ML051960487

Improved Technical Specifications, Volume 6, Revision 0, ITS Section 3.1, Reactivity Control Systems.

ML051960487
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Site:	Monticello
Issue date:	06/29/2005
From:	Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
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Download: ML051960487 (232)

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{{#Wiki_filter:IMPROVED TECHNICAL SPECIFICATIONS MONTICELLO NUCLEAR GENERATING PLANT VOLUME 6 ITS Section 3.1, Reactivity Control Systems Committed to Nuc ear Excellece7

Attachment 1, Volume 6, Rev. 0, Page 1 of 231 ATTACHMENT 1 VOLUME 6 MONTICELLO IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.1 REACTIVITY CONTROL SYSTEMS Revision 0 Attachment 1, Volume 6, Rev. 0, Page 1 of 231

Attachment 1, Volume 6, Rev. 0, Page 2 of 231 LIST OF ATTACHMENTS

1. ITS 3.1.1

2. ITS 3.1.2

3. ITS 3.1.3

4. ITS 3.1.4

5. ITS 3.1.5

6. ITS 3.1.6

7. ITS 3.1.7

8. ITS 3.1.8

9. Relocated/Deleted Current Technical Specifications (CTS)

Attachment 1, Volume 6, Rev. 0, Page 2 of 231

, Volume 6, Rev. 0, Page 3 of 231 ATTACHMENT I ITS 3.1.1, SHUTDOWN MARGIN (SDM) , Volume 6, Rev. 0, Page 3 of 231

Attachment 1, Volume 6, Rev. O,Page 4 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 6, Rev. 0, Page 4 of 231

C C, C ITS 3.1.1 ITS 0 ITS 3.0 LIMmNG CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

                                                                                      *1.

Su 3.3 CONTROL ROD SYSTEM 4.3 CONTROL HQD SYSTEM

                                                                                    .4 I_                                                  _  _ _   _  _  _ _   _ _ _   _                      0)

AnDflc atflc Aepoliabl!r CD a7 0 Appilos to the surveilance Aqurements of the control rori Applies to the o rational status of the control rod sysle system. 01 oblverIe h To assure e abilty of the control rod system to co rol 0 reacrtyi/ To ver fy the ability o1h rcontrol rott system lo control F- reactuty. I 0 0 00

                                                                                                              /                                   Once ithin 4 hours after crftkcafity         0)

IA. Rea"VI/tlumitations I nowrg h.l rnovemrent within the M.3 0 a 3.1.1 ' 1. I Reactivity 5iiiFf Tc--ore loadi-ngl Ireactor pressure vessel or control D 11. ReartitVibur&,in coreI od replacment kodng P LCO 3.1.1 'Tb(q enrA 1nadhwFRhffbe limited t ,( I  ;' be made subcril [durind the ODO with the 0) -o operable control rod in its full-ou to other operable 0 lesubsequent fuel cycle with the stronge .sV control rod fully withdrawn and all other 0 Ca' erds fully Inserted See ITS 1.0 Add proposed SR 3.1.1.1 I

                                                                                                                                 \~

-4' -4' tfirst Freque~ncyM. lAPPLICABILUY: MODES 1,2.3,4. and 5 3.3/4.3 75 1/9/81 Amendment No. 0 Page 1 of 3

( C. C ITS 3.1.1 ITS 0 3.0 UMITING CONDmONS FOR OPERATION l 4.0 SURVEILLANCE REQUIREMENTS I F. Scram Discharge Volume F. Scram Discharge Volume

1. During reactor operation, the scram discharge The scram discharge volume vent and drain valves shall 0 volume vent and drain valves shall be operable, be cyded quarterly. a) except as specified below. Once per operating cyde verify the scram discharge 0 volume vent and drain valves dose within 30 seconds D 2. If any scam discharge volume vent or drain valve Is after receipt of a reactor scram signal and open when CD made or found Inoperable, the Integrity of the scram the scram Lsreset.

discharge volume shall be maintained by either 0

a.

• Verifying dafly, for a period not to exceed 7 days, the operability of the redundant valve(s). 0 00F or 0

b. Maintaining the Inoperable valve(s), or the CD Co 0

-;a M associated redundant valve(s), In the dosed position. Periodically the Inoperable and the redundant valve(s) may both be In the open position to allow draining the scram discharge

                                                                                                                                                                --     See ITS 3.1.8 I        CD 0

An X IV volume. la II 0 If a or b above cannot be met, at least all but one 0) co

;z                               operable control rods (not Induding rods removed to                              per specification 3.10.E or Inoperable rods allowed CD                                                                                                                                                                                            CD by 3.3A2) shall be fully Inserted within ten hours.

4- a) 0 G. Required Actlon 0) 0

-96                                                                                       [(except when the reactor mode                                                                      -16 ACTIONS                 If Specifcatins 3.3A hrough D above are not metfah             Iswvitd Isinthe Refuel position)

A. B, Corderly shutd shall be Initiated and yave reactor in_ -- G and D . the cold shutd condition wIthin 24 )our s, [Add proposed ACTIONS A, B. C.and D M.5 I A 3.314.3 83a 5/1/84 Amendment No. 24 Page 2 of 3

Attachment 1, Volume 6, Rev. 0, Page 7 of 231 ITS 3.1.1 ITS 0 Qt) INSERT A ACTION E 2. If Specification 3.3.A is not met when the reactor mode switch is in the Refuel position, immediately suspend core alterations exce t for fuel assembi ad onro I removal and immediately initiate action to fully ro Ineto insert all insertable control rods in core cells containing one or more fuel assemblies. Add propod ReqPeI fActon D.3 and 0.4 Insert Page 83a Page 3 of 3 Attachment 1, Volume 6, Rev. 0, Page 7 of 231

Attachment 1, Volume 6, Rev. 0, Page 8 of 231 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) 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 3.3.G 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 CTS 3.3.G.2 requires the immediate suspension of core alterations except for "fuel assembly removal" and to "immediately initiate action to fully insert all insertable control rods in core cell containing one or more fuel assemblies" if CTS 3.3.A is not met when the reactor mode switch is in the Refuel position. ITS 3.1.1 ACTION E covers the condition for SDM not met in MODE 5, and in part, requires the immediate suspension of CORE ALTERATIONS except for "control rod insertion and fuel assembly removal" and requires the immediate initiation of action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. This changes the CTS by clarifying that CORE ALTERATIONS that involve "the insertion of control rods" are also excepted. The purpose of CTS 3.3.G.2 is to immediately stop all core alterations that can reduce shutdown margin. The CTS definition (CTS 1.0.A) of "Alteration of the Reactor Core" does not include normal operating functions such as control rod movement using the normal drive mechanism. In ITS 1.1, the "CORE ALTERATIONS" definition includes the movement of control rods as long as the associated core cell contains one or more fuel assemblies. This change is acceptable because CTS 3.3.G.2 specifically requires action to fully insert all insertable control rods in core cells containing one more fuel assemblies. Therefore, the addition of the exception is considered administrative. This change is designated as administrative because it does not represent a technical change to the Technical Specifications. MORE RESTRICTIVE CHANGES M.1 CTS 4.3.A.1 states, in part, reactivity margin of "0.25 per cent Ak" is required. ITS LCO 3.1.1 states SDM shall be: a. 2 0.38% Ak/k, with the highest worth control rod analytically determined; or b. 2 0.28% Ak/k, with the highest worth control rod determined by test. This changes the CTS by replacing the existing SDM limit with two new limits. The purpose of ITS LCO 3.1.1 is to allow flexibility in the determination of SDM. This change is acceptable because the LCO requirements continue to ensure Monticello Page 1 of 6 Attachment 1, Volume 6, Rev. 0, Page 8 of 231

Attachment 1, Volume 6, Rev. 0, Page 9 of 231 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) that the reactor core is maintained consistent with the safety analyses. ITS LCO 3.1.1 provides a SDM of 2 0.38% Ak/k, with the highest worth control rod analytically determined or a SDM limit of 2 0.28% Ak/k, with the highest worth control rod determined by test. The current limit of 2 0.25% Ak/k does not specify. how the strongest control rod is determined. This change is acceptable because for the SDM demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10% Ak/k) must be added to the SDM limit for uncertainties in the calculation. The SDM may be demonstrated during an in sequence control rod withdrawal, in which the highest worth control rod is analytically determined, or during local control rod tests, where the highest worth control rod is determined by testing. The proposed allowances are consistent with the ISTS and the additional margin is considered sufficient based on the uncertainties observed in the calculation methodology. This change is designated as more restrictive since the new limits will require additional SDM in order to satisfy the Specification. M.2 CTS 3.3.A.1 states, in part, that core loading shall be limited to that which can be made subcritical in the most reactive condition during the operating cycle. CTS 4.3.A.1 states, in part, that a test shall be performed to demonstrate that the core can be made subcritical at any time in the subsequent fuel cycle. CTS 3.3.G.1 requires the unit to be in cold shutdown (MODE 4) within 24 hours if CTS 3.3.A.1 is not met.. CTS 3.3.G.2 provides Actions for when the reactor mode switch is in the Refuel position (i.e., MODE 5 in the ITS). ITS LCO 3.1.1 requires SDM to be met during MODES 1, 2, 3, 4, and 5. This changes the CTS by changing the Applicability from MODE 1, 2, and 3 (based on the shutdown requirement of CTS 3.3.G.1) and MODE 5 (based on the reactor mode switch position requirement of CTS 3.3.G.2) to MODES 1, 2, 3, 4, and 5. Changes to the requirements of CTS 3.3.G.1 are discussed in DOC M.5 and changes to the requirements of CTS 3.3.G.2 are discussed in DOCs A.3 and M.6. The purpose of the ITS 3.1.1 Applicability is to ensure SDM is met whenever fuel is in the reactor core. The change is acceptable because the safety analyses assume that SDM is met whenever fuel is in the reactor core. In MODES 1 and 2, SDM must be provided because subcriticality with the highest worth control rod withdrawn Is assumed in the control rod drop accident analysis and other accident and transient analyses. In MODES 3 and 4, SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod. SDM is required in MODE 5 to prevent an open vessel, inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblies or a fuel assembly insertion error. This change is designated as more restrictive because it increases the conditions for when the Specification is required to be met. M.3 CTS 4.3.A.1 states, in part, the reactivity margin demonstration shall be performed "following a refueling outage when core alterations were performed." ITS SR 3.1.1.1 states, verify SDM to be within limits at a Frequency of 'Once within 4 hours after criticality following fuel movement within the reactor pressure vessel or control rod replacement." This changes the CTS by stating a finite time to complete the Surveillance (once within 4 hours after criticality) and requiring the Surveillance to be performed following fuel movement within the reactor Monticello Page 2 of 6 Attachment 1, Volume 6, Rev. 0, Page 9 of 231

Attachment 1, Volume 6, Rev. 0, Page 10 of 231 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) pressure vessel or control rod replacement in lieu of following a "refueling outage" when core alterations were performed. The purpose of CTS 4.3.A.1 is to ensure there Is sufficient reactivity margin designed in the reactor core and that this Is demonstrated after a refueling outage after core alterations are made. The proposed Surveillance Frequency in ITS SR 3.1.1.1 states that a SDM demonstration must be performed "Once within 4 hours after criticality following fuel movement within the reactor pressure vessel or control rod replacement" instead of the current requirement to perform the demonstration "following a refueling outage when core alterations were performed." Therefore, this change effectively places a finite time limit on completing the Surveillance. In addition, the current Surveillance is only required after a refueling outage. The intent of this portion of the Surveillance Frequency is to verify the core reactivity after in-vessel operations that could have altered the core reactivity. During refueling outages, core reactivity is normally significantly altered. However, conditions could arise mid-cycle that require replacing a fuel assembly or control rod, and the CTS would not require this Surveillance to be performed during the subsequent reactor startup. This mid-cycle replacement has the potential of altering core reactivity. The ITS words cover both planned refueling outages and other outages where CORE ALTERATIONS may occur, thus this change is considered acceptable. This change is acceptable since the proposed Frequency ensures SDM is within limits shortly after any fuel movement within the reactor pressure vessel or any control rod replacements have been made. The Frequency of 4 hours after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification. This change is designated as more restrictive because a Surveillance will be performed under more conditions and with a finite time limit for completion under the ITS than under the CTS. M.4 ITS SR 3.1.1.1 requires verification of SDM "Prior to each in vessel fuel movement during fuel loading sequence." Currently, the CTS does not require a SDM verification at this Frequency. This changes the CTS by adding a new Surveillance Frequency for the SDM verification. The purpose of the new Surveillance Frequency in ITS SR 3.1.1.1 (first Frequency) is to ensure SDM is met during the fuel loading sequence. This change adds a requirement to ensure SDM is met "Prior to each in vessel fuel movement during fuel loading sequence." This change is acceptable because the new Surveillance Frequency in ITS SR 3.1.1.1 will ensure the reactor core will not go critical during a fuel loading sequence. During MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of each in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the associated uncertainties. Spiral offload/reload sequences inherently satisfy the SR, provided Monticello Page 3 of 6 Attachment 1, Volume 6, Rev. 0, Page 10 of 231

Attachment 1, Volume 6, Rev. 0, Page 11 of 231 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SDM. This change is designated as more restrictive because it adds the requirement to verify SDM during a fuel loading sequence. M.5 CTS 3.3.G.1 requires the unit to be in cold shutdown (MODE 4) within 24 hours if CTS 3.3.A.1 is not met. ITS 3.1.1 specifies specific ACTIONS for each MODE (MODE 1, 2, 3, 4, and 5). ITS 3.1.1 ACTION A covers the condition for SDM not met in MODES I or 2, and requires the restoration of SDM to within limits within 6 hours. If this is not met, ITS 3.1.1 ACTION B requires the unit to be in MODE 3 in 12 hours. ITS 3.1.1 ACTION C covers the condition for SDM not met in MODE 3, and requires immediate initiation of action to fully insert all insertable control rods. ITS 3.1.1 ACTION D covers the condition for SDM not met in MODE 4, and requires immediate initiation of action to fully insert all insertable control rods, and within 1 hour, to restore secondary containment to OPERABLE status, to restore one standby gas treatment (SGT) subsystem to OPERABLE status, and to restore isolation capability in each required secondary containment penetration flow path not isolated. This changes the CTS by specifying explicit compensatory actions for MODES 1, 2, 3, and 4 in lieu of a single common action for these MODES. The purpose of the ITS 3.1.1 ACTIONS are to ensure the appropriate compensatory actions are taken when SDM is not met. CTS 3.3.G.1 requires the unit to be in cold shutdown (MODE 4) within 24 hours when the SDM requirements are not met. In MODES 1 and 2, the ITS will allow 6 hours to restore the SDM to within limits. If this can not be met, the unit must be in MODE 3 (hot shutdown) within the next 12 hours. This portion of the change is acceptable since it reduces the time the unit must be at a specified condition (from 24 hours to cold shutdown to 18 hours to MODE 3) and places the unit in condition where the core reactivity is reduced. If the unit is brought to MODE 3 there is no requirement to go to MODE 4 (cold shutdown) since in this condition the reactivity of the core may effectively increase due to the reduction in reactor coolant temperature. Therefore, the ITS 3.1.1 compensatory action in MODE 3 (ITS 3.1.1 ACTION C) is acceptable since it will help to reduce the reactivity conditions of the core by requiring the immediate initiation of action to insert all insertable control rods. Although there is no requirement to achieve MODE 4 conditions, the proposed action to stay in MODE 3 is acceptable and appropriate considering the behavior of the reactor core. In MODE 4, the ITS compensatory actions continue to require the reduction of the core reactivity and to help minimize any consequences of an event if an event should occur during the time period when SDM is not met. The compensatory actions proposed for MODE 4 are considered appropriate and acceptable. This change is designated as more restrictive because it adds compensatory actions and reduces the time limit in which the unit must be in a specified condition. M.6 CTS 3.3.G.2 requires the immediate suspension of core alterations except for "fuel assembly removal" and to "immediately initiate action to fully insert all insertable control rods in core cell containing one or more fuel assemblies" if CTS 3.3.A.1 is not met when the reactor mode switch is in the Refuel position. ITS 3.1.1 ACTION E covers the condition for SDM not met in MODE 5, and requires the immediate suspension of CORE ALTERATIONS except for control Monticello Page 4 of 6 Attachment 1, Volume 6, Rev. 0, Page 11 of 231

Attachment 1, Volume 6, Rev. 0, Page 12 of 231 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) rod insertion and fuel assembly removal, immediate initiation of action to fully insert all insertable control rods in core cells containing one or more fuel assemblies, and to initiate action within 1 hour to restore secondary containment to OPERABLE status, restore one standby gas treatment (SGT) subsystem to OPERABLE status, and restore isolation capability in each required secondary containment penetration flow path not isolated. This changes the CTS by adding the explicit compensatory actions associated with the secondary containment functions. The purpose of CTS 3.3.G.2 is to immediately stop all core alterations that can reduce shutdown margin. Actions have been added that require the restoration of the secondary containment, one SGT subsystem, and the isolation capability in each required secondary containment penetration flow path not isolated. These actions are provided for the control of potential radioactive release. In MODE 5, the ITS compensatory actions continue to require the reduction of the core reactivity and to help minimize any consequences of an event if an event should occur during the time period when SDM is not met. The compensatory actions proposed for MODE 5 are considered appropriate and acceptable. This change is designated as more restrictive because it adds compensatory actions. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.l (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.3.A.1 states, in part, that the core loading shall be limited to that which can be made subcritical "in the most reactive condition during the operating cycle." ITS LCO 3.1.1 requires SDM to be met. This changes the CTS by relocating the details that the core loading shall be limited to that which can be made subcritical "in the most reactive condition during the operating cycle" to the ITS Bases in the form of a discussion about how core reactivity varies during the fuel cycle and that the SDM verification should consider this behavior. 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 that the SDM shall be within limits. This is required all times during the operating cycle, including the most reactive condition during the operating cycle. The details of how SDM is calculated does not need to appear in the Specification in order for the requirement to apply. 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 Monticello Page 5 of 6 Attachment 1, Volume 6, Rev. 0, Page 12 of 231

Attachment 1, Volume 6, Rev. 0, Page 13 of 231 DISCUSSION OF CHANGES ITS 3.1.1, SHUTDOWN MARGIN (SDM) as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the CTS. LA.2 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.3.A.1 states, in part, "Sufficient control rods shall be withdrawn ... to demonstrate" reactivity margin is within the specified limit. ITS SR 3.1.1.1 states "Verify SDM to be within limits," but does not provide similar details of how to perform the verification. This changes the CTS by relocating the test method "Sufficient control rods shall be withdrawn ... to demonstrate" reactivity margin 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 that the SDM shall be within limits and to verify the SDM limits are met. The details of how SDM is performed does not need to be stated in the Specification in order for the requirement to apply. 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 CTS. LESS RESTRICTIVE CHANGES None Monticello Page 6 of 6 Attachment 1, Volume 6, Rev. 0, Page 13 of 231

Attachment 1, Volume 6, Rev. 0, Page 14 of 231 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) 0 Attachment 1, Volume 6, Rev. 0, Page 14 of 231

Attachment 1, Volume 6, Rev. 0, Page .15 of 231 SDM 3.1.1 K-) CTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) LCO 3.1.1 SDM shall be: 3.3.A.1, 4.3.A.1

a. 2:0.38j% Ak/k, with the highest worth control rod analytically (D determined or 0

b. 420.28jo% Ak/k, with the highest worth control rod determined by test.

0D 3.3A.1 APPLICABILITY: MODES 1, 2,3,4, and 5. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.3.G.1 A. SDM not within limits in A.1 Restore SDM to within 6 hours MODE 1 or 2. limits. 3.3.G.1 B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time of Condition A not met. 3.3.G.1 C. SDM not within limits in C.1 Initiate action to fully insert Immediately MODE 3. all insertable control rods. 3.3.G.1 D. SDM not within limits in D.1 Initiate action to fully insert Immediately MODE 4. all insertable control rods. AND D.2 Initiate action to restore 1 hour Rsecondari containment to OPERABLE status. 0D AND _ __ __ _ _ _ _ _ _ _ _ _ _ _ I BWR/4 STS 3.1.1-1 Rev. 3.0, 03/31/04 Attachment 1,Volume 6, Rev. 0, Page 15 of 231

Attachment 1, Volume 6, Rev. 0, Page 16 of 231 SDM 3.1.1 CTS ACTIONS (continued) . CONDITION REQUIRED ACTION COMPLETION TIME 3.3.G.1 D.3 Initiate action to restore one 1 hour standby gas treatment (SGT) subsystem to OPERABLE status. AND D.4 Initiate action to restore 1 hour isolation capability in each required~secondarzj 0 containment penetration flow path not isolated. 3.3.G.2 E. SDM not within limits in E.1 Suspend CORE Immediately MODE 5. ALTERATIONS except for control rod insertion and. fuel assembly removal. AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies. AND BWR/4 STS 3.1.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 16 of 231

Attachment 1, Volume 6, Rev. 0, Page 17 of 231 SDM 3.1.1 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.3.G.2 E.3 Initiate action to restore 1 hour Tsecondary containment to OPERABLE status. 0 AND E.4 Initiate action to restore one 1 hour SGT subsystem to OPERABLE status. AND E.5 Initiate action to restore 1 hour isolation capability in each required secondary containment penetra ion 0 flow path not isolated. BWR/4 STS 3.1.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 17 of 231

Attachment 1, Volume 6, Rev. 0, Page 18 of 231 SDM 3.1.1 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.3.A.1 SR 3.1.1.1 Verify SDM to be within limits. Prior to each in vessel fuel movement during fuel loading sequence AND Once within 4 hours after criticality following fuel movement* within the reactor pressure vessel or control rod replacement BWR/4 STS 3.1.1-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 18 of 231

Attachment 1, Volume 6, Rev. 0, Page 19 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1, SHUTDOWN MARGIN (SDM)

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

2. This punctuation correction has been made consistent with the Writer's Guide for the Standard Technical Specifications, NEI 01-03, Section 5.1.3.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 19 of 231

Attachment 1, Volume 6, Rev. 0, Page 20 of 231 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 20 of 231

Attachment 1, Volume 6, Rev. 0, Page 21 of 231 SDM B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) BASES BACKGROUND SDM requirements are specified to ensure: INSERT 1 0

a. The re aor can be made su ritical from all operaf g conditions and transie ts and Design Basis/vents,

b. The r activity transients as ociated with postulat d accident con itions are controllablewithin acceptable Iim's, and

c. Th reactor will be main ned sufficiently sub itical to preclude dveten crtic shtd l USAR, Section3.33.3 These requirement re satisfied by the control rods, as described in IMD= Ref. 1), which can compensate for the reactivity effects of the fuel and water temperature changes experienced during all operating 0D conditions.

APPLICABLE SAFETY [NET2 The control rod drop accident (CRDA) analysis (Refs. 2 and 3) assumes the core is subcritical with the highest worth control rod withdrawn. 0 ANALYSES Typically, the first conFol rod withdrawn has a v ry high reactivity worth and, should the core e critical during the with awal of the first control rod, the consequen s of a CRDA could exce d the fuel damage limits for a CRDA (see B ses for LCO 3.1.6. "Rod attern Control"). Also, SD

                       ' s assumed as an' itial condition fo hnecontrol rod removal error during adequate   refueling l          and fuel assembly insertion error during refueling SOM  anKRef.        accidents. The ana sis of these reativity insertio events operation               the refueling interlockslare CPFRABI when the reactor is in of      the refueling mode of operation. These interlocks prevent the withdrawal of more than one control rod from the core during refueling. (Special 0D consideration and requirements for multiple control rod withdrawal during refueling are covered in Special Operations LCO 3.10.6, "Multiple Control Rod Withdrawal - Refueling.") The anl               assumelthis condition is acceptable since the core will be shut down with the highest worth control rod withdrawn, if adequate SDM has been demonstrated.

l,hreb {g Fl E 0 Preventio or miti ation o reactivity insertio events is necessary to limit energy deposition in the fue preven ficant fuel damage, which could result in undue release of radioactivity. Adequate SDM ensures inadvertent criticalities and potential CRDAs involving high worth control 0 rods( (namely the firProd withdrawn will not cause significant fuel damage. SDM satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). BWR/4 STS B 3.1.1-1 Rev. 3.0, 03/31104 Attachment 1, Volume 6, Rev. 0, Page 21 of 231

Attachment 1, Volume 6, Rev. 0, Page 22 of 231 B 3.1.1 03 INSERT 1

a. The reactor core is designed so that control rod action, with the maximum worth control rod fully withdrawn and unavailable for use, is capable of bringing the reactor core subcritical and maintaining it so from any power level in the operating cycle; and

b. The reactor core and associated systems are designed to accommodate unit operational transients or maneuvers which might be expected without compromising safety and without fuel damage.

0, INSERT 2 Having sufficient SDM assures that the reactor will become and remain subcritical after all design basis accidents and transients. Insert Page B 3.1.1-1 Attachment 1, Volume 6, Rev. 0, Page 22 of 231

Attachment 1, Volume 6, Rev. 0, Page 23 of 231 SDM B 3.1.1 BASES LCO The specified SDM limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test when the highest worth control rod is determined by measurement. When SDM is demonstrated by calculations not associated with a test (e.g., to confirm SDM during the fuel loading sequence), additional margin is included to account for uncertainties in the calculation. To ensure adequate SDM Idurina th n Process a design margiTs included to (E account for uncertainties in the design calculations (Ref. . <1D APPLICABILITY In MODES 1 and 2, SDM must be provided because subcriticality with the

           -NSERTe3       ,worth control rod withdrawn is assumed in the CRDA analysis iiej. 2)In MODES 3 and 4, SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod. SDM is required in MODE 5 to prevent an open vessel, inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assembliesMor a fuel assembly insertion errorl(R           J. 3 ACTIONS           A.1 With SDM not within the limits of the LCO in MODE 1 or 2, SDM must be restored within 6 hours. Failure to meet the specified SDM may be caused by a control rod that cannot be inserted. The allowed Completion Time of 6 hours is acceptable, considering that the reactor can still be shut down, assuming no failures of additional control rods to insert, and the low probability of an event occurring during this interval.

B.1 If the SDM cannot be restored, the plant must be brought to MODE 3 in 12 hours, to prevent the potential for further reductions in available SDM (e.g., additional stuck control rods). The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. C.1 With SDM not within limits in MODE 3, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted. This action results in the least reactive condition for the core. BWR/4 STS B 3.1.1-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 23 of 231

Attachment 1, Volume 6, Rev. 0, Page 24 of 231 B 3.1.1 Q INSERT 3 and other design basis accidents and transients Insert Page B 3.1.1-2 Attachment 1, Volume 6, Rev. 0, Page 24 of 231

Attachment 1, Volume 6, Rev. 0, Page 25 of 231 SDM B 3.1.1 BASES ACTIONS (continued) D.1, D.2, D.3, and D.4 With SDM not within limits in MODE 4, the operator must immediately initiate action to fully insert all insertable control rods. Action must continue until all insertable control rods are fully inserted. This action results in the least reactive condition for the core. Action must also be ( initiated within 1 hour to provide means for control of potential radioactive releases. This includes ensuring secondary containment is OPERABLE;avaizable at least one Standby Gas Treatment (SGT) bsstem is OPERABLE; andgsecondary containmentjisolation capabi .e., at least one secondary containment isolation valve and ascte ntru entatin are OPERABLE, or other acceptable administrative controls to assueI t isolation capability)Fin 'each a'ssociated penetration flow ptn that is assumed to be isolated to mitigate radioactivity release . hi ay be performed as an administrative check, by examining logs or other NSr information, to determine if the components are out of service for maintenance or other reasons. It is not necessary to perform the surveillances needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE. E.1. E.2, E.3. E.4. and E.5 With SDM not within limits in MODE 5, the operator must immediately suspend CORE ALTERATIONS that could reduce SDM (e.g., insertion of fuel in the core or the withdrawal of control rods). Suspension of these activities shall not preclude completion of movement of a component to a safe condition. Inserting control rods or removing fuel from the core will reduce the total reactivity and are therefore excluded from the suspended actions. Action must also be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all insertable control rods in core cells containing one or more fuel assemblies have been fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and therefore do not have to be inserted. BWR/4 STS B 3.1.1-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 25 of 231

Attachment 1, Volume 6, Rev. 0, Page 26 of 231 B 3.1.1 Q INSERT4 These administrative controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated. CD INSERT5 (ensuring components are OPERABLE) Insert Page B 3.1.1-3 Attachment 1, Volume 6, Rev. 0, Page 26 of 231

Attachment 1, Volume 6, Rev. 0, Page 27 of 231 SDM B 3.1.1 BASES ACTIONS (continued) Action must also be initiated within I hour to provide means for control 0fs potential radioactive releases. This includes ensuring seconda available containment is OPERABLE; at least one SGT ubsystem issOPERABLE;, andasecondary containment~isolation capabilit i.e., at least one\J secon ary containment isolation valve an associated instrumentation are OPERABLE, or other acceptable administrative controls to assure isolation capabilitY in each associated penetration flow path not isolatedST that is assumed to be isolated to mitigate radioactivity releases. This may be performed as an administrative check, by examining logs or other INSET 7 information, to determine if the components are out of service for2 maintenance or other reasons. It is not necessary to perform the Surveillances as needed to demonstrate the OPERABILITY of the components. If, however, any required component is inoperable, then it must be restored to OPERABLE status. In this case, SRs may need to be performed to restore the component to OPERABLE status. Action must continue until all required components are OPERABLE. S This can be accomplished by a test, an SURVEILLANCE S R_ 3. 1.1.1 evaluation, or a combination of the two. REQUIREMENTS Adequate SDM must be em ed to ensure tha the reactor can be (Ii) made subcritical from any initial operating condition. {Adequate SDM is 2 demonstrated by testing before or durinq the first startup after fuel moveirentAcontrol rod replacement5Mos =ufinq within the re-a-cTo7 Pressure vesse. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Since core reactivity will vary during the cycle as a function of fuel depletion and poison bumup, the beginning of cycle (BOC) test must also account for changes In core-reactivity during the cycle. Therefore, to obtain the SDM, the initial measured valuevmust be increased by an 1afcUetj adder, "R", which is the difference between the calculated value of 'c ' maximum core reactivity during the operating cycle and the calculated BOC core reactivity. If the value of R is negative (that is, BOC is the most reactive point in the cycle), no correction to the BOC measured value is 2irequir (Re . For the SDM demonstrations that rely solely on calculation of the highest worth control rod, additional margin (0.10% Ak/k) must be added to the SDM limit as specifdihe COLN to 0 account for uncertainties in the calculation. a=, The SDM may be demonstrated during an ire nce control rod withdrawal, in which the highest worth control rod is analytically 0 determined, or during local determined by testing.

                                                        , where the highest worth control rod is 0D lonrlrdtst BWR/4 STS                              B 3.1.1-4                                       Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 27 of 231

Attachment 1, Volume 6, Rev. 0, Page 28 of 231 B 3.1.1 INSERT 6 These administrative controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device. Inthis way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated. Q INSERT 7 (ensuring components are OPERABLE) Insert Page B 3.1.14 Attachment 1, Volume 6, Rev. 0, Page 28 of 231

Attachment 1, Volume 6, Rev. 0, Page 29 of 231 SDM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS (continued) control rod Local tests require the withdrawal of out of sequence control rods. his tesin Thisul therefore require bypassing of the rod worth minimizer to allow the out of sequence withdrawal, and therefore additional o0 requirements must be met (see LCO 3.10.7, "Control Rod Testing - Operating"). analytical calculation of SDM may be used to The Frequency of 4 hours after reaching criticality is allowed to provide a assure the requirements of SR 3.1.1.1 are met. reasonable amount of time to perform the required calculations and have appropriate verification. During MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals. An evaluation of 0 each in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling. This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern. For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence. These bounding analyses include additional margins to the associated uncertainties. Spiral offload/reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle. Removing fuel from the core will always result in an increase in SDM.

