability standpoint. REFERENCES [1[.SAR, ha r 4.o3
~~.C P. FSAR, Ckapter [6]1 BWRJ4 STS B 3.7.2-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 40 of 161~~
~~Attachment 1, Volume 12, Rev. 0, Page 41 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.2 BASES, EMERGENCY SERVICE WATER (ESW) AND ULTIMATE HEAT SINK (UHS)~~
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. The brackets have been removed and the proper plant specific information/value has been provided.

4. The Reviewer's Note is deleted as it is not part of the plant-specific ITS.
Monticello Page 1 of I Attachment 1, Volume 12, Rev. 0, Page 41 of 161
Attachment 1, Volume 12, Rev. 0, Page 42 of 161 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 12, Rev. 0, Page 42 of 161
Attachment 1, Volume 12, Rev. 0, Page 43 of 161 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.2, EMERGENCY SERVICE WATER (ESW) AND ULTIMATE HEAT SINK (UHS) There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 43 of 161
Attachment 1, Volume 12, Rev. 0, Page 44 of 161 ATTACHMENT 3 ITS 3.7.3, Emergency Diesel Generator-Emergency Service Water (EDG-ESW) System Attachment 1, Volume 12, Rev. 0, Page 44 of 161
Attachment 1,Volume 12, Rev. 0, Page 45 of 161 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 12, Rev. 0, Page 45 of 161
~~, Volume 12, Rev. 0, Page 46 of 161 ITS 3.7.3 4- Add iS3-.7.3 Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 46 of 161~~
Attachment 1, Volume 12, Rev. 0, Page 47 of 161 DISCUSSION OF CHANGES ITS 3.7.3, EMERGENCY DIESEL GENERATOR-EMERGENCY SERVICE WATER (EDG-ESW) SYSTEM ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES. M.1 The CTS does not have any specific requirements for the Emergency Diesel Generator-Emergency Service Water (EDG-ESW) System. The EDG-ESW System requirements are governed by the EDG Technical Specifications. ITS LCO 3.7.3 requires two EDG-ESW subsystems to be OPERABLE. An appropriate ACTION and Surveillance Requirements are also provided. This changes the CTS by incorporating the requirements of ITS 3.7.3. The EDG-ESW System is necessary to support the emergency diesel generator requirement following a Design Basis Accident (DBA). The requirement to maintain two EDG-ESW subsystems OPERABLE assures adequate cooling capacity is available for the removal of heat from the emergency diesel generators required for safe shutdown following a DBA. The EDG-ESW System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). This change is acceptable because the ability of the EDG-ESW System to provide adequate cooling to the emergency diesel generators is an implicit assumption for the safety analysis evaluated for a DBA. The ITS restoration actions for when an EDG-ESW subsystem is inoperable is consistent with the EDG CTS Actions, since the ITS requires the associated EDG to be declared inoperable. In addition, specific Surveillance requirements are now specified. 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 12, Rev. 0, Page 47 of 161
Attachment 1, Volume 12, Rev. 0, Page 48 of 161 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 48 of 161
Aftachment 1, Volume 12, Rev. 0, Page 49 of 161 System 3.7.3 0D KJCT 3.7 PLANT SYSTEMS - - 3.7.3 Diesel Generator (G [lgadbi Service Water 6 System 0 subsystems DOC M.A LCO 3.7.3 he DG [ SysteF shall be OPERABLE. 0D [fMO MODES 1. 2,and 3 1 APPLICABILITY: When DG [1 Bl i tobe OPERABL. 0 A I MLGI I k Iel I UN; .. - " ik l NOTE Separate Condition entry is allowed for each EDG-ESW subsystem. 0D CONDITION REQUIRED ACTION COMPLETION TIME DOC M.1 A. B snl Q1t A.1 Align cc ling water to DG 8 hours
/ inoperable. [1 B] fr a Unit (1] plant r GESWservic water (PSW) ubsys t e J sub stem.
~~AND~~
~~. . (D)~~
A.2 erify cooling water s Once per 31 d ys aligned to DG [1 B] rm a Unit [1] PSW sub ystem. AN A Restore DG B] SSW 60 days System to ERABLE status. B. equire Action and Declare EgIB inoperable. Immediately E associ ed Comple Ion Time ot met. ____________________ A BWR/4 STS 3.7.3-1 Rev. 3.0, 03/31/04 Affachment 1, Volume 12, Rev. 0, Page 49 of 161
Attachment 1, Volume 12, Rev. 0, Page 50 of 161 E System 13.7f.3 03 iCTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY eiVEDG-ESW subsystem I DOC M.1 SR 3.7.3.1 Verify eachMGlljS5 b3§fiman 5We-1 31 days I ooerated-adautomaivle ithflw path, that is not locked, sealed, or otherwise secured in position, is in the correct position.
~~,_( each EDG-ESW subsystem I ,-%- 2t4s 00(~~
_ I DOC M.1 SR 3.7.3.2 Verify ktaj D*SW mppump starts automatically when starts and energizes the respective bus. t ( h soiae D BWR/4 STS 3.7.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 50 of 161.
Attachment 1, Volume 12, Rev. 0, Page 51 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.3, EMERGENCY DIESEL GENERATOR-EMERGENCY SERVICE WATER (EDG-ESW) SYSTEM
1. The brackets have been removed and the proper plant specific information/value has been provided.

2. ISTS 3.7.3 Required Actions A.1, A.2 and A.3 have been deleted because they are not applicable to Monticello. The Monticello EDG-ESW System design does not provide for credited alternate cooling sources. In addition, ISTS 3.7.3 Condition B has been deleted to reflect the ISTS 3.7.3 ACTION A requirements for inoperability of one or more EDG-ESW subsystems. The following requirements have been renumbered, where applicable, to reflect this deletion.

3. ISTS 3.7.3 provides the cooling water requirement for the DG I B SSW System. This Specification was a plant-specific Specification, written for the E. I. Hatch Units 1 and 2 design. Hatch Units I and 2 design includes a SSW System for the common DG (1B DG), separate from the cooling water for the remaining DGs. ISTS 3.7.2, Plant Service Water (PSW) System and Ultimate Heat Sink (UHS) provided the cooling water requirement for the remaining DGs. This is shown in the ISTS 3.7.2 Bases, Background section, which states that the PSW System provides cooling water for.
the removal of heat from equipment, such as the DGs. The ISTS 3.7.2 Applicability is MODES 1, 2, and 3. As stated in the ISTS 3.7.2 Bases, Applicability section, the MODES 4 and 5 OPERABILITY requirements of the PSW System and UHS are determined by the systems they support. Thus, for all DGs except the common 1B DG, no specific PSW requirements are provided in the ISTS for MODES other than MODES 1, 2, and 3. In addition, NUREG-1431, Rev.3, the Westinghouse ISTS, the ISTS includes only MODES 1, 2, 3, and 4 (which is equivalent to the BWR MODES 1, 2, and 3) for the DG cooling water Specification (ISTS 3.7.6, Service Water System). The Monticello design includes an EDG-ESW System that only provides cooling water to the EDGs. Therefore, consistent with the ISTS intent, the Applicability for ISTS 3.7.3 (ITS 3.7.3) has been changed to MODES 1, 2, and 3.
4. Each EDG-ESW subsystem includes only manual valves. There are no power operated or automatic valves; therefore, reference to "power operated" and "automatic" valves in ISTS SR 3.7.3.1 has been deleted.

5. An ACTIONS Note has been added to ISTS 3.7.3 (ITS 3.7.3) to allow separate Condition entry for each inoperable EDG-ESW subsystem. This change is intended to ensure that each occurrence of an inoperable EDG-ESW subsystem be assessed in accordance with the applicable Conditions and Required Actions of LCO 3.8.1 for its impact on the EDG capability to function as an AC power source. This change is acceptable since the ITS 3.7.3 Required Action is to declare the associated EDG inoperable.
Monticello Page 1 of I Attachment 1, Volume 12, Rev. 0, Page 51 of 161
Attachment 1, Volume 12, Rev. 0, Page 52 of 161 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 52 of 161
Attachment 1, Volume 12, Rev. 0, Page 53 of 161 E SW-E System Q B 3.7.3 (i B 3.7 PLANT SYSTEMS B 3.7.3 Diesel Generatorl(DG) [1EIStandF Service Water (f e System BASES QM _ -F3,-DG--- BACKGROUND The I System isi designed to~provide cooling water for the the ofheatisrom P the only componentserved 2)
\by the JDG [lWSSWWI System. ' TiDGeSSWpump autostartS upon receipt of ieraor Mississippi start signal when power is available to the pump's electrical bus. EGS 0ing water is pumped from th aa River] by thee y
{E)pumF'to the essentia2,DG components through the header. ter removing heat from the components, the water is discharged to the un I eieenaterWP'iiSChame-ra-l~h chepaapabiexi to Syteto tu lrn e he DG [1 B1j Y eswhn te SWpum ilb el complete description FE -DG -EFSW j o System is presented n eSAR, Se(it)o APPLICABL he ability of the 4 System to provide adequate cooling to theO SAFETY ANALYSES ba is an implicit assumption for the safety analyses presented in the (Refs. 2land 3, ecsive . The ability to 0 (0 I Secton 14.7.2 vide onsite emergency AC power is dependent on the ability of the System to cool the503c (Di The DG WSSY\ System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). () LCO The OPERABIL of the DG [1 B] SS ystem is required to provide N3 coolant sou o ensure effective eration of the DG [1B] in t 6vent o0 an acc nt or transient. Th PERABILITY of the DG [1 SW SysterT ed on having an 9PRABLE pump and an OPERBLE flow path. EDG-ESW lAn adequate suction source is not addressed in this LCO since the minimum net positive suction head of theIDG flBI SSVV pump Is boun ed gby thQSW requirements (LCO 3.7.2, i Service Water (SWi System and EUltimate Heat Sink (UHS t).3 t{E APPLICABILITY The requirements for OPE ITY of the DG [1 B] SSW m are governed by the reP PERABILITY of the DOG (LCO 3.8.1, "AC 0 Sources - Oeng," and LCO 3.8.2, "AC ces - Shutdown"). BWR/4 STS B 3.7.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 53 of 161
Attachment 1, Volume 12, Rev. 0, Page 54 of 161 B 3.7.3 Q INSERT I In addition, the Service Water System to EDG-ESW System isolation valves open upon receipt of an associated EDG start signal. If offsite power is not lost, then the Service Water System will supply cooling water to the associated EDG (i.e., the EDG-ESW and Service Water pumps are in parallel, and the discharge pressure of the Service Water pump is higher than the EDG-ESW pump). However, the Service Water System capability is not required to be OPERABLE as part of this Specification. Q' INSERT IA The EDG-ESW subsystems are independent of each other to the degree that each has separate controls, power supplies, and the operation of one subsystem does not depend on the other subsystem. In the event of a DBA, one EDG-ESW subsystem is required to provide the minimum heat removal capability assumed in the safety analysis for the associated EDG. To ensure this requirement is met, two EDG-ESW subsystems must be OPERABLE. At least one subsystem will operate, if the worst single active failure occurs coincident with the loss of offsite power. A subsystem is considered OPERABLE when it has an OPERABLE pump and an OPERABLE flow path capable of taking suction from the intake structure and transferring the water to the associated EDG. Q INSERT2 In MODES 1,2, and 3, the EDG-ESW System is required to be OPERABLE to support OPERABILITY of the EDGs. Therefore, the EDG-ESW System is required to be OPERABLE in these MODES. In MODES 4 and 5, the OPERABILITY requirements of the EDG-ESW System are determined by the systems it supports. Therefore, the requirements are not the same for all facets of operation in MODES 4 and 5. LCO 3.8.2, 'AC Sources - Shutdown," will govern EDG-ESW System OPERABILITY requirements in MODES 4 and 5. Insert Page B 3.7.3-1 Attachment 1, Volume 12, Rev. 0, Page 54 of 161
Attachment 1, Volume 12, Rev. 0, Page 55 of 161 B9ystem B 3.7.3 0D BASES ACTIONS A.1. A.2, and A.3 If the DG [1B] 55 System is inoperabl ,the OPERABILITY of the DG [1B] is affect due to loss of its co ing source; however, the capability exists/o provide cooling to [1 B] from the PSW Syste of Unit [1]. Conti ued operation is allo d for 60 days if the OPERA LITY of a Unit 1 P System, with respe t to its capability to provide c oling to the DG [1B] can be verified. This's accomplished by aligning c oling water to D [1 B] from the Unit 1 SW System within 8 hours ad verifying is lineup once every 1 days. The 8 hour Complet n Time is based oW the time required to r asonably complete the Req red Action, and th low probability of an ent occurring requiring DG B] during this perio The 31 day verificat n of the Unit [1] PSW lineupfo the DG [1 B] 0D is cqdsistent with the PSW alve lineup SRs. The 60 da Completion Ti e to restore the DG [1E] SSW System to OPERABlI status allows s icient time to repair t e system, yet prevents indefi ite operation with oling water provided rom the Unit [1] PSW Syste LE-i EL 0D With one or more traqde8h e made available to the DG [1 Nithe 8 hour EDG-ESW subsystems If cooling water can Inoperable Completion Tj, ori oln ater cann DIG [1;em a Unit [1] PSIA suZ Xs rquired by the 31 day odt eaindt 0 cad&tion Required ActiortU(1B cannot perform its intended [asocitedEl function and must be immediately declared inoperable. In accordance 0 with LCO 3.0.6, this also requires enteringOiEaIthepXpplicable Conditions J and Required Actions-EdLCO 3 rAddition ly, if the Sources-po~eraing. 1 66 D1 S B ot restored to P E stat within must be immediately declared inoper le. 0D SURVEILLANCE SR 3.7.3.1 REQUIREMENTS Verifying the corrept alignment for manualL power peraled, and fI FAG^. I--
~~-ne -. 1A Acts tl]ti-tSW IocoUW- --.. -. oe_~~
51 AF5wm I valves 47he -DG (IJ51-eM Sys-t-em]flow path provides assurance that the proper flow paths will exist ofPG 1B 1 S steml 0 operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these valves were verified to be in the correct position prior to locking, sealing, or securing. lA vales also 0D BWR/4 STS B 3.7.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 55 of 161
Attachment 1, Volume 12, Rev. 0, Page 56 of 161 B 3.7.3 Q INSERT 3 A Note has been provided to modify the ACTIONS related to EDG-ESW 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 EDG-ESW subsystems provide appropriate compensatory measures for separate inoperable EDG-ESW subsystems. As such, a Note has been provided that allows separate Condition entry for each inoperable EDG-ESW subsystem. Insert Page B 3.7.3-2 Attachment 1, Volume 12, Rev. 0, Page 56 of 161
Attachment 1, Volume 12, Rev. 0, Page 57 of 161
@31tSystem B 3.7.3 0 BASES SURVEILLANCE REQUIREMENTS (continued) allowed to be in the npffaccident position, and yet be coered in the correct position prided it can be automatically rened to its accident 0 position, within We required time This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.
~~The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. SR 3.7.3.2 each EDGt~~
~~.This SR ensures that theD SSWI ystem pump will automatically start to provide required cooling to th 1 when theD 1 B starts (astedEDG J and the respec ive bus is energized.~~
24 Operating experience has.shown that these components usually pass the SR when performed at thsy month Frequency, which is based at the refueling cycle. Therefore, this Frequency is concluded to be acceptable 0D from a reliability standpoint. REFERENCES 1B AR, Section [-E3D 0 0
2. SAR, h er] 63 S 4.72.3.1.1 .0 0

13. FSAR-C6pter [151.1 0D BWR/4 STS B 3.7.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 57 of 161
~~Attachment 1, Volume 12, Rev. 0, Page 58 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.3 BASES, EMERGENCY DIESEL GENERATOR-EMERGENCY SERVICE WATER (EDG-ESW) SYSTEM~~
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. The brackets have been removed and the proper plant specific information/value has been provided.