                                                                                          .. I.-

REFERENCES 1. 10 CFR 50. x A. GDC 2

                                                                                    . I Ub   ' QVIOVF O.J.;.J I (D

C\J '2.SAR, Section 38 7 0

3. NEDE-24011-P-~, "General Electric Standard Application for Reactor Fuel," Supplement for United States, Section S.2.2.3.1ID ISe t r (revIsion specified InSpecification 5.6.3) ]

0'

4. JFSAR, Section [1 .1.13].

5. /FSAR, Section [j/5.1.14]. 0

                            'SA        R, Section A>                   [.4.1                                                                  30 NEDE-24011-P-A4, "General ElectricStandard Application for Reactor Fuel," Section 3.2.4.1giSepte                        I      .

I 0D BWR/4 STS B 3.1.1-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 29 of 231

Attachment 1, Volume 6, Rev. 0, Page 30 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.1 BASES, SHUTDOWN MARGIN (SDM)

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

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

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

4. The Bases have been changed to reflect the Specification.

5. Typographical/grammatical error corrected.

Monticello Page 1 of I Attachment 1, Volume 6, Rev. 0, Page 30 of 231

Attachment 1, Volume 6, Rev. 0, Page 31 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 6, Rev. 0, Page 31 of 231

Attachment 1, Volume 6, Rev. 0, Page 32 of 231 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.1, SHUTDOWN MARGIN (SDM) There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 6, Rev. 0, Page 32 of 231

, Volume 6, Rev. 0, Page 33 of 231 ATTACHMENT 2 ITS 3.1.2, Reactivity Anomalies Attachment 1, Volume 6, Rev. 0, Page 33 of 231

Attachment 1, Volume 6, Rev. 0, Page 34 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 6, Rev. 0, Page 34 of 231

C C, C, 0 ITS 3.1.2 ITS ITS a 3.0 IMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS / Oeqwit un

                                                                                                                                            ,~1ln      '                        owing oriim E. ReactMty Anomalfes                                                 E. Reactivity Anomalies       /                    .

3.1.2 _ a: e emn-- ihn fthe tA1~---

0) St nflde iim IA n ilinl n thRrr~rri- s_ I Dturlgg the stdalst Drourt _q  !!2 to~rpressr VW,=eio eac-~ M.2) 0)

SR 3.1.2.1 actua control rod Inventory will bei perodical compa *If ollowingas aclrodrontrolry oryles h od repiavernent LCO 3.1.2 It -ainormalezA FVDff6U~predction of the Irwtory. R 3.1.2.1 shall be compared to a m .- I CD CD ffe erence e-xcees one per cent. la'k.le W MJ base data for rea mntoing dun su en Dower overatn shall not be permitted un2 the caus ACTION A as been evaluated and a nate co ective action I ower operation rougho ut the hiet cite 0 has been completed.] Add proposed power operamIgconailionsq the actu rdcngato I> ACTIONS A and B L1 will be compared to the configuraition exetdaL I upon nonropfiatelri~ecled pas ail 0 Applicability. 0 E' tD CD 0) tD 0) i 9) to to CD Cfl 0 0

                                                                                                                                                                                                      -9 Cj, 3.3/4.3                                                                                                83               511/84 Amendment No. 24 Page 1 of 1

Attachment 1, Volume 6, Rev. 0, Page 36 of 231 DISCUSSION OF CHANGES ITS 3.1.2, REACTIVITY ANOMALIES 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 4.3.E states, in part, the reactivity anomaly Surveillance must be performed "During the startup test program." ITS SR 3.1.2.1 does not include this requirement. This changes the CTS by deleting the requirement to perform this test "During the startup test program." The Monticello startup test program has been completed and is not required to be performed again. Thus, there is no need to retain this requirement in the ITS. This change is considered a presentation preference change only and, as such, is considered an administrative change. MORE RESTRICTIVE CHANGES M.1 CTS 4.3.E states that the reactivity anomaly Surveillance shall be performed "at each" startup following refueling outages. The ITS SR 3.1.2.1 Surveillance Frequency states that the Surveillance is performed "Once within 24 hours after reaching equilibrium conditions" following startup after fuel movement within the reactor pressure vessel or control rod replacement. This changes the CTS by providing an explicit time period to complete the Surveillance following a startup. This change to the "following refueling outage" portion of the frequency is discussed in DOC M.2. The purpose of CTS 4.3.E is to verify the core reactivity after in-vessel operations, which could have significantly altered the core reactivity. A specific time for completing the reactivity anomaly surveillance CTS 4.3.E is proposed to clarify when "during the first startup" the test must be completed. This test is performed by comparing the difference between the actual control rod inventory and the predicted control rod inventory as a function of cycle exposure while at steady state reactor power conditions. Therefore, 24 hours after reaching these conditions is provided as a reasonable time to perform the required calculations and complete the appropriate verification, and thus this time is considered acceptable. Therefore, this change is considered a more restrictive change since a finite completion time is now provided. M.2 CTS 4.3.E states, in part,.that the reactivity anomaly Surveillance shall be performed "following refueling outages." This Frequency is changed in ITS SR 3.1.2.1 to be "after fuel movement within the reactor pressure vessel or control rod replacement." This changes the CTS by clearly defining the activities after which the reactivity anomaly Surveillance should be performed. Monticello Page 1 of 4 Attachment 1, Volume 6, Rev. 0, Page 36 of 231

Attachment 1, Volume 6, Rev. 0, Page 37 of 231 DISCUSSION OF CHANGES ITS 3.1.2, REACTIVITY ANOMALIES The purpose of CTS 4.3.E is to verify the core reactivity after in-vessel operations that could have altered the core reactivity. During refueling outages, core reactivity is normally significantly altered. However, conditions could arise mid-cycle that require replacing a fuel assembly or control rod, and the CTS would not require this Surveillance to be performed during the subsequent reactor startup. This mid-cycle replacement has the potential of altering core reactivity. The ITS words cover both planned refueling outages and other outages where CORE ALTERATIONS may occur, thus this change is considered acceptable. This change is considered a more restrictive change since the Surveillance will be required under more conditions than is currently required. M.3 CTS 3.3.E requires the reactivity anomaly requirements to be met in the 'reactor power operation" condition. ITS LCO 3.1.2 is Applicable in MODES 1 and 2. This changes the CTS by requiring the reactivity anomaly limit to be met in MODE 2 < 1% RATED THERMAL POWER (RTP). The purpose of CTS 3.3.E is to ensure plant operation is maintained within the assumptions of the safety analyses. This change expands the Applicability to require the reactivity anomaly limit to be met at all times when in MODE 2, instead of when > 1% RTP (the CTS 1.0.0 definition states that Power Operation is when reactor power is > 1%RTP). This change is acceptable since the reactivity anomaly must be met in MODE 2 because control rods are typically being withdrawn during a startup. This change is designated as more restrictive because the LCO will be applicable under more reactor conditions. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.3.E states, in part, "At a specific steady state base condition" the reactor actual control rod inventory will be periodically compared to a "normalized computed" prediction of the inventory. CTS 3.3.E also implies that the reactivity difference shall be shall be within +/- 1%Ak/k. CTS 4.3.E states, in part, the actual rod inventory shall be compared to a "normalized computed" prediction of inventory and that "These comparisons will be used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle." Furthermore, the actual rod configuration will be compared to the configuration expected 'based upon appropriately corrected past data." ITS LCO 3.1.2 states "The reactivity difference between the monitored control rod inventory and the predicted control rod inventory shall be within +/- 1% Ak/k." ITS SR 3.1.2.1 states "Verify core reactivity difference between the monitored control rod inventory and the predicted control rod inventory is within +/- 1% Ak/k." This changes the CTS by relocating these details for performing the reactivity anomaly Surveillance to the ITS Bases. Monticello Page 2 of 4 Attachment 1, Volume 6, Rev. 0, Page 37 of 231

Attachment 1, Volume 6, Rev. 0, Page 38 of 231 DISCUSSION OF CHANGES ITS 3.1.2, REACTIVITY ANOMALIES The removal of these details for evaluating surveillance requirements from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS LCO 3.1.2 still retains the requirement that "The reactivity difference between the monitored control rod inventory and the predicted control rod inventory shall be within +/- 1%Ak/k" and ITS SR 3.1.2.1 still retains the requirement to "Verify core reactivity difference between the monitored control rod inventory and the predicted control rod inventory is within

      +/- 1% Ak/k." Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5.

This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.A (Category 4 - Relaxation of Required Action) CTS 3.3.E states, in part, "If the difference exceeds one per cent, delta k, reactor power operation shall not be permitted until the cause has been evaluated and appropriate corrective action has been completed." This effectively requires an immediate unit shutdown if the reactivity difference is greater than 1%Ak/k. ITS 3.1.2 ACTIONS A and B cover the condition when the reactivity anomaly criterion is not met. ITS 3.1.2 ACTION A requires restoration of the core reactivity difference to within limit in 72 hours. If this Required Action and Completion Time are not met, ITS 3.1.2 ACTION B requires the unit to be in MODE 3 in 12 hours. This changes the CTS by allowing 72 hours to restore the reactivity difference before commencing a shutdown. The purpose of the ITS 3.1.2 ACTIONS is to allow time to confirm that a reactivity anomaly is of no concern or to correct the problem. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the 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 repair period. The ITS 3.1.2 compensatory actions allow 72 hours of plant operations in MODE 1 and 2 before requiring a reactor shutdown. According to ITS 3.1.2 Required Action A.1 Bases restoration to within the limit could be performed by an evaluation of the core design and safety analysis to determine the reason for the anomaly. This evaluation normally reviews the core conditions to determine their consistency with input to design calculations. Measured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions. This change is acceptable Monticello . Page 3 of 4 Attachment 1, Volume 6, Rev. 0, Page 38 of 231

Attachment 1, Volume 6, Rev. 0, Page 39 of 231 DISCUSSION OF CHANGES ITS 3.1.2, REACTIVITY ANOMALIES since the current requirement that does not allow reactor power operation to continue is overly restrictive because in most cases any reactivity anomaly is normally indicative of incorrect analysis inputs or assumptions of fuel reactivity used in the analysis. A determination and explanation of the cause of the anomaly would normally involve a fuel analysis department and the fuel vendor. Contacting and obtaining the necessary input may require a time period much longer than one shift (particularly on weekends and holidays). Since SHUTDOWN MARGIN has typically been demonstrated by test prior to reaching the conditions at which this Surveillance is performed, the safety impact of the extended time for evaluation is negligible. Given these considerations, the. ITS allows this time to be 72 hours. 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 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) The Frequency of the reactivity anomaly Surveillance in CTS 4.3.E is at least every "equivalent full power month" (approximately 611 MWDIT, where T is a short ton), and it is required to be performed "At specific power operating conditions." ITS SR 3.1.2.1 requires this same test to be performed every 1000 MWD/T during operations in MODE 1. This changes the CTS by extending the Surveillance Frequency from 611 MWD/T to 1000 MWDIT, and specifies that the "specific power operating condition" is MODE 1. The purpose of CTS 4.3.E isto verify the reactivity difference iswithin limit. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. This changes extends the Surveillance Frequency for the reactivity anomaly test. This change is acceptable based on the slow rate of core reactivity changes due to fuel depletion and operating experience related to variations in core reactivity. The proposed change is consistent with the ISTS. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. Monticello Page 4 of 4 Attachment 1, Volume 6, Rev. 0, Page 39 of 231

Attachment 1, Volume 6, Rev. 0, Page 40 of 231 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 40 of 231

Attachment 1, Volume 6, Rev. 0, Page 41 of 231 Reactivity Anomalies 3.1.2 CTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Reactivity Anomalies nsi and the 3.3.E LCO 3.1.2 The reactivitygdifferencejbetween thegmonitoredlrod predicted shall be within 1% Ak/k. 0D APPLICABILITY: MODES 1 and 2.

                                                                                                            .i ACTIONS CONDITION                        REQUIRED ACTION               COMPLETION TIME 3.3.E   A. Core reactivity                A.1       Restore core reactivity     72 hours 1differencefnot within limit.

adifferenceMto within limit. 03 3.3.E B. Required Action and B.1 Be in MODE 3. 12 hours associated Completion Time not met. BWR/4 STS 3.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 41 of 231

Attachment 1, Volume 6, Rev. 0, Page 42 of 231 Reactivity Anomalies 3.1;2 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE = FREQUENCY 3.3.E, SR 3.1.2.1 Verify cor actvt[differencejbetweenth Once within 4.3.E Rmonitored is within i 1% and the predicted Ak/k. 24 hours after reaching 0D equilibrium conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 1000 MWD/T thereafter during operations in MODE I BWRI4 STS 3.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 42 of 231

Attachment 1, Volume 6, Rev. 0, Page 43 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.2, REACTIVITY ANOMALIES

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

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 43 of 231

Attachment 1, Volume 6, Rev. 0, Page 44 of 231 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 44 of 231

Attachment 1, Volume 6, Rev. 0, Page 45 of 231 Reactivity Anomalies B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 Reactivity Anomalies BASES .BACKGROUND In accordance wit GOC 26, GDC 28, and GDC 29 (fef. 1), reactivity shall be controlla e such that subcriticality is maintai ed under cold conditions and a~eptable fuel design limits are not Exceeded during 0 norma prto and anticipated operatoa ocurne. erefore, eactivity Anom is used as a measure of the predicted versus measured core reactivity during power operation. The continual 0 Ies confirmation of core reactivity is necessary to ensure that the Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity anomaly could be the result of unanticipated changes in fuel reactivity or control rod worth or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") in assuring the reactor can be brought safely to cold, subcritical conditions. When the reactor core is critical in normal-ower aor a reactivity () balance exists and the net reactivity is zero. A comparison of predicted and measured reactivity is convenient under such a balance, since parameters are being maintained relatively stable under steady state power conditions. The positive reactivity inherent in the core design is balanced by the negative reactivity of the control components, thermal feedback, neutron leakage, and materials in the core that absorb neutrons, such as burnable absorbers, producing zero net reactivity. eg. ga. oinia In order to achieve the required fuel cycle energy output, the uranium enrichment in the new fuel loading and the fuel loaded in the previous cycles provide excess positive reactivity beyond that required to sustain steady state operation at the beginning of cycle (BOC). When the reacto is critical at RTP and operating moderator temperature, the exce positive reactivity is compensated by burnable absorbers if n ), control rods, and whatever neutron poisons (mainly xenon and samarium) are present in the fuel. The predicted core reactivity as represented by control rod er ,is calculated by a 3D core simulator code as a function of cycle exposure. This calculation is performed for projected operating states and conditions throughout the cycle. The core reactivity is determined from control rod deWies for actual plant conditions and is then compared to the predicted value for the cycle exposure.-n3 BWR/4 STS B 3.1.2-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 45 of 231

Attachment 1, Volume 6, Rev. 0, Page 46 of 231 B 3.1.2 0 INSERT 1 In accordance with USAR, Section 1.2.2 (Ref. 1), the reactor core is designed so that control rod action, with the maximum worth control rod fully withdrawn and unavailable for use, is capable of bringing the reactor core subcritical and maintaining it so from any power level in the operating cycle. In addition, the reactor core and associated systems are designed to accommodate unit operational transients or maneuvers that might be expected without compromising safety and without fuel damage. Insert Page B 3.1.2-1 Attachment 1, Volume 6, Rev. 0, Page 46 of 231

Attachment 1, Volume 6, Rev. 0, Page 47 of 231 Reactivity Anomalies B 3.1.2 BASES APPLICABLE SAFETY Accurate prediction of core reactivity isfeitheZ an assumption in the accident analysis evaluations (Ref. 2). oimplicit In particular, 0D ANALYSES SDM and reactivity transients, such as control rod withdrawal accidents or rod drop accidents, are very sensitive to accurate prediction of core reactivity. These accident analysis evaluations rely on computer codes that have been qualified against available test data, operating plant data, and analytical benchmarks. Monitoring reactivity anomaly provides additional assurance that the nuclear methods provide an accurate representation of the core reactivity. The comparison between measured and predicted initial core reactivity provides a normalization for the calculational models used to predict core reactivity. If the measured and predicted rod Ue si or identical core conditions at BOC do not reasonably agree, then the assumptions used in the reload cycle design analysis or the calculation models used to predict rod Reit may not be accurate. If reasonable agreement between Inventory measured and predicted core reactivity exists at BOC, then the prediction may be normalized to the measured value. Thereafter, any significant deviations in the measured od Ke irom the predictedirod de it hat 3 develop during fuel depletion may be an indication that the assumptions of the DBA and transient analyses are no longer valid, or that an inventory unexpected change in core conditions has occurred. . Reactivity nomalies satist riterion 2 of 10 CFR 50.36(c)(2)(ii). 0 LCO The reactivity anomaly limit is established to ensure plant operation is maintained within the assumptions of the safety analyses. Large differences between monitored and predicted core reactivity may indicate that the assumptions of the DBA and transient analyses are no longer valid, or that the uncertainties in the "Nuclear Design Methodology' are than ex ected. A limit on the difference between the monitored and the predictedirod dejityQf +/- 1% Ak/k has been established based on engineering judgment. A > 1% deviation in reactivity from that predicted is larger than expected for normal operation and should iinventory therefore be evaluated. APPLICABILITY In MODE 1, most of the control rods are withdrawn and steady state operation is typically achieved. Under these conditions, the comparison between predicted and monitored core reactivity provides an effective measure of the reactivity anomaly. In MODE 2, control rods are typically being withdrawn during a startup. In MODES 3 and 4, all control rods are fully inserted and therefore the reactor is in the least reactive state, where monitoring core reactivity is not necessary. In MODE 5, fuel loading BWR/4 STS B 3.1.2-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 47 of 231

Attachment 1, Volume 6, Rev. 0, Page 48 of 231 Reactivity Anomalies B 3.1.2 BASES APPLICABILITY (continued) results in a continually changing core reactivity. SDM requirements _ (LCO 3.1.1) ensure that fuel movements are performed within the bounds _eor rof thesafety analysis, and an SDM demonstration is requiredrduring the first startup following operations that could have altered core reactivity (e.g., fuel movement, control rod replacement, shuffling). The SDM test, required by LCO 3.1.1, provides a direct comparison of the predicted and monitored core reactivity at cold conditions; therefore,/eactivityapnomali ( is not required during these conditions. e ACTIONS A.1 Should an anomaly develop between measured and predicted core reactivity, the core reactivity difference must be restored to within the limit to ensure continued operation is within the core design assumptions. Restoration to within the limit could be performed by an evaluation of the core design and safety analysis to determine the reason for the anomaly. This evaluation normally reviews the core conditions to determine their consistency with input to design calculations. Measured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions. The required Completion Time of 72 hours is based on the low probability of a DBA occurring during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the core design and safety analysis. B.1 Ifthe core reactivity cannot be restored to within the 1% Ak/k limit, the plant must be brought to a MODE Inwhich the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Verifying the reactivity difference between the monitored and predicted the assumptions of the DBAan coin Tqloe transien~tanalvseg nirgSyst-e-53calculateshro n rneoy: ( efeco plant instnumentation.A/I oiinotierom

                                                         ~~~~~~~~~edestio
     .I    . .    ~~comparison of the monitored rod                   eprdcdro BWR/4 STS                               B 3.1.2-3                                  Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 48 of 231

Attachment 1, Volume 6, Rev. 0, Page 49 of 231 Reactivity Anomalies B 3.1.2 BASES SURVEILLANCE REQUIREMENTS (continued) the same cycle exposure is used to calculate the reactivity difference. The comparison is required when the core reactivity has potentially changed by a significant amount. This may occur following a refueling in which new fuel assemblies are loaded, fuel assemblies are shuffled within the core, or control rods are replaced or shuffled. Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Also, core reactivity changes during the cycle. The 24 hour interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xeno n c n o t o uch that an accurate comparison between nvent~rY the monitored and predictedkroddefitycan be made. For3the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes) at 2 75% RTP have been obtained. The 1000 MWD/T Frequency was f2 7 3 developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity. This comparison requires the core to be operating at power levels which minimize the uncertainties and measurement errors, in order to obtain meaningful results. Therefore, the comparison is only done when in MODE 1. REFERENCES 1. 110 CFR 50.A ix A.1.22 0D c 2SSAR, Chapter 0 BWRI4 STS B 3.1.2-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 49 of 231

Attachment 1, Volume 6, Rev. 0, Page 50 of 231 B 3.1.2 0 INSERT 2 At a specific steady state base condition the actual control rod inventory will be periodically compared to a normalized computed prediction of the inventory. The comparisons will be used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle. Insert Page B 3.1.24 Attachment 1, Volume 6, Rev. 0, Page 50 of 231

Attachment 1, Volume 6, Rev. 0, Page 51 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.2 BASES, REACTIVITY ANOMALIES

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

2. Typographical/grammatical error corrected.

3. Changes are made to reflect those changes made to the Specification.

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

5. Changes are made to be consistent with the Specification.

6. Editorial change made for clarity.

Monticello Page 1 of I Attachment 1, Volume 6, Rev. 0, Page 51 of 231

Attachment 1, Volume 6, Rev. 0, Page 52 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 6, Rev. 0, Page 52 of 231

Attachment 1, Volume 6, Rev. 0, Page 53 of 231 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.2, REACTIVITY ANOMALIES There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 53 of 231

, Volume 6, Rev. 0, Page 54 of 231 ATTACHMENT 3 ITS 3.1.3, Control Rod OPERABILITY , Volume 6, Rev. 0, Page 54 of 231

Attachment 1, Volume 6, Rev. 0, Page 55 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 6, Rev. 0, Page 55 of 231

C C, C ITS 3.1.3 0 01 0 C) AD

r 0

on discovery of Condition A concrrent on with THERMAL POWER greater than the low power -, setpoint of t RWM C a -U ;U 0) E, X 0CD 0 0

;u ci
                                                                            -ut CD D

to -n Ca 0 N Cs).

                                                                            -4' 3.314.3     77            7112193 Amendment No. 86 Page 1 of 4

C eC C. ITS 3.1.3 ITS 3.0 UMING CONDImONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS B. Control Rod Withdrawal CA SR 3.1.3.5 1. lThe couprng Intotrity shall be verflied for each 0S withdrawn control rod as follows: 0 3 (a~lhe rod Is tull withdrawni Isubseuent lrrriueiinro oul~aWfbev1 0 rD that the drive does rdt go to tereootoe tD posilion: anld/\ aa 0 CD and prior to dedlaring controiiro /

                                                         "OPERABLE afer work on control 3                                                         rod or CRD System that could affect coupling 0

0 P) 3 (0 0 tD lu 0 -41 0) CA) 3.314.3 78 1/9181 Amendment No. 0 Page 2 of 4

c C CI ITS 3.1.3 3.0 UMING CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUiREMENTS (b) when the rod s wthdrawn e first time subsequent to each reu ing outage, observe a) discernible response

                                                                                                                 !'         e nuclear                                      C1)

Instrumentation. H for Initial rods wher response Is not mible, subsequent exercising of th e rods after the reactor Is 0 critical shall b erformed to observe nuclear 0 Instrumentatl'n response. CD

2. The control rod drive housing support system shall be In 2. The control rod drive housing support system shall be 79 CD 0

place during reactor power operation and when the reactor coolant system is pressurized above atmospheric pressure with fuel In the reactor vessel. unless all operable control rods are fully Inserted and Soecification 3.3A1 Is met. I-Inspected after reassembly and the results of the Inspection recorded. See CTS 314.3.3.2

                                                                                                                                                                        }I CD 0

3.(a)Controt rod withdrawal sequences shall be established 3.(a) To consider the rod worth minimizer operable, the -u CD so that the maximum calculated reactivity that could be following steps must be performed: C1) added by dropout of any Increment of any one control (I) The control rod withdrawal sequence forthe rod Co

a blade wig not make the core more than 1.3% AX CD worth minimizer computer shall be verified as supercrtXcal.I correct (O) The rod worth minimizer computer on-line diagnostic -{See ITS 3.32.1}

test shall be successfully completed. QI Proper annunciation of the selection error of at least -0 CA) one out-of-sequence control rod hI each fully CA) Inserted group shal be verified. N f See ITS 3.1.6 } 3.314.3 79 1/9/81 Amendment No. 0 Page 3 of 4

C C C ITS 3.1.3 0 ITS 3.0 LIMmNG CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS F. Scram Discharge Volume F. Scram Discharge Volume

1. During reactor operation, the scram discharge The scram discharge volume vent and drain valves sh s volume vent and drain valves shall be operable. be cydled quarterly.

M except as specified below. One per operating cyde verify the scram discharge 0 CD volume vent and drain valves dose within 30 seconds

2. If any scam dischage volume vent or drain valve Is after receipt of a reactor scram signal and open when CD 3 made or found Inoperable, the Integrity of the scram the scram is reset discharge volume shall be maintained by either

a.

• Verifying da* for a period not to exceed 7 days, the operability of the redundant valve(s),

0 or -{ See ITS 3.1.8 } 0 CD 0 b. Maintaining the Inoperable valve(s), or the -o associated redundant valve(s), Inthe dosed 0 positIon. Periodically the Inoperable and the redundant valve(s) may both be Inthe open positIon to alnow draining the scram discharge volume. CD to a) IU If a or b above cannot be met, at least all but one operable control rods (not including rods removed 0) CD per specification 3.10.E or inoperable rods allowed 0

       / 1\               by 3.3.A.2) shall be fully Inserted within ten hours.

01 to G. Required Action l ' CA) If Specfications 3.3A 'throughD above are otme, n (exceptwhenthereactormode1 -9' ACTION E orderly shutdown shall be initiated and have reactor In swchIs Inthe Refuel position) (A) the cold s di conditinwithin hours. 12\ 3.3!4.3 L 83a 5/1/84 Amendment No. 24 Apd s .d C ce'drt o Page 4 of 4

Attachment 1, Volume 6, Rev. 0, Page 60 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY 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 3/4.3.A.2 provides requirements for stuck control rods. CTS 3/4.3.B.1 provides requirements for control rod coupling. ITS 3.1.3 provides requirements for each control rod. ITS LCO 3.1.3 states "Each control rod shall be OPERABLE." This changes the CTS by combining the OPERABILITY requirements for control rods into one Specification and adding an explicit statement concerning control rod OPERABILITY. Additional aspects of control rod OPERABILITY are also added in accordance with DOC M.4. The purpose of ITS 3.1.3 is to include in one Specification all conditions that can affect the ability of the control rods to provide the necessary reactivity insertion. This change is acceptable because it provides a clear statement concerning the OPERABILITY requirements for each control rod. This change is acceptable since there are no technical changes. Therefore, this change is considered a presentation preference change only and, as such, is considered an administrative change. A.3 CTS 3.3.A.2.(a) states that the directional control valves for inoperable control rods shall be disarmed. CTS 3.3.B.1 states that each control rod shall be coupled to its drive or completely inserted and the directional control valves disarmed. These CTS Actions do not limit the number of control rods to which these Actions apply. ITS 3.1.3 ACTIONS Note states "Separate Condition entry is allowed for each control rod." This changes the CTS by adding an explicit Note for separate condition entry for each control rod. The purpose of CTS 3.3.A.2.(a) and CTS 3.3.B.1, in part, is to provide compensatory actions for an inoperable control rod on an individual basis. This change provides more explicit instructions for proper application of the ACTIONS for Technical Specification compliance. In conjunction with the proposed Specification 1.3, "Completion Times," this Note provides direction consistent with the intent of the existing ACTIONS for inoperable control rods. It is intended

• that each inoperable control rod be allowed a specified period of time in which compliance with certain limits is verified and, when necessary, the control rod is fully inserted and disarmed. Therefore, this change is considered a presentation preference change only and, as such, is considered an administrative change.

A.4 CTS 3.3.A.2.(a) states, in part, "The directional control valves for inoperable control rods shall be disarmed." CTS 3.3.B.1 states, in part, "Each control rod shall be coupled to its drive or completely inserted and the directional control valves disarmed." These compensatory actions are covered in ITS 3.1.3 ACTION A for stuck rods and ITS 3.1.3 ACTION C for coupling inoperabilities. In Monticello Page 1 of 11 Attachment 1, Volume 6, Rev. 0, Page 60 of 231

Attachment 1, Volume 6, Rev. 0, Page 61 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY addition, these ITS 3.1.3 ACTIONS include a Note that states rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block Instrumentation," if required, to allow continued operation. This changes the CTS by adding these clarification Notes. The purpose of the ITS 3.1.3 ACTION A and ITS 3.1.3 Required Action C.1 Notes are to allow continued unit operation with inoperable control rods. This change is acceptable since CTS 3/4.3.B.3.(b) allows the RWM to be bypassed. To complete the associated actions the RWM may be required to be bypassed. This note is informative in that the RWM may be bypassed at any time, provided the proper ACTIONS of CTS 3/4.3.B.3.(b) (ITS 3.3.2.1), the RWM Specification, are taken. This is a human factors consideration to assure clarity of the requirement and allowance. Therefore, this change is considered a presentation preference change only and, as such, is considered an administrative change. A.5 CTS 3.3.A.2 does not explicitly state when the stuck control rod requirements are required to be met. However, CTS 3.3.A.2.(b) states that the reactor should be brought to hot shutdown under certain situations. ITS 3.1.3 is applicable in MODES 1 and 2. This changes the CTS by explicitly stating the Applicability. The purpose of CTS 3.3.A.2 isto limit the number of stuck control rods. This change is acceptable because the proposed MODE is consistent with the current shutdown condition. This change is considered a presentation preference change only and, as such, is considered an administrative change. A.6 These changes to CTS 3.3.G 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. MORE RESTRICTIVE CHANGES M.1 CTS 3.3.A.2.(a) states, in part, "The directional control valves for inoperable control rods shall be disarmed electrically and the rods shall be in such positions that Specification 3.3.A.1 is met." CTS 3.3.B.1 states, in part, "Each control rod coupled to its drive or completely inserted and the directional control valves disarmed." ITS 3.1.3 ACTION A covers the condition of one withdrawn control rod stuck, and requires the immediate verification that the stuck control rod separation criteria is met (Required Action A.1), the disarming of the associated control rod drive within 2 hours (Required Action A.2), and the performance of SR 3.1.1.1 (SHUTDOWN MARGIN verification test) within 72 hours (Required Action A.4). ITS 3.1.3 ACTION C covers the condition of one or more control rods inoperable for reasons other than a stuck control rod, and requires fully inserting an inoperable control rod within 3 hours (Required Action C.1) and disarming the associated control rod drive within 4 hours (Required Action C.2). This changes the CTS by adding finite times to perform the Required Actions and adds a new Required Action to verify stuck control rod separation criteria is met. The purpose of CTS 3.3.A.2.(a) and CTS 3.3.B.1 are to place the unit in a safe condition when control rods are inoperable. This change is acceptable since the Monticello Page 2 of 11 Attachment 1, Volume 6, Rev. 0, Page 61 of 231

Attachment 1, Volume 6, Rev. 0, Page 62 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY proposed Completion Times for performing the stuck control rod separation criteria verification, disarming control rod drives, Inserting inoperable control rods, and for performing a SHUTDOWN MARGIN test are consistent with industry practice and can be safely accomplished. The stuck rod separation criteria is defined in the ITS 3.1.3 Bases. This additional requirement ensures the local scram reactivity will be met with a stuck rod. Disarming a control rod as required by CTS 3.3.A.2.(a) and CTS 3.3.B.1 involves personnel actions by other than control room operating personnel. This process will require coordination of personnel and preparation of equipment, and potentially require anti-contamination "dress-out," in addition to the actual procedure of disarming the control rod. The proposed Completion Times are acceptable in recognition of the potential time required to complete this task. The proposed time to disarm a control rod does not represent a significant safety concern as the control rod is already in an acceptable position and the ACTION to disarm is solely a mechanism for precluding the potential for damage to the control rod drive mechanism. With a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Also, a notch test is required by ITS 3.1.3 Required Action A.3 for each remaining withdrawn control rod to ensure that no additional control rods are stuck. Given these considerations, the time to demonstrate SHUTDOWN MARGIN in ITS 3.1.3 Required Action A.4 is 72 hours. This Completion Time provides a reasonable time to perform the analysis or test. The change has been designated as more restrictive because it adds an explicit Required Action Completion Times and adds a new Required Action to verify stuck rod separation criteria. M.2 CTS 3.3.A.2.(c) allows continued operation with up to six non-fully inserted, inoperable (i.e., stuck) control rods. CTS 4.3.A.2.(c) states "If power operation is continuing with two or more non-fully inserted control rods that are inoperable, each operable fully or partially withdrawn control rod shall be exercised at least one notch every 24 hours." ITS 3.1.3 ACTION B requires the unit to be in MODE 3 with two stuck control rods. This changes the CTS by changing the number of non-fully inserted control rods that can be inoperable (i.e., stuck) and continue operations in MODE 1 and 2 from "six" to "one." The purpose of CTS 3.3.A.2.(c) is to limit the number of non-fully inserted stuck control rods. This change is acceptable since with two or more withdrawn control rods stuck the unit must be brought to MODE 3. The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert. The CTS 4.3.A.2.(c) requirement to test all OPERABLE fully or partially withdrawn control rods is not necessary since operation is not allowed with two or more non-fully inserted stuck control rods. This change is designated as more restrictive because it changes the number of non-fully inserted (i.e., stuck) inoperable control rods in MODE 1 and 2 from "six" to "one." M.3 CTS 3.3.A.2.(c), in part, requires the unit to be in hot shutdown (MODE 3) in within 48 hours. ITS 3.1.3 ACTION B requires the unit to be in MODE 3 within 12 hours. This changes the CTS by changing the time to reach MODE 3 from 48 hours to 12 hours. Monticello Page 3 of 11 Attachment 1, Volume 6, Rev. 0, Page 62 of 231

Attachment 1, Volume 6, Rev. 0, Page 63 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY The purpose of CTS 3.3.A.2.(c), in part, is to provide the appropriate time for the unit to be in MODE 3. This change is acceptable since the allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. This change is designated as more restrictive because it reduces the required time to achieve MODE 2 from 48 hours to 12 hours. M.4 CTS 3/4.3.A.2 provides requirements for stuck control rods. CTS 3/4.3.B.1 provides requirements for control rod coupling. There are no requirements associated with the determination of each control rod position and maximum scram time of the control rods. ITS 3.1.3 includes two Surveillance Requirements to cover these requirements. ITS SR 3.1.3.1 requires the determination of the position of each control rod every 24 hours. ITS SR 3.1.3.4 requires the verification that each control rod scram time from the fully withdrawn position to notch position 06 is within limit (i.e. < 7 seconds) in accordance with SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. This changes the CTS by adding two additional OPERABILITY requirements for the control rods (i.e., maximum scram insertion time, and control rod position). The purpose of the new Surveillance Requirements is to ensure important aspects of control rod OPERABILITY are monitored on a regular basis. This change is acceptable because it provides additional assurance that the control rods will provide its scram function (i.e., scram insertion time and its position will be known). This change is designated as more restrictive because it adds two new Surveillance Requirements to the CTS. M.5 CTS 4.3.A.2.(a) requires each fully or partially withdrawn operable control rod to be "exercised" at least one notch. CTS 4.3.A.2.(b) requires the same testing when a control rod is found to be stuck. ITS SR 3.1.3.2, ITS SR 3.1.3.3, and ITS 3.1.3 Required Action A.3 requires the same testing however the control rods must be "inserted" in lieu of "exercised." This changes the CTS by requiring the OPERABLE withdrawn control rods to be "inserted" one notch instead of "exercised" one notch. The purpose of CTS 4.3.A.2.(a) and CTS 4.3.A.2.(b) are to periodically verify that each withdrawn OPERABLE control rod is not stuck and is free to insert on a scram signal. This change is acceptable because it provides additional assurance that the control rods will provide their scram function. The existing requirement to exercise the control rod could be met by control rod withdrawal. It is conceivable that a mechanism causing binding of the control rod that prevents insertion could exist and that a withdrawal test would not detect the problem. Since the purpose of the test isto assure scram insertion capability, restricting the test to control rod insertion provides an increased likelihood of this test detecting a problem that impacts insertion capability. This change is designated as more restrictive because it changes the CTS acceptance criteria. M.6 CTS 3.3.A.2 provides requirements for stuck control rods. CTS 3.3.B.1 provides requirements for control rod coupling. ITS 3.1.3 ACTION D provides an additional restriction for when two or more inoperable control rods are not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods and reactor power is Monticello Page 4 of 11 Attachment 1, Volume 6, Rev. 0, Page 63 of 231

Attachment 1, Volume 6, Rev. 0, Page 64 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY

      < 10% RTP. In this condition, ITS 3.1.3 ACTION D requires within 4 hours either the restoring of compliance with BPWS or the restoring of a control rod to OPERABLE status. This changes the CTS by adding an explicit ACTION for inoperable control rods under certain conditions when reactor power is < 10%

RTP. The purpose of ITS 3.1.3 ACTION D is to provide control rod operational restrictions at s 10% RTP. This change is acceptable because out of sequence control rods may increase the potential reactivity worth of a dropped control rod during a control rod drop accident. At s 10% RTP, the generic BPWS analysis requires inserted control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the diagonal. Therefore, if two or more inoperable control rods are not in compliance with BPWS and not separated by at least two OPERABLE control rods, action must be taken to restore compliance with BPWS or restore the control rods to OPERABLE status. This change is designated as more restrictive because it adds a new ACTION to the CTS. M.7 CTS 3/4.3.8.1 does not place a limitation of the number of inoperable control rods. ITS 3.1.3 ACTION E (second part of Condition E) covers the condition for nine or more inoperable control rods, and requires the unit to be in MODE 3 in 12 hours. This changes the CTS by adding an explicit ACTION for nine or more inoperable control rods. The purpose of ITS 3.1.3 Condition E (second part) is to limit the number of inoperable control rods. The change is acceptable since this condition (with nine or more inoperable control rods) could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems. This change is more restrictive since a limitation on the number of inoperable control rods has been added to the CTS. M.8 CTS 4.3.A.2.(b), which requires a periodic exercise test of the remaining fully and partially withdrawn OPERABLE control rods when a control rod is found to be stuck, states "This surveillance is not required if it has been confirmed that control rod drive collet housing failure is not the cause of the immovable control rod." The ITS does not maintain this allowance. ITS 3.1.3 Required Action A.3 will require a similar test when a control rod is found to be stuck, regardless of the reason for the stuck control rod. This changes the CTS by requiring an insertion test of remaining fully and partially withdrawn OPERABLE control rods when a stuck rod is found, regardless of the reason the rod is stuck. The purpose of CTS 4.3.A.2.(b) is to verify that each control rod is not stuck and is free to insert on a scram signal. This change is considered acceptable since the test will now be required regardless of the reason the rod is stuck. This will ensure that all remaining fully and partially withdrawn OPERABLE control rods are not also stuck. This change is a more restrictive change since the ITS will require a test under more conditions than currently required in the CTS. Monticello Page 5 of 11 Attachment 1, Volume 6, Rev. 0, Page 64 of 231

Attachment 1, Volume 6, Rev. 0, Page 65 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY M.9 CTS 4.3.B.1.(a) states that "when the rod is fully withdrawn the first time subsequent to each refueling outage," observe that the drive does not go to the overtravel position. ITS SR 3.1.3.5 requires the same verification, however, it must be performed each time the control rod is withdrawn to the full out position and prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling. This changes the CTS by changing the requirement to perform the coupling verification from "when the rod is fully withdrawn the first time subsequent to each refueling outage" to "Each time the control rod is withdrawn to full out position" and by adding the new Frequency of "Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling." The purpose of CTS 4.3.B.1.(a) is to ensure each control rod is coupled to its associated drive. The requirement to perform the coupling verification "when the rod is fully withdrawn the first time subsequent to each refueling outage" has been changed to "Each time the control rod is withdrawn to "full out" position" and a new Frequency, "Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling," has been added. This change is acceptable because a coupling check Is necessary after any work is performed on a control rod or Control Rod Drive System that could affect coupling. In addition, the requirement to perform the Surveillance each time the control rod is withdrawn is acceptable since a control rod could uncouple from its drive whenever a control rod is moved, not just after the first time it is fully withdrawn subsequent to each refueling outage. If a control rod is inserted one notch or more and then returned to the "full out" position during the performance of ITS SR 3.1.3.2 or for some other reason, a coupling verification can be easily performed since the verification only requires a check to make sure the control rod does no go to the withdrawn overtravel position. This change is designated as more restrictive because it requires the control rod coupling test to be verified more often. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS.3.3.A.2.(a) states, in part, the "directional control valves" for inoperable control rods shall be disarmed "electrically." CTS 3.3.B.1 states, in part, each control rod shall be coupled to its drive or completely inserted and the "directional control valves" disarmed "electrically." ITS 3.1.3 ACTION A covers the condition of one withdrawn control rod stuck. ITS 3.1.3 Required Action A.2 states "Disarm the associated control rod drive (CRD)." ITS 3.1.3 ACTION C covers the condition of one or more control rods inoperable for reasons other than a stuck rod. ITS 3.1.3 Required Action C.2 states "Disarm the associated CRD." Neither of these two Required Actions provides specific details of how to disarm the CRD. This changes the CTS by relocating the Monticello Page 6 of 11 Attachment 1, Volume 6, Rev. 0, Page 65 of 231

Attachment 1, Volume 6, Rev. 0, Page 66 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY details that the "directional control valves" are disarmed "electrically" to the ITS Bases. The removal of these details for performing Required Actions from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS 3.1.3 Required Actions A.2 and C.2 still retain the requirement to disarm the CRD. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 1- Relaxation of LCO Requirements) CTS 3.3.A.2.(a) states, in part, that control rod drives which cannot be moved "with control rod drive pressure" shall be considered inoperable. ITS 3.1.3 does not include this specific requirement. ITS 3.1.3 requires each control rod to be OPERABLE. A rod is considered OPERABLE, with respect to motion, if it can be inserted at least one notch using either scram pressure or normal control rod drive pressure (ITS SR 3.1.3.2 and SR 3.1.3.3) and, if it can be scrammed within < 7.0 seconds (ITS SR 3.1.3.4). This changes the CTS by deleting the requirement to consider a control rod inoperable if it cannot be moved by control rod drive pressure alone. The purpose of the ITS 3.1.3 is to base the OPERABILITY of an individual control rod on a combination of factors, including the scram insertion times.- As long as a control rod can be scrammed within < 7 seconds it should be considered to be OPERABLE. The control rod can satisfy this requirement with either accumulator pressure (scram pressure), control rod drive pressure, or a combination of the two. Accumulator OPERABILITY is addressed by LCO 3.1.5, "Control Rod Accumulators." The associated scram accumulator status for a control rod only affects the scram insertion times; therefore, an inoperable accumulator does not immediately require declaring a control rod inoperable. Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the design basis accident and transient analyses. This reactivity control requirement is monitored in LCO 3.1.4, "Control Rod Scram Times." This change is acceptable since the proposed requirements will continue to ensure the reactivity control is maintained. This change is designated as less restrictive because a control rod will not be considered inoperable if it cannot be moved by control rod drive pressure alone. L.2 (Category 4 - Reiaxation of Required Action) CTS 3.3.A.2.(b) requires, in part, the unit to be in hot shutdown within 48 hours if it is confirmed that a control rod drive collet housing failure is the cause of the stuck control rod. ITS 3.1.3 Monticello Page 7 of 11 Attachment 1, Volume 6, Rev. 0, Page 66 of 231

Attachment 1, Volume 6, Rev. 0, Page 67 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY ACTION A covers the condition for one stuck control rod. Continuous operation is allowed regardless of the reason for the control rod being stuck. This changes the CTS by allowing continuous operation with any type of stuck rod even as a result of a control rod drive collet housing failure. The purpose of CTS 3.3.A.2.(b) isto allow continuous unit operation with stuck rods as long as the cause of the failure is not a control rod drive collet housing failure. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken In response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the 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 design basis accident occurring during the repair period. This change will allow continuous operation with one stuck rod regardless of the cause of the failure. ITS 3.1.3 ACTION A will allow continuous operation as long as it is verified that the stuck control rod separation criteria are met, the stuck rod is disarmed, the other control rods are confirmed to not be stuck, and SHUTDOWN MARGIN is met with the stuck rod. The control rod separation criteria is described in the ITS 3.1.3 Bases. The separation criteria are not met if: a) the stuck control rod occupies a location adjacent (face or diagonal) to two "slow" control rods, b)the stuck control rod occupies a location adjacent to one "slow" control rod, and the one "slow" control rod is also adjacent to another "slow" control rod, or c) if the stuck control rod occupies a location adjacent to one "slow" control rod when there is another pair of "slow" control rods adjacent to one another. A "slow" control rod is described in the ITS 3.1.4 Bases. These proposed Required Actions are acceptable because they support continuous operation regardless of the type of failure since they continue to ensure SHUTDOWN MARGIN (SDM) can be met, other control rods are not stuck, and there is sufficient reactivity insertion capability to support the accident and transient analysis. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. L.3 (Category 7- Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.3.A.2.(a) requires each fully or partially withdrawn operable control rod to be exercised at least one notch "each week." ITS SR 3.1.3.2 requires a similar Surveillance for fully withdrawn control rods and ITS SR 3.1.3.3 requires a similar Surveillance for partially withdrawn control rods, however the Surveillance Frequency for ITS SR 3.1.3.3 is every 31 days. In addition, each Surveillance contains a Note that allows the performance of the Surveillance to be delayed for a certain time after the control rod is withdrawn and THERMAL POWER is greater than the low power setpoint (LPSP) of the rod worth minimizer (RWM). ITS SR 3.1.3.2 may be delayed for 7 days while ITS SR 3.1.3.3 may be delayed 31 days. This changes the CTS by extending the Surveillance Frequency from 7 days to 31 days for control rods that are partially withdrawn and provides a delay period for initial performance of the Surveillance after a control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM. Monticello Page 8 of 11 Attachment 1, Volume 6, Rev. 0, Page 67 of 231

Attachment 1, Volume 6, Rev. 0, Page 68 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY The purpose of CTS 4.3.A.2.(a) is to periodically verify that each withdrawn control rod is not stuck and is free to insert on a scram signal. This change is acceptable because the Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. Decreasing the Frequency of control rod exercise test for partially withdrawn control rods is acceptable based on the potential power reduction required to allow the control rod movement and considering the operating experience related to changes in control rod drive performance. These Surveillances are not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the Banked Position Withdrawal Sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). This portion of the change is acceptable since the unit does not normally operate for extended periods below the LPSP and since during a startup the control rods are withdrawn which helps to verify the control rod is not stuck since control rods are being withdrawn. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.4 (Category 3 - Relaxation of Completion Time) CTS 4.3.A.2.(b) states, in part, "each fully or partially withdrawn operable control rod shall be exercised at least one notch every 24 hour period" when a control rod is found to be stuck. When a control rod is stuck, ITS 3.1.3 Required Action A.3 states to perform SR 3.1.3.2 and SR 3.1.3.3 (the control rod insertion Surveillances for fully and partially withdrawn control rods) for each withdrawn OPERABLE control rod "24 hours from discovery of the stuck rod concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM." This changes the CTS by only requiring the test to be performed one time, and allows the test to be delayed up to 24 hours from discovery of the stuck rod concurrent with THERMAL POWER greater than the LPSP of the RWM. The purpose of CTS 4.3.A.2.(b) is to periodically verify that each withdrawn control rod is not stuck and is free to insert on a scram signal. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the 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 design basis accident occurring during the repair period. This change, in part, only requires the test to be performed one time at the accelerated Frequency. This is acceptable since performing this test one time ensures the control rods are not stuck and are free to insert on a scram signal. ITS SR 3.1.3.2 and ITS SR 3.1.3.3, the control rod insertion Surveillances, are still required to be performed at the specified Surveillance Frequency when the unit is operating with a stuck rod. These tests will ensure the control rods are OPERABLE. This change also allows the tests to be delayed up to 24 hours from discovery of the stuck rod concurrent with THERMAL POWER greater than the LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the Banked Position Withdrawal Sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). This portion of the change is acceptable since the Monticello Page 9 of 11 Attachment 1, Volume 6, Rev. 0, Page 68 of 231

Attachment 1, Volume 6, Rev. 0, Page 69 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY unit does not normally operate for extended periods below the LPSP and since during a startup the control rods are withdrawn which helps to verify the control rod is not stuck. This change is designated as less restrictive because the test only has to be performed once and the test may be delayed up to 24 hours from discovery of the stuck rod concurrent with THERMAL POWER greater than the LPSP of the RWM. L.5 (Category 5 - Deletion of Surveillance Requirement) CTS 4.3.3.1 .(b) states "when the rod is withdrawn the first time subsequent to each refueling outage, observe discernible response of the nuclear instrumentation. However, for initial rods when response is not discernible, subsequent exercising of these rods after the reactor is critical shall be performed to observe nuclear instrumentation response." ITS 3.1.3 does not include this requirement. This changes the CTS by eliminating the Surveillance Requirement to verify discernible nuclear instrumentation response when the rod is withdrawn. The purpose of CTS 4.3.B.1.(b) isto ensure the control rod is coupled to its drive during the withdrawal of a control rod. This change is acceptable because the deleted Surveillance Requirement is not necessary to ensure the control rods are coupled to their drives. Coupling verification is performed to ensure each control rod is connected to its drive so that it will perform its intended function when necessary. ITS SR 3.1.3.5 requires verifying a control rod does not go to the withdrawn overtravel position. The overtravel position feature provides a positive check on the coupling integrity since only an uncoupled control rod drive can reach the overtravel position. The verification is required to be performed any time a control rod iswithdrawn to the "full out" position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events. Observation of nuclear instrumentation indication will still normally occur during control rod withdrawal since nuclear instrumentation indication is close to the controls used to withdraw control rods. This change is designated as less restrictive because a Surveillance that is required in the CTS will not be required in the ITS. L.6 (Category 2 - Relaxation of Applicability) CTS 3.3.8.1 does not explicitly state when the control rod coupling requirements are required to be met, however it does state that the requirement is not applicable when moving a control rod drive for inspection as long as the reactor is in the refueling mode. CTS 3.3.G.1 requires the unit to be in cold shutdown (MODE 4) within 24 hours when the requirements of CTS 3/4.3.B.1 are not met. Thus, the implication is that CTS 3.3.B.1 is applicable in MODES 1, 2, and 3. ITS 3.1.3 states that the control rods must be OPERABLE in MODES 1 and 2 and ITS 3.1.3 ACTION E only requires the unit to be in MODE 3 (hot shutdown) within 12 hours when the actions are not met. This changes the CTS by only requiring the control rod coupling requirements to be met in MODES 1 and 2 and, concurrently, changes the shutdown action condition from cold shutdown (MODE 4) in 24 hours to hot shutdown (MODE 3) in 12 hours. The purpose of CTS 3.3.8.1 is to ensure each control rod is coupled to its drive prior to control rod withdrawal to help ensure a control rod drop accident does not Monticello Page 10 of 11 Attachment 1, Volume 6, Rev. 0, Page 69 of 231

Attachment 1, Volume 6, Rev. 0, Page 70 of 231 DISCUSSION OF CHANGES ITS 3.1.3, CONTROL ROD OPERABILITY occur during plant operation. This change is acceptable because the requirements continue to ensure that the control rods are maintained in the MODES and other specified conditions assumed in the safety analyses. The control rods are not required to be OPERABLE in MODES 3 and 4 since the Reactor Manual Control System places a rod withdrawal block when the mode switch is placed in shutdown so that no control rods can be withdrawn. In this condition, all control rods will be inserted, therefore coupling requirements are not necessary since the potential for a rod drop accident is highly unlikely. During refueling (MODE 5), with the MODE switch in the refueling position, coupling requirements are not necessary since only one rod can be withdrawn at a time. The CTS 3.3.G.1 requirement to be in cold shutdown has been replaced with the requirement to be in MODE 3. Since the OPERABILITY requirements are only necessary in MODES 1 and 2, the necessary shutdown action condition is MODE 3. Consistent with other actions to be in MODE 3; 12 hours is provided to reach this MODE. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS. Monticello Page 11 of 11 Attachment 1, Volume 6, Rev. 0, Page 70 of 231

Attachment 1, Volume 6, Rev. 0, Page 71 of 231 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) If Attachment 1, Volume 6, Rev. 0, Page 71 of 231

Attachment 1, Volume 6, Rev. O,Page 72 of 231 Control Rod OPERABILITY 3.1.3 CTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY 3.3.A.2, LCO 3.1.3 Each control rod shall be OPERABLE. 3.3.B.1 oCL6 APPLICABILITY: MODES 1 and 2. ACTIONS I-DOC Separate Condition entry is allowed for each control rod. A.3 CONDITION REQUIRED ACTION COMPLETION TIME A. One withdrawn control ------- -- NOTE------- 3.3A.2.(a). rod stuck. Rod worth minimizer (RWM) may 3.3A.2.(b) be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block Instrumentation," if required, to allow continued operation. A.1 Verify stuck control rod Immediately separation criteria are met. AND A.2 Disarm the associated 2 hours control rod drive (CRD). AND BWR/4 STS 3.1.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 72 of 231

Attachment 1,Volume 6, Rev. 0, Page 73 of 231 Control Rod OPERABILITY 3.1.3 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 4.3A.2.(b) A.3 Perform SR 3.1.3.2 and 24 hours from SR 3.1.3.3 for each discovery of withdrawn OPERABLE Condition A control rod. concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM AND 3.3A2.(a) A.4 Perform SR 3.1.1.1. 72 hours

                                     -                                       4-3.3A.2.(c)  B. Two or more withdrawn    B.1     Be in MODE 3.                    12 hours control rods stuck.
                                     -                                       4-3.3.8.1  C. One or more control      C.1       ---     NOTE-----

rods inoperable for RWM may be bypassed as reasons other than allowed by LCO 3.3.2.1, if Condition A or B. required, to allow insertion of inoperable control rod and continued operation. Fully insert inoperable 3 hours control rod. AND C.2 Disarm the associated 4 hours CRD. BWR/4 STS 3.1.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 73 of 231

Attachment 1, Volume 6, Rev. 0, Page 74 of 231 Control Rod OPERABILITY 3.1.3 CTS ACTIONS (continued) CONDITION REQUIRED ACTION - COMPLETION TIME D.. Retr copinewt DOC D. -----NOTE----- DA1 Restore compliance with 4 hours M.6 Not applicable when BPWS. THERMAL POWER

            > 1Or°/. RTP.              OR                                                           0D D.2     Restore control rod to       4 hours Two or more inoperable             OPERABLE status.

control rods not in compliance with banked position withdrawal sequence (BPWS) and not separated by two or more OPERABLE control rods. E. -- NO E---- E.1 Rest re control rod to 4hour- [Not applicible when OP RABLE status. THERMA POWER

            > [10]%      P.                                                                         0D One or ore groups with four or ore inoperable contro rods.

I I 3.3.G.1 Required Action and Be in MODE 3. 12 hours

                                                                                                    .0 associated Completion Time of Condition A, C, I      Enot met.

OR DOC Nine or more control M.7 rods inoperable. BWR/4 STS 3.1.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 74 of 231

Attachment 1, Volume 6, Rev. 0, Page 75 of 231 Control Rod OPERABILITY 3.1.3 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY MAs4 SR 3.1.3.1 Determine the position of each control rod. 24 hours 4 SR 3.1.3.2 ----- NOTES-- 4.3A.2.(a) Not required to be performed until 7 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of RWM. Insert each fully withdrawn control rod at least one 7 days notch.

                                                                                    .1.

SR 3.1.3.3 ------------ NOTE- - -- 4.3A.2.(a) Not required to be performed until 31 days after the control rod is withdrawn and THERMAL POWER is greater than the LPSP of the RWM. Insert each partially withdrawn control rod at least 31 days one notch. DOC SR 3.1.3.4 Verify each control rod scram time from fully In accordance 0 M.4 withdrawn to notch position 106l is

• 7 seconds. with SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4 4.3.B.1 SR 3.1.3.5 Verify each control rod does not go to the withdrawn Each time the overtravel position. control rod is withdrawn to "full out" position AND Prior to declaring control rod OPERABLE after work on control rod or CRD System that could affect coupling BWR/4 STS 3.1.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 75 of 231

Attachment 1, Volume 6, Rev. 0, Page 76 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3, CONTROL ROD OPERABILITY

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

2. As stated in the ISTS Bases, ISTS 3.1.3 ACTION E is applicable to plants with ANF fuel. Monticello does not have this type of fuel. Consequently, this ACTION is not applicable to Monticello and has been deleted. As a result of this deletion, the following ACTION has been renumbered.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 76 of 231

Attachment 1, Volume 6, Rev. 0, Page 77 of 231 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 77 of 231

Attachment 1, Volume 6, Rev. 0, Page 78 of 231 Control Rod OPERABILITY B 3.1.3 OB 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod OPERABILITY BASES BACKGROUND Control rods are components of the Control Rod Drive (CRD) System, which is the primary reactivity control system for the reactor. In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including anticipated operational occurrences, that specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System. The CRD System is designed to satisfy the requirements lof GDC 26, G 7,GDC 28, and 20 (Ref. 1). USAR. Sectlon 1.2.2 0D The CRD System consists of locking piston control rod drive mechanisms (CRDMs) and a hydraulic control unit for each drive 0D mechanism. The locking piston type CRDM is a double acting hydraulic piston, which uses condensate water as the operating fluid. Accumulators provide additional energy for scram. An index tube and piston, coupled to the control rod, are locked at fixed increments by a. collet mechanism. The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion. ,_., land LCO 3.1.6. 'Rod Pattern Control. This Specification, along with LCO 3.1.4 "Control Rod Scram Times," a(3 LCO 3.1.5, "Control Rod Scram Accumulators,' nsure that the I performance of the control rods in the event of a Design Basis Accident (DBA) or transient meets the assumptions used in the safety analyses of References 2, 3, and 4. APPLICABLE The analytical methods and assumptions used in the evaluations SAFETY involving control rods are presented in References 2, 3, and 4. The ANALYSES control rods provide the primary means for rapid reactivity control (reactor scram), for maintaining the reactor subcritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System. The capability to insert the control rods provides assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated. Since the SDM ensures the reactor will be subcritical with the highest worth control rod withdrawn (assumed single failure), the BWR/4 STS B 3.1.3-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 78 of 231

Attachment 1, Volume 6, Rev. 0, Page 79 of 231 Control Rod OPERABILITY B 3.1.3 BASES APPLICABLE SAFETY ANALYSES (continued) additional failure of a second control rod to insert, if required, could invalidate the demonstrated SDM and potentially limit the ability of the CRD System to hold the reactor subcritical. If the control rod is stuck at an inserted position and becomes decoupled from the CRD, a control rod drop accident (CRDA) can possibly occur. Therefore, the requirement that all control rods be OPERABLE ensures the CRD System can perform its intended function. The control rods also protect the fuel from damage which could result in release of radioactivity. The limits protected are the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"), the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE.(APLHGR)," and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), and the fuel a 0 limit (see Bases for LCO 3.1.6 "Rod Pas Control ') during reactivity ( insertion events. The negative reactivity insertion (scram) provided by the CRD System provides the analytical basis for determination of plant thermal limits and provides protection agains fuelaa imits during a CRDA. The Bases for LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6 discuss in more detail how the SLs are protected by the CRD System. d Control rod OPERABILITY satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO The OPERABILITY of an individual control rod is based on a combination of factors, primarily, the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position. Accumulator OPERABILITY is addressed by LCO 3.1.5. The associated scram accumulator status for a control rod only affects the scram insertion times; therefore, an inoperable accumulator does not immediately require declaring a control rod inoperable. Although not all control rods are required to be OPERABLE to satisfy the intended reactivity control requirements, strict control over the number and distribution of inoperable control rods is required to satisfy the assumptions of the DBA and transient analyses. BWR/4 STS B 3.1.3-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 79 of 231

Attachment 1, Volume 6, Rev. 0, Page 80 of 231 B 3.1.3 INSERT I OPERABILITY requirements for control rods also include correct assembly of the CRD housing supports. Insert Page B 3.1.3-2 Attachment 1, Volume 6, Rev. 0, Page 80 of 231

Attachment 1, Volume 6, Rev. 0, Page 81 of 231 Control Rod OPERABILITY B 3.1.3 BASES APPLICABILITY In MODES 1 and 2, the control rods are assumed to function during a DBA or transient and are therefore required to be OPERABLE in these MODES. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod OPERABILITY during these conditions. Control rod requirements in MODE 5 are located in LCO 3.9.5, "Control Rod OPERABILITY - Refueling." ACTIONS The ACTIONS Table is modified by a Note indicating that a separate Condition entry is allowed for each control rod. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable control rod. Complying with the Required Actions may allow for continued operation, and subsequent inoperable control rods are governed by subsequent Condition entry and application of associated Required Actions. A.1. A.2. A.3. and A.4 A control rod is considered stuck if it will not insert by either CRD drive water or scram pressure. With a fully inserted control rod stuck, no actions are required as long as the control rod remains fully inserted. The Required Actions are modified by a Note, which allows the rod worth minimizer (RWM) to be bypassed if required to allow continued operation. LCO 3.3.2.1, "Control Rod Block Instrumentation," provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis. With one withdrawn control rod stuck, the local scram reactivity rate assumptions may not be met if the stuck control rod separation criteria are not met. Therefore, a verification that the 4 fi}j separation criteria are met must be performed immediately. The or separation criteria are not met if: a) the stuck control rod occupies a ldaonal) §location adjacenIto two "slow" control rods, b) the stuck control rod occupies a location adjacent to one "slow" control rod, and the one "slow" control rod is also adjacent to another "slow" control rod, or c) if the stuck i the coren control rod occupies a location adjacent to one "slow" control rod when there is another pair of "slow" control rods adjacent to one another. The description of "slow" control rods is provided in LCO 3.1.4, "Control Rod Scram Times." In addition, the associated control rod drive must be disarmed in 2 hours. The allowed Completion Time of 2 hours is acceptable, considering the reactor can still be shut down, assuming noNR additional control rods fail to insert, and provides a reasonable time to perform the Required Action in an orderly manner. Isolating the control rod from scram prevents damage to the CRDM. The control rod G Iisolated from scram andwn onserth draw pressure yet still l cooling water to the CRQ. and normal Insert isolation mehod and withdraw also ensure pressure BWR/4 STS B 3.1.3-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 81 of 231

Attachment 1, Volume 6, Rev. 0, Page 82 of 231 B3.1.3 CD3 INSERT2 The control rod must be isolated from both scram and normal insert and withdraw pressure.. Insert Page B 3.1.3-3 Attachment 1, Volume 6, Rev. 0, Page 82 of 231

Attachment 1, Volume 6, Rev. 0, Page 83 of 231 Control Rod OPERABILITY B 3.1.3 BASES ACTIONS (continued) Monitoring of the insertion capability of each withdrawn control rod must also be performed within 24 hours from discovery of Condition A concurrent with THERMAL POWER greater than the low power setpoint (LPSP) of the RWM. SR 3.1.3.2 and SR 3.1.3.3 perform periodic tests of the control rod insertion capability of withdrawn control rods. Testing each withdrawn control rod ensures that a generic problem does not exist. This Completion Time allows for an exception to the normal "time ero" for beginning the allowed outage time "clock." The Required Action Al- Completion Time only begins upon discovery of Condition A Q) concurrent with THERMAL POWER greater than the actual LPSP of the RWM since the notch insertions may not be compatible with the requirements of rod pattern control (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The allowed Completion Time of 24 hours from discovery of Condition A, concurrent with THERMAL POWER greater than the LPSP of the RWM, provides a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests. To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours. Should a DBA or transient require a shutdown, to preserve the single failure criterion, an additional control rod would have to be assumed to fail to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be evaluated (by measurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod assumed to be fully withdrawn. The allowed Completion Time of 72 hours to verify SDM is adequate, considering that with a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity. Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a required scram. Even with the postulated additional single failure of an adjacent control rod to insert, sufficient reactivity control remains to reach land intain MODE 3 conditions ( BWR/4 STS B 3.1.3-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 83 of 231

Attachment 1, Volume 6, Rev. 0, Page 84 of 231 Control Rod OPERABILITY B 3.1.3 BASES ACTIONS (continued) B.1 With two or more withdrawn control rods stuck, the plant must be brought to MODE 3 within 12 hours. The occurrence of more than one control rod stuck at a withdrawn position increases the probability that the reactor cannot be shut down if required. Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

                - C.1 and C.2 With one or more control rods inoperable for reasons other than being s                   withdrawn position, operation may continue, provided the control rods are fully inserted within 3 hours and disarmed (electrically or hydraulically) within 4 hours. Inserting a control rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. Required Action C.1 is modified by a Note, which allows the RWM to be bypassed if required to allow insertion of the inoperable control rods and continued operation. LCO 3.3.2.1 provides additional requirements when the RWM is bypassed to ensure compliance with the CRDA analysis.