4. Changes have been made to reflect changes made to other Specifications.

5. Typographical error corrected.
Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 58 of 161
Attachment 1, Volume 12, Rev. 0, Page 59 of 161 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 12, Rev. 0, Page 59 of 161
Attachment 1, Volume 12, Rev. 0, Page 60 of 161 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.3, EMERGENCY DIESEL GENERATOR-EMERGENCY SERVICE WATER (EDG-ESW) SYSTEM There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 60 of 161
Attachment 1, Volume 12, Rev. 0, Page 61 of 161 ATTACHMENT 4, ITS 3.7.4, Control Room Emergency Filtration (CREF) System Attachment 1, Volume 12, Rev. 0, Page 61 of 161
Attachment 1,Volume 12, Rev. 0, Page 62 of 161 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 12, Rev. 0, Page 62 of 161
C, CI ITS 3..4 ITS ITS 3.0 LIMmNG CONDITiONS FOR OPERATION 4.0 SURVEILLANCE RECUIREMENTS .0) 3.a With both control room ventilation trains Inoperable, L) W restore at least one train to operable status wvthin 24 0 hours. 0 3 3.b It 3.a Is not met, then be in hot shutdown within the next 12 hours and In cold shutdown within 24 hours following { See ITS 3.7.5} 0 the 12 hours.
-U 0 3.c 1t3.a Is not met during movement of Irradiated fuel assemblies In the secondary contninment, core 0 -o 3
~~0 alterations, or activities having the potential for draining the reactorvessel then Immediately suspend those~~
) 3 activiies. Li 0 3.7.4 ddLC B. Control Room Emergency Filtration System t.A1 :a t.D B. Control Room Emergency Filtration System l
t. At least once per month, initiate CD LCO 3.7.4 (D I 1. Exceptassprcifiedin3.17.i.1.alhroughdbelowtrol control room emexency filtration system filter trainA l looocTa+/- I o%1 flow hrounh both trans of the ' 2) CD emergency filtration treatment system. The system shall

0) operate for at least 10 hours with the heaters M
co 0 Applicability 0
-9, to -9' CD CD
0) ,m W ,
0
~~,0 3.1714.17 229v 8118/00 89 Amendment No. GS, , 112 Page 1 of 5~~
C C C ITS 3.7.4 ITS ACTION A I CD 3-2 0 ACTION C tD rD 3 0 ACTION B 0 3 0 CD ACTION C
. With one control room emergency filtration em ;u filter train Inoperable during movement o rradiated tD ACTION A fuel assemblies In the second containment , K) -o or acttles hav ng the potential for 6 -6 draining the reactor vessel, restore the Iopera. tD a) train to operable status within 7 d or immediately tot aner the 7 ays iniuate andm the operab CD ACTION D emergency filtration system filter train In the la pressurization mode or Immediately suspend these 0
~~0) activities.~~
~~-4 6,. ACTION F 6,~~
I3.17/4.17 229vv 818100 1 Amendment No. 112 Page 2 of 5
C C C 0 ITS 3.7.4 l Add proposed SR 3.7.4.2 7-~ 3.0 UMmNG CONDIONS FOR OPERATiON --- l 4.0 SURVEIULANCE REQUIREMENTS 1 I 2. Performance Requirement Test
2. Performance Requirements The in-place performance testing of HEPA filter banks and charcoal adsorber banks shall be conducted In

a. Acceptance Criteria - Periodic Requirements accordance with Sections 10 and 11 of ASME 0 (1) The results of the In-pace DOP tests at 1000 N510-1989. The carbon sample test for methyl Iodide 0 0) cfm (+/- 10%) shall show s 1% DOP penetration shali be conducted In accordance with ASTM cX D on each IndMdual HEPA filter and shall show D 3803-1989. Sample removal shall be In accordance 0 S0.05% DOP penetration on the combined with Regulatory Position C.6.b of Regulatory Guide 1.52.
HEPA filters. Revision 2, March 1978. See ITS 5.5 } (2) The results of In-place halogenated a. At least once per operating cycle, but not to exceed 0 0 hydrocarbon tests at 1000 cfm (+/- 10%) shall 18 months; or following painting. fire, or chemical show s 1% penetration on each Individual release while the system Is operating that could I charcoal adsorber and shall show s0.05% penetration on the combined charcoal banks. contaminate the HEPA filters or charcoal adsorbers, perform the following: 0 (3) The results of laboratory carbon sample (1) In-place DOP test the HEPA filter banks. 0
-U I analysis shall show s0.5% methyl Iodide penetration when tested at 30-C and 95% (2) In-place test the charcoal adsorber banks with ;U relative humidity. halogenated hydrocarbon tracer. 0) tD (3) Remove one carbon test sample from each i0 Co charcoal adsorber bank- Subject thi smple to tD CD ;U a laboratory analysis to verify methyl Iodide i 0 removal efficiency.
t) IV 0 la (4) Initiate from the control room 1000 dCn (+/-t 10%) -4' to flow through both trains of the emergency to filtration treatment system. CD 0) 3.17/4.17 229w 8118/00 Amendment No. 65,4 0 -14 0 8 r 112 Page 3 of 5
C C C ITS 3.7.4 0. 3.0 IUMING CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS I b. Acceptance Criteria - System Operation I b. At least once per 720 hours ofstem operathn, Requirements remove one carbon test sample from each charcoal Se 5 t adsorber bank Subject this sample to a laboratory I SI5 0: The results of laboratory carbon sample analysis analysis to verify methyl Iodide removal eficdency. 0) I shall show s0.5% methyl Iodide penetration when 0 tested at 300C and 95% relative humidity. 3 D 0 0 a0 0 3
~~-o 0 ;0 X~~
to Co CD D
~~0) 0)~~
~~-co a) 0 0 -4' C?~~
3.17/4.17 229ww 818(1w0 Amendment No. 408 112 Page 4 of 5
C C C ITS 3.7.4 ITS ITS (See ITS 5.5} 3.0 UMmNG CONDITIONS FOR OPERATiON I 4.l 4.0 SURVEILLANCE REQUIREMENTS
~~c. The system shall be shown to be operable withh ISR 3.7.4.3, c. At least once perropera~lng cycle, but not4o excee~~
= n~nth the following conditions shall be on a STAGGERED (1) Combined filter pressure drop S8 Inches water. SR 3.7.4.4 demonstrated for each emergency flit ion system ITEST BASIS for L, 0:
(2) Inlet heater power output 5kw 10%. 0 2 (1) Pressure drop across the combined filters of S / ~each tran shall be measured at 10W0 cfrnm . (3) Automatic In i ti a t h (+/- 10%) flow rate. See 3 SR 3.7.4.3
~~)"ctEuoar. X ~~(2) Operability of Inlet heater at niominali rated J 0~~
SJ power shall be verified. C (3) Verify that on a'simuIatedritWradion signal, 0 SR 3.7.4.3, the train switches to the pressurization mode of SR 3.7.4.4 operation and the control room Is maintained at 3 a positive pressure with respect to adjacent 0 01 CD areas at the desian flow rate aD D r3 rT7 , 10r Njj
~~-t ;U~~
-U 3. Post Maintenance Requirements n r> IngI ross15mailnenr11celusu T J~oatre-n lI -4 i tD a. After any maintenance or testing that could affect a. After any maintenance or testing that could affect 0 the HEPA filter or HEPA filter mounting frame leak the leak tight integrity of the HEPA filters, perform K1 X tight Integrity, the results of the In-place DOP tests 6 at 1000 cfm (+/- 10%) shall show s I% DOP In-place DOP tests on the HEPA filters.
-u CD penetration on each Individual HEPA filter and shalt b. After any maintenance or testing that could affect I show s 0.05% DOP penetration on the combined the leak tight integrity of the charcoal adsorber to la HEPA filters. -{ See ITS 5.5} la i3 banks. perform halogenated hydrocarbon tests or fJ b. After any maintenance or testing that could affect the charcoal adsorbers. CY) -4 the charcoal adsorber leak tight integrity; the results CD 0
of In-place halogenated hydrocarbon tests at 1000 cfm (+/-10%) shall show s 1% penetration on each 0) Individual charcoal adsorber and shall show a) I 50.05% penetration on the combined charcoal adsorber banks. 3.17/4.17 229x 8/181W0 Amendment No. 65,SO0,4O8r 112 Page 5 of 5
Attachment 1, Volume 12, Rev. 0, Page 68 of 161 DISCUSSION OF CHANGES ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM 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 4.1 7.B.1 states to operate each CREF subsystem for at least 10 hours with the heaters "operable." ITS SR 3.7.4.1 requires the CREF System to operate with the heaters "operating." This changes the CTS by requiring the CREF heaters to be "operating" in lieu of being "operable" during the test. The purpose of CTS 4.17.B.1 is to ensure each CREF subsystem remains OPERABLE to support the safety analyses. This change requires the CREF heaters to be "operating" in lieu of being "operable" during the test. Moisture may accumulate on the high efficiency particulate air filters and the charcoal adsorber when the system is in standby. Performing this Surveillance on a periodic frequency will remove the moisture since it is necessary for the heaters to actually operate to reduce moisture from the adsorbers and HEPA filters. Therefore it is necessary that the heaters are actually operating during the test. This change is consistent with current practice. The heaters are still required to be OPERABLE, as required by ITS SR 3.7.4.2 and ITS 5.5.6, Ventilation Filter Testing Program (VFTP). This change is designated as administrative because it does not result in technical changes to the CTS. A.3 Under certain conditions, CTS 3.17.B.1.c and 3.17.B.1 .d, in part, require the immediate suspension of movement of recently irradiated fuel assemblies in the secondary containment. ITS 3.7.4 ACTIONS D and F include the same requirement, however a Note has been added that states that LCO 3.0.3 is not applicable. This changes the CTS by adding this Note. The purpose of CTS 3.17.B.1.c and 3.17.B.1.d, in part, is to provide the appropriate actions when one or two CREF subsystems, as applicable, are inoperable while moving recently irradiated fuel in the secondary containment. This change adds a Note that states LCO 3.0.3 is not applicable. This Note has been added because ITS LCO 3.0.3 has been added to ITS Section 3.0 in accordance with ITS Section 3.0 DOC M.1. This Note is necessary because if moving recently irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving recently irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of recently irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. Since ITS LCO 3.0.3 is not currently included in the CTS this change is considered administrative. This change is designated as administrative because it does not result in technical changes to the CTS. Monticello Page 1 of 7 Attachment 1, Volume 12, Rev. 0, Page 68 of 161
Attachment 1, Volume 12, Rev. 0, Page 69 of 161 DISCUSSION OF CHANGES ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM A.4 CTS 3/4.17.B.2 specifies the performance requirements for the CREF subsystems while CTS 3/4.17.B.3 specifies the post maintenance requirements for the CREF subsystems. ITS SR 3.7.4.2 requires the performance of the required CREF filter testing in accordance with the Ventilation Filter Testing Program (VFTP). CTS 314.17.B does not include a VFTP, but the requirements that make up the program are being moved to ITS 5.5. This changes the CTS by requiring testing in accordance with the VFTP, whose requirements are being moved to ITS 5.5. This change is acceptable because filter testing requirements are being moved to the VFTP as part of ITS 5.5 and ITS SR 3.7.4.2 references the VFTP for performing these tests. This change is.designated as administrative because it does not result in technical changes to the CTS. A.5 CTS 4.17.B.2.c requires verification of the OPERABILITY of each CREF subsystem each "operating cycle." ITS SR 3.7.4.3 and ITS SR 3.7.4.4 require the same testing however the Surveillances are required to be performed every "24 months." This changes the CTS by changing the Frequency from "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 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. 4.17.B.2.c was included in this evaluation. This change is designated as administrative because it does not result in any technical changes to the CTS. A.6 These changes to CTS 3.17.B.1, CTS 3.17.B.1.c, and CTS 3.17.B.1.d are provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-04-023, from Thomas J. Palmisano (NMC) to USNRC, dated April 29, 2004. As such, these changes are administrative. A.7 This change to CTS 4.17.B.2.c is 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, this change is administrative. MORE RESTRICTIVE CHANGES M.1 CTS 3.17.B.1, is applicable, in part, whenever irradiated fuel is in the reactor vessel and reactor water temperature is greater than 2120 F. ITS LCO 3.7.4 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring the CREF System to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 212 0 F. Monticello Page 2 of 7 Attachment 1, Volume 12, Rev. 0, Page 69 of 161
Attachment 1, Volume 12, Rev. 0, Page 70 of 161 DISCUSSION OF CHANGES ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM The purpose of CTS 3.17.B.1 is to ensure the CREF System is OPERABLE to mitigate the consequences of a design basis accident. The CREF System is required to be OPERABLE during MODES 1, 2, and 3 when there is considerable energy in the reactor core or there is a potential for the reactor core to become critical and the CREF System would be required to mitigate the consequences of a design basis accident. 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. Therefore, it is necessary and acceptable to require the CREF System to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 CTS 3.17.B.1.a allows 7 days to restore an inoperable CREF subsystem. If this cannot be met, it requires the unit to be in hot shutdown (i.e., MODE 3) in 12 hours and to be below 2120 F (i.e., be in MODE 4) or initiate and maintain the OPERABLE CREF subsystem in the pressurization mode within the following 24 hours. ITS 3.7.4 ACTION C does not include the option to place the OPERABLE CREF subsystem in operation in lieu of being in MODE 4. This changes the CTS by deleting the allowance'to place the OPERABLE CREF subsystem in operation in lieu of achieving MODE 4 conditions. The purpose of the 3.17.B.1.a, in part, is to place the unit in a MODE in which the LCO is no longer applicable in a reasonable amount of time. ITS 3.7.4 ACTION C does not include the option to place the OPERABLE CREF subsystem in operation. ITS 3.7.4 ACTION C requires the unit to be in MODE 3 in 12 hours and MODE 4 in 36 hours. This change deletes the allowance to place the OPERABLE CREF subsystem in operation and continue to operate in MODE 3. CTS 3.17.B.1 allows the unit to operate in MODE 3 conditions indefinitely with an inoperable CREF subsystem as long as the OPERABLE CREF subsystem is operating. With the system operating, automatic functions of the systems may not be required since the system is operating, however, there is a potential that an operating component can still fail if required (for example, power is not available). In MODE 3 conditions, the potential for a design basis accident still exists, therefore the allowance has been deleted since the failure of the operating subsystem may result in a failure to satisfy the requirements of the system. This change is designated as more restrictive since the option to start the OPERABLE CREF subsystem has been deleted. M.3 When both CREF subsystems are inoperable, CTS 3.17.B.1.b allows 24 hours to restore an inoperable CREF subsystem to OPERABLE status prior to initiating a reactor shutdown. When both CREF subsystems are inoperable due to an inoperable control room boundary, ITS 3.7.4 ACTION B allows 24 hours to restore the control room boundary to OPERABLE status. When both CREF subsystems are inoperable for reasons other than an inoperable control room boundary, ITS 3.7.4 ACTION E requires immediate entry into LCO 3.0.3. This will require the unit to initiate action within 1 hour to place the unit in MODE 2 within 7 hours, MODE 3 within 13 hours, and MODE 4 within 37 hours. This changes the CTS by requiring an immediate entry into LCO 3.0.3 when two Monticello Page 3 of 7 Attachment 1, Volume 12, Rev. 0, Page 70 of 161
Attachment 1, Volume 12, Rev. 0, Page 71 of 161 DISCUSSION OF CHANGES ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM CREF subsystems are inoperable for any reason other than the control room boundary being inoperable. The purpose of the CtS 3.17.B.1.b is to provide appropriate compensatory action when both CREF subsystems are inoperable. This change requires immediate entry into LCO 3.0.3 when two CREF subsystems are inoperable for any reason other than the control room boundary being inoperable. Under the same conditions in the CTS, 24 hours is allowed to restore one CREF subsystem prior to initiating a reactor shutdown. This change is acceptable because it does not allow the unit to operate for 24 hours with two inoperable CREF subsystems in MODES 1, 2, and 3. This is necessary because if there is an accident, the CREF subsystem will not be able to pressurize the control room boundary and ensure the control room dose remains within the safety analyses dose calculations. This change is designated as more restrictive because with two CREF subsystems inoperable in MODES 1, 2, and 3 the reactor will have to be shutdown sooner in the ITS than is currently allowed in the CTS. RELOCATED SPECIFICATIONS None -REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.17.B.1 states to "initiate from the control room" flow through CREF subsystem and operate for at least 10 hours. ITS SR 3.7.4.1 includes the same requirement, however, the statement to "initiate from the control room" is not included. This changes the CTS by moving the requirement to "initiate from the control room" from the CTS to the ITS Bases. 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 the requirement to operate the CREF subsystem for > 10 hours in ITS SR 3.7.4.1. 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. LA.2 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.17.B.2.c.(3) and CTS 4.17.B.2.c.(3) require each CREF subsystem be shown to be OPERABLE with automatic initiation upon receipt of a "high radiation" signal. ITS SR 3.7.4.3 requires verification that each CREF subsystem actuates on an initiation signal. This changes the CTS by moving the specific type of actuation signal to the ITS Bases. Monticello Page 4 of 7 Attachment 1, Volume 12, Rev. 0, Page 71 of 161
Attachment 1, Volume 12, Rev. 0, Page 72 of 161 DISCUSSION OF CHANGES ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM The removal'of these details, which are 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. The ITS still retains the requirement to verify each CREF subsystem actuates on an initiation signal. 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. LESS RESTRICTIVE CHANGES L.i (Category 1 - Relaxation of LCO Requirement) CTS 3.17.B.1 requires two CREF subsystems to be OPERABLE, but does not allow the main control room boundary to be opened intermittently under administrative controls. If it is opened, both CREF subsystems are inoperable. ITS LCO 3.7.4 also requires the two CREF subsystems to be OPERABLE, however a Note to the LCO is included that allows the main control room boundary to be opened intermittently under administrative controls. This changes the CTS by allowing the main control room boundary to be opened intermittently under administrative controls and not consider the CREF System to be inoperable. The purpose of CTS 3.17.B.1 is to ensure the CREF System remains OPERABLE to support the safety analyses. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses. The LCO is modified by a Note allowing the main control room boundary to be opened intermittently under administrative controls. The control room boundary is often opened intermittently to allow entry and exit and also periodically opened to perform maintenance. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the main control room. This individual will have a method to rapidly close the opening when a need for main control room isolation is indicated. This change is acceptable because the ITS Bases requires these administrative controls to be met in order to utilize the proposed LCO Note. 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 (Category6 - Relaxation Of Surveillance Requirement Acceptance Criteria) CTS 4.17.B.1 requires that each CREF subsystem be initiated with 1000 cfm (+/- 10%) of flow and operated for 10 hours. ITS SR 3.7.4.1 includes a similar requirement, except the flow rate is not specified. This changes the CTS by deleting the flow rate requirement from the Surveillance acceptance criteria. The purpose of 4.17.B.1 is to ensure the each CREF subsystem remains OPERABLE to support the safety analyses. This change is acceptable because it Monticello Page 5 of 7 Attachment 1, Volume 12, Rev. 0, Page 72 of 161
Attachment 1, Volume 12, Rev. 0, Page 73 of 161 DISCUSSION OF CHANGES ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM has been determined that the relaxed Surveillance Requirement acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its required functions. This change deletes the flow rate requirement from the Surveillance acceptance criteria. Moisture may accumulate on the high efficiency particulate air filters and the charcoal adsorber when the system is in standby. Performing this Surveillance on a periodic frequency will remove the moisture since it is necessary for the heaters to actually operate (cycle properly when required) to reduce moisture from the adsorbers and HEPA filters. The actual flow rate is not specifically required to remove the moisture. Therefore, the explicit flow rate requirement has been deleted. Each CREF subsystem is still required to be tested at the specified flow rate as required by ITS SR 3.7.4.2 and ITS 5.5.6. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. L.3 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria) CTS 3.17.B.2.c.(3) and 4.17.B.2.c.(3) require verification of the automatic actuation of each CREF subsystem upon a receipt of the specified inputs (i.e., simulated signal). ITS SR 3.7.4.3 specifies that the signal may be from either an "actual" or "simulated" initiation signal. This changes the CTS by explicitly allowing the use of an actual signal for the test. The purpose of CTS 3.17.B.2.c.(3) and 4.17.B.2.c.(3) is to ensure that each CREF subsystem operates correctly upon receipt of an actuation signal. This change is acceptable because it has been determined that the relaxed Surveillance Requirement acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its required functions. Equipment cannot discriminate between an "actual" or "simulated" signal and, therefore, the results of the testing are unaffected by the type of signal used to initiate the test. This change allows taking credit for unplanned actuation if sufficient information is collected to satisfy the Surveillance test requirements. The change also allows a simulated signal to be used, if necessary. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. L.4 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.17.B.2.c.(3) requires each CREF subsystem to be operating in the pressurization mode of operation while maintaining the control room at a positive pressure with respect to adjacent areas, at least once per operating cycle. ITS SR 3.7.4.4 requires this same test, however it is required to be performed every 24 months "on a STAGGERED TEST BASIS." This changes the CTS by requiring the test to be performed using each CREF subsystem at least once per 48 months. The purpose of the CTS 4.17.B.2.c.(3), in part, is to ensure the integrity of the control room boundary. This change Is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. The change is acceptable since the proposed Surveillance Frequency will continue to require performance of the test every 24 months. This will ensure the control room boundary integrity is maintained. The status of the integrity of the control room boundary can be determined with either CREF Monticello Page 6 of 7 Attachment 1, Volume 12, Rev. 0, Page 73 of 161
Attachment 1, Volume 12, Rev. 0, Page 74 of 161 DISCUSSION OF CHANGES ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM subsystem. The secondary purpose of this test will now also serve to ensure that each CREF subsystem is being tested every 48 months (based on the definition of STAGGERED TEST BASIS). ITS SR 3.7.4.3 requires the performance of a test to ensure each CREF subsystem actuates on an actual or simulated initiation signal. Therefore, each subsystem will continue to be tested to ensure it can be automatically aligned to the pressurization mode, however the pressurization verification will only be required with one subsystem in operation. This change is designated as less restrictive because the pressurization Surveillance will only be required to be performed on one CREF subsystem each Surveillance interval instead of on both CREF subsystems. L.5 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria) CTS 4.17.B.2.c.(3) requires each CREF subsystem to be operating in the pressurization mode of operation while maintaining the control room at a positive pressure with respect to adjacent areas with a flow rate of 1000 cfm (+/- 10%). ITS SR 3.7.4.4 requires this same test, however the flow rate limit is specified to be < 1100 cfm. This changes the CTS by deleting the minimum flow rate limit. The purpose of the CTS 4.17.B.2.c.(3), in part, is to ensure the integrity of the control room boundary. This change is acceptable because it has been determined that the relaxed Surveillance Requirement acceptance criteria are not necessary for verification that the equipment used to meet the LCO can perform its required functions. The CREF flow rate limit range has been changed to a maximum flow rate limit. The CTS requires the CREF flow rate to be from 900 cfm to 1100 cfm during the control room boundary integrity test. In the ITS, the same test may be performed at < 1100 cfm. This change is acceptable because if the control room boundary is positive with a lower flow rate, the control room boundary is intact. This change is designated as less restrictive because less stringent Surveillance Requirements are being applied in the ITS than were applied in the CTS. Monticello Page 7 of 7 Attachment 1, Volume 12, Rev. 0, Page 74 of 161
Attachment 1, Volume 12, Rev. 0, Page 75 of 161 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 75 of 161
Attachment 1, Volume 12, Rev. 0, Page 76 of 161 F[M>EC]l System 3.7.4 (0 CTS _EmIoergenc 3.7 PLANT SYSTEMS (CREF)_ 3.17.9 3.7.4 [ in Control Room Environmen C rol (MCREC)] System 0D 3.17.8.1 LCO 3.7.4 TwoM t subsystems shall be OPERABLE. 0 K1rTI> J I ----- ________ DOC LI The main control room boundary may be opened intermittently under administrative control. 3.173..1 APPLICABILITY: MODES 1, 2, and 3, During movement of Mrecentlt irradiated fuel assemblies in the tsecondaryM containment, During operations with a potential for draining the reactor vessel
~~} 0D (OPDRVs).~~
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.17.B. 1 a, 3.17.B.1.c A. O nejMEC subsystem inoperable. A.1 Restore subsystem to OPERABLE 7 days 0D status. 3.17.B.1.b B. Two
,--FRF-I L B.1 Restore control room 24 hours 0D subsystems inoperable boundary to OPERABLE due to inoperable control status.
~~room boundary in MODE 1, 2, or 3.~~
-9 3.17.B.1.a, C. Required Action and C.1 Be in MODE 3. 12 hours 3.17.B.1.b associated Completion Time of Condition A or B AND not met in MODE 1, 2, or3. C.2 Be in MODE 4. 36 hours
____ ___ ____ ___ ___I BWR/4 STS 3.7.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 76 of 161
Attachment 1, Volume 12, Rev. 0, Page 77 of 161 F[M-a 6CI System 3.7.4 (0 \
~~- ACTIONS (continued)~~
CONDITION REQUIRED ACTION COMPLETION TIME 3.17.B.1.c D. Required Action and CO0.i---NOT E.--- associated Completion LCO 3.0.3 is not applicable. Time of Condition A not _ __ _ _ _ _ _ _ met during movement of ecentlyM irradiated fuel D.1 - - -NOTE-- -- assemblies in the [Place in toxic gas rsecondaryl containment protec on mode if or during OPDRVs. autorm tic transfer to oxic gas p tection mode s inope ble. ] Place OPERABLE E Immediately [IN isystem in TpressurizationM mode. 0 OR D.2.1 Suspend movement of Immediately MrecentiA irradiated fuel assemblies in the 0 Secondaryj containment. AND D.2.2 Initiate action to suspend Immediately OPDRVs. 3A7.B.1.b E. Two subsystems inoperable E.1 Enter LCO 3.0.3. Immediately 0 in MODE 1, 2, or 3 for reasons other than Condition B. BWR/4 STS 3.7.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 77 of 161
Attachment 1, Volume 12, Rev. 0, Page 78 of 161 l[`M 4
~~-E3C System 0~~
IS- ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.17.B.l.d F. Twoj[M(;;RECIl 0 NOTE--. subsystems inoperable LCO 3.0.3 is not applicable. during movement of
~~.- ecentlyE irradiated fuel assemblies in the~~
[becondaryl containment F.1 Suspend movement of [recentlyj irradiated fuel Immediately 0 (D or during OPDRVs. assemblies in the IsecondaryT containment. AND F.2 Initiate action to suspend Immediately OPDRVs. SURVEILLANCE REQUIREMENTS SURVEILLANCE l FREQUENCY 4.17.9.1 SR 3.7.4.1 Operate each C] subsystem for 10 continuous hours wit the heaters operating or or systems witho res) 15 minutes . 31 days
~~} (0 4.17.B.2.~~
~~4.17.B.3 SR 3.7.4.2 Perform required tMC!ECf filter testing in accordance with the Ventilation Filter Testing Program (VFTPA. In accordance with the [VFTPI _Is24~~
~~} 0D 3.17.B.2.c.(3),~~
4.17.B.2.c, 4.17.B.2.c.(3) SR 3.7.4.3 Verify each [M C jsubsystem actuates on an actual or simulated initiation signal. 0 4.17.B.2.c, SR 3.7.4.4 f Verify eachl[5 Jj subsystem can maintain a Mfm onth's on a 4.17.B.2.c.(3) positive pressured of 2 [0.11 jC ter gau STAGGERED 0 relative tolthe [tu ul ing during the TEST BASIS I Mpressurizatiorn modeof operation at a flow rate of SL40 c m . BWR/4 STS 3.7.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 78 of 161
Attachment 1, Volume 12, Rev. 0, Page 79 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM
1. The brackets have been removed and the proper plant specific information/value has been provided.