The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods, and provide time to insert and disarm the control rods in an orderly manner and without challenging plant systems. D.1 and D.2 Out of sequence control rods may increase the potential reactivity worth of a dropped control rod during a CRDA. At s 10% RTP, the generic banked position withdrawal sequence (BPWS) analysis (Ref. 5) requires inserted control rods not in compliance with BPWS to be separated by at least two OPERABLE control rods in all directions, including the diagonal. (E Therefore, if two or more inoperable control rods are not in compliance (i.e.. all other control rods in a five-by-fi ra cnee ton the Inoperable control rod are OPERABLE) BWR/4 STS B 3.1.3-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 84 of 231

Attachment 1, Volume 6, Rev. 0, Page 85 of 231 Control Rod OPERABILITY B 3.1.3 BASES ACTIONS (continued) with BPWS and not separated by at least two OPERABLE control rods, action must be taken to restore compliance with BPWS or restore the 0 control rods to OPERABLE status. Condition D is modified by a Note indicating that the Condition is not applicable when > 10% RTP, since the BPWS is not required to be followed under these conditions, as described in the Bases for LCO 3.1.6. The allowed Completion Time of 4 hours is acceptable, considering the low probability of a CRDA occurring. E.1 In addition t the separation require ents for inoperable control ods, an assumption n the CRDA analysis fo ANF fuel is that no more t an three inoperable ontrol rods are allowed n any one BPWS group. T erefore, with one o more BPWS groups ha ing four or more inoperabl control rods, the ntrol rods must be rest red to OPERABLE status. Required 0 Action E. is modified by a Note i dicating that the Condition not applicabl when THERMAL PO R is> 10% RTP since th BPWS is not requ ed to be followed undethese conditions, as descried in the Bases rLCO 3.1.6. The allowI d Completion Time of4 hors is acceptq le, considering the low/probability of a CRDA occurng. 0 If any Required Action and associated Completion Time of Condition A, C 9EIM 4 are not met, or there are nine or more inoperable control rods, the plant must be brought to a MODE in which the LCO does not apply. 0 To achieve this status, the plant must be brought to MODE 3 within 12 hours. This ensures all insertable control rods are inserted and places the reactor in a condition that does not require the active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when operating above 10% RTP (e.g., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power in an orderly manner and without challenging plant systems. BWR/4 STS B 3.1.3-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 85 of 231

Attachment 1, Volume 6, Rev. 0, Page 86 of 231 Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE SR 3.1.3.1 REQUIREMENTS The position of each control rod must be determined to ensure adequate information on control rod position is available to the operator for Id) determining OPERABILITY and controlling rod patterns. Control rod position may be determined by the use of OPERABLE position indicators, by moving control rods to a position with an OPERABLE indicator, or by the use of other appropriate methods. The 24 hour Frequency of this SR is based on operating experience related to expected changes in control rod position and the availability of control rod position indications in the control room. SR 3.1.3.2and SR 3.1.3.3 Control rod insertion capability isdemonstrated by inserting each partially or fully withdrawn control rod at least one notch and observing that the control rod moves. The control rod may then be returned to its original position. This ensures the control rod is not stuck and is free'to insert on a scram signal. These Surveillances are not required when THERMAL POWER is less than or equal to the actual LPSP of the RWM, since the notch insertions may not be compatible with the requirements of the Banked Position Withdrawal Sequence (BPWS) (LCO 3.1.6) and the RWM (LCO 3.3.2.1). The 7 day Frequency of SR 3.1.3.2 is based on operating experience related to the changes in CRD performance and the ease of performing notch testing for fully withdrawn control rods. Partially withdrawn control rods are tested at a 31 day Frequency, based on the potential power reduction required to allow the control rod movement and considering the large testing sample of SR 3.1.3.2. Furthermore, the 31 day Frequency takes into account operating experience related to changes in CRD performance. At any time, if a control rod is immovable, a determination of that control rod's trippability (OPERABILITY) must be made and appropriate action taken.

                                                               -NSERT3 SR 3.1.3.4 Verifying that the scram time for each control rod to notch position 06 is s 7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function. This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4. The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor BWR/4 STS                        B 3.1.3-7                                Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 86 of 231

Attachment 1, Volume 6, Rev. 0, Page 87 of 231 B 3.1.3 03 INSERT 3 These SRs are modified by Notes that allow 7 days and 31 days respectively, after withdrawal of the control rod and increasing power to above the LPSP, to perform the Surveillance. This acknowledges that the control rod must be first withdrawn and THERMAL POWER must be increased to above the LPSP before performance of the Surveillance, and therefore, the Notes avoid potential conflicts with SR 3.0.1 and SR 3.0.4. Insert Page B 3.1.3-7 Attachment 1, Volume 6, Rev. 0, Page 87 of 231

Attachment 1, Volume 6, Rev. 0, Page 88 of 231 Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE REQUIREMENTS (continued) Protection System (RPS) Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (SDV) Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function. The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle. SR 3.1.3.5 Coupling verification is performed to ensure the control rod is connected to the CRDM and will perform its intended function when necessary. The/ Surveillance requires verifying'a control rod does not go to the withdrawn (U overtravel positio. The overtravel position feature provides a positive (D

                          ;ln   pling integrity since only an uncoupled CRD can reach the overtravel position. The verification is required to be performed any time a control rod is withdrawn to the "full out" position (notch position 48) or prior to declaring the control rod OPERABLE after work on the control rod or CRD System that could affect coupling. This includes control rods inserted one notch and then returned to the "full out" position during the performance of SR 3.1.3.2. This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events.

REFERENCES 1. 110 (E9.n ix 27 29. (9 2A R , SSectioa

                                         ... 3.2.2.4 SAR,    ecti      A.4 ASectio,         5.13

5. NEDO-21231, "Banked Position Withdrawal Sequence," Section 7.2, January 1977.

BWR14 STS B 3.1.3-8 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 88 of 231

Attachment 1, Volume 6, Rev. 0, Page 89 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.3 BASES, CONTROL ROD OPERABILITY

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

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

3. Change made to be consistent with the Specification.

4. Changes are made to reflect changes made to the Specification.

5. Typographical/grammatical error corrected.

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

Monticello Pagel of 1 Attachment 1, Volume 6, Rev. 0, Page 89 of 231

Attachment 1,Volume 6, Rev. 0, Page 90 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 6, Rev. 0, Page 90 of 231

Attachment 1, Volume 6, Rev. 0, Page 91 of 231 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.3, CONTROL ROD OPERABILITY There are no specific NSHC discussions for this Specification. Monticello Page Iof 1 Attachment 1, Volume 6, Rev. 0, Page 91 of 231

, Volume 6, Rev. 0, Page 92 of 231 ATTACHMENT 4 ITS 3.1.4, Control Rod Scram Times , Volume 6, Rev. 0, Page 92 of 231

Attachment 1, Volume 6, Rev. 0, Page 93 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 6, Rev. 0, Page 93 of 231

C c C ITS 3.1.4 0 ITS ITS

                                                         %U I

0 0 Z., 0 0 -g -u CD CD CD CD CD CD 0 0 -4 -4 3.314.3 81 3/27/81 Amendment No. 3 Page 1 of 2

c C. C ITS 3.1.4 J ITS 3.0 UMTNG CONDImONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS _ _ _ . _I_._._..i F. Scram Discharge Volume F. Scram Discharge Volume a) S

1. DurIng reactor operation, the scram discharge The scram discharge volume vent and drain valves shall volume vent and drain valves shall be operable, be cycled quarterly. 1W except as specified below. Once per operating cycle verify the scram discharge a) 01) volume vent and drain valves dose within 30 seconds 0

C) 2. If any scam discharge volume vent or drain valve Is after receipt of a reactor scram signal and open when made or found Inoperable, the Integrity of the scram the scram Isreset. 3 discharge volume shall be maintained by elther 0

a.

• Verifying daily, for a period not to exceed 7 days, the operabiTity of the redundant valve(s).

CD or

.I CD b. Maintaining the Inoperable valve(s), or the ro.

0) associated redundant valve(s). Inthe dosed position. Periodically the Inoperable and the a{i See ITS 3.1.8 } CD a) redundant valve(s) may both be In the open 0 position to allow draining the scram discharge CD volume.

If a or b above cannot be met, at least all but one -o operable control rods (not Including rods removed CD CD per specification 3.10.E or Inoperable rods allowed CD U' by 3.3A2) shall be fully inserted within ten hours. 0 0 G. Required Action IS3 A 800 psig, to notch positions 46, 36, 26, and 06. ITS Table 3.1.4-1 Note 1 states that OPERABLE control rods with scram times not within the limits of this Table are considered "slow." ITS Table 3.1.4-1 Note 2 states "Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times > 7 seconds to notch position 06. These control rods are inoperable, in accordance with ITS SR 3.1.3.4, and are not considered "slow." ITS Table 3.1.4-1 footnote (b)states "Scram times as a function of reactor steam dome pressure when < 800 psig are Monticello Page 1 of 6 Attachment 1, Volume 6, Rev. 0, Page 96 of 231

Attachment 1, Volume 6, Rev. 0, Page 97 of 231 DISCUSSION OF CHANGES ITS 3.1.4, CONTROL ROD SCRAM TIMES within established limits." This changes the CTS by specifying control rod scram time for each individual control rod as a function of reactor steam dome pressure instead of the current scram time requirements based on the average scram insertion time of all OPERABLE control rods and for the average scram insertion time for the three fastest control rods of all groups of four control rods in a two by two array. In addition, criteria has been established for no more than 8 "slow" OPERABLE control rods and no more than 2 "slow" OPERABLE control rods occupying adjacent locations. The purpose of the control rod scram time LCOs (CTS 3.3.C.1 and 3.3.C.2) is to ensure the negative scram reactivity is consistent with those values assumed in the accident and transient analysis. CTS 3.3.C.1 and 3.3.C.2 place requirements on the average sdram times and local scram times (four control rod group). Because of the methodology used in the design basis transient analysis (one-dimensional neutronics), all control rods are assumed to scram at the same speed, which is the analytical scram time requirement. Performing an evaluation assuming all control rods scram at the analytical limit results in the generation of a scram reactivity versus time curve, the analytical scram reactivity curve. The purpose of the scram time LCO is to ensure that, under allowed unit conditions, this analytical scram reactivity will be met. Since scram reactivity cannot be readily measured at the unit, the safety analyses use appropriately conservative scram reactivity versus insertion fraction curves to account for the variation in scram reactivity during a cycle. Therefore, the Technical Specifications must only ensure the scram times are satisfied. The first obvious result is that, if all control rods scram at least as fast as the analytical limit, the analytical scram reactivity curve will be met. However, a distribution of scram times (some slower and some faster than the analytical limit) can also provide adequate scram reactivity. By definition, for a situation where all control rods do not satisfy the analytical scram time limits, the condition is acceptable if the resulting scram reactivity meets or exceeds the analytical scram reactivity curve. This can be evaluated using models that allow for a distribution of scram speeds. It follows that the more control rods that scram slower than the analytical limit, the faster the remaining control rods must scram to compensate for the reduced scram reactivity rate of the slower control rods. ITS 3.1.4 incorporates this philosophy by specifying scram time limits for each individual control rod instead of limits on the average of all control rods and the average of three fastest rods in all four control rod groups. This philosophy has been endorsed by the BWR Owners' Group and is described in EAS-46-0487, "Revised Reactivity Control Systems Technical Specifications." The scram time limits listed in ITS Table 3.1.4-1 have margin to the analytical scram time limits listed in EAS-46-0487, Table 3-4 to allow for a specified number and distribution of slow control rods, a single stuck control rod and an assumed single failure. Therefore, if all control rods met the scram time limits found in ITS Table 3.1.4-1, the analytical scram reactivity assumptions are satisfied. If control rods do not meet the scram time limits, ITS LCO 3.1.4 specifies the number and distribution of these "slow" control rods to ensure the analytical scram reactivity assumptions are still satisfied. If the number of slow rods is more than 8 or the rods do not meet the separation requirements, the unit must be shutdown. This change is designated as more restrictive because explicit requirements have been included in the Technical Specifications to cover conditions not currently addressed in the CTS. ITS 3.1.4 specifies limitations on scram times for each individual control rod. That is, Monticello Page. 2 of 6 Attachment 1, Volume 6, Rev. 0, Page 97 of 231

Attachment 1, Volume 6, Rev. 0, Page 98 of 231 DISCUSSION OF CHANGES ITS 3.1.4, CONTROL ROD SCRAM TIMES currently, the "average time" of all rods or a group can be improved by a few fast scramming rods, even when there may be more than 8 slow rods, as defined in the proposed Specification. Therefore, ITS 3.1.4 limits the number of slow rods to 8 and ensures no more than 2 slow rods occupy adjacent locations. The maximum scram time requirement has been added to the ITS (see Discussion of Changes for ITS 3.1.3) for the purpose of defining the threshold between a slow control rod and an inoperable control rod even though the analyses to determine the LCO scram time limits assumed slow control rods did not scram. Note 2 to ITS Table 3.1.4-1 ensures that a control rod is not inadvertently considered "slow" when the scram time exceeds 7 seconds. This change is designated as more restrictive because explicit requirements have been included in the Technical Specifications to cover conditions not currently addressed in the CTS. M.2 CTS 4.3.C requires each OPERABLE rod to be scram time tested during each operating cycle, however, it also states that if testing is not accomplished during reactor power operation, the measured scram time may be extrapolated to the reacto& power operation condition. ITS SR 3.1.4.1 requires verification that each control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure > 800 psig prior to exceeding 40% RTP after each reactor shutdown > 120 days. ITS SR 3.1.4.2 requires verification that, for a representative sample, each tested control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure > 800 psig every 200 days of cumulative operation in MODE 1. ITS SR 3.1.4.3 requires verification that each affected control rod scram time is within the limits of Table 3.1.4-1 with any reactor steam dome pressure prior to declaring a control rod OPERABLE after work on control rod or CRD System that could affect scram time. ITS SR 3.1.4.4 requires verification that each affected control rod scram time is within the limits of Table 3.1.4-1 with reactor steam dome pressure 2 800 psig prior to exceeding 40% RTP after fuel movement within the affected core cell and prior to exceeding 40% RTP after work on a control rod or the CRD System that could affect the scram time. In addition, a Surveillance Note has been added that states "During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be isolated from the associated scram accumulator." This changes the CTS by requiring a scram time test to be performed prior to declaring a control rod OPERABLE after work on control rod or CRD System that could affect scram time. It also requires the unit to complete scram time testing of affected control rods prior to exceeding 40% RTP after fuel movement within the affected cell and after work on a control rod or the CRD System that could affect the scram time. In addition, if the reactor is shutdown for > 120 days, a scram time test of each control rod is required to be performed prior to exceeding 40% RTP, and, after every 200 days of cumulative operation in MODE 1, a representative sample of control rods must be scram time tested. Finally the change requires the single control rod scram time Surveillance to be performed with the CRD pumps isolated form the associated scram accumulator. The purpose of CTS 4.3.C is to ensure the control rods can meet the scram reactivity insertion requirements to support the unit safety analysis. This change provides more explicit control rod scram time testing requirements than in the' CTS to ensure control rods are OPERABLE prior to entering MODE 2 when work has been performed on the control rod or CRD System that could affect its scram time. Soon after entering MODE 2 (and prior to exceeding 40% RTP), scram Monticello Page 3 of 6 Attachment 1, Volume 6, Rev. 0, Page 98 of 231

Attachment 1, Volume 6, Rev. 0, Page 99 of 231 DISCUSSION OF CHANGES ITS 3.1.4, CONTROL ROD SCRAM TIMES time tests are required at steam dome pressures of > 800 psig (after fuel movement within the affected cell and after work on control rod or CRD System that could affect scram time) to confirm the scram time performance at the most limiting conditions. In addition, if the reactor has been shutdown for > 120 days, each control rod must be tested prior to exceeding 40% RTP. Furthermore, every 200 days of cumulative operation in MODE 1, a representative sample of control rods must be scram time tested to ensure the limits of Table 3.1.4-1 are met. The scram time criteria at < 800 psig will be lower than the scram time values specified in Table 3.1.4-1 for > 800 psig. The criterion is established based on previously determined correlations. Satisfying the test at these conditions (<800 psig) will almost guarantee with a high probability the acceptance criteria at > 800 psig will be satisfied. This low pressure testing is required when work has been performed on a control rod or CRD System that could affect scram time. Affected control rods cannot be declared OPERABLE until this test is performed. The test at > 800 psig is necessary because at approximately 800 psig there are competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure > 800 psig ensures that the measured scram times will be within the specified limits at higher reactor pressures. The first Frequency of ITS SR 3.1.4.4 is prior to exceeding 40% RTP after fuel movement within the affected core cell. This Frequency will basically require a high pressure scram test for each control rod after a refueling outage. CTS 4.3.C requires a test during each Operating Cycle, which is defined in CTS 1.0.N as the interval between the end of one refueling outage and the end of the next subsequent refueling outage. This Surveillance Frequency in ITS SR 3.1.4.4 will ensure that the necessary scram testing is performed shortly after MODE 2 is entered after a refueling outage (i.e., prior to exceeding 40% RTP). This Frequency is necessary since control rod scram performance is necessary in establishing MCPR operating limits and to ensure the MCPR Safety Limit is met during a unit transient. The second Frequency of ITS SR 3.1.4.4 is prior to exceeding 40% RTP after work on a control rod or CRD System that could affect scram time. This Frequency will basically only require a high pressure scram test for each affected control rod after a non-refueling outage if work has been performed on the control rod or CRD System. Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve In the piping required for scram. ITS SR 3.1.4.1 has been added to ensure that if the unit has been shutdown for a long period of time (> 120 days), the control rods are scram timed to ensure compliance with the acceptance criteria. This is necessary to ensure that any maintenance activity over this shutdown period has not affected the control rod scram capabilities and due to the fact that the control rods are normally not exercised during shutdown conditions. ITS SR 3.1.4.2 has been added to ensure a representative sample of control rods are periodically tested (every 200 days of cumulative operation in MODE 1) to ensure the scram times remain within the limits of Table 3.1.4-1 throughout the cycle. The four SRs of this LCO are modified by a Note stating that during a single control rod scram time Surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated (i.e., charging valve closed), the influence of the CRD pump head will affect the single control rod scram times. Monticello Page 4 of 6 Attachment 1, Volume 6, Rev. 0, Page 99 of 231

Attachment 1, Volume 6, Rev. 0, Page 100 of 231 DISCUSSION OF CHANGES ITS 3.1.4, CONTROL ROD SCRAM TIMES This Note restriction is not necessary during a full core scram since the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times. This change is designated as more restrictive because the Surveillances prescribe more frequent Surveillance Frequencies than are required by the CTS. M.3 CTS 3.3.G.1 requires the unit to be in cold shutdown (MODE 4) within 24 hours if CTS 3.3.C is not met. ITS 3.1.4 ACTION A requires the unit to be in MODE 3 in 12 hours when ITS LCO 3.1.4 is not met. This changes the CTS by requiring the unit to be in MODE 3 in 12 hours instead of 24 hours. The change to the unit condition required to be achieved (MODE 3 versus MODE 4) is discussed in DOC A.2. The purpose of CTS 3.3.G.1 is to place the unit in a condition where the scram time limits of the control rods are not required. This change is acceptable because the allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging unit systems. The requirement to be in MODE 3 in 12 hours is designated as more restrictive since MODE 3 must be achieved in a faster time limit than is currently required. M.4 CTS 3.3.C.1 requires the scram times to be within the limits in the "reactor power operation condition." ITS LCO 3.1.4 is Applicable in MODES I and 2. This changes the CTS by requiring the scram time limits to be met in MODE 2

     <1% RATED THERMAL POWER (RTP).

The purpose of CTS 3.3.C.1 is to ensure the negative scram reactivity is consistent with those values assumed in the accident and transient analysis. This change expands the Applicability to require the scram time limits to be met at all times when in MODE 2, instead of when > 1% RTP (the CTS 1.0.0 definition states that Power Operation is when reactor power is > 1% RTP). This change is acceptable since the scram time limits must be met in MODE 2 because the reactor is critical or control rods are withdrawn (thus the need for the control rods to be capable of properly scramming in the assumed time exists). This change is designated as more restrictive because the LCO will be applicable under more reactor conditions. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None Monticello Page 5 of 6 Attachment 1, Volume 6, Rev. 0, Page 100 of 231

Attachment 1, Volume 6, Rev. 0, Page 101 of 231 DISCUSSION OF CHANGES ITS 3.1.4, CONTROL ROD SCRAM TIMES LESS RESTRICTIVE CHANGES None Monticello Page 6 of 6 Attachment 1, Volume 6, Rev. 0, Page 101 of 231

Attachment 1, Volume 6, Rev. 0, Page 102 of 231 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 102 of 231

Attachment 1, Volume 6, Rev. 0, Page 103 of 231 Control Rod Scram Times 3.1.4 CTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Control Rod Scram Times 3.3.C.1, 3.3.C.2 LCO 3.1.4 a. No more thanM)XOPERABLE control rods shall be "slow," in 0 accordance with Table 3.1.4-1, and

b. No more than 2 OPERABLE control rods that are "slow" shall occupy adjacent locations.

3.3.C.1 APPLICABILITY: MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.3.G.1 A. Requirements of the A.1 Be in MODE 3. 12 hours LCO not met. SURVEILLANCE REQUIREMENTS A-----------------NOTE--a DOC During single control rod scram time Surveillances, the control rod drive (CRD) pumps shall be M.2 isolated from the associated scram accumulator. SURVEILLANCE FREQUENCY 4.3.C SR 3.1.4.1 Verify each control rod scram time is within the limits Prior to exceeding of Table 3.1.4-1 with reactor steam dome pressure 40% RTP after 2 800J psig. each reactor shutdown 0

                                                                                         ,- eofn 4-.,-

4.3.C SR 3.1.4.2 Verify, for a representative sample, each tested control rod scram time is within the limits of cumulative Table 3.1.4-1 with reactor steam dome pressure operation in

                                ŽID800 psig.                                             MODE I                (

BWR/4 STS 3.1.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 103 of 231

Attachment 1, Volume 6, Rev. 0, Page 104 of 231 Control Rod Scram Times 3.1.4 CTS SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY 4.3.C SR 3.1.4.3 Verify each affected control rod scram time is within Prior to declaring the limits of Table 3.1.4-1 with any reactor steam control rod dome pressure. OPERABLE after work on control rod or CRD System that could affect scram time 4.3.C SR 3.1.4.4 Verify each affected control rod scram time is within Prior to exceeding the limits of Table 3.1.4-1 with reactor steam dome 40% RTP after pressure 4E0clMpsig. fuel movement within the affected core cell AND Prior to exceeding 40% RTP after work on control rod or CRD System that could affect scram time BWRI4 STS 3.1.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 104 of 231

Attachment 1, Volume 6, Rev. 0, Page 105 of 231 Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1) Control Rod Scram Times CTS kIe<%TCC _ _ _ _ ._____ - -. t _ DOC 1. OPERABLE control rods with scram times not within the limits of this Table are considered M.1 "slow." DoC 2. Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod M.A OPERABILITY," for control rods with scram times > 7 seconds to notch positionM06J These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered 0

             "slow."

SCRAM TIMES(aXb) (seconds) WHEN REACTOR STEAM DOME PRESSURE 3.3.C.1, 3.3.C.2 NOTCH POSITION 2f800mpsig 0 IN R36M Mugl.Q R26M Ml 1.82M R06M - R3.3o 3.3.C.1. (a) Maximum scram time from fully withdrawn positiorjJbased on de-energization of scram pilot ( 4.3.C valve solenoids at time zero. DOC M.1 (b) Scram times as a function of reactor steam dome pressuregwhen < 800 psig are within established limits. 0D BWR/4 STS 3.1.4-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 105 of 231

Attachment 1, Volume 6, Rev. 0, Page 106 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4, CONTROL ROD SCRAM TIMES

1. The brackets are removed and the proper plant specific information/value is provided.

2. Typographical/grammatical error corrected.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 106 of 231

Attachment 1, Volume 6, Rev. 0, Page 107 of 231 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 107 of 231

Attachment 1, Volume 6, Rev. 0, Page 108 of 231 Control Rod Scram Times B 3.1.4 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.4 Control Rod Scram Times BASES BACKGROUND The scram function of the Control Rod Dnive (CRD) systems~ reactivity ctpes-uring abnormal operati~~sients to ensure thall ispec~ifiegcceptable-fuel design limit,%a6not exceeded (Ref. 1). The control rods are scrammed by positive means using hydraulic pressure exerted on the CRD piston. When a scram signal is initiated, control air is vented from the scram valves, allowing them to open by spring action. Opening the exhaust valve reduces the pressure above the main drive piston to atmospheric pressure, and opening the inlet valve applies the accumulator or reactor pressure to the bottom of the piston. Since the notches in the index tube are tapered on the lower edge, the collet fingers are forced open by cam action, allowing the index tube to move upward without restriction because of the high differential pressure across the piston. As the drive moves upward and the accumulator pressure reduces below the reactor pressure, a ball check valve opens, letting the reactor pressure complete the scram action. If the reactor pressure is low, such as during startup, the accumulator will fully insert the control rod in the required time without assistance from reactor pressure. APPLICABLE The analytical methods and assumptions used in evaluating the control SAFETY rod scram function are presented in References 2, 3, and 4. The Design ANALYSES Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. The resulting negative scram reactivity forms the basis for the determination of plant thermal limits (e.g., the MCPR). Other distributions of scram times (e.g., several control rods scramming slower than the average time with several control rods scramming faster than the average time) can also provide sufficient scram reactivity. Surveillance of each individual control rod's scram time ensures the scram reactivity assumed in the DBA and transient analyses can be met. The scram function of the CRD System protects the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and the 1% cladding lastic strain fuel design limit (see Bases for LCO 3.2.1, "AV-RAGE LANAR LINEAR HEAT GENERATION RATE (APLHGR)'), which=+/- nsure that no fuel damage will occur if these limits are noexceeded. 0 800 psig, the scram function is designe negative 0D larid LCO 3.2.3, LINEAR HEAT GENERATION RATE (LHGR) I BWR/4 STS B 3.1.4-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 108 of 231

Attachment 1, Volume 6, Rev. 0, Page 109 of 231 B 3.1.4 0 INSERT 1 is designed to accommodate plant operational transients or maneuvers which might be expected without compromising safety and without fuel damage Insert Page B 3.1.4-1 Attachment 1, Volume 6, Rev. 0, Page 109 of 231

Attachment 1, Volume 6, Rev. 0, Page 110 of 231 Control Rod Scram Times B 3.1.4 BASES APPLICABLE SAFETY ANALYSES (continued) reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power transient. Below 800 psig, the scram function is assumed to perform design during the control rod drop accident (Ref. 5) and, therefore, also provides protection against violating fuellda agefimits during reactivity insertion (,I) accidents (see Bases for LCO 3.1.6, "Rod Pattern Control"). For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits. Control rod scram times satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO The scram times specified in Table 3.1.4-1 1(in the ac ning LCO) (i) are required to ensure that the scram reactivity assumed in the DBA and transient analysis is met (Ref. 6). To account for single failures and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis analysis. The scram times have a margin that allows up to approximately 7% of the control 3 rods (e.g. 7x 7 °/W~ to have scram times exceeding the specified ( limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times. The scram times are specified relative to measurenients based on reed switch positions, which provide the control rod position indication. The reed switch closes ("pickup") when the Index tube passes a specific location and then opens ("dropout") as the index tube travels upward. Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the "dropout" times. To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed "slow" control rods mav occupy adjacent locations. I(face or diagVonal) Table 3.1.4-1 is modified by two Notes which state that control rods with scram times not within the limits of the Table are considered "slow" and that control rods with scram times > 7 seconds are considered inoperable as required by SR 3.1.3.4. This LCO applies only to OPERABLE control rods since inoperable control rods will be inserted and disarmed (LCO 3.1.3). Slow scramming control rods may be conservatively declared inoperable and not accounted for as "slow" control rods. BWR/4 STS B 3.1.4-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 110 of 231

Attachment 1, Volume 6, Rev. 0, Page 111 of 231 Control Rod Scram Times B 3.1.4 BASES APPLICABILITY In MODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are assumed to occur during startup and power operation; therefore, the scram function of the control rods is required during these MODES. In MODES 3 and 4, the control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate requirements for control rod scram capability during these conditions. Scram requirements in MODE 5 are contained in LCO 3.9.5, "Control Rod OPERABILITY - Refueling." ACTIONS A.1 When the requirements of this LCO are not met, the rate of negative reactivity insertion during a scram may not be within the assumptions of' the safety analyses. Therefore, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE The four SRs of this LCO are modified by a Note stating that during a REQUIREMENTS single control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated, (i.e., charging valve closed) the influence of the CRD pump head does not affect the single control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect on the scram insertion times. SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on an assumed control rod scram time. Measurement of the scram times with reactor steam dome pressure 2 800 psig demonstrates acceptable scram times for the transients analyzed in ReferencesrS-and3. 0 Maximum scram insertion times occur at a reactor steam dome pressure of approximately 800 psig because of the competing effects of reactor steam dome pressure and stored accumulator energy. Therefore, demonstration of adequate scram times at reactor steam dome pressure 2 800 psig ensures that the measured scram times will be within the BWR/4 STS B 3.1.4-3 . Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 111 of 231

Attachment 1, Volume 6, Rev. 0, Page 112 of 231 Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE REQUIREMENTS (continued) specified limits at higher pressures. Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed. To ensure that scram time testing is performed within a reasonable time following a shutdown 2 120 days or longer, control rods are required to be tested before exceeding 40% RTP following the shutdown. This Frequency Is acceptable considering the additional Surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by fuel movement within the associated core cell and by work on control rods or the CRD System. SR 3.1.4.2 Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A Age representative sample contains at least 10% of the control rods. The 460 sample remains representative if no more than 1t% of the control rods the sample tested are determined to be "slow." With more than of the sample declared to be "slow" per the criteria in Table 3.1.4-1 additional control rods are tested until this criterion (e.g., of the entire sample size) is satisfied, or until the total number of "slow" control rods (throughout the core, from all Surveillances) exceeds the LCO limit. For planned testing, the control rods selected for the sample should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data may have been previously tested in a sample. The f'J-"1I day Frequency is based on operating experience that has shown (TSTF->4 control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable based on the additional Surveillances done on the CRIs at more frequent intervals in accordance with LCO 3.1.3 and LCO 3.1.5, "Control Rod Scram Accumulators." SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures fro zero mimum )the BWR/4 STS B 3.1.4-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 112 of 231

Attachment 1, Volume 6, Rev. 0, Page 113 of 231 Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE REQUIREMENTS (continued) lscram tme I found In the Technical Requirements Manual (Ref. 7) and are issie pr sure. The scram testing must be performed once before/ declaring the control rod OPERABLE. The required scram time testing / must demonstrate the affected control rod is still within acceptable limits./ Th imits for reactor pressures < 800 psig arebased high probability of meeting the acceptance criteria at reactor pressures on a -D(D 2 800 psig. Limits for 2 800 psig are found in Table 3.1.4-1. If testing demonstrates the affected control rod does not meet these limits, but is within the 7-second limit of Table 3.1.4-1, Note 2, the control rod can be declared OPERABLE and "slow." Specific examples of work that could affect the scram times are (but are not limited to) the following: removal of any CRD for maintenance or modification; replacement of a control rod; and maintenance or modification of a scram solenoid pilot valve, scram valve, accumulator, isolation valve or check valve in the piping required for scram. The Frequency of once prior to declaring the affected control rod OPERABLE is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. SR 3.1.4.4 When work that could affect the scram insertion time is performed on a control rod or CRD System, or when fuel movement within the reactor pressure vessel occurs, testing must be done to demonstrate each affected control rod is still within the limits of Table 3.1.4-1 with the reactor steam dome pressure 2 800 psig. Where work has been performed at high reactor pressure, the requirements of SR 3.1.4.3 and SR 3.1.4.4 can be satisfied with one test. For a control rod affected by work performed while shut down, however, a zero pressure and high pressure test may be required. This testing ensures that, prior to withdrawing the control rod for continued operation, the control rod scram performance is acceptable for operating reactor pressure conditions. Alternatively, a control rod scram test during hydrostatic pressure testing could also satisfy both criteria. When fuel movement within the reactor pressure vessel occurs, only those control rods associated with the core cells affected by the fuel movement are required to be scram time tested. During a routine refueling outage, it is expected that all control rods will be affected. BWR/4 STS B 3.1.4-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 113 of 231

Attachment 1, Volume 6, Rev. 0, Page 114 of 231 Control Rod Scram Times B 3.1.4 BASES SURVEILLANCE REQUIREMENTS (continued) The Frequency of once prior to exceeding 40% RTP is acceptable because of the capability to test the control rod over a range of operating conditions and the more frequent surveillances on other aspects of control rod OPERABILITY. REFERENCES 1. 0CFR50, xA, GDC1_

2. SARJ Sectio .4.3.2.2.4 O(RSAR, _ci 4 SAR, Sectj 15.111 .

5. NEDE-24011-P-Q, "General Electric Standard Application for Reactor Fuelm"ISection 3.2. tember 198

                                                       ..      l (-regaisn as specified in Specification 5.6.3) l

6. Letter from R.F. Janecek (BWROG) to R.W. Starostecki (NRC),

                    "BWR Owners Group Revised Reactivity Control System Technical Specifications," BWROG-8754, September 17, 1987.

7. ecniclRqieet aul a0 BWR/4 STS B 3.1.4-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 114 of 231

Attachment 1, Volume .6, Rev. 0, Page 115 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.4 BASES, CONTROL ROD SCRAM TIMES

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

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

3. Typographical/grammatical error corrected.

4. The brackets have been removed and the proper plant specific information has been provided.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 115 of 231

Attachment 1, Volume 6, Rev. 0, Page 116 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1,Volume 6, Rev. 0, Page 116 of 231

Attachment 1, Volume 6, Rev. 0, Page 117 of 231 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.4, CONTROL ROD SCRAM TIMES There are no specific NSHC discussions for this Specification. Monticello Page I of 1 Attachment 1, Volume 6, Rev. 0, Page 117 of 231

Attachment 1, Volume 6, Rev. 0, Page 118 of 231 ATTACHMENT 5 ITS 3.1.5, Control Rod Scram Accumulators Attachment 1, Volume 6, Rev. 0, Page 118 of 231

Attachment 1, Volume 6, Rev. 0, Page'119 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 6, Rev. 0, Page 119 of 231

c c C ITS 3.1.5 ITS ITS CD 0 LCO 3.1.5 0 CD X, 0 0 -o ACTIONS A, B, and C

                                                                                            *0 C)

CD C) Co ri3 CD 0 0 0 0u 3.314.3 82 10/26101 Amendment No. 5,11, 13, 54, 63, 140, 123 Page 1 of 2

C C ITS 3.1.5 C ITS 0 3.0 UMING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS i F. Scram Discharge Volume F. Scram Discharge Volume

1. During reactor operation, the scram discharge The scram discharge volume vent and drain valves shall a) volume vent and drain valves shall be operable, be cycled quarterly. 0 except as specified below. Once per operating cycle verify the scram discharge volume vent and drain valves dose within 30 seconds El CD 0 2. Ifany scam discharge volume vent or drain valve Is after receipt of a reactor scram signal and open when made or found Inoperable, the Integrity of the scram the scram is reset.

discharge volume shall be maintained by either 0 CD

a.

• Verifying daily, for a period not to exceed 7 days, the operability of the redundant valve(s),

or 0 b. Maintaining the Inoperable valve(s), or the associated redundant valve(s), In the dosed -- i See ITS 3.1.81 CD position. Periodically the Inoperable and the 0) redundant valve(s) may both be Inthe open CD position to allow draining the scram discharge -o volume. CD CD l ~If a or b above cannot be met, at least anl but one

           <                   operable control rods (not Inchuing rods removed

0) 1. .

I perspedlation3.1 .Eorinoperableredsalowed by 3.3A2) shall be fully Inserted within ten hours. 03

                                                                                                                                                                                    -9' ACTION D        N    .Required Action IfHSpecil cations 3.4<<rug9             bv are notmea          (except when the reactor mode (A)                        orderly shutdown shall        initiated and have reactor In   switch Is In the Refuel position)                                                           CA) the cold shutdown ndltion within 24 hours.

_ dd proposed ACTION D L3 83a 5/1/84 Amendment No. 24 Page 2 of 2

Attachment 1, Volume 6, Rev. 0, Page 122 of 231 DISCUSSION OF CHANGES ITS 3.1.5, CONTROL ROD SCRAM ACCUMULATORS 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 3.3.D states, in part, that if "a" control rod with an inoperable accumulator is inserted full-in and either its directional control valves are electrically disarmed or is hydraulically isolated, it shall not be considered to have an inoperable accumulator. CTS 3.3.D.1 states, in part, that "a" rod accumulator may be inoperable provided that no other control rod within two control rod cells in any direction has an inoperable accumulator or directional control valve are electrically disarmed while in a non-fully inserted position. These CTS Actions do not limit the number of accumulators to which these Actions apply. ITS 3.1.5 ACTIONS Note allows separate Condition entry for each control rod scram accumulator. This changes the CTS by adding an explicit Note for separate Condition entry for each control rod scram accumulator. The purpose of CTS 3.3.D and CTS 3.3.D.1, in part, is to provide compensatory actions for an inoperable scram accumulator on an individual basis. ITS 3.1.5 ACTION Note "Separate Condition entry is allowed for each control rod scram accumulator" has been added and provides more explicit instructions for proper application of the ACTIONS for Technical Specifications compliance. In conjunction with proposed Specification 1.3, "Completion Times," this Note provides direction consistent with the intent of the existing CTS ACTIONS for inoperable control rod accumulators. Upon discovery of each inoperable accumulator, each specified ACTION is applied, regardless of previous application to other inoperable accumulators. This change is designated as administrative because it does not result in technical changes to the CTS. A.3 These changes to CTS 3.3.G 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. MORE RESTRICTIVE CHANGES M.1 CTS 4.3.0 requires a check of the accumulator pressure alarm located in the control room. ITS SR 3.1.5.1 requires a verification that each control rod scram accumulator pressure is > 940 psig. This changes the CTS by providing an explicit value for control rod accumulator pressure, in lieu of specifying the alarm in the control room must be checked. The purpose of CTS 4.3.D is to ensure that each control rod scram accumulator is OPERABLE. ITS SR 3.1.5.1 includes the acceptance criteria for accumulator Monticello Page 1 of 7 Attachment 1, Volume 6, Rev. 0, Page 122 of 231

Attachment 1, Volume 6, Rev. 0, Page 123 of 231 DISCUSSION OF CHANGES ITS 3.1.5, CONTROL ROD SCRAM ACCUMULATORS pressure (2 940 psig) consistent with current Monticello practice, and requires verification that each accumulator meets this pressure criterion. Although this change is consistent with current practice, adding this acceptance criterion and verification requirement in ITS SR 3.1.5.1 is an additional restriction on unit operation since control of this requirement will now be governed by Technical Specifications. This change is designated as more restrictive because it adds an explicit Surveillance limit that does not appear in the CTS. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 3.3.D states, in part, that if a control rod with an inoperable accumulator is inserted full-in and either its "directional control valves are electrically "disarmed" or it is hydraulically isolated," it shall not be considered to have an inoperable accumulator. ITS 3.1.3 ACTION C covers the compensatory actions for one or more inoperable control rods (control rods inoperable as a result of an inoperable accumulator is covered by this condition when declared inoperable) ITS 3.1.3 Required Action C.2 states "Disarm the associated CRD," but does not provide the specific details of how to disarm the CRD. This changes the CTS by relocating the details that the "directional control valves are electrically" disarmed "or it is hydraulically isolated" to the ITS Bases. The removal of these details for performing Required Actions from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. ITS 3.1.3 Required Action C.2 still retains the requirement to disarm the CRD. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 7 - Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.3.D requires a check of the status in the control room of the required OPERABLE accumulator every 12 hours. ITS SR 3.1.5.1 requires a similar verification that the pressure in each accumulator is > 940 psig every 7 days. This changes the CTS extending the Surveillance Frequency from once every 12 hours to every 7 days. Monticello Page 2 of 7 Attachment 1, Volume 6, Rev. 0, Page 123 of 231

Attachment 1, Volume 6, Rev. 0, Page 124 of 231 DISCUSSION OF CHANGES ITS 3.1.5, CONTROL ROD SCRAM ACCUMULATORS The purpose of CTS 4.3.0 is to ensure the control rod scram accumulators are OPERABLE to support the associated control rod scram function. This change is acceptable because the new Surveillance Frequency has been evaluated to ensure that it provides an acceptable level of equipment reliability. This change allows the unit to perform the Surveillance every 7 days instead of every 12 hours. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications (i.e., alarm) available in the control room. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS. L.2 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria) CTS 4.3.0 requires, in part, the check of the status in the control room of the required OPERABLE accumulator level alarm. The ITS does not include this requirement. This changes the CTS by deleting the requirement to verify the alarm for accumulator level in the control room. The purpose of CTS 4.3.0 is to ensure each control rod scram accumulator is OPERABLE to support the associated control rod scram function. 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. ITS SR 3.1.5.1 requires verification that the accumulator pressure is within the pressure limit for each accumulator. The actual limit has been added as described in DOC M.1. This change deletes the requirement to verify OPERABILITY of the control rod accumulators via the accumulator level alarm in the control room. The ISTS do not specify OPERABILITY requirements for equipment that only provides indication to support OPERABILITY of a system or component. The control rod scram accumulator level alarm does not necessarily relate directly to accumulator OPERABILITY. Control of the availability of, and necessary compensatory activities, for alarms, are addressed by unit procedures and policies. The requirement to verify control rod scram accumulator pressure (which does relate directly to accumulator OPERABILITY) iswithin limits is still maintained in ITS SR 3.1.5.1. Therefore, the requirements associated with the control rod accumulator level alarm are proposed to be removed from the Technical Specifications. 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 4 - Relaxation of Required Action) CTS 3.3.0 states, in part, that if a control rod with an inoperable accumulator is inserted full-in and is disarmed, it shall not be considered to have an inoperable accumulator. CTS 3.3.D.1 also states a control rod scram accumulator may be inoperable provided that no other control rod within two control rod cells in any direction has an inoperable accumulator or a directional control valve electrically disarmed while in a non-fully inserted position. CTS 3.3.G.1 states, in part, that if Specification 3.3.0 is not met, an orderly shutdown shall be initiated and the reactor shall be placed in the cold shutdown (MODE 4) condition within 24 hours. CTS 3.3.0 and CTS 3.3.D.1 do not provide any time to Insert control rods associated with inoperable control rod accumulators, therefore as soon as it is determined that a control rod accumulator is inoperable and the provisions of CTS 3.3.D.1 are not Monticello Page 3 of 7 Attachment 1, Volume 6, Rev. 0, Page 124 of 231

Attachment 1, Volume 6, Rev. 0, Page 125 of 231 DISCUSSION OF CHANGES ITS 3.1.5, CONTROL ROD SCRAM ACCUMULATORS met, CTS 3.3.G.1 must be immediately entered. ITS 3.1.5 ACTION A covers the condition of one control rod scram accumulator inoperable with reactor steam dome pressure 2 900 psig, and requires the declaration within 8 hours that either the associated control rod scram time is slow (ITS 3.1.5 Required Action A.1) or the associated control rod is inoperable (ITS 3.1.5 Required Action A.2). The requirement to declare the associated control rod slow is only applicable if the associated control rod scram time was within limits during the last scram time' Surveillance. ITS 3.1.5 ACTION B covers the Condition for two or more control rod scram accumulators inoperable with reactor steam dome pressure 2 900 psig, and requires the restoration of charging water header pressure to 2 940 psig within 20 minutes from discovery of Condition B (i.e., two or more control rod scram accumulators inoperable with steam dome pressure

      > 900 psig) concurrent with charging water header pressure < 940 psig (ITS 3.1.5 Required Action B.1) and within 1 hour to either declare the associated control rod scram time slow (ITS 3.1.5 Required Action B.2.1) or declare the associated control rod inoperable (ITS 3.1.5 Required Action B.2.2). The requirement to declare the associated control rod scram time slow is only applicable if the associated control rod scram time was within limits during the last scram time Surveillance. ITS 3.1.5 ACTION C covers the condition for one or more control rod scram accumulators inoperable with reactor steam dome pressure < 900 psig, and requires the immediate verification that all control rods associated with inoperable accumulators are fully inserted upon discovery of charging water header pressure < 940 psig (ITS 3.1.5 Required Action C.1) and the declaration within 1 hour that the associated control rod is inoperable (ITS 3.1.5 Required Action C.2). ITS 3.1.5 ACTION D covers the condition when Required Action B.1 or C.1 and associated Completion Time is not met, and requires the immediate placement of the reactor mode switch in the shutdown position (Required Action D.1). This Required Action is not applicable if all inoperable control rod scram accumulators are associated with fully inserted control rods. This changes the CTS in several ways, some administrative, some more restrictive, and some less restrictive. However, all these changes are discussed in this single less restrictive change discussion for clarity. The individual changes and their justification and categorization are as follows:

• The ITS 3.1.5 ACTIONS for inoperable control rods are configured based upon the reactor steam dome pressure (i.e., > 900 psig and < 900 psig). At reactor pressures < 900 psig and with control rod scram accumulators inoperable, the resultant control rod scram time is not expected to satisfy the minimum scram time requirement of ITS SR 3.1.3.4 (i.e., 7 second scram time requirement). ITS 3.1.5 ACTION C covers the condition of one or more control rod scram accumulators inoperable with reactor steam dome pressure

          < 900 psig. When ACTION C is entered and it is discovered that charging water header pressure is < 940 psig, an immediate verification is required to ensure that all control rods associated with inoperable scram accumulators are fully inserted. With one or more control rod scram accumulators inoperable and the reactor steam dome pressure < 900 psig, the pressure supplied to the charging water header must be adequate to ensure that accumulators remain charged. With the reactor steam dome pressure
          < 900 psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become Monticello                              Page 4 of 7 Attachment 1, Volume 6, Rev. 0, Page 125 of 231

Attachment 1, Volume 6, Rev. 0, Page 126 of 231 DISCUSSION OF CHANGES ITS 3.1.5, CONTROL ROD SCRAM ACCUMULATORS severely degraded during a depressurization event or at low reactor pressures. Therefore, immediately upon discovery of charging water header pressure < 940 psig, concurrent with Condition C, all control rods associated with inoperable accumulators must be verified to be fully inserted. If ITS 3.1.5 Required Action C.1 and its associated Completion Time cannot be met ITS 3.1.4 ACTION D requires the immediate placement of the reactor mode switch to the shutdown position. This will ensure all control rods are inserted into the reactor core and that the reactor is in a condition that does not require the active function (i.e., scram) of the control rods. In addition, within 1 hour the associated control rod must be declared inoperable. Withdrawn control rods with inoperable accumulators may fail to scram under these low pressure conditions. When a control rod is declared inoperable, ITS 3.1.3 ACTION C requires the insertion of the control rod within 3 hours and the disarming of the control rod drive in 4 hours. CTS 3.3.D does not provide any additional restrictions for inoperable control rod scram accumulators at low reactor steam dome pressures. Thus, CTS allows rods to remain not-fully inserted at low reactor pressures (< 900 psig) with inoperable accumulators. These changes are acceptable because they place the control rods in a safer condition under the specified conditions. This change is more restrictive because ITS 3.1.5 ACTION C requires the control rod to be declared inoperable and ITS 3.1.3 ACTION C will require the rod to be inserted and the control rod drive mechanism disarmed whereas in the CTS the control rod may remain withdrawn as long as the criteria in CTS 3.3.D.1.(a) and (b)are met. In addition, it is also more restrictive as a result of the addition of the explicit requirements for when charging water header pressure is not within limit and the requirement to place the reactor mode switch in the shutdown condition under certain conditions.

• The ITS 3.1.5 ACTIONS for inoperable control rods are configured based on the reactor steam dome pressure (i.e., > 900 psig and < 900 psig). At reactor pressures > 900 psig and with a control rod scram accumulators inoperable, the resultant scram time of the associated rod may still satisfy the minimum scram time requirements of ITS SR 3.1.3.4 (i.e., 7 second scram time requirement) and the scram time criteria of ITS Table 3.1.4-1. Therefore, ITS 3.1.5 ACTION A allows the associated control rods to be declared either inoperable or slow. This declaration must be performed within 8 hours. If two or more control rod scram accumulators are Inoperable with reactor steam dome pressure > 900 psig, ITS 3.1.5 ACTION B requires the associated control rods to also be declared either inoperable or slow. However, this declaration must be performed within 1 hour.: If during this condition (i.e., in Condition B), it is found that the charging water header pressure is

          < 940 psig, it must be restored to > 940 psig within 20 minutes. If this cannot be met the reactor mode switch must be placed in the shutdown position. If a control rod has an inoperable accumulator in the CTS, it must be inserted and disarmed or it may be allowed to remain withdrawn as long as the criteria of CTS 3.3.D.1.(a) and (b)are met. The CTS essentially allows 24 hours to satisfy the requirements of CTS 3.3.D or 3.3.D.1, since CTS 3.3.G.1 allows 24 hours to place the reactor in cold shutdown. In the ITS, 8 hours is allowed to declare the rod inoperable or slow if one control rod scram accumulator is inoperable with reactor steam dome pressure > 900 psig. If declared inoperable, ITS 3.1.3 ACTION C allows 3 hours to fully insert the rod and Monticello                                Page 5 of 7 Attachment 1, Volume 6, Rev. 0, Page 126 of 231

Attachment 1, Volume 6, Rev. 0, Page 127 of 231 DISCUSSION OF CHANGES ITS 3.1.5, CONTROL ROD SCRAM ACCUMULATORS 4 hours to disarm it. Therefore, with one control rod scram accumulator inoperable with reactor steam dome pressure > 900 psig, the ITS requires the same operations to be completed in 12 hours (8 hours to declare the control rod inoperable and 4 hours to insert and disarm it). With one or more control rod scram accumulators inoperable with reactor steam dome pressure

          > 900 psig, the ITS requires the same operations to be completed in 5 hours (1 hour to declare the control rod inoperable and 4 hours to insert and disarm it). This change is acceptable since it places the reactor in a safer condition sooner under the same conditions and prescribes explicit requirements for when the charging water header pressure requirements cannot be met. This portion of the change is more restrictive since less time is provided to perform the same actions (insert and disarm control rods) and provides more explicit action for when charging water header pressure is not within limits.

• CTS 3.3.D.1 includes a provision that allows control rods with inoperable control rod scram accumulators to remain withdrawn as long as there is no other inoperable control rod (i.e., inoperable accumulator, or directional control valve electrically disarmed while in a non-fully inserted position) within two control rod cells in any direction. ITS 3.1.5 ACTION A includes a similar allowance only if the control rod is declared slow. When a rod is declared slow, an evaluation is normally performed to ensure LCO 3.1.4 is met. ITS LCO 3.1.4.a includes a requirement that limits the total number of OPERABLE control rods that are "slow" to 8 and ITS LCO 3.1.4.b includes a requirement that allows no more than 2 OPERABLE control rods that are slow to occupy adjacent locations. This changes the CTS by effectively limiting the total number of withdrawn control rods with inoperable control rod scram accumulators to 8. This also changes the CTS by allowing 2 OPERABLE control rods that are slow (i.e., the control rod has an inoperable accumulator and is not fully inserted) to occupy adjacent locations and allows other slow control rods to only be separated by a single OPERABLE control rod. This change is acceptable since an analysis has been performed to ensure the control rod scram reactivity can be met when in this configuration.

The scram times specified in ITS Table 3.1.4-1 are required to ensure that the scram reactivity assumed in the design basis accident and transient analyses is met. To account for single failures and slow scramming control rods (control rods with inoperable scram accumulators), the scram times specified in ITS Table 3.1.4-1 are faster than'those assumed in the design basis analyses. The scram times have a margin that allows'up to approximately 7% of the control rods (i.e., 8)to have scram times exceeding the specified limits (i.e., slow control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod OPERABILITY") and an additional control rod failing to scram per the single failure criterion. To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed slow control rods may occupy adjacent locations. This change is acceptable because the limitations placed on the inoperable control rod scram accumulators will ensure the safety analyses will be met. The change limiting the total number of slow control rods to 8 is more restrictive than the CTS while the changes related to separation criteria are less restrictive than the CTS. Monticello Page 6 of 7 Attachment 1,Volume 6, Rev. 0, Page 127 of 231

Attachment 1, Volume 6, Rev. 0, Page 128 of 231 DISCUSSION OF CHANGES ITS 3.1.5, CONTROL ROD SCRAM ACCUMULATORS

• CTS 3.3.G.1 states, in part, that if Specification 3.3.D is not met, an orderly shutdown shall be initiated and the reactor shall be placed in the cold shutdown (MODE 4) condition within 24 hours. ITS 3.1.5 ACTION D covers the condition when Required Action B.1 or C.1 and associated Completion Time is not met, and requires the immediate placement of the reactor mode switch in the shutdown position. Placing the reactor mode switch in shutdown places the reactor in hot shutdown (MODE 3). This change is considered acceptable since CTS 3.3.D, in part, is applicable in the Startup and Run conditions, i.e., MODES 1 and 2. Thus, once MODE 3 is achieved, continuation to MODE 4 is no longer required. Therefore, this change is considered administrative since the technical requirements are not being changed; the change is being made to ensure the shutdown actions are consistent with the current Applicability.

Monticello Page 7 of 7 Attachment 1, Volume 6, Rev. 0, Page 128 of 231

Attachment 1, Volume 6, Rev. 0, Page 129 of 231 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 129 of 231

Attachment 1, Volume 6, Rev. 0, Page 130 of 231 Control Rod Scram Accumulators 3.1.5 CTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Rod Scram Accumulators 3.3.D LCO 3.1.5 Each control rod scram accumulator shall be OPERABLE. APPLICABILITY: MODES I and 2. ACTIONS

                     -----------------------                     'JL KI T.r%

I r- - -- -- - DOC Separate Condition entry is allowed for each control rod scram accumulator. A.2 CONDITION REQUIRED ACTION l COMPLETION TIME 3.3.D, A. One control rod scram A.1 ------- NOTE----- 3.3.D.1 accumulator inoperable Only applicable if the with reactor steam dome associated control rod pressure 2:M90O0psig. scram time was within the 0 limits of Table 3.1.4-1 during the last scram time Surveillance. Declare the associated 8 hours control rod scram time "slow." OR A.2 Declare the associated 8 hours control rod inoperable.

                                                    -4                                            4 3.3.0D    B. Two or more control rod                   B.1      Restore charging water              20 minutes from 3.3.D.1       scram accumulators                                header pressure to                  discovery of inoperable with reactor                           2R9401psig.                         Condition B steam dome pressure                                                                   concurrent with 21900mpsig.                                                                           charging water header pressure
                                                                                                                             }0
                                                                                                    <M940M psig AND BWR/4 STS                                                    3.1.5-1                             Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 130 of 231

Attachment 1, Volume 6, Rev. 0, Page 131 of 231 Control Rod Scram Accumulators 3.1.5 CTS ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.3.D. 3.3.D.1 B.2.1 ------- NOTES-- Only applicable if the associated control rod scram time was within the limits of Table 3.1.4-1 during the last scram time Surveillance. Declare the associated 1 hour control rod scram time "slow." OR B.2.2 Declare the associated 1 hour control rod inoperable. \ )' 3.3.D, C. One or more control rod C.1 Verify all control rods Immediately upon 3.3.D.1 scram accumulators associated with inoperable discovery of charging inoperable with reactor accumulators are fully water header steam dome pressure

                   < MOO0( psig.

inserted. pressure <M94OJpsig 0D AND C.2 Declare the associated 1 hour control rod inoperable. 3.3.G.1 D. Required Actio and D.1 ----- NOTE-- associated Completion Time Lot Re x Actionl Not applicable If all inoperable control rod 0 or C. not met. scram accumulators are associated with fully inserted control rods. Place the reactor mode Immediately switch in the shutdown position. BWR/4 STS 3.1.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 131 of 231

Attachment 1, Volume 6, Rev. 0, Page 132 of 231 Control Rod Scram Accumulators 3.1.5 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 4.3.D SR 3.1.5.1 Verify each control rod scram accumulator pressure 7 days is 494Clpsig. 0D BWR/4 STS 3.1.5-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 132 of 231

Attachment 1, Volume 6, Rev. 0, Page 133 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5, CONTROL ROD SCRAM CONTROL ROD SCRAM ACCUMULATORS

1. The brackets are removed and the proper plant specific information/value is provided.

2. The wording of the Condition has been made to be consistent with a similar type of requirement in another Specification in NUREG-1433, Rev. 3 (i.e., ISTS 3.5.2 Condition D). This change was also approved in the ITS conversion for the four most recently approved BWR conversions (Quad Cities 1 and 2, Dresden 2 and 3, LaSalle 1 and 2, and FitzPatrick).

Monticello Page 1of 1 Attachment 1, Volume 6, Rev. 0, Page 133 of 231

Attachment 1, Volume 6, Rev. 0, Page 134 of 231 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 134 of 231

Attachment 1, Volume 6, Rev. 0, Page 135 of 231 Control Rod Scram Accumulators B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Control Rod Scram Accumulators BASES BACKGROUND The control rod scram accumulators are part of the Control Rod Drive (CRD) System and are provided to ensure that the control rods scram under varying reactor conditions. The control rod scram accumulators store sufficient energy to fully insert a control rod at any reactor vessel pressure. The accumulator is a hydraulic cylinder with a free floating piston. The piston separates the water used to scram the control rods from the nitrogen, which provides the required energy. The scram accumulators are necessary to scram the control rods within the required insertion times of LCO 3.1.4, "Control Rod Scram Times." APPLICABLE The analytical methods and assumptions used in evaluating the control SAFETY rod scram function are presented in References 1, 2, and 3. The Design ANALYSES Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate. OPERABILITY of each individual control rod scram accumulator, along with LCO 3.1.3, "Control Rod OPERABILITY," and LCO 3.1.4, ensures that the scram reactivity assumed in the DBA and transient analyses can be met. The existence of an inoperable accumulator may invalidate prior scram time measurements for the associated control rod. The scram function of the CRD System, and therefore the OPERABILITY of the accumulators, protects the MCPR Safety Limit (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"), which ensure that no fuel damage will occur if these limits are not exceeded (see Bases for LCO 3.1.4). In addition, the scram function at low reactor vessel pressure (i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control"). Control rod scram accumulators satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO The OPERABILITY of the control rod scram accumulators is required to ensure that adequate scram insertion capability exists when needed over the entire range of reactor pressures. The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure. BWR/4 STS B 3.1.5-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 135 of 231

Attachment 1,Volume 6, Rev. 0, Page 136 of 231 Control Rod Scram Accumulators B 3.1.5 BASES APPLICABILITY In MODES 1 and 2, the scram function is required for mitigation of DBAs and transients, and therefore the scram accumulators must be OPERABLE to support the scram function. In MODES 3 and 4, control qr-ods-`argo w to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. This provides adequate 0D requirements for control rod scram accumulator OPERABILITY during these conditions. Requirements for scram accumulators in MODE 5 are contained in LCO 3.9.5, "Control Rod OPERABILITY - Refueling." ACTIONS The ACTIONS/able is modified by.a Note indicating that a separate 0 Condition entry is allowed for each control rod scram accumulator. This is acceptable since the Required Actions for each Condition provide appropriate compensatory actions for eact000 accumulator. e C i teRequired Actions may allow for continued operation and subsequen laiie accumulators governed by subsequent Condition entry and application of associated Required Actions. 0 A.1 and A.2 With one control rod scram accumulator inoperable and the reactor steam dome pressure > 900 psig, the control rod may be declared "slow," since the control rod will still scram at the reactor operating pressure but may not satisfy the required scram times in Table 3.1.4-1. Required Action A.1 is modified by a Note indicating that declarin the controro F 7) "slow" only applies if the associated contro scram time was within th Surveillance limits of T ble 3.1.4-1 during the last scram timels. Otherwise, the

             -    ~control li)      ro w Id already be considered "slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times. In this event, the associated control rod is declared inoperable (Required Action A.2) and LCO 3.1.3 is entered.

This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function, in accordance with ACTIONS of LCO 3.1.3. The allowed Completion Time of 8 hours is reasonable, based on the large number of control rods available to provide the scram function and the ability of the affected control rod to scram only with reactor pressure at high reactor pressures. BWR/4 STS B 3.1.5-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 136 of 231

Attachment 1, Volume 6, Rev. 0, Page 137 of 231 Control Rod Scram Accumulators B 3.1.5 BASES ACTIONS (continued) B.1, B.2.1, and B.2.2 With two or more control rod scram accumulators inoperable and reactor steam dome pressure 2 900 psig, adequate pressure must be supplied to the charging water header. With inadequate charging water pressure, all of the accumulators could become inoperable, resulting in a potentially severe degradation of the scram performance. Therefore, within 20 minutes from discovery of charging water header pressure < 940 psig concurrent with Condition B, adequate charging water header pressure must be restored. The allowed Completion Time of 20 minutes is reasonable, to place a CRD pump into service to restore the charging header pressure, if required. This Completion Time is based on the ability of the reactor pressure alone to fully insert all control rods. The control rod may be declared "slow," since the control rod will still scram using only reactor pressure, but may not satisfy the times in Table 3.1.4-1. Required Action B.2.1 is modified by a Note indicating that 2 d declaring the control rod "slow" only applies if the associated control ma scram time is within the limits of Table 3.1.4-1 during the last scram time Otherwise, the control rod ready e considered "slow" and the further degradation of scram performance with an inoperable accumulator could result in excessive scram times. In this event, the [theACTIONS o associated control rod is declared inoperable (Required Action B.2.2) and

                     ,LCO 3.1.3 entered. This would result in requiring the affected control rod to be fully inserted and disarmed, thereby satisfying its intended function in accordance with ACTIONS of LCO 3.1.3.