2. Required Action D.1 is modified by a Note that states to place the CREF System in the toxic gas protection mode if automatic transfer to toxic gas protection modes is inoperable. The Monticello CREF design does not include an automatic transfer to the toxic gas protection mode. Therefore, the Note is deleted.

3. ISTS SR 3.7.4.4 specifies a bracketed positive pressure criterion of 0.1 inches water gauge relative to the turbine building. ITS SR 3.7.4.4 only requires maintaining a positive pressure relative to adjacent areas. This difference was accepted by the NRC in a letter dated May 30,1989 from John F. Stepano (NRC) to Mr. Musolf (NSPC) and discussed in Section 2, page 7, of the associated Safety Evaluation.
Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 79 of 161
Attachment 1, Volume 12, Rev. 0, Page 80 of 161 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 80 of 161
Attnr-hmant '1 vJntua 12 fRav, fn Pqano i nf iRi
* ~~System B 3.7.4 0
B 3.7 PLANT SYSTEMS /1 3 B 3.7.4 Control Rooml Environmeaentrol (MCREC System ( BASES ,CE BACKGROUND The [M ECJ System provides a radiologically controlled environment from w ic I e unit can be safely operated following a Design Basis Accident (DBA). F CREF l The safety related function oo System includes two independent () I and redundant high efficiency air i lra ion subsystems for emergency fIlter treatment o outside supply air. Each subsystem consists of [an electric heater a r i te a high efficiency M. lanemergencyfilter [anexhausti _ _ _.articulate air (HEPA) filter.5R activated cnarcoa- adsorber section, a second HEPA filte=x fan, an air handling unit (excluding the condensing unit)+and the associated ductwork and dampers. I l o Iremove water ets-tM the alrstreag.r P ters and HEPA filters Low:efficiencyfilters remove paricu ate mater, Whic may e radioactive. The charcoal adsorbers provide a holdup period for gaseous iodine, allowing time for decay. I ary 2 The [IfEC System is a standby system, parts of which also operate 0 aControlRoomAirnet dur n lunit operations to maintain the control room environment. Radiation-iHigh piinitiation signagJ (indicative of conditions that could result in radiation exposure to control room personnel), th[( boundary(the main System automatically switches to the pressurization mode of oeration to portions oftanefirt INSERT 1 revent infiltration of contaminatevd air into the control roomM A system of and secondfoors of dampers isolates the control rooRyandas part Uzhe-rerlrculaeai Sl Filtration Train (EFT) Irouted through ei erigifteo filter subsystems. Outside air is taken in at the normal ventilation intake and isImixed wit ircu a e air Ibefobeingm passed through one of the charcoal adsorber filter subsystems for removal of airborne radioactive particles The System is designed to maintain the control room boundary environment or a 30 day continuous occupancy after a DBA without7 exceeding 5 rem whole body dose or its equivalent to any part of the, S ubZZZE Asyingl EC subsystem will pressurize the control room (1) (0) Iou . aer au to prevent infiltration of air from i_ System operation in maintaining control (j) room habitability is discussed in theSAR, ahaters and.,M respectively). B 3314.7.6.3.2, STect3ons4 o1 14.7.2 (R 0 efs 51 (i 0 Li4.7.32.4, and i... ! BWR/4 STS B 3.7.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 81 of 161
Attachment 1, Volume 12, Rev. 0, Page 82 of 161 B 3.7.4 Q boundary from untreated outside air. INSERT I O INSERT 2 This air isthen combined with return air from the control room boundary and passed through an exhaust/recirculation fan, which is then passed through the air handling unit into the control room boundary. Insert Page B 3.7.4-1 Attachment 1, Volume 12, Rev. 0, Page 82 of 161
Attachment 1, Volume 12, Rev. 0, Page 83 of 161 System Q B 3.7.4 BASES 2 CR3 0-APPLICABLE (D The ability of the [M(!ZEC] System to maintain the habitability of the SAFETY boundary control room is an explicit assumption for the safety analyses presented ANALYSES inathnSA aer an (Refs.n resp-cively). The (i) pesurization mode of thej[M9RE jl;ystem Isassum-edto olperated
/following a loss of coolant accidnfe hadig cietnvlngc Sections 14.7.24.3, handling recently irradiated fuel (i.e., fuel that has occupied part ofa24o and 14.7.1.6, critical reactor core within the previousiq2!ggV,, main steam line break 1i .and con trol rod drop accident, as discussed in theMMAR-t 23an(E 56Sectioro6.4.2.2l(Re.4d c pesnnla a resutOfnBn-fb~rS-tq are suffarffe IRM -res~petivey [{
~~Referince 3] No single active or sive failure will Sq o ofC butside or recircullated~tirlr= the cnrl°°peethtCREF System m]~~
~~. X 1 lpRrEoT ng Its safteyfunctin The[ECg1System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).~~
r 11 -- LCO Two redundant subsystems of the [M System are required to be ( OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a DBA. The [MCE]System is considered OPERABLE when the individual 0 components necessary to control operator exposure are OPERABLE in both subsystems. A subsystem is considered OPERABLE when its associated: , exaus~ecirulaon fan, and air handigui cond*enser Emerg/encdilgther
a. y Fan OPERABLEC, unit) are 0 0

b. HEPA filtand charcoal adsorbers are not excessively restricting flow and are capable of performing their filtration functions . D

c. Heater, ductwork, va es, and dampers are OPERABLE, and air circulation can be maintained.
Inaddition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors. The LCO is modified by a Note allowing the main control room boundary to be opened intermittently under administrative controls. For entry and exit through doors, the administrative control of the opening is performed by the person(s) entering or exiting the area. For other openings, these controls consist of stationing a dedicated individual at the opening who is in continuous communication with the main control room. This individual will have a method to rapidly close the opening when a need for main control roorrisolation is indicated. 0 BWR/4 STS B 3.7.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 83 of 161
Attachment 1, Volume 12, Rev. 0, Page 84 of 161 F[M-EC1lSystem B 3.7.4 0D BASES l CREFFl APPLICABILITY In MODES 1, 2, and the [M C] System must be OPERABLE to control operator exposure uring an following a DBA, since the DBA 0D could lead to a fission product release. In MODES 4 and 5, the probability and consequences of a DBA are CREF }reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaiin i[System OPERABLE is not required in MODE 4 or 5, except for the following situations under which 0D significant radioactive releases can be postulated:
a. During operations with potential for draining the reactor vessel (OPDRVs) and rEn 0

b. During movement of jrecentlAj irradiated fuel assemblies in the IsecondarA containment. aDue to radioactive decay, the mE (24 }
System is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied 0D LhouJ part of a critical reactor core within the previousl [X ay ).] ACTIONS A.1 With one Is ystem inoperable, the inoperable 011( subsystem Mus e restored to OPERABLE status within 7 days. With the unit in this condition, the remaining OPERABLffI[MQREC]I subsystem is adequate to perform control room radiation protection. However, the 0D CREF overall reliability is reduced because a single failure in the OPERABLE subsystem could result in reducedj[MRECJ System capability. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period, and that the remaining subsystem can provide the required capabilities. B.1 _____- -- REVIEWER'S NOTE---_ Adoption of Condi on B is dependent on a commit ent from the licensee to have written p cedures available describing co pensatory measures to be taken in th event of an intentional or unintee tional entry into
~~-oD Condition B.~~
I
~~/ l BWR14 STS B 3.7.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 84 of 161~~
Attachment 1, Volume 12, Rev. 0, Page 85 of 161 M _[ t B 3.7.4 BASES ACTIONS (continued) If the main control room boundary is inoperable in MODE 1, 2, or 3, the subsystems cannot perform their intended functions. Actions must be taken to restore an OPERABLE main control room boundary within 24 hours. During the period that the main control room boundary is 10 CFR 50. inoperable, appropriate compensatory measures (consistent with the AnintoGDC 19) should be utilized to protect control room operators 0 from potential hazards such as radioactive contamination, toxic chemicals, smoke, temperature and relative humidity, and physical security. Preplanned measures should be available to address these concerns for intentional and unintentional entry into the condition. The 24 hour Completion Time is reasonable based on the low probability of a DBA occurring during this time period, and the use of compensatory measures. The 24 hour Completion Time is a typically reasonable time to diagnose, plan and possibly repair, and test most problems with the main control room boundary. C.i and C.2 In MODE 1, 2, or 3, if the inoperables ystem or control room ( boundary cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems. D.1. D.2.11and 0.2.2 The Required Actions of Condition D are modified by a Note indicating that LCO 3.0.3 does not apply. If movingarecentlyi irradiated fuel ( 3 assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of mrecentlyM irradiated fuel assemblies is not sufficient reason to require a reactor shutdown. BWR/4 STS B 3.7.4-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 85 of 161
Attachment 1, Volume 12, Rev. 0, Page 86 of 161 3Estemn B 3.7.4 BASES ACTIONS (continued) During movement offrecentl irradiated fuel assemblies in the y'E [secondaryI containment or during OPDRVs, if the inoperable EC (7' CREF subsystem cannot be restored to OPERABLE status within the required Completion Time, the OPERABLEI[MCREC]Isubsystem may be placed in the pressurization mode. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent automatic actuation will occur, and that any active failure will be readily detected. Required Action p1 is modified by a Note alertingae operator to [place the system in tp4 toxic gas protection mode if thtoxic gas automatic ( transfer cap;1ility is inoperable]. An alternative to Required Action D.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control roor,. This places the unit in a condition that minimizes risk. If applicable, movement ofarecentlA irradiated fuel assemblies in the T( IsecondarA containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. E.1 If both C] su systems are inoperable in MODE 1, 2, or 3 for CREF lreasons o n an inoperable control room boundary (i.e., Condition B C]j System may not be capable of performing the intended function and te unit is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately. F.1 and F.2 The Required Actions of Condition F are modified by a Note indicating that LCO 3.0.3 does not apply. If movingirecentli irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of LrecentlA irradiated fuel assemblies is not sufficient reason to require a reactor shutdown. BWR/4 STS B 3.7.4-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 86 of 161
Attachment 1, Volume 12, Rev. 0, Page 87 of 161 [M ECB S7stem. B 3.7.4 0D BASES ACTIONS (continued) During movement of jrecentlIj irradiated fuel assemblies in the jsecondarA containment or during OPDRVs, with twol[MgE ] yCRF J 0 subsystems inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control roon,. This places the unit in a condition that minimizes risk. 1 0 If applicable, movement of VrecentlA irradiated fuel assemblies in the TsecondarA containment must be suspended immediately. Suspension olf these activities shall not preclude completion of movement of a component to a safe position. If applicable, actions must be initiated immediately to suspend OPDVRs-to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. SURVEILLANCE SR 3.7.4.1 REQUIREMENTS (ii 'This SR verifies that a subsystem in a standby mode starts on demand control[ Lr...n and continues to operate. Standby systems should be checked periodically to ensure that they start and function properly. As the 0D environmental and normal operating conditions of this system are not severe, testing each subsystem once every month provides an adequate check on this system. Monthly heater operation dries out any moisture that has accumulated in the charcoal as a result of humidity in the ambient air.. jSystems with heaters must be operated for 2 10 continuous hours with the heaters energized.jSystems without eaters need only b'-' operated for 2 15 mo utes to demonstrate the functin of the system.1 / 0D Furthermore, the 31 day Frequency is based on the known reliability of ) the equipment and the two subsystem redundancy available. SR 3.7.4.2 This SR verifies that the requiredl[5: itesting is performed in accordance with the EVentilation Filter Testing Program (VFTPl. The EVFTPM includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations). 0 Specific test frequencies and additional information are discussed in detail in the MVFTFS BWR/4 STS B 3.7.4-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 87 of 161
Attachment 1, Volume 12, Rev. 0, Page 88 of 161 l[MC System B 3.7.4 0D BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.7.4.3 l cF This SR verifies that on an actual or simulated initiation signal, each LCO3.3.7.1, ' subsystem starts and operates. The LOGIC SYSTEM Control Room FUNCTIONAL TEST Ig]SR aa7.f. overlaps this SR to provide complete ( 3 Filtration (CREF) testing of the safety function. rhe 11 month cy is specified in5 Instrumentaton,- I NSERT 3 j SR 3.7.4.4 0D This SR verifies the integrity of the control room the assumed inleakape rates of potentially contaminated air. The control I boundary e roorrlFpositive pressure, with respect to potentially contaminated adjacent CREF areas lte tur uiin], is periodically tested to verify proper function
~~-IZIIllE QSystem. During the emergency mode of operation, the designed to slightly pressurize the control roomy_~~
d t12 .1Iinches watereaugg-positive pressurelwith respect tobeo nabine aesito prevent unfiltered inleakage. The [M Cjst E Ft designed to maintain this positive pressure at a flow rate of s g,9lcfmto ZJ ? 2(boundary the control room in the pressurization mode. The Frequency o mjoljn months on a STAGGERED TEST BASIS is consistent with industry Add practice and other filtration systems SRs.j J K REFERENCES TS-~F SAR, Cha er 0D0D
2. SAR, jChapfer 9 I Section14.7.6.3.2 ] 0 0

3. fSAR, hat [1 5 Scin4.7.32. 0 0D EN+SAR, Section [6.4.2.2 0 0

15. Regulat~q1-Gtde1.52, Rev. [21.1 0 BWR/4 STS B 3.7.4-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 88 of 161
Attachment 1, Volume 12, Rev. 0, Page 89 of 161 B 3.7.4 INSERT 3 Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint. Insert Page B 3.7.4-7 Attachment 1, Volume 12, Rev. 0, Page 89 of 161
Attachment 1, Volume 12, Rev. 0, Page 90 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.4 BASES, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM
1. The brackets have been removed and the proper plant specific information/value has been provided.

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. The Reviewer's Note is deleted as it is not part of the plant-specific ITS.

4. Changes made to be consistent with changes made to the Specification.

5. These changes have been made for consistency with similar phrases in other parts of the Bases and/or to be consistent with the Specification.