The allowed Completion Time of 1 hour is reasonable, based on the ability of only the reactor pressure to scram the control rods and the low probability of a DBA or transient occurring while the affected accumulators are inoperable. C.1 and C.2 With one or more control rod scram accumulators inoperable and the reactor steam dome pressure < 900 psig, the pressure supplied to the charging water header must be adequate to ensure that accumulators remain charged. With the reactor steam dome pressure < 900 psig, the function of the accumulators in providing the scram force becomes much more important since the scram function could become severely BWR/4 STS B 3.1.5-3 Rev. 3.0, 03/31/04 Attachment 1,Volume 6, Rev. 0, Page 137 of 231

Attachment 1, Volume 6, Rev. 0, Page 138 of 231 Control Rod Scram Accumulators B 3.1.5 BASES ACTIONS (continued) degraded during a depressurization event or at low reactor pressures. Therefore, immediately upon discovery of charging water header pressure < 940 psig, concurrent with Condition C, all control rods associated with inoperable accumulators must be verified to be fully inserted. Withdrawn control rods with inoperable accumulators may fail to scram under these low pressure conditions. The associated control rods must also be declared inoperable within 1 hour. The allowed Completion Time of 1 hour is reasonable for Required Action C.2, considering the low probability of a DBA or transient occurring during the time that the accumulator is inoperable. D.1 The reactor mode switch must be immediately placed in the shutdown position if either Required Action and associated Completion Time associated with loss of the CRD0c in pump (Required Actions B.1 and C.1) cannot be met. This ensures that all insertable control rods are 0 inserted and that the reactor is in a condition that does not require the active function (i.e., scram) of the control rods. This Required Action is modified by a Note stating that the action is not applicable if all control rods associated with the inoperable scram accumulators are fully inserted, since the function of the control rods has been performed. SURVEILLANCE SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force. The primary indicator of accumulator OPERABILITY is the accumulator pressure. A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable. The minimum accumulator pressure of 940 psig is well below the expected pressure of 1100 psigl(R. Declaring the accumulator inoperable when the,minimum pressure is not maintained ensures that significant degradation in scram times does not occur. The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room. REFERENCES AR,Sectiop .3.2-2-.24. AR, Sectio, A.4.3. 1 3.AR, ec . BWRI4 STS B 3.1.5-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 138 of 231

Attachment 1, Volume 6, Rev. 0, Page 139 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.5 BASES, CONTROL ROD SCRAM ACCUMULATORS

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

2. Typographical/grammatical error corrected.

3. The brackets are removed and the proper plant specific information/value is provided.

4. Changes are made (additions, deletions, and/or changes) to the ISTS Bases, which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 139 of 231

Attachment 1, Volume 6, Rev. 0, Page 140 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 6, Rev. 0, Page 140 of 231

Attachment 1, Volume 6, Rev. 0, Page 141 of 231 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.5, CONTROL ROD SCRAM ACCUMULATORS There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 141 of 231

, Volume 6, Rev. 0, Page 142 of 231 ATTACHMENT 6 ITS 3.1.6, Rod Pattern Control , Volume 6, Rev. 0, Page 142 of 231

Attachment 1, Volume 6, Rev. 0, Page 143 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1,Volume 6, Rev. 0, Page 143 of 231

C C CI ITS 3.1.6 ITS T 3.0 UMmNG CONDITIONS FOR OPERATION I 4.0 SURVEILi.ANCE REOUIREMENTS (b) when the rod Is withdrawn the first time subsequent to each refueling outage. obsraerve 0). J) discernible response of the nuclear C, 0 Instrumentation. However, for Initial rods iNhen See ITS 3.1.3 } 0 response Is not discernible, subsequent exercising of these rods after the reactor i rD D critical shall be performed to observe nud ear Instrumentation response. 0 0- r I -

2. The control rod drive housing support system shall be

2. The control rod drive housing support system shall be In 0

-o place during reactor power operation and when the Inspected after reassembly and the results of the Inspection recorded. See CTS 0 reactor coolant system Is pressurized above atmospheric pressure with fuel In the reactor vessel, 314.3.B.2 J aw 0t unless all operable control rods wre fully Inserted and 0 Specification 3.3A1 Is met

                                                                                                                                                                                            -U 0

3.(a 1onrol rod zthdrawal sequences shall be e~tablishedI 3.(a) To consider the rod worth minimizer operable, the following steps must be perfomned: CD CD LCO 3.1.6 o that the faxdmum calculated reactivty tpt could be 0 dded by *ropout of any Increment of any 6ne control N 1) The control rod withdrawal sequence for the rod CD o blade wnotmaehethe conemore thani1 .% akr worth minimizer computer shall be verified as

up cel./ correct. { See ITS 3.3.21}

0 -i The rod worth minimizer computer on-line diagnostic O control rods sha comp 5L_ fPERABLE the I test shall be successfully completed. to 6 tD requirements of the banked position withdrwal to sequence (BPWS) 777 I (MI Procer annunciation of the selection error of at least one out-of-sequence control rod in each tunly Inserted group shanl be verifie. 1J l <-l Adpooed SR 3.1.6.1 l lAdd proposed Appllcabitity CD 3.3/4.3 79 1/9/81 Amendment No. 0 Page 1 of 2

C C C' ITS 3.1.6 ITS 3.0 UMmNG CONDITIONS FOR OPERATION l 4.0 SURVEILLANCE REQUIREMENTS F. Scram Discharge Volume F. Scram Discharge Volume

1. During reactor operation, the scram discharge The scram discharge volume vent and drain valves shall

0) be cycled quarterly. 0) volume vent and drain valves shall be operable, C, except as specified below. Once per operating cycle verIfy the scram discharge 3 3 volume vent and drain valves dose within 30 seconds 0

2. If any scam discharge volume vent or drain valve Is after receipt of a reactor scram signal and open when made or found Inoperable, the Integrity of the scram the scram Is reset discharge volume shal be maintained by elther 0

                                                                                                                                                                            -U CD 0                        a.

• Verifying daNy, for a period not to exceed 7 0 days. the operability of the redundant valve(s),

3 or 3

0) a)

-o b. Maintaining the Inoperable valve(s), or the associated redundant valve(s), In the dosed position. Periodically the Inoperable and the See ITS 3.1.8 } 0 redundant valve(s) may both be In the open position to allow draining the scram discharge volume. If a or b above cannot be met, at least anlbut one Co operable control rods (not Including rods removed CD

                                                                                                                                                                            ;0 per specification 3.10.E or Inoperable rods allowed CA,                      by 3.3A2) shall be fully Inserted within ten hours.

U' CA) G. Required Action 0 -9' If Specifications 3.3 thmu aove are not met. an (except when the reactor mode ACTIONS orderly shutdown shall initiated and have reactor In 1 switch IsInthe Refuel positin) A and B the cold shutd ndition within 24 hours. _ I Add proposed ACTIONS A and B F I 3.3/4.3 83a 5/1/84 I a Amendment No. 24 Page 2 of 2

Attachment 1, Volume 6, Rev. 0, Page 146 of 231 DISCUSSION OF CHANGES ITS 3.1.6, ROD PATTERN CONTROL 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 3.3.B.3.(a) states "Control rod withdrawal sequences shall be established so that the maximum calculated reactivity that could be added by dropout of any increment of any one control blade will not make the core more than 1.3% Ak supercritical." Implicit in this requirement is that once the control rod withdrawal sequence is established it will be maintained. ITS LCO 3.1.6 states "OPERABLE control rods shall comply with the requirements of the banked position withdrawal sequence (BPWS)." This changes the CTS by requiring a control rod withdrawal sequence to be continuously met by clarifying the actual control rod withdrawal sequence being used at Monticello. The change that relocates the details of the system design of control rod withdrawal sequences is discussed in DOC LA.1. The purpose of ITS LCO 3.1.6 is to provide the explicit requirements of the actual required control rod withdrawal sequence that must be used at Monticello. The change is acceptable because the Monticello USAR currently assumes the unit is utilizing the BPWS in the control rod drop accident analysis. This change is designated as administrative because it does not result in any technical changes to the CTS. A.3 These changes to CTS 3.3.G 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. MORE RESTRICTIVE CHANGES M.1 CTS 4.3.B.3.(a) does not require any verification of proper control rod sequence. ITS SR 3.1.6.1 requires verification that all OPERABLE control rods comply with bank position withdrawal sequence (BPWS) every 24 hours. This changes the CTS by adding a Surveillance requirement to verify all OPERABLE control rods comply with BPWS. This change is acceptable because it requires a verification to ensure all OPERABLE control rods comply with BPWS. This verification gives additional confidence that the control rod withdrawal sequence iswithin the bounds assumed in the control rod drop accident. This change is designated as more restrictive because it adds a Surveillance Requirement that is not required in the CTS. Monticello Page 1 of 4 Attachment 1, Volume 6, Rev. 0, Page 146 of 231

Attachment 1, Volume 6, Rev. 0, Page 147 of 231 DISCUSSION OF CHANGES ITS 3.1.6, ROD PATTERN CONTROL RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 1 - Removing Details of System Design and System Description, Including Design Limits) CTS 3.3.B.3.(a) states "Control rod withdrawal sequences shall be established so that the maximum calculated reactivity that could be added by dropout of any increment of any one control blade will not make the core more than 1.3% Ak supercritical." ITS LCO 3.1.6 states "OPERABLE control rods shall comply with the requirements of the banked position withdrawal sequence (BPWS)." This changes the CTS by relocating the details of the system design of control rod withdrawal sequences to the USAR. The removal of this detail, which is related to system design, from the Technical Specifications is acceptable because this type of information is not necessary to be included in the Technical Specifications to provide adequate protection of public health and safety. The ITS still retains that OPERABLE control rods be in compliance with the BPWS. Compliance with the BPWS will ensure the maximum reactivity limit of 1.3% Ak is met. Also, this change is acceptable because the removed information will be adequately controlled in the USAR. The USAR is controlled under 10 CFR 50.59 or 10 CFR 50.71(e), which ensures changes are properly evaluated. 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 2 - Relaxation of Applicability) CTS 3.3.B.3.(a) requires the control rod withdrawal sequences to be established but does not explicitly specify the Applicability of the control rod withdrawal sequences. However, CTS 3.3.G.1 requires the unit to be in cold shutdown (MODE 4) within 24 hours if CTS 3.3.B.3.(a) is not met. Thus this implicitly requires the control rod withdrawal sequence to be met in MODES 1, 2, and 3. ITS LCO 3.1.6 requires all OPERABLE control rods to be in compliance with the bank position withdrawal sequence in MODES 1 and 2 with THERMAL POWER < 10% RTP. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS. The purpose of CTS 3.3.B.3.(a) is to ensure the control rod withdrawal sequences are established so that the consequences of a control rod drop accident are within the bounds of the safety analysis. This change is acceptable because the control rod drop accident (CRDA) is relevant at THERMAL POWER

       < 10% RTP. CTS 3.3.B.3.(a) implies the Applicablity includes MODES 1, 2, and 3 since the default action (CTS 3.3.G.1) requires a shutdown to cold shutdown (MODE 4). At THERMAL POWER > 10% RTP, there is no credible control rod configuration that results in a control rod worth that could exceed the 280 cal/gm fuel design limit during a CRDA. In MODES 3, 4, and 5, since the reactor is shut Monticello                               Page 2 of 4 Attachment 1, Volume 6, Rev. 0, Page 147 of 231

Attachment 1, Volume 6, Rev. 0, Page 148 of 231 DISCUSSION OF CHANGES ITS 3.1.6, ROD PATTERN CONTROL down and, at most, only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable (because the reactor will remain subcritical with a single control rod withdrawn). This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS. L.2 (Category 4 - Relaxation of Required Action) CTS 3.3.G.1 requires the unit to be in cold shutdown (MODE 4) within 24 hours if the requirement of CTS 3.3.B.3.(a) (control rod withdrawal sequence requirement) is not met. ITS 3.1.6 ACTION A covers the condition when one or more OPERABLE control rods are not in compliance with BPWS, and requires the associated control rod(s) to be moved to the correct position or to declare the associated control rod(s) inoperable within 8 hours. A Note is included for ITS 3.1.6 Required Action A.1 that states the rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block Instrumentation." ITS 3.1.6 ACTION B covers the condition when nine or more OPERABLE control rods are not in compliance with BPWS, and requires the immediate suspension of control rod withdrawal and requires the reactor mode switch to be placed in the shutdown position within 1 hour. A Note similar to the one for ITS 3.1.6 Required Action A.1 is included for ITS 3.1.6 Required Action B.1. This changes the CTS by adding specific ACTIONS for OPERABLE control rods not in compliance with BPWS, in lieu of a shutdown to MODE 4. The purpose of CTS 3.3.G.1 is to place the unit in a condition in which the LCO does not apply. However, the Applicability of CTS 3.3.B.3.(a) was changed as described in DOC L.1. The purpose of the ITS 3.1.6 ACTIONS is to provide a short period of time to comply with BPWS or to declare the rods inoperable. In addition, the ITS 3.1.6 ACTIONS limit the total number of control rods not in compliance with BPWS, and if this number is exceeded, will also require exiting the Applicability of the LCO. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABILITY status of the redundant systems of required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of an accident occurring during the repair period. The ITS 3.1.6 ACTIONS provide a short period of time to comply with BPWS or to declare the rods inoperable. In addition, the ITS 3.1.6 ACTIONS limit the total number of control rods not in compliance with BPWS. With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours. Noncompliance with the prescribed sequence may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to s 10% RTP before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantly deviates from the prescribed sequence. ITS 3.1.6 Required Action A.1 is modified by a Note that allows the RWM to be bypassed to allow Monticello Page 3 of 4 Attachment 1, Volume 6, Rev. 0, Page 148 of 231

Attachment 1, Volume 6, Rev. 0, Page 149 of 231 DISCUSSION OF CHANGES ITS 3.1.6, ROD PATTERN CONTROL the affected control rods to be returned to their correct position. ITS LCO 3.3.2.1 requires verification of control rod movement by a second qualified member of the technical staff. This ensures that the control rods will be moved to the correct position. A control rod not in compliance with the prescribed sequence is not considered inoperable except as required by ITS 3.1.6 Required Action A.2. OPERABILITY of control rods is determined by compliance with LCO 3.1.3,

     'Control Rod OPERABILITY," LCO 3.1.4, "Control Rod Scram Times," and LCO 3.1.5, "Control Rod Scram Accumulators." The allowed Completion Time of 8 hours is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence. If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence.

Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod worth than withdrawals have. ITS 3.1.6 Required Action B.1 is also modified by a Note similar to the Note for ITS 3.1.6 Required Action A.1 and is acceptable for the same reasons. When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring. This change isdesignated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. Monticello Page 4 of 4 Attachment 1, Volume 6, Rev. 0, Page 149 of 231

Attachment 1,Volume 6, Rev. 0, Page 150 of 231 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 150 of 231

Attachment 1, Volume 6, Rev. 0, Page 151 of 231 Rod Pattern Control 3.1.6 crs 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Rod Pattern Control LCO 3.1.6 OPERABLE control rods shall comply with the requirements of the 3.3.B.3.(a) Rbanked position withdrawal sequence (BPWSJ 0 DOCLI APPLICABILITY: MODES I and 2 with THERMAL POWER sj1 Or'/O RTP. 0 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 3.3.G.1 A. One or more A.1 ----- NOTE---- OPERABLE control rods Rod worth minimizer not in compliance with (RWM) may be bypassed RBPWS. as allowed by LCO 3.3.2.1, "Control Rod Block 0D Instrumentation."- Move associated control 8 hours rod(s) to correct position. OR A.2 Declare associated control 8 hours

• rod(s) inoperable.

3.3.G.1 B. Nine or more B.1 - -- NOTE- --- OPERABLE control rods Rod worth minimizer not in compliance with (RWM) may be bypassed RBPWJ. as allowed by LCO 3.3.2.1. 0D Suspend withdrawal of Immediately control rods. AND BWR/4 STS 3.1.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 151 of 231

Attachment 1, Volume 6, Rev. 0, Page 152 of 231 Rod Pattern Control 3.1.6 CTrs ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME 3.3.G.1 B.2 Place the reactor mode 1 hour switch in the shutdown position. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC SR 3.1.6.1 Verify all OPERABLE control rods comply with 24 hours M.1 RBPWS. 0D BWR/4 STS 3.1.6-2 Rev. 3.0, 03131/04 Attachment 1, Volume 6, Rev. 0, Page 152 of 231

Attachment 1, Volume 6, Rev. 0, Page 153 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6, ROD PATTERN CONTROL

1. The brackets are removed and the proper plant specific information/value is provided.

Monticello Page 1 of I Attachment 1, Volume 6, Rev. 0, Page 153 of 231

Attachment 1, Volume 6, Rev. 0, Page 154 of 231 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 154 of 231

Attachment 1, Volume 6, Rev. 0, Page 155 of 231 Rod Pattern Control B 3.1.6 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.6 Rod Pattern Control BASES BACKGROUND Control rod patterns during startup conditions are controlled by the operator and the rod worth minimizer (RWM) (LCO 3.3.2.1, "Control Rod Block Instrumentation"), so that only specified control rod sequences and relative positions are allowed over the operating range of all control rods inserted to11 CV/o RTP. The sequences limit the potential amount of reactivity addition that could occur in the event of a Control Rod Drop 0D Accident (CRDA). This Specification assures that the control rod patterns are consistent with theassumptionsoftheCRDAanalysesofReferencesa dB12 [I (E) [7} , an 31 J-- APPLICABLE The analytical methods an assumptions used in evaluating the CRDA SAFETY are summarized in References a d CRDA analyses assume that ANALYSES the reactor operator follows prescribed withdrawal sequences. These sequences define the potential initial conditions for the CRDA analysis. The RWM (LCO 3.3.2.1) provides backup to operator control of the withdrawal sequences to ensure that the initial conditions of the CRDA analysis are not violated. Prevention or mitigation of positive reactivity insertion events is necessary design to limit the energy deposition in the fuel, thereby preventing significant fuel damage which could result in the undue release of radioactivity. Since the failure consequences for U02 have been shown to be y insignificant below fuel energy depositions of 300 cal/gm (Ref. A; the fuel {dapiagX limit of 280 cal/gm provides a margin of safety from significant f core damage which would result in release of radioactivity (Refs.. F 5H1 5). Generic evaluations (Refs. ass ( a CRDA resulting in a peak fuel energy deposition of 280 cal gm ave shown' that if the peak fuel enthalpy remains below 280 cal/gm, then the mi reactor pressure will be less than the required ASME Code limits (Re. and the calculated offsite doses will be well within the required limits (Ref.

                                                                                                      }Id Control rod patterns analyzed in Reference I follow the banked position withdrawal sequence (BPWS). The BPWS is applicable from the condition of all control rods fully inserted to@IC4°/o RTP (Ref. 2). For the BPWS, the control rods are required to be moved in groups, with all 0D control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12). The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation.

BWR/4 STS B 3.1.6-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 155 of 231

Attachment 1, Volume 6, Rev. 0, Page 156 of 231 Rod Pattern Control B 3.1.6 BASES APPLICABLE SAFETY ANALYSES (continued) Genic analysis of the BPWS (Ref. 1) has demonstrated that the E 280 ca a a limit will not be violated during a CRDA while following the BPWSF1OFDof operation. The generic BPWS analysis 02 (Ref. also evaluates the effect of fully inserted, inoperable control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, Inoperable control rods. Rod pattern control satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS. This LCO only applies to OPERABLE control rods. For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, 'Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS. APPLICABILITY In MODES 1 and 2, when THERMAL POWER is <11 Or/. RTP, the CRDA 0 is a Design Basis Accident and, therefore, compliance with the assumptions of the safety analysis is required. When THERMAL ( POWER is >10r/o RTP, there is no credible control rod configuration that (0 results in a control rod worth that could exceed the 280 cal/am fuel INSERT 1 des limit during a CRDA (Ref. 2).1 In MODERI 5, since the 0 reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn. ACTIONS A.1 and A.2 With one or more OPERABLE control rods not in compliance with the prescribed control rod sequence, actions may be taken to either correct the control rod pattern or declare the associated control rods inoperable within 8 hours. Noncompliance with the prescribed sequence may be the result of "double notching," drifting from a control rod drive cooling water transient, leaking scram valves, or a power reduction to <l10 '/oRTP 0 before establishing the correct control rod pattern. The number of OPERABLE control rods not in compliance with the prescribed sequence is limited to eight, to prevent the operator from attempting to correct a control rod pattern that significantl deviates from the rescribed sequence. trol rod pattern is not in compliance with the prescribed se ence, all control rod movement should/be stopped except (3) for moves ne ed to correct the rod pattern, or scram/if warranted. BWR/4 STS B 3.1.6-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 156 of 231

Attachment 1, Volume 6, Rev. 0, Page 157 of 231 B 3.1.6 0 INSERT I In MODES 3 and 4, the reactor is shut down and the control rods are not able to be withdrawn since the reactor mode switch is in the shutdown position and a control rod block is applied, therefore a CRDA is not postulated to occur. Insert Page B 3.1.6-2 Attachment 1, Volume 6, Rev. 0, Page 157 of 231

Attachment 1, Volume 6, Rev. 0, Page 158 of 231 Rod Pattern Control B 3.1.6 BASES ACTIONS (continued) NR Required Action A.1 is modified by a Note which allows the RWM to be shelpsbypassed to allow the affected control rods to be returned to their correct osition. LCO 3.3.2.1 requires verification of/control rod movement by a ET eJ member of the technical sta Thi ensureg that the control rods ill be moved to the correct position. A control rod not in compliance with e prescribed sequence is not considered inoperable except as required by Required Action A.2. {OPERABILITY of control rods~ determined by comphlance/with LCO 3.1.3, "Control Rod OPERABILI ," LCO 3.1.4, "Control Rqsd Scram Times," and LCO 3.1.5, "Control EtdScram / Accumulators."/ The allowed Completion Time of 8 hours is reasonable, considering the restrictions on the number of allowed out of sequence control rods and the low probability of a CRDA occurring during the time the control rods are out of sequence. B.1 and B.2 If nine or more OPERABLE control rods are out of sequence, the control rod pattern significantly deviates from the prescribed sequence. Control rod withdrawal should be suspended immediately to prevent the potential for further deviation from the prescribed sequence. Control rod insertion to correct control rods withdrawn beyond their allowed position is allowed since, in general, insertion of control rods has less impact on control rod Q worth than withdrawals have.. Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position. LCO 3.3.2.1 requires verification f control rod movement by *qualified member of the technical staff. lISET31 hen nine or more OPERABLE control rods are not in compliance with the reactor mode switch must be placed in the shutdown position LBPWS, within 1 hour. With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO. The allowed Completion Time of 1 hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence. SURVEILLANCE SR 3.1.6.1 REQUIREMENTS The control rod pattern is verified to be in compliance with the BPWS at a 24 hour Frequency to ensure the assumptions of the CRDA analyses are met. The 24 hour Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence and is required to be OPERABLE when operating at s5 l Or%RTP. 0D BWR/4 STS B 3.1.6-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 158 of 231

Attachment 1, Volume 6, Rev. 0, Page 159 of 231 B 3.1.6 CD INSERT 2 second licensed operator (Operator or Senior Operator) or by a Q INSERT 3 (e.g., engineer) Insert Page B 3.1.6-3 Attachment 1, Volume 6, Rev. 0, Page 159 of 231

Attachment 1, Volume 6, Rev. 0, Page 160 of 231 Rod Pattern Control B 3.1.6 BASES REFERENCES 1. NEDE-24011-P-P~ , "General Electric Standard Application for

              .D--Ractor Fuel        pplement for              "Secin2.2 lINSERT 41 0

2. "Modifications to the Requireme 6rControl Rod Drop Accident 02)

         ,zz23.

Mitigating System," BW ners Group, NUREG-0979, Section 4.2.1.3.2, April 1983. July 1986. NUREG-0800, Section 15.4.9, Revision 2, July 1981.

3. USAR. Secton 14.7.1.l 0

10CFR 100.11. [

6. NEDO-21778-A, "Transient Pressure Rises Affected Fracture Toughness Requirements for Boiling Water Reactors,"

December 1978.

               .4i
                                                               -I ] NSRT5l ASME, Boiler and Pressure Vessel Code.

NEDO-21231, "Banked Position Withdrawal Sequence," January 1977. BWR/4 STS B 3.1.6-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 160 of 231

Attachment 1,Volume 6, Rev. 0, Page 161 of 231 B 3.1.6 OJ INSERT4 Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC), "Amendment 17 to General Electric Licensing Topical Report NEDE-2401 1-P-A," BWROG-8644, August 15,1986. O JINSERT5

7. NEDO-1 0527, 'Rod Drop Accident Analysis for Large BWRs," (including Supplements 1 and 2), March 1972.

Insert Page B 3.1.6-4 Attachment 1, Volume 6, Rev. 0, Page 161 of 231

Attachment 1, Volume 6, Rev. 0, Page 162 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.6 BASES, ROD PATTERN CONTROL

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

4. This requirement has been deleted since ACTION A does not require that all rod movement (except for the moves needed to correct the rod pattern or a scram) be suspended. 1

5. Changes have been made to more clearly match the requirements of ITS 3.3.2.1 Required Action C.2.2.

6. A reference to the location where control rod OPERABILITY is determined has been deleted from the Bases for Required Actions A.1 and A.2 of ITS 3.1.6. This section is discussing under what conditions related to control rod sequence to declare a control rod inoperable - not determination of OPERABILITY per the other LCOs. As such, the reference is not applicable and could be interpreted as requiring an action that is not in the actual ITS 3.1.6 ACTION A.

7. Typographical error corrected.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 162 of 231

Attachment 1, Volume 6, Rev. 0, Page 163 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 6, Rev. 0, Page 163 of 231

Attachment 1, Volume 6, Rev. 0, Page 164 of 231 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.6, ROD PATTERN CONTROL There are no specific NSHC discussions for this Specification. Monticello Page I of 1 Attachment 1, Volume 6, Rev. 0, Page 164 of 231

Attachment 1, Volume 6, Rev. 0, Page 165 of 231 ATTACHMENT 7 ITS 3.1.7, Standby Liquid Control (SLC) System Attachment 1, Volume 6, Rev. 0, Page 165 of 231

Attachment 1, Volume 6, Rev. 0, Page 166 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 6, Rev. 0, Page 166 of 231

C itC C ITS 3.1.7 ITS ITS Ap CD

                                                                                      /bi                                                                                     C) s Applies to the o      aing status of the standby liquid contr             Applies to the periodic test!     requirements for the standby system.                                                                   liquid control system.

r-s 0 Di 0 To assure t e avallability of an Independent reactM control To verify the operablif of the standby liquid control system. mecanis CD SpecificatIon: 0 A.2 - A. The operability of the standby liquid control system shall A. System Operation be verged by performance of the following tests: aD CD ED Ar

        *C 3.1.7at

[MODES 1and 2 1/ The standby Id control system shall be operable all times wen e is inthe reactor and th o not shut down specifIed In 3.4.A.2. etr xcept as .7.7 l1- edit least egnce oer Aer - l Pump minimum flow rate of 24 gpm shall be verified against a system head of 1275 psig when tested In I Q CD

                                                                                                                                                                              ;U accordance with the Inprvlo Tesing Fora               I 0  ACTION B               2. From and after the date that a redundant                                    Compaison of the me sured pump flow rate o                             component Is made or found to be Inoperable.                               Ianalr equation 2 of paragraph 3.4 .1 shall be reactor operailn Is permissible only during the 0                              following 7 day provided that the redundant

-9 component Is operable. CD CD [Add proposed Required Action B.1 second Completion Time to CD

                                                                                                                                                                               -4

-4 0

                                                                                                                                                                              -9, 3.414.4                                                                                                      93             08/01/01 Amendment No. 66         ,-f4 Si 122                         M Page 1 of 6

C enC C ITS 3.1.7 3.0 LIMING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS

b. xhonoftopinrasmln 9) hescn de sebyhtqeploston v9 C, 0

ED 0 I 0 t. 0 Z., ax 0 0 CD 0) -o CD CD la 0 0) -4' -9' 3.414.4 94 10/1199 Amendment No. 56, :Z. 106 Page 2 of 6

C C C ITS 3.1.7 ITS ITS LCO 3.1.7 0 I A) 0 3-0 SR 3.1.7.1,  :.I CD SR 3.1.7.5. 0 SR 3.1.7.10 0 0 Equation 1 iOriqinal Den Bas) v bo1x+ ( T.05 Xs 0.99g V (* 4 1(101.E

                                                            - ) (-- ) (12

-1

                         + 128 gal Equation 2 1ATWS DenBasis C 2- 8.28 j()       (g                                               Boron concentration shall be determined. In SR 3.1.7.1.                                                                               additlon, the boron concentration shag be /

0 SR 3.1.7.5. Table 3.1.7-1 where: determinter or boron are adde V - indicated Boron solution tank volume (gal) the solution temperatured ie the lint -U E - measured Boron solution onrichment (atom%) specified by Figur 3.4.2717 C - measured Boron solution concentration (wt%) Ca 0 - measured pump flow rate (gpm) at 1275 psig o (D CD fEqualton I is satisfied but Equalion 2 cannot be -0 ACnON A mn continued plant operation is permissible, -, provided that: CD to

a. Compliance with Equation 2 is demonstrated SR 3.1.7.2 b. The solution temperature shalf be checked.

co - wnthin 7 daysE[ 4

b. Td Commision be notiied a aspoda SR 3.1.7.3 c. The room temperature shall be checked in the 0 CD reart provided oining the ecti taken and vlanity ot tho standby liquid control system compliancennd edulo witsc the for dlowstrating A1WS Desln Basis. W~

SR 3.1.7.2 2. The temperature shall not be less than the solution dA.1 second Completon lime M-, t4 temperature presented in Figure 3.4.2. SR 3.1.7.3 3. Tho heat tracing on the pump suction tines shall be operable whenever the room temperature Is less than the solution temperature presented in Figure 3.4.2. 3.414.4 95 2/15/91 Amendment No. G0, 7. 77 Page 3 of 6

C eC CI ITS 3.1.7 ITS 3.0 LMMNG CONDIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS VP C. It Speedfication 3.4A through B are not met, an orderly shutdown shall be Initiated mnd the reactor shall be In 0) ACTION D 0) 0 Hot Shutdown within 12 hours. 9 0C, D

0) aw 0 0 0

CD at -U CD ID 0

                                                                                                                                         .p 0

0)

                                                                                                                                          ;U p                                                                                                                                         to CD CD to                                                                                                                                         -4 0

0

                                                                                                                                          -9' 3.414.4                                                                                     96        2/15/91 Amendment No. 06, 77 Page 4 of 6

Attachment 1, Volume 6, Rev. 0, Page 171 of 231 ITS 3.1.7 ITS Figure 3.17-1 o i .. .. .. ... ,.,._... .. Allowed .... .. on

                                                                                                       .   .. rB-In

100 10 -: 1 1 0 1-0 10 1 19'-00 20. I D 1 :DXo . *e+ +*,**,**,*................

1 . . . . . . . .. .*.' Eo . .L- '..1 . .' . . . '. '''''. A mnm°en t .57 N . .* . ~...

                                                                                       .en               .           Aed12-             No      7 CDf                 A        s        /                          ................                .......             .. I ..   ... @.....

1O0W 11l 0 1200 1300 1400 l S00 l1600 l1700 1800 1900 2000 Indicated Tank Volume (Gallons) @Z Amendment No. 57 Figure 3.4-1 Sodium Pentaborate Solution Volume Concentration Requirements

                                               -97                                                                                          9/23188 Amendment No. 57 Page 5 of 6 Attachment 1, Volume 6, Rev. 0, Page 171 of 231

Attachment 1, Volume 6, Rev. 0, Page 172 of 231 0e ITS 3.1.7 ITS Figure 3.1.7-2

                                                ~~~.........                                                                ................                 *
                                       ~~~~.........................................................................

90 - .. .. .. .. .. .. .. .. . .. . . . . .......... U-lL ... . . ... . . . ..

                                                                                  ...                      . . .                .. . . ... . . . .. . . ;.i 0                             ..............                .        .           .   .    . .            .        .     .       .S V

c-

                                                  ~~~~~~~.........                                                          .........::::://...      e 80 -   ......... .... ,...... ........ __...,

0I-1 03 V E w 0 70 -

                  .0 U,

U, 60 - eU e 50- -  ;.. -; 40 - 5 10 t2 1 .14 16 18 Weight Percent Sodium Pentaborate In Solution wlo Na2 B 10 O1 6-H 2 0 Figure 3.4-2 Sodium Pentaborate Solution Temperature Requirements

                                                                                                                                              - 98                   12/11/87 Amendment No. 56 Page 6 of 6 Attachment 1, Volume 6, Rev. 0, Page 172 of 231

Attachment 1, Volume 6, Rev. 0, Page 173 of 231 DISCUSSION OF CHANGES ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) 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 3.4.A.1 requires the Standby Liquid Control (SLC) System to be OPERABLE at all times when fuel is in the reactor and the reactor is not shut down by control rods. ITS LCO 3.1.7 requires the SLC System to be OPERABLE in MODES 1 and 2. This changes the CTS by explicitly stating the applicable MODES in which the SLC System must be OPERABLE. The purpose of the CTS 3.4.A.1 is to ensure the SLC System is available to shutdown the reactor core whenever it is not shut down (i.e., multiple control rods are withdrawn). ITS 3.1.7 only requires the SLC System to be OPERABLE in MODES 1 and 2. This change is acceptable because MODES 1 and 2 are the only MODES in which the reactor is not shut down by control rods. In MODES 3 and 4, control rods cannot be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied. In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies, and LCO 3.1.1, "SHUTDOWN MARGIN (SDM)," ensures that the reactor will remain in shutdown by use of control rods. This clarification is also consistent with CTS 3.4.C, which requires a unit shutdown to Hot Shutdown (MODE 3) if CTS 3.4.A or B is not met. This change is designated as administrative since it does not result in any technical changes to the CTS. A.3 CTS 4.4.A.1 specifies those Surveillance Requirements that must be performed "At least once per quarter." CTS 4.4.A.1 only requires the performance of a SLC System flow test. However, CTS 4.4.A.1 also states that the SLC System flow test must be performed "in accordance with the Inservice Testing Program." ITS SR 3.1.7.7 requires the same test to be performed 'in accordance with the Inservice Testing Program." This changes the CTS by deleting the duplicative information associated with the testing Frequency. The purpose of the CTS 4.4.A.1 is to perform the SLC System flow test in accordance with the Inservice Testing Program. This changes the CTS by deleting the duplicative information associated with the testing Frequency. This change is acceptable since currently the Frequency of pump tests in the Inservice Testing Program is 92 days. This change simply deletes duplicative testing Frequencies from the CTS. Therefore, this change is considered a presentation preference change only and, as such, is considered an administrative change. A.4 CTS 4.4.8.1 requires a determination of boron enrichment, but does not specify the actual limit. The design limit for Monticello is 55.0 atom percent, as stated in CTS Figure 3.4-1. ITS SR 3.1.7.10 requires verification that the sodium Monticello Page 1 of 7 Attachment 1, Volume 6, Rev. 0, Page 173 of 231

Attachment 1, Volume 6, Rev. 0, Page 174 of 231 DISCUSSION OF CHANGES ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) SYSTEM pentaborate enrichment is 2 55.0 atom percent. This changes the CTS by specifying the actual limit in the sodium pentaborate enrichment verification Surveillance. The purpose of CTS 4.4.8.1 is to verify the sodium pentaborate enrichment is within the design limit so that Figure 3.4-1, which is based on a boron enrichment of 55.0 atom percent, can be used. Therefore, this change is acceptable since the limit specified in CTS Figure 3.4-1 is being added to the appropriate Surveillance. This change is considered a presentation preference change only and, as such, is considered an administrative change. A.5 CTS 4.4.A.2 requires the performance of a SLC System test at least once "during each operating cycle." ITS SR 3.1.7.8 requires performance of an SLC test every "24 months" on a STAGGERED TEST BASIS. This changes the CTS by changing the Frequency from "during each 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. CTS 4.4.A.2 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 CTS 4.4.A.2.a requires the performance of a SLC subsystem test to verify flow can be injected into the reactor vessel. During this test SLC pump capacity must be verified. ITS SR 3.1.7.8 requires the performance of the same test, however the requirement to verify pump capacity has nbt been included. This changes the CTS by deleting the specific requirement to verify SLC pump capacity during the SLC subsystem reactor vessel injection test. - The purpose of CTS 4.4.A.2.a is to ensure the flow path from the pump to the reactor vessel is not obstructed. This change deletes the specific requirement to verify pump capacity during the SLC subsystem reactor vessel injection test. This change is acceptable because SLC pump capacity is verified more frequently as required by CTS 4.4.A.1, the quarterly pump capacity test. This Surveillance is maintained in the ITS as SR 3.1.7.7 at a Frequency in accordance with the Inservice Testing Program (currently every 92 days). This test will ensure that the SLC pump capacity is adequate to perform its safety function. ITS SR 3.1.7.8 requires the verification of flow through one SLC subsystem from pump into the reactor pressure vessel, and is sufficient to ensure the piping from the pump to the reactor vessel is not obstructed. The requirement to verify SLC pump capacity is duplicative and is therefore deleted from CTS 4.4.A.2.a. As such, this change is considered a presentation preference change only and is therefore designated as an administrative change. Monticello Page 2 of 7 Attachment 1, Volume 6, Rev. 0, Page 174 of 231

Attachment 1, Volume 6, Rev. 0, Page 175 of 231 DISCUSSION OF CHANGES ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) SYSTEM MORE RESTRICTIVE CHANGES M.1 CTS 3.4.A.2 allows 7 days to restore a redundant component (i.e., one SLC subsystem) to OPERABLE status. CTS 3.4.B.1 allows 7 days to comply with the boron solution concentration requirements of Equation 2 (ATWS Design Basis). The CTS does not limit the time both an SLC subsystem is inoperable and the boron concentration-is not within limits. ITS 3.1.7 ACTION A covers the condition when the concentration of boron in solution is not within the limits of Equation 2 of ITS Table 3.1.7-1 and ITS 3.1.7 ACTION B covers the condition when one SLC subsystem is inoperable for reasons other than Condition A (i.e., boron concentration not within limits). The second Completion Time for ITS 3.1.7 Required Actions A.1 and B.1 requires the boron concentration to be restored to within limits and the SLC subsystem to be restored to OPERABLE status, respectively, within "14 days from discovery of failure to meet the LCO." This changes the CTS by limiting the maximum time allowed for any combination of boron solution concentration out of limit or inoperable SLC subsystems during any single contiguous occurrence of failing to meet the LCO. The purpose of CTS 3.4.A.2 is to cover inoperabilities associated with redundant SLC subsystem components while the purpose of CTS 3.4.B.1 is to cover inoperabilities associated with the ATWS Design Basis boron solution concentration requirements. This change limits the maximum time allowed (14 days) for any combination of ATWS Design Basis boron solution concentration out of limit or inoperable SLC subsystems during any single contiguous occurrence of failing to meet the LCO. The change isacceptable since the 14 day Completion Time is an appropriate limitation for failure to meet the SLC System LCO. This change is designated as more restrictive because it limits the time the SLC System ACTIONS can be entered in any one single contiguous occurrence of failing to meet the LCO. M.2 ITS SR 3.1.7.4 requires the verification of the continuity of the explosive charge. ITS SR 3.1.7.6 requires verification that each SLC subsystem manual 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. ITS SR 3.1.7.9 requires verification that all heat traced piping between storage tank and pump suction is unblocked. The CTS does not include these Surveillance Requirements. This changes the CTS by adding these new Surveillances. This change is acceptable because the new Surveillance Requirements will help ensure the SLC System is OPERABLE. These verifications give additional confidence that the explosive valves are OPERABLE, the SLC manual valves are aligned correctly (or can be aligned), and that the heat traced piping between the storage tank and pump suction is unblocked. This change is designated as more restrictive because it adds Surveillance Requirements that are not required in the CTS. M.3 With boron concentration limits of CTS 3.4.B.1 not met, CTS 3.4.B.1.a requires compliance with Equation 2 to be demonstrated within 7 days. If compliance with Equation 2 is not demonstrated within 7 days, CTS 3.4.B.1.b requires the Commission to be notified and a special report provided outlining the actions taken and the plans and schedule for demonstrating compliance with the ATWS Monticello Page 3 of 7 Attachment 1, Volume 6, Rev. 0, Page 175 of 231

Attachment 1, Volume 6, Rev. 0, Page 176 of 231 DISCUSSION OF CHANGES ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) SYSTEM Design Basis. ITS 3.1.7 ACTION A maintains 7 days to establish the appropriate conditions to satisfy the ATWS Design Basis, but if Equation 2 is not satisfied within the 7 day period, ITS 3.1.7 ACTION D requires a shutdown to MODE 3 within 12 hours. This changes the CTS by deleting the option to notify the Commission and continuing to operate with Equation 2 not met. The purpose of CTS 3.4.B.1.b isto provide an outline of the actions and schedule to comply with the ATWS Design Basis. The 7 day Completion Time in CTS 3.4.B.1.a (ITS 3.1.7 Required Action A.1) is considered an acceptable amount of time to restore all normal problems associated with the boron solution. If the 7 day Completion Time is not satisfied, ITS 3.1.7 ACTION D will require the unit to be in MODE 3 in 12 hours. This is the current shutdown action in CTS 3.4.C. The change has been designated as more restrictive because it will require the unit to be in MODE 3 in 12 hours instead of allowing operations to continue indefinitely with the ATWS Design Basis not met. M.4 CTS 4.4.B.2 requires the boron concentration to be determined anytime water or boron is added to the solution or if the solution temperature drops below the limits specified in Figure 3.4-2. However, no finite time to complete performance of this Surveillance is provided. ITS SR 3.1.7.5 requires the same Surveillance; however, a requirement has been added to require the Surveillance to be completed once "within 24 hours" after water or boron is added to the solution and once "within 24 hours after solution temperature is restored" within the limits of Figure 3.1.7-2. This changes the CTS by placing a time limit of 24 hours to perform the Surveillance. The purpose of CTS 4.4.B.2 is to ensure the boron concentration is within limits. This change places a time limitation for performing the boron concentration verification after adding water, boron, or after the temperature falls below the temperature limit and is subsequently restored. This change is acceptable because the time limit of 24 hours issufficient to notify the appropriate personnel to take a sample, send the sample to the laboratory, analyze the sample, and evaluate the results of the chemical analysis. This ensures that any potential change to the boron concentration is quickly evaluated. Also, the second Frequency ensures that the boron concentration is verified "after" the temperature is restored to within limits. The change has been designated as more restrictive because it explicitly limits the time to perform the Surveillance. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.1 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.4.A.2.a states "Manually initiate" one of the two standby liquid control systems "and pump demineralized water" into the reactor vessel. It further states that "This test checks explosion of the charge associated with the tested system, proper operation of the valves" and that both SLC Monticello Page 4 of 7 Attachment 1, Volume 6, Rev. 0, Page 176 of 231

Attachment 1, Volume 6, Rev. 0, Page 177 of 231 DISCUSSION OF CHANGES ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) SYSTEM subsystems shall be tested "and inspected, including each explosion valve." CTS 4.4.A.2.b states "Explode one of the primer assemblies manufactured in the same batch to verify proper function. Then install, as a replacement, the second primer assembly in the explosion valve of the system tested for operation." ITS SR 3.1.7.8 requires verification of flow through one SLC subsystem from pump into reactor pressure vessel. This changes the CTS by relocating the above procedural details concerning performance of the flow path test 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 verify flow through one SLC subsystem from pump into reactor pressure vessel. Also, this change is acceptable because this type 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 a procedural detail for meeting Technical Specification requirements is being removed from the Technical Specifications. LA.2 (Type I - Removing Details of System Design and System Description, Including Design Limits) CTS 3.4.B.1 states Equation I is consistent with the "Original Design Basis" and Equation 2 is consistent with the "ATWS Design Basis." CTS Figure 3.4-1 specifies that the curves are based on "B-1 0 Enrichments Greater than 55.0%." ITS Figure 3.1.7-1 includes the same requirements as CTS Figure 3.4-1, except the detail concerning the B-10 enrichment. ITS Table 3.1.7-1 includes Equation 1 and Equation 2, however the statements concerning the "Original Design Basis" and "ATWS Design Basis" are not included. This changes the CTS by relocating these details to the ITS Bases. The removal of these details, which are related to the system design capabilities, 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 same Figure and equations that are in the CTS. The details on the design basis details of the equations do not need to appear in the specification in order for the requirement to apply. Also, this change is acceptable because the removed information will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because information relating to system design is being removed from the Technical Specifications. LA.3 (Type 3 - Removing Procedural Details for Meeting TS Requirements or Reporting Requirements) CTS 4.4.B.1 requires the boron enrichment to be determined by "laboratory analysis." ITS SR 3.1.7.10 does not specify the method that shall be used to determine the B-10 enrichment. This changes the CTS by relocating the procedure detail "laboratory analysis" to the ITS Bases. Monticello Page 5 of 7 Attachment 1, Volume 6, Rev. 0, Page 177 of 231

Attachment 1, Volume 6, Rev. 0, Page 178 of 231 DISCUSSION OF CHANGES ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) SYSTEM 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 verify boron enrichment is within limits. Also, this change is acceptable because these types of procedural details will be adequately controlled in the ITS Bases. Changes to the Bases are controlled by the Technical Specification Bases Control Program in Chapter 5. This program provides for the evaluation of changes to ensure the Bases are properly controlled. This change is designated as a less restrictive removal of detail change because procedural details for meeting Technical Specification requirements are being removed from the Technical Specifications. LESS RESTRICTIVE CHANGES L.1 (Category 3 - Relaxation of Completion Time) CTS 3.4.A does not provide actions for when two SLC subsystems are inoperable, thus CTS 3.4.C must be entered and the unit must be placed in hot shutdown. ITS 3.1.7 ACTION C covers the condition when two SLC subsystems are inoperable for reasons other than Condition A (i.e., boron concentration not within limits), and requires the restoration of one SLC subsystem to OPERABLE status within 8 hours. This changes the CTS by providing 8 hours to restore one SLC subsystem to OPERABLE status when it is discovered that both SLC subsystems are inoperable prior to requiring a unit shutdown. The purpose of ITS 3.1.7 ACTION C isto allow 8 hours to restore one SLC subsystem to OPERABLE status when both are inoperable. This change is acceptable because the Completion Time is consistent with safe operation under the specified condition, considering the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a design basis accident occurring during the allowed Completion Time. The ITS 3.1.7 Completion Time of 8 hours is considered acceptable given the low probability of a design basis accident or transient occurring concurrent with the failure of the control rods to shut down the reactor. This change is designated as less restrictive because additional time is allowed to restore a SLC subsystem to OPERABLE status than was allowed in the CTS. L.2 (Category 5- Deletion of Surveillance Requirement) CTS 4.4.A.1 requires the performance of a SLC pump test. It also states "Comparison of the measured pump flow rate against equation 2 of paragraph 3.4.B.1 shall be made to demonstrate operability of the system in accordance with the ATWS Design Basis." ITS SR 3.1.7.7 requires the SLC pump test, but does not include the requirement about the demonstration of the OPERABILITY of the system in accordance with the ATWS Design Basis. This changes the CTS by deleting the requirement to perform this comparison. The purpose of CTS 4.4.A.1, in part, is to ensure the ATWS Design Basis is met after a flow test is performed. This change is acceptable because the deleted Surveillance Requirement is not necessary to verify that the equipment used to meet the LCO can perform its required functions. Thus, appropriate equipment Monticello Page 6 of 7 Attachment 1, Volume 6, Rev. 0, Page 178 of 231

Attachment 1, Volume 6, Rev. 0, Page 179 of 231 DISCUSSION OF CHANGES ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) SYSTEM continues to be tested in a manner and at a Frequency necessary to give confidence that the equipment can perform its assumed safety function. This change deletes the explicit CTS requirement to demonstrate operability of the system in accordance with the ATWS Design Basis after the performance of the SLC pump test. This change is acceptable since there are other Surveillances that are performed more frequently which confirm the ATWS Design Basis. ITS SR 3.1.7.5 requires the verification of the concentration of boron in solution is within the limits of Figure 3.1.7-1 or within the'limits of Equation 2 (ATWS Design Basis) of Table 3.1.7-1 every 31 days, once within 24 hours after water or boron is added to solution, and once within 24 hours after solution temperature is restored within limits of Figure 3.1.7-2. The ATWS Design Basis will always be met if the SLC system flow rate is > 24 gpm, the sodium pentaborate concentration is within the limits of ITS Figure 3.1.7-1 (CTS Figure 3.4-1), and B-10 enrichment is > 55.0%. If the sodium pentaborate concentration is not within the limits of the Figure or if Boron enrichment is < 55.0%, then it is necessary to determine whether the concentration limits of Equation 2 of Table 3.1.7-1 (ATWS Design Basis) is met, as required by ITS SR 3.1.7.5. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS. L.3 (Category 6 - Relaxation Of Surveillance Requirement Acceptance Criteria) CTS 4.4.B.1 requires the boron enrichment to be determined at least once per cycle. The laboratory analysis to determine enrichment shall be obtained within 30 days of sampling or chemical addition. ITS SR 3.1.7.10 requires the determination of B-1 0 enrichment is > 55.0 atom percent B-1 0 prior to addition to the SLC tank. This changes the CTS by deleting the requirement to verify the storage tank enrichment every cycle and replaces it with a requirement to verify that the solution added to the SLC storage tank is at the proper B-10 enrichment. The purpose of CTS 4.4.8.1 is to ensure the B-10 enrichment in the boron solution tank is within the appropriate limits. 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. Enriched sodium pentaborate solution is made by mixing granular, enriched sodium pentaborate with water. Isotopic tests on the granular sodium pentaborate in the storage container to verify the actual B-1 0 enrichment must be performed prior to addition to the SLC tank in order to ensure that the proper B-10 atom percentage is being used. This change is acceptable since no deterioration of the B-10 enrichment level should occur to the B-10 while it is stored in its storage container. In addition, the deletion of the requirement to obtain the test results (i.e., the laboratory analysis) within 30 days of sampling or chemical addition is acceptable since the granular B-10 cannot be added to the SLC storage tank until the results of the analysis are known (i.e., the Frequency of ITS SR 3.1.7.10 requires performance 'prior to addition"). This change is designed 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 6, Rev. 0, Page 179 of 231

Attachment 1, Volume 6, Rev. 0, Page 180 of 231 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 180 of 231

Attachment 1, Volume 6, Rev. 0, Page 181 of 231 SLC System 3.1.7 CTS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Standby Liquid Control (SLC) System 3.4Aj1, LCO 3.1.7 Two SLC subsystems shall be OPERABLE.

3.4.8 APPLICABILITY

MODES 1 and 2. ACTIONS A_ CONDITION REQUIRED ACTION __ _ _ _ COMPLETION TIME - sodium pentaborate 3.4.B.1, A. I Concentration of& i'- A.1 estore concentration of 3.4.B.1.a in solution not within o in solution to within limitsb t limits. avallable volume of sodium pentaborate solution is 3.1.7-1 Equation 2. wittin limits of Table 3.1.7-1 Equation 1 l days from discovery of failure to meet the LCOM

                                                                                                       +

3.4.A.2 B. One SLC subsystem inoperable Ifor reasons B.1 Restore SLC subsystem to OPERABLE status. 7 days 0D other than Condition Al AND = days from discovery of failure to meet the LCOC 4 + DOC C. Two SLC subsystems C.1 Restore one SLC 8 hours LA inoperable for reasons other than Condition Al subsystem to OPERABLE status. 0

• 4 3.4.C D. Required Action and D.1 Be in MODE 3. 12 hours associated Completion Time not met.

                                               -1 BWR/4 STS                                                  3.1.7-1                                  Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 181 of 231

Attachment 1, Volume 6, Rev. 0, Page 182 of 231 SLC System 3.1.7 CTS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 3.4.1.1, SR 3.1.7.1 Verify available volume of sodium pentaborate 24 hours 4.4.B.3.a solution is [within the limits of Figure 3.1.7-1 or 12 45 allon. on 3 0D 3.4.B.2. 4.4.B.3.b SR 3.1.7.2 IVerify temperature of sodium pentaborate solution is within the limits of Figure 3.1.7-2. 24 hours 0 ____________________________________I I solution temperature I 3.4.81.3. 4.4.B.3.c SR 3.1.7.3 \ or verify SLC pump suctionines Verifytemperature o pump limits of MFigure 3.1.7-2. npipinl iswithin i vicinity of 24 hours j 0D wheat tracing Is OPERABLE __theSLCmpump DOCM.2 SR 3.1.7.4 Verify continuity of explosive charge. 31 days 3.4.B.1, 4.4.B.2 SR 3.1.7.5 Verify the concentration of lil Twithin the limits of Figure 3.1.7-1 . solution is 31 days 0D or within the limits of Equation 2 of Table 3.1.7-1 AND sodium pentaborate Once within 24 hours after water O'diibaIs added to solution AND Once within 24 hours after solution temperature is restored within the limits of 1- [Figure 3.1.7-22 0D DOC M.2 SR 3.1.7.6 Verify each SLC subsystem manual loperatedarautomaticvalveg in the flow path that er 31 days 0 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 3.1.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 182 of 231

Attachment 1, Volume 6, Rev. 0, Page 183 of 231 SLC System 3.1.7 CTS SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY 4.4AA1 SR 3.1.7.7 Verify each pump develops a flow rate at a discharge pressure RES Psig. [ ] gpm Tin accordance with the Inservice 0D Testing Program

                                                      -t                     or 9,ays]I
                                                                          -t 4.4.A2     SR 3.1.7.8   Verify flow through one SLC subsystem from pump into reactor pressure vessel.

months on a STAGGERED 0 TEST BASIS DOC SR 3.1.7.9 M.2 4.4.B.1 SR 3.1.7.10 Verify sodium pentaborate enrichment is 2 [. atom percent B-10. Prior to addition to SLC tankI 0D BWR/4 STS 3.1.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 183 of 231

Attachment 1, Volume 6, Rev. 0, Page 184 of 231 6 SLC System INSERT 1 3.1.7 Q.) I (1420 gal, 13%) I I 13 This figure for ill istration only. Do not use for o eration. 12 I I[ I WI_ I _ 11 - KS I I I I 0 2, O u)

        .a z._

2 10 - z co Z

     °.0

w. =

CES

   = I-ca)

E 9 k I ACCE TABLE O 0Cn C0 a) 8 0 W 0) 7 I I W I r V I I (28E gal, 6.2%) I8-- If (38( 0 gal, 6.2%) NOT ACCEPTAB E 1I00 I I I I I I I I 000o 1400I 1800 2200 2600 000 3400 3800 GROSS V LUME OF SOLUTION IN T NK (gallons) Fig re 3.1.7-1 (page 1 of 1) Sodium entaborate Solution Vol me Versus oncentration Requirem nts BWR/4 STS 3.1.7-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 184 of 231

Attachment 1, Volume 6, Rev. 0, Page 185 of 231 3.1.7 K>_ CTS INSERT I Figure 3.4-1

                                  --ABlowedOperation[

14- _ a, (1167,13.0) a)

          -0C 13-                     7 a,

U C: a) 0 a, 0 12 -

          .0 wc a)

E 0 C') f I /0If /0 If I 8 A (1404,10.7) I/xr I ft I S /0 X I 10 1000 1100 1200 ht1300 1400 1500 IR/0' 1700 I,1800Ilgg///7X 1600

                                                                                                /yxy.1900 2000 Indicated Tank Volume (Gallons)

Figure 3.1.7-1 (page 1 of 1) Sodium Pentaborate Solution Volume Versus Concentration Requirements Insert Page 3.1.7-4 Attachment 1, Volume 6, Rev. 0, Page 185 of 231

Attachment 1, Volume 6, Rev. 0, Page 186 of 231 0 BWR/4 STS 3.1.7-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 186 of 231

Affachment 1, Volume 6, Rev. 0, Page 187 of 231 CTS 3.1.7 INSERT 2 Figure lC 3.4-2 10- =.-___ _= 9 8 1 [Allowed Operatoiih 0L L.. CL. E 7'

        .C S..

U CO, El a) 61o- \\\\\

                         ===_\\ \\ \\\ N\\t\

54 0 -_ 41 8 10 12 14 16 18 20 Sodium Pentaborate in Solution (Weight Percent, wt%) Figure 3.1.7-2 (page 1 of 1) Sodium Pentaborate Solution Temperature Versus Concentration Requirements Insert Page 3.1.7-5a Attachment 1, Volume 6, Rev. 0, Page 187 of 231

Attachment 1, Volume 6, Rev. 0, Page 188 of 231 Q INSERT 3 3.1.7 CTS Table 3.1.7-1 (page I of 1) Equations for Required Sodium Pentaborate Tank Volume and Concentration 3.4.B.1 Equation I V Ž2( o o7>8 V 71.18 9 8) (0.005lxC +0.998)( ECi )&(i" )+128 gala

                                               )1 1+ 14821 )1-)1°+2 1101-Et E )

Where: C = measured boron solution concentration (wt%) E = measured boron solution enrichment (atom%) V = indicated boron solution tank volume (gal) Equation 2 C 2 8.2886)(19.8) Where: C = measured boron solution concentration (wt%) E = measured boron solution enrichment (atom%) Q = measured pump flow rate (gpm) at 1275 psig Insert Page 3.1.7-5b Attachment 1, Volume 6, Rev. 0, Page 188 of 231

Attachment 1, Volume 6, Rev. 0, Page 189 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) SYSTEM

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

2. The ITS 3.1.7 Required Action A.1 first Completion Time has been extended from 72 hours to 7 days, consistent with the current licensing basis (CTS 3.4.B.1 .a). In addition, the ITS 3.1.7 Required Action A.1 and B.1 second Completion Time has been changed from 10 days to 14 days. The ISTS 3.1.7 10 day Completion Time was calculated based on the sum of the first Completion Times in ISTS 3.1.7 Required Actions A.1 and B.1, as described in the Bases of ISTS 3.1.7 ACTIONS A.1 and B.1. Therefore, the second Completion Time in ITS 3.1.7 Required Actions A.1 and B.1 has been set at 14 days, which is the sum of the first Completion Times in ITS 3.1.7 Required Actions A.1 and B.1.

3. The proper Monticello nomenclature has been used (CTS Figures 3.4.-1 and 3.4-2).

This is also consistent with the nomenclature used in SR 3.1.7.1 and SR 3.1.7.2.

4. The changes in ITS SR 3.1.7.6 are made since there are no automatic valves in the SLC System and there are no power operated valves other than the explosive valves in the SLC System, and these are not checked as part of this Surveillance (as described in the ISTS Bases for this SR). Explosive valves are tested by other Surveillances in this Specification.

5. ISTS SR 3.1.7.9 requires a verification that all heat traced piping is unblocked once within 24 hours after solution temperature is restored within the limits of Figure 3.1.7-2. A change in solution temperature in the tank does not necessarily have an impact on the piping temperature, as long as the piping heat trace circuit is functioning properly. The intent of the second Frequency is to ensure that, if the heat tracing is inoperable such that piping temperature falls below the specified minimum temperature, after the heat tracing is restored to OPERABLE status'and the piping temperature is greater than or equal to the specified minimum temperature the piping is still unblocked. This is supported by the ISTS SR 3.1.7.9 Bases description for this second Frequency, which describes the requirement as required to be performed after piping temperature is restored. However, since the Monticello design does not include temperature indication on the suction piping, the plant-specific requirement for determining piping temperature, by measuring the room temperature in the vicinity of the SLC pumps, will be used in the Frequency. Thus the second Frequency has been changed to once within 24 hours after "room temperature in the vicinity of the SLC pumps" is restored within the "solution temperature" limits of Figure 3.1.7-2. This plant-specific requirement concerning the room temperature is shown in CTS 3.4.B.3.c and 4.4.B.3.c. Furthermore, a Note has been added stating that the second Frequency is only required if the SLC pump suction lines heat tracing is inoperable, consistent with the above discussed intent.

6. The following changes have been made to reflect the current licensing basis requirements.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 189 of 231

Attachment 1, Volume 6, Rev. 0, Page 190 of 231 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 190 of 231

Attachment 1, Volume 6, Rev. 0, Page 191 of 231 SLC System B 3.1.7 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive, xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref. 1) on anticipated transient without scram,. 0 The SLC System consists of a boron solution storage tank, two positive displacement pumps, two explosive valves that are provided in parallel for redundancy, and associated piping and valves used to transfer borated water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged near the bottom of the core shroud, where it then mixes with the cooling water rising through the core. A smaller tank containing demineralized water is provided for testing purposes. APPLICABLE The SLC System is manually initiated from the main control room, as ,detemines SAFETY directed by the emergency operating procedures, if the operator Ie es ANALYSES the reactor cannot be shut down, or kept shut down, with the control rods. The SLC System is used in the event that enough control rods cannot be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to add negative reactivity to compensate for all of the various reactivity effects that could occur during plant operations. To meet this objective, it is t necessary to inject a quantity of boron roduces a concentration of 660 ppm of natural bororn in the reactor coolant at 680 F. To allow for potential leakage and imperfect mixirig in the reactor system, an amount of boron equal to 25% of the amount cited above is added (Ref. 2). The volume versus concentration limits in Figure 3.1.7-1 and the temperature versus concentration limits in Figure 3.1.7-2 are calculated such that the required concentration is achieved accounting for dilution in the RPV with normal water level and including the water volume in the residual heat removal shutdown cooling piping and in the recirculation loop piping. s quantity of borated solution is the amount that is above the pump suction su el in tirs srution storage tang. No credit is taken for the portion of the tank volume that cannot be injected. nozzle and accountsforwide range instrumenteacuracy andvwth B-10 enrichmentof> 55.0atom percent The SLC System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). BWR/4 STS B 3.1.7-1 Rev. 3.0, 03/31/04 Attachment 1,Volume 6, Rev. 0, Page 191 of 231

Attachment 1, Volume 6, Rev. 0, Page 192 of 231 SLC System B 3.1.7 BASES LCO The OPERABILITY of the SLC System provides backup capability for reactivity control independent of normal reactivity control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Two SLC subsystems are required to be OPERABLE; each contains an OPERABLE pump, an explosive valve, and associated piping, valves, and instruments and controls to ensure an' OPERABLE flow path. APPLICABILITY In MODES 1 and 2, shutdown capability is required. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block Is applied. This provides adequate controls to ensure that the reactor remains subcritical. In MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Demonstration of adequate SDM (LCO 3.1.1, "SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE when only a single control rod can be withdrawn. I concentration of sodium pentaborate in I Ar.Ti flMq A I

                                                         /         [no-t    w-thiL      J o Figure 3.1.7-1 and Table 3.1.7-1 Equation 2l (ATWS design             If the         lutionlconc raicoisess                   n t1                          iilm              a  Areuire basis) Jor                 b99greater        than the congenuaietrequred for colasuow)                                      (

available volume of sodium o ice in basis), the concentration must be restored to wi T~atble 3.1.7-1 withn iaamaitseo~f ms inl7Kr It is not necessary under these conditions to enter .Equation I Condition C for both SLC subsystems inoperable since they are capable es g of performing their original design basis function. Because of the low probability of an event and the fact that the SLC System capability still exists for vessel injection under these conditions, the allowed Completion Time ofD Iis acceptable and provides adequate time to restore concentration to within limits. t3 (1 The second Completion Time for Required Action A.1 establishes a limit on the maximum time allowed for any combination of concentration out of limits or inoperable SLC subsystems during any single contiguous to occurrence of failing to meet the LCO. If Condition A is entered while, for instance, an SLC subsystem is inoperable and that subsystem is I statusJ subsequently returnedtb OPERABLE,1the LCO may already have been To

                            ,9 not met for up to 7 days. This situation could ead to a total duration oft days (7days in Condition B, followed by[ldays in Condition A),

failur~the LCO, to restore the SLC System. Then an SLC th initial

                                               ;could be found inoperable again, and concentration could be restored to within limits. This could continue indefinitely.                                              (:)

BWR14 STS B 3.1.7-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 192 of 231

Attachment 1, Volume 6, Rev. 0, Page 193 of 231 SLC System B 3.1.7 BASES ACTIONS (continued) This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock," resulting in establishing the 14 "time zero" at the time the LC9 was initially not met instead of at the time Condition A was entered. The t]day Completion Time is an acceptable (i) limitation on this potential to fail to meet the LCO indefinitely. B.1 If one SLC subsystem is inoperable for reasons other than Condition A, the inoperable subsystem must be restored to OPERABLE status within _7 ATWS design basis days. In this condition, the remaining OPERABLE subsystem is

                        .adequate to perform t[es ~dow function. However, the overall reliability is reduced because a single failure in the remaining OPERABLE subsystem could result in reduced SLC System shutdown capability. The 7 day Completion Time is based on the availability of an OPERABLE subsystem capable of performing the intended SLC System function and the low probability of a Design Basis Accident (DBA) or severe transient occurring concurrent with the failure of the Control Rod Drive (CRD)

System to shut down the plant. 0 The second Completion Time for Required Action B.1 establishes a limit on the maximum time allowed for any combination of concentration out of limits or inoperable SLC subsystems during any single contiguous occurrence of failing to meet the LCO. If Condition B is entered while, for instance, concentration is out of limits, and is subsequently returned t within limits, the LCO may already have been no met for up to ays. This situation could lead to a total duration of W-days Nda s in ( frm Condition A, followed by 7 days in Condition B) lsi ceinitial failure he LCO, to restore the SLC System. en concentration could be found out of limits again, and the SLC subsystem could be restored to OPERABLE. 0 This could continue indefinitely. This Completion Time allows for an exception to the normal "time zero" for beginning the allowed outage time "clock," resulting in establishing the "time zero" at the time the LCO was initially not met instead of at the tim e1 Condition B was entered. The WDfay Completion Time is an acceptable limitation on this potential to fail to meet the LCO indefinitely. BWR/4 STS B 3.1.7-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 193 of 231