6. 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.~~
7. ISTS SR 3.7.4.3 verifies that each [MCREC] subsystem (changed to CREF subsystem in ITS SR 3.7.4.3) actuates on an actual or simulated initiation signal every 18 months. The justification for the 18 month Frequency is that it is specified in Regulatory Guide 1.52. Regulatory Guide 1.52 addresses filtration requirements.
The Surveillance verifies mechanical requirements and the Bases have been modified to correctly state the basis of the Frequency. Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 90 of 161
Attachment 1, Volume 12, Rev. 0, Page 91 of 161 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 12, Rev. 0, Page 91 of 161
Attachment 1, Volume 12, Rev. 0, Page 92 of 161 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.4, CONTROL ROOM EMERGENCY FILTRATION (CREF) SYSTEM There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 92 of 161
Attachment 1, Volume 12, Rev. 0, Page 93 of 161 ATTACHMENT 5 ITS 3.7.5, Control Room Ventilation System Attachment 1, Volume 12, Rev. 0, Page 93 of 161
Attachment 1, Volume 12, Rev. 0, Page 94 of 161 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 12, Rev. 0, Page 94 of 161
C C C ITS 3.7.5 ITS 0 3.0 LIMING CONDIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS
.17 CONTROL ROOMABTAUILITY A 4.17 CONTROL ROO A~plicatift~r. Applicabilitr 0 0i W
Applies to the ntrol room ventilation system equlpmqt Applies to the periodic sting requirements of systems W C, necessary toIntsi habiabilty, required to maintain ntrol room habitability. CD 0
~~,as:~vl Oblectives:~~
~~0 To assur the control room Is habitable both und normal and acci ent conditions. To verify the oper bilty of equipment related to control room~~
~~-A habitabirdy. 0~~
-A ED 3.7.5 A. Control Room Ventilation System 0 ED A. Control Room Ventilation System
1. Except as specified In 3.17.A.2 and 3.17.A.3 below, both LCO 3.7.5 l trains of the control room ventilation system shall be operable ever irraialed uel is in the reactor 1. 0(ceper4l2P rscheck rorolom I-0 r IA

a d reaor coolant tem Ad p os Sk .emperat -T
/
CD 21FEor during movement of Irradiated fuel assemblies 0 in the secondary containment core3lteration's or Applicability Ivitles having the potential for draining the reactor vessel. e_ w 0CD ACTION A 2.a With one control room ventilation train Inoperable, U' restore the Inoperable train to operable status within 30 to 0 days. to CD D En ACTION B 2.b If 2.a Is not met, then be In hot shutdown within the next Cn 12 hours following the 30 days and In cold shutdown II LCO 3.0.3 Is not applicable 0o within 24 hours following the 12 hours. . 0 I - aL ACTION C 2.c If 2.a Is not met during movement of irradiated fuel -L assembliesInthe scondary conta men llalt onslor actvit1eshaving the potential for draining the reactor vessel then immediately place the operable control room ventilation train in operation or Immediately suspend these activities. 3.17/4.17 229u 1Z24/98 Amendment No. 6 889,104 Page 1 of 2
C, CC C ITS 3.7.5 0 ITS 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS ACTION EI 3.a With both control room ventilation tratns Inoperable. [reslore at leassrone train to operrble slalids within 241 a)
0) 0 ACTION E 3.b iso 113.
n ot t htn the next t2 our atn cid hutownwitin 4 hours followingl lt lWnh bcnh ventilation onblrom :-.I ACTION E 3.c 3 mot pturing movenent ofiradinted tuel is not (D nssemb~es In the secondary conlainrnn 0 0 ktE~i r atiite having the poternial for drainin the reactor vessel then Immnediately s;uspenct these F', activities. 0 t, D. Control Room Emergency Filtration System B. Control Room Emergency Fltration System 0
1. Except as specified in 3.17.13. 1. through d below. Iwo 1. At least once per month. Initiate from the control room CD control room emergency filtration system filter trains 1000 ctm (+/- 10%) flow through both trains of the shall be operable whenever irradiated fuel is In the emergency filtrrtion treatment system. The system shall -o 0 operate for at least 10 hours with the heaters operable. - fSee ITS 3.7.4}
reactor vessel and reactor coolant temperature Is 0 I greater than 212F or during movement of Irradiated 0 (D fuel assemblies In the secondary containment, core CD alteratbons or acdives having the potential for drahning 0) to the reactor vessel. tD 0 0-4 -16 a, 3.1714.17 229v 8/18/00 5 8 Amendment No. 6 rt 9t 112 Page 2 of 2
Attachment 1, Volume 12, Rev. 0, Page 97 of 161 DISCUSSION OF CHANGES ITS 3.7.5, CONTROL ROOM VENTILATION SYSTEM 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 Under certain conditions, CTS 3.17.A.2.c and CTS 3.17.A.3.c, in part, require immediate suspension of movement of irradiated fuel assemblies in the secondary containment. ITS 3.7.5 ACTIONS C and E include the same requirement, however a Note has been added that states that LCO 3.0.3 is not applicable. This changes the CTS by adding this Note. The purpose of CTS 3.17.A.2.a and CTS 3.17.A.3.c, in part, is to provide the appropriate actions when one or two control room ventilation subsystems, as applicable, are inoperable while moving irradiated fuel in the secondary containment. This change adds a Note that states LCO 3.0.3 is not applicable. This Note has been added because ITS LCO 3.0.3 has been added to ITS Section 3.0 in accordance with ITS Section 3.0 DOC M.1. This Note is necessary because if moving irradiated fuel assemblies while in MODE 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement Is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. Since ITS LCO 3.0.3 is not currently included in the CTS this change is considered administrative. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 These changes to CTS 3.17.A.1, CTS 3.17.A.2.c, and CTS 3.17.A.3.c are provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the NRC for approval in NMC letter L-MT-04-023, from Thomas J. Palmisano (NMC) to USNRC, dated April 29, 2004. As such, these changes are administrative. MORE RESTRICTIVE CHANGES M.1 CTS 3.17.A.1 is applicable, in part, whenever irradiated fuel is in the reactor vessel and reactor water temperature is greater than 21 20 F. ITS LCO 3.7.5 is applicable in MODES 1, 2, and 3. This changes the CTS by requiring the Control Room Ventilation System to be OPERABLE in MODE 2 when reactor water temperature is less than or equal to 212 0F. The purpose of CTS 3.17.A.1 is to ensure the Control Room Ventilation System is OPERABLE to mitigate the consequences of a design basis accident. The Control Room Ventilation System is required to be OPERABLE during MODES 1, Monticello Page 1 of 3 Attachment 1, Volume 12, Rev. 0, Page 97 of 161
Attachment 1, Volume 12, Rev. 0, Page 98 of 161 DISCUSSION OF CHANGES ITS 3.7.5, CONTROL ROOM VENTILATION SYSTEM 2, and 3 when there is considerable energy in the reactor core or there is the potential for the reactor core to become critical and the Control Room Ventilation System would be required to mitigate the consequences of a design basis event. In MODES 1 and 3, the reactor coolant temperature will always be above 212 0 F. 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. Therefore, it is necessary and acceptable to require the Control Room Ventilation System to be OPERABLE. This change is designated as more restrictive because the LCO will be applicable under more reactor operating conditions than in the CTS. M.2 CTS 4.17.A does not provide a requirement to verify the capability of the Control Room Ventilation System to remove the assumed heat load. ITS 3.7.5 includes a Surveillance Requirement to cover this requirement. ITS SR 3.7.5.1 requires verification that each control room ventilation subsystem has the capability to remove the assumed heat load every 24 months. This changes the CTS by adding an additional OPERABILITY requirement for the Control Room Ventilation System. The purpose of the new Surveillance Requirement is to ensure that the heat removal capability of the system is sufficient to remove the control room boundary heat load assumed in the safety analyses. The SR consists of a combination of testing and calculation. The 24 month Frequency is appropriate since significant degradation of the Control Room Ventilation System is not expected over this time period. This change is acceptable because it provides additional assurance that the Control Room Ventilation System will perform correctly during a design basis event. This change is designated as more restrictive because it adds a new Surveillance Requirement to the CTS. M.3 When both control room ventilation subsystems are inoperable, CTS 3.17.A.3.a allows 24 hours to restore an inoperable control room ventilation subsystem to OPERABLE status. If CTS 3.17.A.3.a is not met, CTS 3.17.A.3.b requires the unit to be in MODE 3 in 12 hours and in MODE 4 within the following 24 hours. ITS 3.7.5 ACTION D requires immediate entry into LCO 3.0.3 under the same conditions. This will require the unit to initiate action within 1 hour to place the unit in MODE 2 within 7 hours, MODE 3 within 13 hours, and MODE 4 within 37 hours. This changes the CTS by requiring an immediate entry into LCO 3.0.3 when two control room ventilation subsystems are inoperable. The purpose of the CTS 3.17.A.3.a and CTS 3.17.A.3.b is to provide appropriate compensatory actions when both control room ventilation subsystems are inoperable and place the unit in a MODE in which the LCO is no longer applicable in a reasonable amount of time if at least one subsystem is not restored to OPERABLE status. This change requires immediate entry into LCO 3.0.3 when two control room ventilation subsystems are inoperable. ITS 3.7.5 ACTION D does not include any specific time for restoration and requires immediate entry into LCO 3.0.3. This change is acceptable because the equipment in the control room boundary may not function correctly if a design basis event were to occur. This change places the unit outside of the applicability of the Specification where the system is not required to mitigate the consequences of those design basis events postulated during operating Monticello Page 2 of 3 Attachment 1, Volume 12, Rev. 0, Page 98 of 161
Attachment 1, Volume 12, Rev. 0, Page 99 of 161 DISCUSSION OF CHANGES ITS 3.7.5, CONTROL ROOM VENTILATION SYSTEM conditions. This change is designated as more restrictive since the restoration time of 24 hours has been deleted. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 5- Deletion of Surveillance Requirement) CTS 4.17.A.1 requires verification that the control room temperature is within limit. ITS 3.7.5 does not include this requirement. This changes the CTS by eliminating the Surveillance Requirement to verify control room temperature. The purpose of CTS 4.17.A.1 is to ensure the continuous duty rating for the instrumentation and equipment cooled by this system is not exceeded. This change is acceptable because the deleted Surveillance Requirement is not necessary to ensure the Control Room Ventilation System can perform its safety function. ITS SR 3.7.5.1 has been added in accordance with DOC M.2 to verify each control room ventilation subsystem has the capability to remove the assumed heat load. This SR will ensure the Control Room Ventilation System can perform its safety function. Temperature is not always the appropriate method to verify the system capability to remove its design basis heat load because the conditions in the control room boundary do not always reflect the assumptions of the accident (e.g., personnel assumed to be in the control room boundary during an accident, the system does not normally operate in the pressurization mode of operation). ITS SR 3.7.5.1 will ensure each control room ventilation subsystem has sufficient cooling capability to meet the safety analyses assumptions. This change is designated as less restrictive because a Surveillance that is required in the CTS will not be required in the ITS. Monticello Page 3 of 3 Attachment 1, Volume 12, Rev. 0, Page 99 of 161
Attachment 1, Volume 12, Rev. 0, Page 100 of .161 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 100 of 161
Attachment 1, Volume 12, Rev. 0, Page 101 of lControl Roomp System 3.7.5 0D CTS 3.7 PLANT SYSTEMS 3.17A 3.7.5 gContr 'ofRoom lir Con flning (AC) System 0 3.17A-1 LCO 3.7.5 Two control room subsystems shall be OPERABLE. 0 3.17A1 APPLICABILITY: MODES 1, 2, and 3, During movement of re t irradiated fuel assemblies in the [secondary; containment, } 0 During operations with a potential for draining the reactor vessel (OPDRVs). ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.17A.2.a A. One control room subsystem inopera e. A.1 Resteontrolroom subsystem to OPERABLE 30 days 0 status. 3.17A.2.b B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not AND met in MODE 1, 2, or 3. B.2 Be in MODE 4. 36 hours
~~-+ 4-3.17A.2.c C. Required Action and -- NOTE-associated Completion LCO 3.0.3 is not applicable.~~
Time of Condition A not met during movement of r-e t irradiated fuel assemblies in the Esecondaryl containment or during OPDRVs. C.1 OR Place OPERABLE jontrol room asubsystem in operation. i Immediately
~~} 0D BWR/4 STS 3.7.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 101 of 161~~
Attachment 1, Volume 12, Rev. 0, Page 102 of 161 WControl RoomWJSystem 0 3.7.5 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.17.A.2.c C.2.1 Suspend movement of Immediately [frecety1 irradiated fuel assemblies in the Msecondary containment. } 0 AND C.2.2 Initiate action to suspend Immediately OPDRVs.
+ 4 3.17.A.3.a, 3.17.A.3.b D. room ]j subsystems inoperabe D.1 Enter LCO 3.0.3. Immediately 0D in MODE 1, 2, or 3.
3.17.A.3.c E. Two room L33NOTE . subsystems inoperable LC O 3.0.3 isnot applicable. durinq movement of [recntly] irradiated fuel assemblies in the E.1 Suspend movement of Immediately 0D Esecondaryi containment lc irradiated fuel or during OPDRVs. assemblies in the secondarA containment. AND E.2 Initiate actions to suspend Immediately OPDRVs.
~~.1.~~
~~SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY~~
~~- 1 -~~
__-venwaionI DOC M2 SR 3.7.5.1 Verify each rcontrol roomg sulsystem has the capability to remove the assumed heat load. I6mnths 0 BWR/4 STS 3.7.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 102 of 161
Attachment 1, Volume 12, Rev. 0, Page 103 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.5, CONTROL ROOM VENTILATION SYSTEM
~~1. The brackets have been removed and the proper plant specific information/value has been provided.~~
Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 103 of 161
Attachment 1, Volume 12, Rev. 0, Page 104 of 161 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 104 of 161
Attachment 1, Volume 12, Rev. 0, Page 105 of 161 Control Room B 3.. B 3.7 PLANT SYSTEMS B 3.7.5 IlControl Room Air Co n gystem The control room boundary Includes the main control room and portions of the first and second floors of the (D Emergency Filtration Train (EFT) building. BASES ai-BACKG ROUND The Control RoomriESystem provides temperature control for the control roo following isolation of the control room. (D bond -VentlaIonj TheMControl Roomr System consists of two independent, redundant (D boundary subsystems that provide cooling and heating of recirculated control room ndoutside air. Each subsystem consists of heating coils, cooling coils, fans, c rs ( nOtPEPBIrr compressors, ductwork, dampers, and instrumentation and controls to provide for control roomitemperature control.inr Ventiation The Control Room KJ-System is designed to provide a controlled environment under both normal and accident conditions. A single cfD subsystem provides the required temperature control to maintain a
~~&-,,suitable control roorlenvironment for a sustained occupancy of boundary I INSERT 9 ersons. he design conditio ro om environment are D 0~~
76 F and 50 are ive humidit . ThegControl Room ysytem etao ( operation in maintaining the control roo temperature is discussed inI u1mSAR, Secti ~ io[,Ref. 1).-L6 bnde >(latio] 0 APPLICAE 3LE
~~-j &~~
The design basis of thelControl RoomJystem is to maintain the ( D SAFETY ANALYSE control roo temperature for a 30 day continuous occupancy. 0O The TControl Room System components are arranged in redundant safety related subsystems. During emergency operation, the mcontrol ! Roo System maintains a habitable environment and ensures the / f co onents in the control roomn.A single failure of a 0 component of the Control 0oo0 System, assuming a loss of offsite power, does not impair the ability of the system to perform its design function. Redundant detectors and controls are provided for control room, emperature control. TheaControl Roomng System is designed inboundz ventla on accordance with Seismic Category I requirements. The MControl Room System is capable of removing sensible and latent heat loads from the control roorr, including consideration of equipment heat loads and personne occupancy requirements to ensure equipment OPERABILITY. 0 Ven-iation The WControl RoomZ1FSystem satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). 0 BWR/4 STS B 3.7.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 105 of 161
Attachment 1, Volume 12, Rev. 0, Page 106 of 161 B 3.7.5 0 INSERT I The system is designed to maintain the control room boundary at 780 F during the summer and 720 F in the winter. The maximum design condition in the control room and most of the EFT is 1040F. Insert Page B 3.7.5-1 Attachment 1, Volume 12, Rev. 0, Page 106 of 161
Attachment 1, Volume 12, Rev. 0, Page 107 of 161 R [Control Room M Sy~stem5
~ControlB 3.7.5 BASES atian LCO Two independent and redundant subsystems of theaControl Roomyd System are required to be OPERABLE to ensure that at least one is available, assuming a single failure disables the other subsystem. Total system failure could result in the equipment operating temperature exceeding limits.
TheRControl Room&System is considered OPERABLE when the (i) individual components necessary to maintain the control roomybudr temperature are OPERABLE in both subsystems. These components include the cooling coils, fans, c rs compressors, ductwork, dampers, and associated instrumentation and controls. Ventilabon APPLICABILITY In MODE 1,2, or 3, thelControl RoomE&System must be OPERABLE to ensure that the 'control roomrtemperature will not exceed equipment i) 0 lbounda OPERABILITY limits following control room isolation. In MODES 4 and 5, the probability and consequences of a Design Basis Accident are reduced due to the pressure and temperature limitations in these MODES. Therefore, maintaining the lControl Room stm OPERABLE is not required in MODE 4 or 5, except for the following n; 0i bd situations under wich significant radioactive releases can be postulated:
a. During operations with a potential for draining the reactor vessel (OPDRVs),and f

b. During movement of irradiated fuel assemblies in the Irentl secondarA containment. [Due to rdioactive decay, the Co ol oom ys em is only require to be OPERABLE duri fuel han ng involving handling re ntly irradiated fuel (i.e. el that has 0
o upied part of a critical r ctor core within the previus [X1 days).1 ACTIONS A.1 venblaton With enTan one control room 9subsystem inoperable, the inoperable controlf roomMl subsystem must be restored to OPERABLE status within unit in this condition, the remaining OPERABLE1 boundary I room C]lsubsystem is adequate to perform the control roomAir( J D conditioning function. However, the overall reliability is reduced because a single failure in the OPERABLE subsystem could result in loss of the control roornair conditioning function. The 30 day Completion Time is e low probability of an event occurring requiring control room isolation, the consideration that the remaining subsystem can provide the required protection, and the availability of alternate safety and nonsafety cooling methods. BWRI4 STS B 3.7.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 107 of 161
Attachment 1, Volume 12, Rev. 0, Page 108 of 161 NControl Room System 0 B 3.7.5 BASES ACTIONS (continued) B.1 and B.2 In MODE 1, 2, or 3, if the inoperablepcontrol room subsystem cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems. C.1. C.2.1. and C.2.2 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving [re ntly irradiated fuel (i) assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of [recntly irradiated fuel assemblies is not sufficient reason to require a (i reactor shutdown. During movement of re ntl irradiated fuel assemblies in the Osecondari, containment or during OPDRVs, if Required Action A.1
~o ant ecmlted within the required Completion Time, the OPERABLE I~control room'M subsystem may be placed immediately in Operation. This action ensures that the remaining subsystem is OPERABLE, that no failures that would prevent actuation will occur, and that any active failure will be readily detected.
An alternative to Required Action C.1 is to immediately suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. If applicable, movement o [re ntl irradiated fuel assemblies in the 14Th TsecondaryM containment must be suspended immediately. Suspension o these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. BWR/4 STS B 3.7.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 108 of 161
Attachment 1, Volume 12, Rev. 0, Page 109 of 161 Room gControl RoomPAISystem 3.7.5 (i
~~~ControIB BASES ACTIONS (continued)~~
D.1 If both gcontrol room subsystems are inoperable in MODE 1, 2, or 3,1 the mControl Room gystem may not be capable of performing the intended function. Therefore, LCO 3.0.3 must be entered immediately. E.1 and E.2 The Required Actions of Condition E are modified by a Note indicating that LCO 3.0.3 does not apply. If moving [repcintly irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of [recently irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown. I ebabonl During movement of [re htly] irradiated fuel assemblies in the V(1 lsecondar containmentor uring OPDRVs, with two control roomin subsystems inoperable, action must be taken immediately to suspend activities that present a potential for releasing radioactivity that might require isolation of the control room. This places the unit in a condition that minimizes risk. If applicable, handling offjic~tly irradiated fuel in the Isecondary( containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended. SURVEILLANCE SR 3.7.5.1 REQUIREMENTS This SR verifies that the heat removal capability of the system is sufficient batoremove the control roomeat load assumed in the Msafety analyse4 ( i The SR consists of a combination of testing and calculation. The 1~ month Frequency is appropriate since significant degradation of the C[rntiabn 1 Control RoorIMjll System is not expected over this time period.ei REFERENCES 41TMSAR, SectiontnAI 0 0 BWR/4 STS B 3.7.5-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 109 of 161
Attachment 1, Volume 12, Rev. 0, Page 110 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.5 BASES, CONTROL ROOM VENTILATION SYSTEM
1. The brackets have been removed and the proper plant specific information/value has been provided.

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. 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 I Attachment 1, Volume 12, Rev. 0, Page 110 of 161
Attachment 1,Volume 12, Rev. 0, Page 111 of 161 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 12, Rev. 0, Page 111 of 161
Attachment 1, Volume 12, Rev. 0, Page 112 of 161 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.5, CONTROL ROOM VENTILATION SYSTEM There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 12, Rev. 0, Page 112 of 161
~~, Volume 12, Rev. 0, Page 113 of 161 ATTACHMENT 6 ITS 3.7.6, Main Condenser Offgas ,Volume 12, Rev. 0, Page 113 of 161~~
Attachment 1, Volume 12, Rev. 0, Page 114 of 161 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 12, Rev. 0, Page 114 of 161
( C ITS 3.7.6 ITS ITS 3.0 LIMING CONDMONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 3.8 MAIN CONDENSOR OFFGAS 7 4.8 MAIN CONDENSE F fi W ADoliCabnif / ADDllCabill 0
.I

0) 3 Applies lo the dloactive release rate from the main Applies to the samp ng nd monitoring of radioactive 0 I condenser effluents discharged the main condenser offgas. l 3
3 to Objective: //Obiective: 0 0 To limit th doses received at the site boundary ir main To limit Ihe doses ceived at the site boundary from main condens r offgas In the event that effluent is di arged with condenser offgas the event that effluent Is discharged with less tha full treatment less than full tree ent. - Co X Spccificatlon: 0 ED
-o Secit: f any main steam -le A2 not isolated and J 0 0 3.7.6 A. Main Condenser Offgas Activity A2 A. Main Condenser Offgas Activity la
1. Wheneverth Stea J Atr SJAEs are E: Not required to be performed until 31 days afte rthe 2U Applicability r the gross gamma activity rate of the QI SJAEs are In operation. I -C, CD e gases measured at the main condenser 0 LCO 3.7.6 offgas system pretreatment monitor station shall be 1. Verity the gross gamma acdihy rate of the noble 0 5s2.6 x 10 CpCIsecond alter a decay of 30 minutes. gases Is :2.6 x 10s pi/Csecond after a decay of 30 to minutes:I to CD: ACTION A 2. When the gross gamma act"ty rate of the noble n gases is not within the limit of 3.8.A.1 above, restore a. Once every month. -L gross gamma activity rate of the noble gases to -L 0 within the limit within 72 hours. b. 4 hours after a 2 50% Increase In the nominal I Ul steady state fission gas release after factoring a), out increases due to changes In thermal power level. I SR 3.7.6.1 3.8/4.8 192 07/25/01 Amendment No. 4 4O120, 05, 121 Page 1 of 2
C C C ITS 3.7.6 ITS 3.0 UMITING CONDIONS FOR OPERATION 4.0 SURVEILtLANCE REQUIREMENTS 4. ACTION B 3. When 3.8.A.2 cannot be met, either. 03 a. Isolate all maln steam tines within 12 hours; or CS) 0 b. Isolate the SJAEs within 12 hours; or 0 D c Be In hot shutdown within 12 hours and cold a shutdown within the following 24 hours. 3 C D E, -o0 0
~~;13 w 0)~~
Co 0 to (D 0 0 CD to (D 02 3.814.8 NEXT PAGE IS 198 193 07125/01 Amendment No. 16, t4O,120,121 Page 2 of 2
Attachment 1, Volume 12, Rev. 0, Page 117 of 161 DISCUSSION OF CHANGES ITS 3.7.6, MAIN CONDENSER OFFGAS 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 3.8.A.1 requires the main condenser offgas activity to be within limit "Whenever the Steam Jet Air Ejectors (SJAEs) are in operation." CTS 4.8.A requires the main condenser offgas activity Surveillance to be performed "after the SJAEs are in operation." ITS LCO 3.7.6 also requires the main condenser offgas activity to be within limit, however the Applicability is MODE 1, and MODES 2 and 3 with any main steam line not isolated and steam jet air ejector (SJAE) in operation. ITS SR 3.7.6.1 includes the same Surveillance requirement to verify the main condenser offgas activity, however a Note has been included that requires the Surveillance to be performed "after any main steam line is not isolated and SJAE in operation." This changes the CTS by clarifying that the LCO is always applicable in MODE 1, and only in MODES 2 and 3 when any main steam line is opened and a SJAE is in operation, and also allows the Surveillance to be performed only after both a main steam line is opened and a SJAE is in service. The purpose of CTS 3.8.A.1, in part, is to state when the main condenser offgas activity is applicable. The purpose of CTS 4.8.A, in part, is to state when the Surveillance is required to be performed. This change clarifies that the LCO is always applicable in MODE 1, and only in MODES 2 and 3 when any main steam line is opened and a SJAE is in operation, and also allows the Surveillance to be performed only after both a main steam line is opened and a SJAE is in service. The SJAEs cannot be placed in service without main steam pressure (i.e., any main steam line not isolated). This proposed Applicability is consistent with the CTS 3.8.A.3 default action to isolate the main steam lines, isolate the SJAEs, or be in cold shutdown. The Surveillance will be required to be performed at the same time as in the CTS since the SJAE cannot be placed in service without steam line pressure. This change is also acceptable because it matches the Actions and Applicability. In addition, the main condenser offgas activity limit is only necessary when the main steam is directed to the main condenser and the offgas system is in operation since this iswhen there is a potential for the Main Condenser Offgas System to rupture. This change is designated as administrative because it does not result in technical changes to the CTS. MORE RESTRICTIVE CHANGES None Monticello Page 1 of 2 Attachment 1, Volume 12, Rev. 0, Page 117 of 161
Attachment 1, Volume 12, Rev. 0, Page 118 of 161 DISCUSSION OF CHANGES ITS 3.7.6, MAIN CONDENSER OFFGAS RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 2 of 2 Attachment 1, Volume 12, Rev. 0, Page .118 of 161
Attachment 1, Volume 12, Rev. 0, Page 119 of 161 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 119 of 161
Attachment 1, Volume 12, Rev. 0, Page 120 of 161 Main Condenser Offgas 3.7.6 CTS 3.7 PLANT SYSTEMS 3.8.A 3.7.6 Main Condenser Offgas 3.8A.1 LCO 3.7.6 The gross gamma activity rate of the noble gases measured atithe maiR or Hsystem pretreatment monitor statiore shall be
<(2 0 mi/secondt[afer decay of 30 minutes. J 3.8A1 APPLICAB ILITY: MODE 1, MODES 2 and 3 with anylmain steam line not isolated ancd steam jet air ejector (SJAE) in operation.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.8.A2 A. Gross gamma activity A.1 Restore gross gamma 72 hours rate of the noble gases activity rate of the noble not within limit. gases to within limit. 3.8-A.3 . B. Required Action and associated Completion B.1 I Isolate all main steam lines. 12 hours I 0D Time not met. OR B.2 Isolate SJAE. 12 hours OR B.3.1 Be in MODE 3. 12 hours AND B.3.2 Be in MODE 4. 36 hours BWR/4 STS 3.7.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 120 of 161
Attachment 1, Volume 12, Rev. 0, Page 121 of 161 Main Condenser Offgas 3.7.6 Y-) CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.8.A SR 3.7.6.1 Not required to be performed until 31 days after any [main steam line not isolated andESJAE in (0 operation. Verify the gross gamma activity rate of the noble 31 days gases is <[ 30 minute4 mCi/second Tafter decay of AND 0D Once within 4 hours after a 2 50% increase in the nominal steady state fission gas release after factoring out increases due to changes in THERMAL POWER level BWR/4 STS 3.7.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 121 of 161
Attachment 1, Volume 12, Rev. 0, Page 122 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.6, MAIN CONDENSER OFFGAS
~~1. The brackets have been removed and the proper plant specific information/value has been provided.~~
Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 122 of 161
Attachment 1, Volume 12, Rev. 0, Page 123 of 161 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 123 of 161
Attachment 1, Volume 12, Rev. 0, Page 124 of 161 Main Condenser Offgas B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Main Condenser Offgas BASES BACKGROUND During unit operation, steam from the low pressure turbine is exhausted directly into the condenser. Air and noncondensible gases are collected in the condenser, then exhausted through the steam jet air ejectors (SJAEs) to the Main Condenser Offgas System. The offgas from the main condenser normally includes radioactive gases. The Main Condenser Offgas System has been incorporated into the unit design to reduce the gaseous radwaste emission. This system uses a catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser; the water and condensibles are stripped out by the offgas condenser a d moistu epara o. The radioactivity of the remaining gaseous mixture m I (i.e., the offgaeombiner euen is monitored Idownsem o Imoistur araos prior to entering the holdup line. at the outlet of the cffgas condenser APPLICABLE The main condenser offgas gross gamma activity rate is an initial INSER I SAFETY condition ofithe Main Condenser aflure event, discusse ANALYSES in tle,SAR, Section [15.1.35] (Re. 1). -The analysis assumes gross failu e in the Main Condenser Off as System that results in thvrupture ofl th Main Condenser Off as S em pressure bounda .y The gross J gamma activity rate is controlled to ensure that, during the event, the calculated offsite doses will be well within the limits of 10 CFR 100
~~)lfMor the NRC staffpr licensing basi The main condenser offgas limits satisfy Criterion 2 of F1RS-RT 2~ QW 10 CFR 50.36(c)(2)(ii).~~
LCO To ensure compliance with the assumptions I Oflqas Systenifaftrevent (Ref. 1 the fission product release rate shoul be consistent with a noble Us release to the reactor co ant of 1004 Ci/MWt-second after decay f 30 minutes. The LCO isCtablished [ must be<2607 ctnsistent with this repuiremen MWt x 100 uCi/MWtgecond
~~-m mCi/seconrda 30 -(i24361 0~~
APPLICABILITY The LCO is applicable when steam is being exhausted to the main condenser and the resulting noncondensibles are being processed via the Main Condenser Offgas System. This occurs during MODE 1, and during MODES 2 and 3 with any main steam line not isolated and] the SJAE in operation. In MODES 4 and 5, steam is not being exhausted to the main 0D condenser and the requirements are not applicable. BWR/4 STS B 3.7.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 124 of 161
Attachment 1, Volume 12, Rev. 0, Page 125 of 161 B 3.7.6 0 INSERT 1 an event that inadvertently releases the main condenser effluent directly to the environment without treatment. 0 INSERT 2 an event that inadvertently releases the main condenser effluent directly to the environment without treatment, Insert Page B 3.7.6-1 Attachment 1, Volume 12, Rev. 0, Page 125 of 161
Attachment 1, Volume 12, Rev. 0, Page 126 of 161 Main Condenser Offgas B 3.7.6 BASES ACTIONS A.1 If the offgas radioactivity rate limit is exceeded, 72 hours is allowed to restore the gross gamma activity rate to within the limit. The 72 hour Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins associated with permissible dose and exposure limits, and the low probability of IMa in Conde nse fasSystiem rup . NsERT3 B.1, B.2, B.3.1. and B.3.2 If the gross gamma activity rate is not restored to within the limits in the associated Completion Time,fNall main steam lines orl the SJAE must be isolated. This isolates the Main Condenser Offgas System from the source of the radioactive steam. The main steam lines are considered isolated if at least one main steam isolation valve in each main steam line is closed, and at least one main steam line drain valve in each drain line is closed. The 12 hour Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems. An altemative to Required Actions B.1 and B.2 is to place the unit in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed 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 unit conditions from full power conditions in an orderly manner and without challenging unit systems. SURVEILLANCE SR 3.7.6.1 REQUIREMENTS This SR, on a 31 day Frequency, requires an isotopic analysis of an offgas sample to ensure that the required limits are satisfied. The noble gases to be sampled are Xe-1 33, Xe-1 35, Xe-1 38, Kr-85, Kr-87, and Kr-88. If the measured rate of radioactivity increases significantly (by 2 50% after correcting for expected increases due to changes in THERMAL POWER), an isotopic analysis is also performed within 4 hours after the increase is noted, to ensure that the increase is not indicative of a sustained increase in the radioactivity rate. The 31 day Frequency is adequate in view of other instrumentation that continuously monitor the offgas, and is acceptable, based on operating experience. This SR is modified by a Note indicating that the SR is not required to be performed until 31 days after anyljmain steam line is not isolated andc the 0 SJAE is in operation. Only in this condition can radioactive fission gases be in the Main Condenser Offgas System at significant rates. BWR/4 STS B 3.7.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 126 of 161
Attachment 1, Volume 12, Rev. 0, Page 127 of 161 B 3.7.6 Q INSERT3 an event that inadvertently releases the main condenser effluent directly to the environment without treatment I Insert Page B 3.7.6-2 Attachment 1, Volume 12, Rev. 0, Page 127 of 161
Attachment 1, Volume 12, Rev. 0, Page 128 of 161 Main Condenser Offgas B 3.7.6 BASES .. REFERENCES 1. FSAR ion 15.1.35]. l E<. 10 CFR 100. 0 BWR/4 STS B 3.7.6-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 128 of 161
Attachment 1, Volume 12, Rev. 0, Page 129 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.6 BASES, MAIN CONDENSER OFFGAS
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.
Monticello Page 1 of I Attachment 1, Volume 12, Rev. 0, Page 129 of 161
Attachment 1,Volume 12, Rev. 0, Page 130 of 161 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 12, Rev. 0, Page 130 of 161
Attachment 1, Volume 12, Rev. 0, Page 131 of 161 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.6, MAIN CONDENSER OFFGAS There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 131 of 161
~~, Volume 12, Rev. 0, Page 132 of 161 ATTACHMENT 7 ITS 3.7.7, Main Turbine Bypass System , Volume 12, Rev. 0, Page 132 of 161~~
~~Attachment 1, Volume 12, Rev. 0, Page 133 of 161 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 12, Rev. 0, Page 133 of 161~~
~~, Volume 12, Rev. 0, Page 134 of 161 ITS 3.7.7 4 -Idd posIS3.7.7 Page 1 of 1 , Volume 12, Rev. 0, Page 134 of 161~~
Attachment 1, Volume 12, Rev. 0, Page 135 of 161 DISCUSSION OF CHANGES ITS 3.7.7, MAIN TURBINE BYPASS SYSTEM ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any requirements for Main Turbine Bypass System. ITS LCO 3.7.7 requires the Main Turbine Bypass System to be OPERABLE. This changes the CTS by incorporating the requirements of ITS 3.7.7. Appropriate ACTIONS and Surveillance Requirements are also provided. The Monticello transient analyses credit the opening of the turbine bypass valves during certain events. If the Main Turbine Bypass System is inoperable, the safety analyses are not met. This change is acceptable since the Monticello transient analyses credit the Main Turbine Bypass System; therefore, the LCO ensures that the assumptions of the safety analyses 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 I Attachment 1, Volume 12, Rev. 0, Page 135 of 161
Attachment 1, Volume 12, Rev. 0, Page 136 of 161 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1,Volume 12, Rev. 0, Page 136 of 161
Attachment 1, Volume 12, Rev. 0, Page 137 of 161 Main Turbine Bypass System 3.7.7 CTS 3.7 PLANT SYSTEMS 3.7.7 WeMain Turbine Bypass System (0 DOC M.1 LCO 3.7.7 The Main Turbine Bypass System shall be OPERABLE. OR The following limits ar ade applicable: [a. LCO 3.2.1, RAGE PLANAR LINEAR HEAT ENERATION RATE (APL GR)," limits for an inoperable Mai Turbine Bypass 0D System specified in the [COLR]; and] [b. LCO .2.2, "MINIMUM CRITICAL POW RATIO (MCPR)," limits for n inoperable Main Turbine Bypas System, as specified in the [0OLR]. ] I APPLICABILITY: THERMAL POWER 2 25% RTP. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC M.1 A. l[Requirements , A.1 [Satify the requirenents of 2 hours the'LCO odj/estore Main I LCnot met Main Turbine Bypass System Turbine Bypass System to 0) inoperablef OPERABLE status].
~~- t DOC M.1 B. Required Action and B.1 Reduce THERMAL 4 hours associated Completion POWER to < 25% RTP.~~
Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC M.1 SR 3.7.7.1 Verify one complete cycle of each main turbine bypass valve. ['days~ 0D BWR/4 STS 3.7.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 137 of 161
Attachment 1, Volume 12, Rev. 0, Page 138 of 161 Main Turbine Bypass System 3.7.7 CI - 1 Ir1, I Al fl- II f-II- A51L rr n ;n!.... OUKVWILLMI'JUI NUUINrIVIINI IO kcon11nueU) . SURVEILLANCE FREQUENCY DOC M. I SR 3.7.7.2 Perform a system functional test. [Mmonths 0 DOC M. 1 SR 3.7.7.3 Verify the TURBINE BYPASS SYSTEM m 10 RESPONSE TIME is within limits. BWR/4 STS 3.7.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 138 of 161
Attachment 1, Volume 12, Rev. 0, Page 139 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.7, MAIN TURBINE BYPASS SYSTEM
1. Editorial change made for enhanced clarity or to be consistent with similar statements in other places in the Bases.

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

3. ISTS SR 3.7.7.1 has a Frequency of 31 days. The Frequency is being changed to 92 days in ITS SR 3.7.7.1. The Monticello CTS does not require this test to be performed, however, the turbine bypass valves are currently cycled every 92 days in accordance with plant procedures, and this 92 day Frequency has been shown to be acceptable.

4. Monticello is not currently analyzed to operate with the Main Turbine Bypass System inoperable. Therefore, this bracketed allowance has not been adopted.
Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 139 of 161
Attachment 1, Volume 12, Rev. 0, Page 140 of 161 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1,Volume 12, Rev. 0, Page 140 of 161
Attachment 1, Volume 12, Rev. 0, Page 141 of 161 Main Turbine Bypass System B 3.7.7 B 3.7 PLANT SYSTEMS B 3.7.7 Main Turbine Bypass System BASES BACKGROUND The Main Turbine Bypass System is designed to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown. It allows excess steam r flow from the reactor to the condenser without going through the turbine. a The bypass capacity of the system is% of the Nuclear Steam Supply A) System rated steam flow. Sudden load reductions within the capacity of the steam bypass can be accommodated without reactor scram. The Main Turbine Bypass System consists t e valves connected to the Q m steam lines between the main steam isolation valves and the turbine
'Electrical Pressure stop valve bypass valve chest. Each of th se re valves is operated by Mechanical Pressure hydraulic cylinders. The bypass valves are controlled by the pressure Regulator regulation function of the Turbine6lectro Hydrauliontrol Systen , as discussed in th SAR, Sec7ioQ [7 .4] (Ref. 1). The bypass valves are ii.norm-a y cosed, and the pressure regulator controls the turbine control valves that direct all steam flow to the turbine. If the speed governor or the load limiter restricts steam flow to the turbine, the pressure regulator controls the system pressure by opening the bypass valves. When the bypass valves open, the steam flows from the bypass chest, through connecting piping, to the pressurelbreakdowrl assemblies, where s ie lof orifices are urther reduce the steam pressure before the steam enters the condenser. isreduced APPLICABLE E The Main Turbine Bypass System is assumed to function during the SAFETY turbine eereoad reiection transient as discussed in thefgAR, ANALYSES Section [15.1. e. 2 Opening the bypass valves during the eedwaterontrlerven mitigates the ase in reactor vessel pr failure (maximum which affects the MCPR during the event. n opdrable Mainur /
demand)andpneuraticn By ass System mayp sult in APLHGR and MCPR penalties s14.4.4and turbine trip with bypass - 4and (Rfs) reduced scram speeds The Main Turbine Bypass System satisfies Criterion 3 of respectvely 10 CFR 50.36(c)(2)(ii). LCO The Main Turbine Bypass System is required to be OPERABLE to limit peak pressure in the main steam lines and maintain reactor pressure within acceptable limits during events that cause rapid pressurization, so that the Safety Limit MCPR is not exceeded. [With the ain Turbine 0 Bypass Syst" inoperable, modifications to the APL R limits (LCO 3.2.1,/AVERAGE PLANAR LINEAR HEAT G ERATION RATE (APLHGRY) and the MCPR limits (LCO 3.2.2, "MIN UM CRITICAL BWR/4 STS B 3.7.7-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 141 of 161
Attachment 1, Volume 12, Rev. 0, Page 142 of 161 Main Turbine Bypass System B 3.7.7 BASES LCO (continued) POWER 10 (MCPR)") may be applied to allow t s LCO to be met.] The APLIH R and MCPR limits for the inoperable in Turbine Byass 0 System a specified in the COLR. An OPERABLE Main Turbine Bypass System requires the bypass valves to open in response to increasing main steam line pressure. This response is within the assumptions of the
-re' n-al7 (Retj..a ";r aplica APPLICABILITY The Main Turbine Bypass System is required to be OPERABLE at feedwater controller 2 25% RTP to ensure that the fuel cladding integrity Safety Limit and the failure (maximum demand) and pneumatic _
system degradation, cladding 1% plastic strain limit are not violated during the ur ne enerat re ectio:5 transient As discussed in the Bases for , 0 turbine ripeWes LCO 3.2.1 and LCO 3.2.2, sufficient margin to these limits exists at
~~< 25% RTP. Therefore, these requirements are only necessary when operating at or above this power level.~~
ACTIONS l A.1i If the Main Turbine Bypass System is inoperable (one or more bypass valves inoperable),/or the APLHGR and MCPR limits fovan inoperable IMain Turbine Oypass System, as specified in the COL , are not applied, the assumptions of the design basis transient analysis may not be met. Under such circumstances, prompt action should be taken to restore the Main Turbine Bypass Svstem to OPERABLE statuslor afwt thelII APLHQR and MCPR limit, accordin IThe 2 hour Completion Time is reasonable, based on the time to complete the Required Action and the low probability of an event occurring during this period requiring the Main Turbine Bypass System. 1 B.1 If the Main Turbine Bypass System cannot be restored to OPERABLE status or the LHGR and MCPR limits for an inoperable Main Turbin 1B3ass Sys m are not applied, THERMAL POWER must be reduced to
< 25% RTP. As discussed in the Applicability section, operation at feedwater controller ' < 25% RTP results in sufficient margin to the required limits, and the Main demand) and pneumatic Turbine Bypass System is not required to protect fuel integrity during the , _
turbinem pdwith bypasen turbine oad rectior transientL The 4 hour Completion Time isl" (J) reduced scram speeds reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems. BWR/4 STS B 3.7.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 142 of 161
Attachment 1, Volume 12, Rev. 0, Page 143 of 161 Main Turbine Bypass System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.1 REQUIREMENTS Cycling each main turbine bypass valve through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will function when required. The'n day Frequency is based on (7 engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. 92 Operating experience has shown that these components usually pass the SR when performe d at day Frequency. Therefore, the Frequency is acceptable from a reliability standpoint. SR 3.7.7.2 The Main Turbine Bypass System is required to actuate automatically to perform its design function. This SR demonstrates that, with the required system ini n signals, the valves will actuate to their required position. C' month Frequency is based on the need to perform this ( Surveillance under the conditions that apply during a unit outage and because of the potential for an unplanned transient if the Surveillance
, were performed with the reactor at power. Operating experience has show t[ month Frequency, which is based on the refueling cycl ( i / acceptable from a reliability standpoint. I. Therefore, Frequency Fthat these components usually pass (was conclud to be Ithe SR when performed atul SR 3.7.7.3 Ie IThis SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME is in compliance with the assumptions of the aropriate safety 24 anals. The response time limits are specified inj funit sikecific m
[tha ion. The month Frequency is based on the need to eeillance under the conditions that apply during a unit outage and because of the potential for an unplanned transient if the that these Surveillance were performed with the reactor at power. Operating Feuny components usually pass the SR when erience has show thef(( month Frequency, which is based on the (D 1 performed at refueling cycl cceptable from a reliability standpoints. Therefore, the Frequency 1was concluded to be0 1 REFERENCES .ThSAR,Section SectiSoon 3.AR, 0 BWRI4 STS B 3.7.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 143 of 161
Attachment 1, Volume 12, Rev. 0, Page 144 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.7 BASES, MAIN TURBINE BYPASS SYSTEM
1. The brackets have been removed and the proper plant specific information/value has been provided.

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. Change made to reflect changes made to the Specification.
Monticello Page 1 of I Attachment 1, Volume 12, Rev. 0, Page 144 of 161
Attachment 1, Volume 12, Rev. 0, Page 145 of 161 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 12, Rev. 0, Page 145 of 161
Attachment 1, Volume 12, Rev. 0, Page 1i46 of 161 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.7, MAIN TURBINE BYPASS SYSTEM There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 12, Rev. 0, Page 146 of 161
Attachment 1, Volume 12, Rev. 0, Page 147 of 161 ATTACHMENT 8 ITS 3.7.8, Spent Fuel Storage Pool Water Level Attachment 1,Volume 12, Rev. 0, Page 147 of 161
Attachment 1, Volume 12, Rev. 0, Page 148 of 161 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1,Volume 12, Rev. 0, Page 148 of 161
( C ITS 3.7.8 ITS ITS 3.0 LIMITING CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 4-B. Core Monitoring B. Core Monitoring See ITS 3.3.1.2} Su MI
0) During core alterations two SRM's shaltbe operable, one In and one adjacent to any core quadrant where fuel Prior to making any alterations to the core while rnore than two fuel assemblies are present In any reactor 0
or control rods are being moved. For an SRM to be quadrant, the SRM's shall be functionally tested and CD considered operable, the following conditions shall be checked for neutron response. Thereafter, the S RMs CD 3 satisfied: will be checked daily for response. f. 0 1. *TheSRM shall be Inserted to the normal operating Z-level. (Use of special moveable. dunking type detectors during Initial fuel loading and major core 0
~~%31 alterations Is permissible as long as the detector is connected Into the normal SRM circuit.)~~
~~is 2. The SRM shall have a minimum of 3 CPS with alt rods fully Inserted In the core except when both of~~
~~/--\ I the following conditions are fulfilled:~~
0 (D At a. No more than two fuel assemblies are present In the core quadrant ae;sodated with the SRM, 6
-o spent 1b. While In core, these fuiat assemblies are In CD locations adjacent to t0ie SRM. 0) to
CD CD

0) 3.7..8 C. Fuel Storage Pool Water Level 3.7.8 C. Fuel Storage Pool Water Level) at
-to Whenever Irradated fuel is stored in the CDk Co Whenever Irriaated fuel Is stored In th uel storage Ielstorage pool the pI6 water level shall be ra tamed at a level pool the popl"evol shall be recorded fly. 0 of greatFor equal to 33 feet. -9' D. The reactor shall be shutdown for a minimum of 24 hours prior to movement of fuel within the reactor. See CTS 3.10.- In1 ITS Section 3.9 3.10/4.10 207 10I26101 Amendment No. 2orm23 Page 1 of 2
Attachment 1, Volume 12, Rev. 0, Page 150 of 161 ITS 3.7.8 ITS ( INSERT I Applicability--[During movement of irradiated fuel assembliesrthe spent LCO 3.7.8 fuel storage pool water level shall be maintained > 37 feet aqbove the bottom of the spent fuel storage pool. ACIO A LO303is not applcbe ACTION A If the spent fuel storage pool water level is made or found not to be within limits, immediately suspend movement of. irradiated fuel assemblies. O INSERT 2 SR 3.7.8.1 i Verify that the spent fuel storage pool water level is
~~> 37 feet above the bottom of the spent fuel storage pool:~~
1. Once every 4 our , during movement if irradiated fuel assemb ies, or

2. Once eve 7 days, when irradiated fu assemblies are sto d in the spent fuel storage ol. A.
Insert Page 207 Page 2 of 2 Attachment 1, Volume 12, Rev. 0, Page 150 of 161
Attachment 1, Volume 12, Rev. 0, Page 151 of 161 v>DISCUSSION OF CHANGES. ITS 3.7.8, SPENT FUEL STORAGE POOL WATER LEVEL 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 These changes to CTS 3.10.C and CTS 4.10.C are provided in the Monticello ITS consistent with the Technical Specifications Change Request submitted to the USNRC for approval in NMC letter L-MT-05-013, from Thomas J. Palmisano (NMC) to USNRC, dated April 12, 2005. As such, these changes are administrative. A.3 Under certain conditions, CTS 3.10.C, in part, requires immediate suspension of movement of irradiated fuel assemblies. ITS 3.7.8 ACTION A includes the same requirement, however a Note has been added that states that LCO 3.0.3 is not applicable. This changes the CTS by adding this Note. The purpose of CTS 3.10.C, in part, is to provide the appropriate actions when the spent fuel storage pool water level is not within limit. This change adds a Note that states LCO 3.0.3 is not applicable. This Note has been added because ITS LCO 3.0.3 has been added to ITS Section 3.0 in accordance with ITS Section 3.0 DOC M.1. This Note is necessary because if moving irradiated fuel assemblies while in MODE 4 or 5, ITS LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement of irradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. Since ITS LCO 3.0.3 is not currently included in the CTS, this change is considered administrative. This change is designated as administrative because it does not result in any technical changes to the CTS. A.4 CTS 4.10.C.2 requires verification that the spent fuel storage pool water level is within limit once every 7 days when irradiated fuel assemblies are stored in the spent fuel storage pool. This Surveillance is not included in ITS 3.7.8. This changes the CTS by deleting this Surveillance. The purpose of CTS 4.10.C.2 is to ensure there is sufficient water in the spent fuel storage pool when irradiated fuel assemblies are stored in the spent fuel storage pool. The applicability specified in CTS 3.10.C is "during movement of irradiated fuel assemblies," not "when irradiated fuel assemblies are stored in the spent fuel storage pool." CTS 4.10.C.1 requires verification that the spent fuel storage pool water level is within limit once every 24 hours during movement of irradiated fuel assemblies. CTS 4.0.C states that whenever the plant condition is such that a system or component is not required to be OPERABLE, the Surveillance testing associated with that system or component may be discontinued. This allowance is retained in SR 3.0.1. Therefore, CTS 4.1 0.C.2 is not required to be met or performed. The requirements in CTS 3/4.1 0.C have Monticello Page 1 of 2 Attachment 1, Volume 12, Rev. 0, Page 151 of 161
Attachment 1, Volume 12, Rev. 0, Page 152 of 161
- DISCUSSION OF CHANGES ITS 3.7.8, SPENT FUEL STORAGE POOL WATER LEVEL recently been proposed to be modified (as discussed in DOC A.2) to ensure that when irradiated fuel assemblies are being moved over the spent fuel storage pool, the spent fuel storage pool level is adequate to ensure the consequences of fuel handling accident over the spent fuel pool will be bounded by the safety analysis. This specific Surveillance should not have been added to the Technical Specifications as part of this Amendment request. Therefore, this Surveillance is not required and is deleted. This change is designated as administrative because it does not result in technical changes to the CTS.
MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 7 - Relaxation of Surveillance Frequency, Non-24 Month Type Change) CTS 4.1 O.C.1 requires a verification every 24 hours that the spent fuel storage pool water level is within the limit. ITS SR 3.7.8.1 requires verifying the spent fuel storage pool water level is within the limit every 7 days. This changes the CTS by extending the Surveillance Frequency from 24 hours to 7 days. The purpose of CTS 4.1 O.C.1 is to ensure the spent fuel storage pool water level is within limit. This change extends the Surveillance Frequency from 24 hours to 7 days. 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 24 hours to 7 days maintains a Frequency that is adequate to trend changes in the spent fuel storage pool water level. In addition, a control room alarm exists to alert the operator to a decreasing water level. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. Monticello Page 2 of 2 Attachment 1, Volume 12, Rev. 0, Page 152 of 161
Attachment 1, Volume 12, Rev. 0, Page 153 of 161 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 153 of 161
Attachment 1, Volume 12, Rev. 0, Page 154 of 161 Spent Fuel Storage Pool Water Level 3.7.8 CTS 3.7 PLANT SYSTEMS 3.10.C 3.7.8 Spent Fuel Storage Pool Water Level 3.10.c LCO 3.7.8 The spent fuel storage pool water level shall be 2 ft veto Fiirted fuel assembl~isseated id the spent seorage pool1 APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel storage pool. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC L.1 A. Spent fuel storage pool A.1 ----- NOTE---- water level not within LCO 3.0.3 is not applicable. limit. Suspend movement of Immediately irradiated fuel assemblies in the spent fuel storage pool. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.10.C SR 3.7.8.1 gKeerify the spent fuel storage pool water level is 7 days 2_J ft ver the too of irradiated fuel ass bliesl the spent fuel storage pool rikB. above the bottom ofl BWR/4 STS 3.7.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 154 of 161
Attachment 1, Volume 12, Rev. 0, Page 155 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.8, SPENT FUEL STORAGE POOL WATER LEVEL
1. The brackets have been removed and the proper plant specific information/value has been provided.

2. ISTS LCO 3.7.8 requires the spent fuel storage water level to be 23 ft over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks and ISTS SR 3.7.8.1 requires verification that the spent fuel storage water level to be 23 ft over the top of irradiated fuel assemblies seated in the spent fuel storage pool racks. The.
value of 23 ft is bracketed. ITS LCO 3.7.8 requires the spent fuel storage water level to be > 37 ft above the bottom of the spent fuel storage pool and ITS SR 3.7.8.1 requires verification that the spent fuel storage level is > 37 ft above the bottom of the spent fuel storage pool. The proposed level will ensure there is > 21 ft 4 inches of water over the top of irradiated fuel assemblies seated in the spent storage pool racks. The proposed level is acceptable because the safety analyses shows that the radiological consequences of a refueling accident occurring over the core bounds the same accident over the spent fuel storage pool. The value and method of referencing the spent fuel storage pool level in ITS LCO 3.7.8 and ITS SR 3.7.8.1 is consistent with a recent NMC License Amendment Request, as provided in NMC letter L-MT-05-013, from Thomas J. Palisano (NMC) to USNRC, dated April 12, 2005. Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 155 of 161
Attachment 1, Volume 12, Rev. 0, Page 156 of 161 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 12, Rev. 0, Page 156 of 161
Attachment 1, Volume 12, Rev. 0, Page 157 of 161 Spent Fuel Storage Pool Water Level B 3.7.8 B 3.7 PLANT SYSTEMS B 3.7.8 Spent Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the spent fuel storage pool meets the assumptions of iodine decontamination factors following a fuel handling accident. i A general description of the spent fuel storage pool design is found in the 3S6AR, SectionS (Ref. 1 The assumptions of the fuel handling UJ (3 (E) are found in thel3SA Setio[15 .4(Ref. 2). accident 14..6 ltotal effective dose equivalent DE APPI -ICABLE The water level above the irradiated fuel assemblies is an explici SAFEETY assumption of the fuel handling accident. A fuel handling accident is ANAI LYSES evaluated to ensure that the radiological consequences (calculawec- E I body and 01foldose: at the exclusion area and low population zone boun aries) are 25% c FR 1001(Ref. 3) exposure guidelines
/ RE - Ref. 4)1. A fuel handling accident could release a fraction of the fission product inventory by breaching the fuel rod cladding as wthin 10 CFR 50.67 discussed in the Regulatory Guide 1.25 (Ref. R_ (
The fuel handling accident is evaluated for the dropping of an irradiated fuel assembly onto the reactor core. The consequences of a fuel handling accident over the spent fuel storage pool are no more severe m than those of the fuel handling accident over the reactor core, ask 43 discussed in the SAR, Sectio (9.1 .2.2] (Ref.l.-The water leveTn the(D i) spent fuel storage pool provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water before being released to the secondary containment atmosphere. This absorption and transport delay reduces the potential radioactivity of the release during a fuel handling accident. The spent fuel storage pool water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii). LCO The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel movement within the spent fuel storage pool. APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel storage pool since the potential for a release of fission products exists. BWR14 STS B 3.7.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 157 of 161
Attachment 1, Volume 12, Rev. 0, Page 158 of 161 Spent Fuel Storage Pool Water Level B 3.7.8 BASES ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not a sufficient reason to require a reactor shutdown. When the initial conditions for an accident cannot be met, action must be taken to preclude the accident from occurring. If the spent fuel storage pool level is less than required, the movement of irradiated fuel assemblies in the spent fuel storage pool is suspended immediately. Suspension of this activity shall not preclude completion of movement of an irradiated fuel assembly to a safe position. This effectively precludes a spent fuel handling accident from occurring. SURVEILLANCE SR 3.7.8.1 REQUIREMENTS This SR verifies that sufficient water is available in the event of a fuel handling accident. The water level in the spent fuel storage pool must be checked periodically. The 7 day Frequency is acceptable, based on operating experience, considering that the water volume in the pool is normally stable, and all water level changes are controlled by unit procedures. REFERENCES £ A AR, Section - q3.NUSAR Section , 3ecC . e Jl .67 (0
3. INUREG-08F0, SectjnSh.-.,evision 1, July 18l 00

14. 1GCF+R 100. l
~~.1~Regulatory Guide 1.25, March 1972.~~
2-@SAR, Section 9.f2.:2 . 1..3. 0-BWR/4 STS B 3.7.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 12, Rev. 0, Page 158 of 161
Attachment 1, Volume 12, Rev. 0, Page 159 of 161 JUSTIFICATION FOR DEVIATIONS ITS 3.7.8 BASES, SPENT FUEL STORAGE POOL WATER LEVEL
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 Monticello Page 1of 1 Attachment 1, Volume 12, Rev. 0, Page 159 of 161
Attachment 1, Volume 12, Rev. 0, Page 160 of 161 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 12, Rev. 0, Page 160 of 161
Attachment 1, Volume 12, Rev. 0, Page 161 of 161 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.7.8, SPENT FUEL STORAGE POOL WATER LEVEL There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 12, Rev. 0, Page 161 of 161}}

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