Attachment 1, Volume 6, Rev. 0, Page 194 of 231 SLC System B 3.1.7 BASES ACTIONS (continued) C.1 If both SLC subsystems are inoperable for reasons other than Condition A, at least one subsystem must be restored to OPERABLE status within 8 hours. The allowed Completion Time of 8 hours is considered acceptable given the low probability of a DBA or transient occurring concurrent with the failure of the control rods to shut down the reactor. D.1 If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.7.1m*R .1 S 0772 (i) REQUIREMENTS 2 Sr SR 3.1.7 are 24 hour Surveillances verifying certain and characteristics of the SLC System (e.g., the volume and temperature of the borated solution in the storage tank), thereby ensuring SLC System OPERABILITY without disturbing normal plant operation. These Surveillances ensure that the proper borated solution volume and temperature( including the temum suction S iin are (i maintained. Maintaining a minimum specified borated solution temperature is important in ensuring that the boron remains in solution and does not precipitate out in the storage tank rithnpscon< 0gffffg.'The temperature versus concentration curve of Figure 3.1.7-2 ensures that a~F margin will be maintained above the saturation 0 temperature The 24 hour Frequency is based on operating experience anUas shown there are relatively slow variations in the measured IINSERT 1 parameters of volume and temperature. BWR/4 STS B 3.1.7-4 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 194 of 231

Attachment 1, Volume 6, Rev. 0, Page 195 of 231 B 3.1.7 Q\hi INSERT I The volume of sodium pentaborate solution requirements in Figure 3.1.7-1 and Table 3.1.7-1 Equation I will ensure both the original design basis and the ATWS design basis are met. Figure 3.1.7-1 can only be used if the B-10 enrichment in the storage tank is > 55.0 atom percent. If the volume requirement of Table 3.1.7-1 Equation 1 is utilized for verification of volume requirements the concentration requirements for the original design basis can also be considered to be met. However, to verify the ATWS design basis requirements are met, Table 3.1.7-1 Equation 2 must be used to verify the concentration of sodium pentaborate solution requirements are met. 0 INSERT 2 SR 3.1.7.3 SR 3.1.7.3 is a 24 hour Surveillance that requires the verification that the room temperature in the vicinity of the SLC pumps is within the solution temperature limits of Figure 3.1.7-2 or that the SLC pump suction lines heat tracing is OPERABLE. This Surveillance will help ensure that the proper borated solution temperature of the pump suction piping is maintained. Maintaining a minimum specified room temperature is important in ensuring that the boron remains in solution and does not precipitate out in the pump suction piping. The temperature versus concentration curve of Figure 3.1.7-2 ensures that a 50F margin will be maintained above the saturation temperature. An acceptable alternate requirement isto verify the pump suction lines heat tracing is OPERABLE. The heat tracing is sized to maintain the pump suction above 700F when the room temperature is 450F. OPERABILITY of the heat tracing is confirmed by verifying the light associated with each controller is on, or by depressing the toggle switch and ensuring the light is on. The 24 hour Frequency is based on operating experience and has shown there are relatively slow variations in the measured room temperature. Insert Page B 3.1.7-4 Attachment 1, Volume 6, Rev. 0, Page 195 of 231

Attachment 1, Volume 6, Rev. 0, Page 196 of 231 SLC System B 3.1.7 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.1.7.4 and SR 3.1.7.6 SR 3.1.7.4 verifies the continuity of the explosive charges in the injection valves to ensure that proper operation will occur if required. Other administrative controls, such as those that limit the shelf life of the explosive charges, must be followed. The 31 day Frequency is based on operating experience and has demonstrated the reliability of the explosive charge continuity. SR 3.1.7.6 verifies that each valve in the system is in its correct position, but does not apply to the squib (i.e., explosive) valves. Verifying the correct alignment for manualL power ope nd automativalves in 4 the SLC System flow path provides assurance that the proper flow paths will exist for system operation. A valve is also allowed to be in the nonaccident position provided it can be aligned to the accident position from the control room, or locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system. This Surveillance also does not apply to valves that are locked, sealed, or otherwise secured in position since they are verified to be in the correct position prior to locking, sealing, or securing. This verification of valve alignment 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 and is consistent with the procedural controls governing valve operation that ensures correct valve positions. SR 3.1.7.5 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper _e rexists in the storage tank. 4SR 3.1.7.5 must be

            \     erformed anyim              or water is added to the storage tank solution to l
                \ determine that th       onsolution concentration is within the specified limits. SR 3.1.7.5 must also be performed anytime the temperature is restored to within the limits of Figure 3.1.7-2, to ensure that no significant boron precipitation occurred. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation o fibo rn concentration between ,Aurveillances.                                              ()

BWR/4 STS B 3.1.7-5 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 196 of 231

Attachment 1, Volume 6, Rev. 0, Page 197 of 231 B 3.1.7 0 INSERT 3 The concentration of sodium pentaborate in solution required in Figure 3.1.7-1 will ensure the original design basis and the ATWS design basis are met. Figure 3.1.7-1 can only be used if the B-10 enrichment in the storage tank is > 55.0 atom percent and as long as the flow rate requirements of SR 3.1.7.7 are met. Equation 2 of Table 3.1.7-1 ensures both the original design basis and ATWS design basis are satisfied. If the volume requirement of Equation 1 of Table 3.1.7-1 is utilized for verification of volume requirements the concentration requirements for the original design basis can also be considered to be met. However, to verify the ATWS requirements are met, Equation 2 of Table 3.1.7-1 must be used to verify the concentration of sodium pentaborate solution requirements are met. Insert Page B 3.1.7-5 Attachment 1, Volume 6, Rev. 0, Page 197 of 231

Attachment 1, Volume 6, Rev. 0, Page 198 of 231 SLC System B 3.1.7 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.1.7.7 24 Demonstratin gthat each SLC System pump develops a flow rate2 2 @ 2pm at a discharge pressure ;-fjE1sig ensures that pump performance has not degraded during the fuel cycle. This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the ump design curve and is indicative of overall performance. Such inservicinsp ionS confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this Surveillance isQin accordance with the Inservice Testing Program or 9 a s " SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both i241 complete flow paths are tested everyg months at alternating month intervals. The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV isto pump demineralized water from a~24 SLC subsystem and into the RPV. The ERointh Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the Inonth Frequency; therefore, (3 the Frequency was concluded to be acceptable from a reliability 4 standpoint. Demonstrating that all heat traced piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution. An acceptable method for verifying that the suction piping is unblocked is to pump from the storage tank to the test tank. BWR/4 STS B 3.1.7-6 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 198 of 231

Attachment 1, Volume 6, Rev. 0, Page 199 of 231 SLC System B 3.1.7 BASES SURVEILLANCE REQUIREMENTS (continued) and the SLC pump suction lines heat tracing Is Inoperable i() E- Thei month Frequency is acceptable since there is a low probability (3 that the subject piping will be blocked due to precipitation of the boron from solution in the heat traced piping. This is especially true in light of the temperature verification of this piping required by SR 3.1.7.3. However, if, in performing SR 3.1.7.3, it is determined that the temperature of this piping has fallen below the specified minimum, lsolution temperath A-., SR 3.1.7.9 must be performed once within 24 hours after th e y

                     -J     temperature is restored to within thehlimits of Figure 3.1.7-2.

tin the vicinity of the SLC pumpsj SR 3.1.7.10 311(laboratoy nales) Enriched sodium pentaborate solution is made by mixin granular, l enriched sodium pentaborate with water. Isotopic teststin the-granular sodium pentaborate to verify the actual B-10 enrichment must be 0 performed prior to addition to the SLC tank in order to ensure that the proper B-10 atom percentage is being used. REFERENCES 1. 10 CFR 50.62. . T-*SAR, Section4. 43 00-(D( BWR/4 STS B 3.1.7-7 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 199 of 231

Attachment 1, Volume 6, Rev. 0, Page 200 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.7 BASES, STANDBY LIQUID CONTROL (SLC) SYSTEM

1. Editorial change made for enhanced clarity.

2. Changes are made (additions, deletions, and/or changes) to the ISTS Bases, which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

3. Changes have been made to reflect the Specification.

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

5. Typographical/grammatical error corrected.

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

Monticello Page 1 of I Attachment 1, Volume 6, Rev. 0, Page 200 of 231

Attachment 1,Volume 6, Rev. 0, Page 201 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 6, Rev. 0, Page 201 of 231

Attachment 1, Volume 6, Rev. 0, Page 202 of 231

                             .i                             I DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.7, STANDBY LIQUID CONTROL (SLC) SYSTEM There are no specific NSHC discussions for this Specification.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 202 of 231

Attachment 1, Volume 6, Rev. 0, Page 203 of 231 ATTACHMENT 8 ITS 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves Attachment 1, Volume 6, Rev. 0, Page 203 of 231

Attachment 1,Volume 6, Rev. 0, Page 204 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 6, Rev. 0, Page 204 of 231

C C ITS 3.1.8 ITS TS 3.0 UMING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS F. Scram DIscharge Volume (M 2 Li F. Scram Discharge Volume [Add proposed SR 3..8.1 M.1 LCO 3.1.8 D ascram dis charge SR 3.1.8.2 The scram discharge volume vent end drain valves shall nC) volume vent and drain vahs shall be cperable, b e cycled quarterly. 24mo 0 except as spedifed below. 0 SR 3.1.8.3 FOnce per aFtiNa clfy the scram discharge 0D

• Note 1 volume vent and drain valves dose withh 30 seconds ACTION A, 2. If any scam discharge volume vent or d[rain valve Is after receipt of a Leactor scram signal and open when 3 00 ACTION B made or found Inoperable, the Integrity of the scram the s rose.

Icram discharge volume shafl be maintained tby either

                                                                           -y tI                                                             acual or simulated'  A.4 ACTION A                 a. Leyjng

[ Oily~for aperiod not to e nd 7 l CD davs.1thRoperability of the redun~it ntW vle(s) 0 F 0 3 {Isciate the ACTION A, b. Itajnr~figh~oealv~/s.o h associated A.2 _ !!Lth!.1nopea N pi Uneo 0 ACTION B 0 Inusggntw runualil vervnw, Irl 1llv lunwl_ 0 Ipgs~tionerlrodically the Inoperable and the CD Add prposed Required L CU CD F ACTIONS r un ant valvets) may both be hI the open tArctopn B- CompletnT m , aw NOTE 2 posion to allow draining the scram discharge 0) 0 volume. ts2 jU 0

                                                                                                                                                                                    -I

- W ACTION C If a or b above carmot be met, at least I but one la operable cbntroI rod9Tsnot Indud'ng r removed to0 per specification 3.10.E or Inoperab awed CD w) LG. 1by 3.3.A!2) shan eflvhetd Required Action wt If Specifications 3.3A through D above are not met, an See ITS 3.1. 1, ITS 3.1.3, C) CA 0 orderly shutdown shall be Initiated and have reactor in ITS 3.1.4. the cold shutdown condition within 24 hours. ITS 3.1.5. I - ITS 3.1.6, ITS 3.3.1.2. ITS 3.9.5 3.314.3 83a 511184 Amendment No. 24 Page 1 of 1

Attachment 1, Volume 6, Rev. 0, Page 206 of 231 DISCUSSION OF CHANGES ITS 3.1.8, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES 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 3.3.F.2.b states, in part, "Maintaining the inoperable valves(s), or the associated redundant valve(s), in the closed position" if the inoperable valve is not restored to OPERABLE status in 7 days. ITS 3.1.8 Required Actions A.1 and B.1 state "Isolate the associated line." This changes the CTS by simplifying the Required Action by requiring isolation of the associated line instead of explicitly stating which valves to use to perform the isolation (i.e., inoperable valve(s) or the associated redundant valves(s)). The purpose of CTS 3.3.F.2.b is to isolate the affected SDV vent or drain line. This change is acceptable since the proposed Required Action also requires isolation of the associated line, and the only valves capable of isolating the SDV vent and drain lines are'the required SDV vent and drain valves. This change is designated as administrative because it does not result in any technical changes to the CTS. A.3 CTS 3.3.F.2 states, in part, "If a or b above cannot be met, at least all but one operable control rods (not including rods removed per specification 3.1 O.E or inoperable rods allowed by 3.3.A.2) shall be fully inserted." ITS 3.1.8 ACTION C, under the same conditions requires the unit to be in MODE 3. This changes the CTS by more clearly defining the all rods in condition as MODE 3. The purpose of CTS 3.3.F.2, when the requirements of CTS 3.3.F.2.a and b are not met, is to insert all operable control rods, which essentially ensures an inoperable SDV vent or drain valve cannot prevent a reactor scram. This change is acceptable because when the unit is in MODE 3, by definition, the reactor mode switch is in the shutdown condition and by design all OPERABLE rods will be inserted. The cross references to CTS 3.3.A.2 (Reactivity Margin-Stuck Control Rods) and CTS 3.10.E (Extended Core and Control Rod Drive Maintenance) are not necessary. CTS 3.3.A.2, which covers stuck control rods, is only applicable in MODES 1 and 2. CTS 3.10.E is only applicable during an outage. Therefore it is not necessary to include these cross references. This change is designated as administrative because it does not result in technical changes to the CTS. A.4 CTS 3.3.F states, in part, "verify the scram discharge volume vent and drain valves close within 30 seconds after receipt of a reactor scram signal and open when the scram is reset." ITS SR 3.1.8.3 requires the same test however the proposed Surveillance states that the reactor scram signal may be an "actual or simulated" signal. This changes the CTS by clarifying that the reactor scram signal may be either an "actual or simulated" reactor scram signal. Monticello Page 1 of 5 Attachment 1, Volume 6, Rev. 0, Page 206 of 231

Attachment 1, Volume 6, Rev. 0, Page 207 of 231 DISCUSSION OF CHANGES ITS 3.1.8, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES The purpose of the test is to verify that the valves close and open on the specified signal. OPERABILITY is adequately demonstrated in either case since the SDV vent and drain valves cannot discriminate between "actual" or simulated" signals. In addition, the CTS does not prohibit the signal from being an "actual" or "simulated" reactor scram signal. This change only clarifies the type of signal that may be used to perform the Surveillance Requirement. This change is designated as administrative because it does not result in any technical changes to the CTS. A.5 CTS 4.3.F requires a SDV vent and drain valve test to be performed "Once per operating cycle." ITS SR 3.1.8.3 requires performance of an SDV vent and drain valve test every "24 months." This changes the CTS by changing the Frequency from "Once per operating cycle" to "24 months." This change is acceptable because the current "operating cycle" is "24 months." In letter L-MT-04-036, from Thomas J. Palmisano (NMC) to the USNRC, dated June 30, 2004, NMC has proposed to extend the fuel cycle from 18 to 24 months and the same time has performed an evaluation in accordance with Generic Letter 91-04 to extend the unit Surveillance Requirements from 18 months to 24 months. CTS 4.3.F was included in this evaluation. This change is designated as administrative because it does not result in any technical changes to the CTS. MORE RESTRICTIVE CHANGES M.1 ITS SR 3.1.8.1 requires the verification that each SDV vent and drain valve is open. A Note is included that states that this Surveillance is not required to be met on vent and drain valves closed during performance of SR 3.1.8.2. The CTS does not contain a similar requirement. This changes the CTS by adding a new Surveillance Requirement for the SDV vent and drain valves. This change is acceptable because it helps to ensure the SDV is capable of performing its intended safety function. During normal operation, the SDV vent and drain valves should be in the open position to allow for drainage of the SDV piping. The Surveillance includes a Note to allow the SDV vent and drain valves to be cycled in accordance with ITS SR 3.1.8.2 without declaring the associated valves inoperable. This verification gives additional confidence that the SDV is available to receive and contain all the water discharge by the control rod drives during a scram. This change is designated as more restrictive because it adds a Surveillance Requirement that does not appear in the CTS. M.2 CTS 3.3.F requires the scram discharge volume vent and drain valve requirements to be met in the "reactor operation" condition. ITS LCO 3.1.8 is Applicable in MODES 1 and 2. This changes the CTS by requiring the scram discharge volume vent and drain valve requirements to be met in MODE 2

       < 1% RATED THERMAL POWER (RTP).

The purpose of CTS 3.3.F is to ensure the scram discharge volume vent and drain valve requirements are met to ensure the negative scram reactivity is consistent with those values assumed in the accident and transient analysis. Monticello Page 2 of 5 Attachment 1, Volume 6, Rev. 0, Page 207 of 231

Attachment 1, Volume 6, Rev. 0, Page 208 of 231 DISCUSSION OF CHANGES ITS 3.1.8, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES This change expands the Applicability to require the scram discharge volume vent and drain valve requirements to be met at all times when in MODE 2, instead of when > 1% RTP (the CTS 1.0.0 definition states that Power Operation is when reactor power is > 1%RTP). This change is acceptable since the control rods must be capable of properly scramming in MODE 2 because the reactor is critical or control rods are withdrawn (thus the need exists for the scram discharge volume vent and drain valves to be OPERABLE). This change is designated as more restrictive because the LCO will be applicable under more reactor conditions. RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 4 - Relaxation of RequiredAction) CTS 3.3.F.2.a allows 7 days of continuous operation with any number of SDV drain or vent valves inoperable as long as the redundant valve (i.e., the one in the same line) is verified to be OPERABLE on a daily basis. After the 7 day period, CTS 3.3.F.2.b requires that either the inoperable valve(s) or the associated redundant valve(s) be closed. However, if one valve has been inoperable for greater than 7 days and the valve or its redundant valve is closed, and another valve in a different line becomes inoperable, the CTS does not allow a separate 7 day time to restore the valve; the second inoperable valve or its redundant valve must be closed immediately in order to meet the requirements of CTS 3.3.F.2.b. ITS 3.1.8 ACTIONS are modified by a Note I that states "Separate Condition entry is allowed for each SDV vent and drain line." ITS 3.1.8 ACTION A covers inoperabilities for one or more SDV vent or drain lines with one valve inoperable. ITS 3.1.8 ACTION B covers inoperabilities for one or more SDV vent or drain lines with both valves inoperable. This changes the CTS by allowing separate Condition entry for each inoperable SDV vent or drain line. That is, under the same scenario described above, the second inoperable valve will get a 7 day restoration time before the associated line must be isolated. Other modifications associated with CTS 3.3.F.2.a and CTS 3.3.F.2.b are discussed in DOCs A.2, L.2, and L.3. The purpose of CTS 3.3.F.2.a isto allow 7 days of operation when any number of SDV vent or drain valves are inoperable as long as the associated redundant valve on the same line is operable. The purpose of CTS 3.3.F.2.b is to require immediate closure of the inoperable valves if both valves on a SDV vent or drain line are inoperable and to also require isolation of the affected penetration after 7 days of operation with any SDV vent or drain valve inoperable. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to Monticello Page 3 of 5 Attachment 1, Volume 6, Rev. 0, Page 208 of 231

Attachment 1, Volume 6, Rev. 0, Page 209 of 231 DISCUSSION OF CHANGES ITS 3.1.8, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the 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 repair period. This change will allow separate Condition entry for each SDV vent and drain line. This change will effectively allow 7 days to isolate the affected line when one valve in the line is discovered to be inoperable. This is acceptable since the since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SDV line. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SDV lines are governed by subsequent Condition entry and application of associated Required Actions. 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 4 - Relaxation of Required Action) When any scram discharge volume vent or drain valve is made or found inoperable and the associated line is not isolated, CTS 3.3.F.2.a requires daily verification of the OPERABILITY of the redundant valve(s). ITS 3.1.8 ACTION A covers the condition when one SDV vent or drain valve is inoperable in one or more SDV vent or drain lines, but does not require daily verification of the OPERABILITY of the redundant valve in the associated line if the line is not isolated. This changes the CTS by deleting the requirement to verify the OPERABILITY of the redundant valve(s) on a daily basis if the associated line is not isolated. The purpose of CTS 3.3.F.2.a is to provide compensatory actions for inoperable SDV vent or drain valves. This change is acceptable because the Required Actions are used to establish remedial measures that must be taken in response to the degraded conditions in order to minimize risk associated with continued operation while providing time to repair inoperable features. The Required Actions are consistent with safe operation under the specified Condition, considering the OPERABLE status of the 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 repair period. This change deletes the requirement to verify on a daily basis the OPERABILITY of the redundant valve. ITS SR 3.1.8.2 requires that each SDV vent and drain valve to be cycled to the fully closed and fully open position every 31 days. ITS SR 3.1.8.3 requires the verification that each valves actuates as required on a scram signal every 24 months. These Surveillances and associated Frequencies are considered sufficient to determine whether or not a SDV vent or drain valve is OPERABLE. As stated in the Bases of ITS SR 3.0.3, it is recognized that the most probable result of any particular Surveillance being performed isthe verification of conformance with the requirements. This change is acceptable since the normal Surveillances and associated Frequencies are considered acceptable with respect to determining the status of a SDV vent or drain valve. This change is designated as less restrictive because less stringent Required Actions are being applied in the ITS than were applied in the CTS. Monticello Page 4 of 5 Attachment 1, Volume 6, Rev. 0, Page 209 of 231

Attachment 1, Volume 6, Rev. 0, Page 210 of 231 DISCUSSION OF CHANGES ITS 3.1.8, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES L.3 (Category 3- Relaxation of Completion Time): When any scram discharge volume vent or drain valve is made or found inoperable, CTS 3.3.F.2.a allows, for a period not to exceed 7 days, the associated line to remain unisolated provided the redundant valve in the line is OPERABLE. If both valves in a SDV line are inoperable, CTS 3.3.F.2.b requires "maintaining" the inoperable valve(s) or the associated redundant valve(s) in the closed position. This effectively means that if both valves in a SDV line are inoperable, the line must be isolated immediately. ITS 3.1.8 ACTION B covers the condition when both valves are inoperable in one or more SDV vent or drain lines. ITS 3.1.8 Required Action B.1 requires isolation of the associated line within 8 hours. This changes the CTS by allowing 8 hours to isolate a vent or drain line in lieu of requiring it to be isolated immediately when both valves are determined to be inoperable. The purpose of CTS 3.3.F.2 is to provide compensatory actions for inoperable SDV vent or drain valves. This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. This change extends the time from immediately to 8 hours to isolate a SDV vent or drain line when it is determined that both valves associated with the same SDV vent or drain line are inoperable. If both valves in a line are inoperable, the line must be isolated to contain the reactor coolant during a scram. The 8 hour Completion Time to isolate the line is acceptable based on the low probability of a scram occurring while the line is not isolated and unlikelihood of significant CRD seal leakage. This change is designated as less restrictive because additional time is allowed to isolate the SDV line than was allowed in the CTS. L.4 (Category 3 - Relaxation of Completion Time) CTS 3.3.F.2 requires the insertion of all OPERABLE control rods within ten hours if the compensatory actions of CTS 3.3.F.2.a and b cannot be met. ITS 3.1.8 ACTION C requires the unit to be in MODE 3 in 12 hours. This change increases the time to insert all OPERABLE control rods (i.e., to be in MODE 3 as discussed in DOC A.3) from 10 hours to 12 hours. The purpose of the action in CTS 3.3.F.2 is to insert all OPERABLE control rods (which ensures an inoperable SDV vent or drain valve cannot prevent a reactor scram) in an acceptable time frame. This change is acceptable because the Completion Time is consistent with safe operation under the specified Condition, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, and the low probability of a DBA occurring during the allowed Completion Time. The Completion Time of 12 hours is needed to give the operator sufficient time to accomplish an orderly power reduction without challenging unit systems. This proposed Completion Time is consistent with the Completion Times'to achieve MODE 3 in all other ITS Specifications. The inoperabilities of SDV vent and drain valves should not require the unit to reach MODE 3 any faster than other Specification requiring entry into this same MODE. This change is designated as less restrictive because additional time is allowed to insert all OPERABLE control rods than was allowed in the CTS. Monticello Page 5 of 5 Attachment 1, Volume 6, Rev. 0, Page 210 of 231

Attachment 1, Volume 6, Rev. 0, Page 211 of 231 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 211 of 231

Attachment 1,Volume 6, Rev. 0, Page 212 of 231 SDV Vent and Drain Valves 3.1.8 CIS 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves 3.3.F.1 LCO 3.1.8 Each SDV vent and drain valve shall be OPERABLE. 3.3.F.1 APPLICABILITY: MODES 1 and 2. ACTIONS ~--NOTES DOC 1. Separate Condition entry is allowed for each SDV vent and drain line. LA 33.F.2.b 2. An isolated line may be unisolated under administrative control to allow draining and venting of the SDV. CONDITION REQUIRED ACTION J COMPLETION TIME A. One or more SDV vent A.1 Isolate the associated line. 7 days 3.3.F.2. or drain lines with one 3.3 F.2 a a 3.3.F.2b' valve inoperable.

• B. One or more SDV vent B.1 Isolate the associated line. 8 hours 3.3.F.2. or drain lines with both 3.3.F.2.b valves inoperable.

I 3.3.F.2 C. Required Action and C.1 Be in MODE 3. 12 hours associated Completion Time not met.

                                            .1.                                      J BWR/4 STS                                     3.1.8-1                            Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 212 of 231

Attachment 1, Volume 6, Rev. 0, Page 213 of 231 SDV Vent and Drain Valves 3.1.8 03 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC SR 3.1.8.1 --- NOTE---- - MA1 Not required to be met on vent and drain valves closed during performance of SR 3.1.8.2. Verify each SDV vent and drain valve is open. 31 days 4.3.F SR 3.1.8.2 Cycle each SDV vent and drain valve to the fully 92 days closed and fully open position. 4-4.3.F SR 3.1.8.3 Verify each SDV vent and drain valve: jFJNmonths 0D

a. Closes in receipt of an Visecond~1er actual or simulated scram signal and 0D 0

b. Opens when the actual or simulated scram signal is reset.

BWR/4 STS 3.1.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 213 of 231

Attachment 1, Volume 6, Rev. 0, Page 214 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES

1. The brackets are removed and the proper plant specific information/value is provided.

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

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 214 of 231

Attachment 1, Volume 6, Rev. 0, Page 215 of 231 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs) Attachment 1, Volume 6, Rev. 0, Page 215 of 231

Attachment 1,Volume 6, Rev. 0, Page 216 of 231 SDV Vent and Drain Valves B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Scram Discharge Volume (SDV) Vent and Drain Valves BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram. During a scram, the SDV vent and drain valves close to contain reactor water. The SDV is a volume of header piping that connects to each hydraulic control unit (HCU) and drains into an instrument volume. There are two SDVs Each (headers) and two instrument volumes, each receiving a proximatel on f me of control roddrive (CRD) discharges. i si volumefoeft connected to a(CRD) drain line t tw vntuent Each header is connected to a co on vent line with two valves in seriesJ for a total of four vent valves. The header piping is sized to receive and contain all the water discharged by the CRDs during a scram. The design and functions of the SDV are described in Reference 1. Iforatotaloffourdrainvalves APPLICABLE The Design Basis Accident and transient analyses assume all of the SAFETY control rods are capable of scramming. The acceptance criteria for the ANALYSES SDV vent and drain valves are that they operate automatically to:

a. Close during scram to limit the amount of reactor coolant discharged so that adequate core cooling is maintained and offsite doses remain within the limits of 10 CFR 100 (Ref. 2) and

b. Open on scram reset to maintain the SDV vent and drain path open so that there is sufficient volume to accept the reactor coolant discharged during a scram.

Isolation of the SDV can also be accomplished by manual closure of the SDV valves. Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves. For a bounding leakage case, the offsite doses are well within the limits of 10 CFR 100 (Ref. 2), and adequate core cooling is maintained (Ref. 3). The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation to ensure that the SDV has sufficient capacity to contain the reactor coolant discharge during a full core scram. To automatically ensure this capacity, a reactor scram (LCO 3.3.1.1, "Reactor Protection System (RPS) Instrumentation") is initiated if the SDV water level in the instrument volume exceeds a specified setpoint. The setpoint is chosen so that all control rods are inserted before the SDV has insufficient volume to accept a full scram. SDV vent and drain valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). BWR14 STS B 3.1.8-1 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 216 of 231

Attachment 1, Volume 6, Rev. 0, Page 217 of 231 SDV Vent and Drain Valves B 3.1.8 BASES LCO The OPERABILITY of all SDV vent and drain valves ensures that the SDV vent and drain valves will close during a scram to contain reactor water discharged to the SDV piping. Since the vent and drain lines are provided with two valves in series, the single failure of one valve in the open position will not impair the isolation function of the system. Additionally, the valves are required to open on scram reset to ensure that a path is available for the SDV piping to drain freely at other times. APPLICABILITY In MODES land 2, scram may be required; therefore, the SDV vent and 3 drain valves must be OPERABLE. In MODES 3 and 4, control rods are not able to be withdrawn since the reactor mode switch is in shutdown and a control rod block is applied.FThis ppvides adequatp, 6ontrols tol lensure that offlva single control rod carrbe withdrawn. IAlso, during MODE 5, only a single control rod can be withdrawn from a core cell containing fuel assemblies. Therefore, the SDV vent and drain valves are not required to be OPERABLE in these MODES since the reactor is subcritical and only one rod may be withdrawn and subject to scram. ACTIONS. The ACTIONS table is modified by Note 1.indicating that a separate Condition entry is allowed for each SDV vent and drain line. This is acceptable, since the Required Actions for each Condition provide appropriate compensatory actions for each inoperable SDV line. Complying with the Required Actions may allow for continued operation, and subsequent inoperable SDV lines are governed by subsequent Condition entry and application of associated Required Actions. [ R When a line is isolated, the potential for an inadvertent scram due to high SDV level is increased. During these periods, the line may be unisolated under administrative control. This allows any accumulated water in the line to be drained, to preclude a reactor scram on SDV high level. This is acceptable since the administrative controls ensure the valve can be closed quickly, by a dedicated operator, if a scram occurs with the valve open. A.1 When one SDV vent or drain valve is inoperable in one or more lines, the associated line must be isolated to contain the reactor coo dng a scram. The 7 day Completion Time is reasonable, given the level of redundancy in the lineJand the low probability of a scram occurring whiled 3) t -the valveS eginoperable and the line is not isolated. The SDV is still J isolable since the redundant valve in the affected line is OPERABLE. During these periods, the single failure criterion may not be preserved, and a higher risk exists to allow reactor water out of the primary system during a scram. BWR/4 STS B 3.1.8-2 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 217 of 231

Attachment 1, Volume 6, Rev. 0, Page 218 of 231 B 3.1.8 Q' INSERT I The ACTIONS Table is modified by a second Note stating that an isolated line may be unisolated under administrative control to allow draining and venting of the SDV. Insert Page B 3.1.8-2 Attachment 1, Volume 6, Rev. 0, Page 218 of 231

Attachment 1, Volume 6, Rev. 0, Page 219 of 231 SDV Vent and Drain Valves B 3.1.8 BASES ACTIONS (continued) B.1 If both valves in a line are inoperable, the line must be isolated to contain the reactor coolant during a scram. he 8 hour Completion Time to isolate the line is based on the low probability of a scram occurring while the line is not isolated and unlikelihood of significant CRD seal leakage. C.1 If any Required Action and associated Completion Time is not met, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at ast MODE 3 within 12 hours. The allowed Completion Time of 12 hours is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. SURVEILLANCE SR 3.1.8.1 REQUIREMENTS During normal operation, the SDV vent and drain valves should be in the open position (except when performing SR 3.1.8.2) to allow for drainage of the SDV piping. Verifying that each valve is in the open position ensures that the SDV vent and drain valves will perform their intended functions during normal operation. This SR does not require any testing or valve manipulation; rather, it involves verification that the valves are in the correct position. The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation, which ensure correct valve positions. SR 3.1.8.2 During a scram, the SDV vent and drain valves should close to contain the reactor water discharged to the SDV piping. Cycling each valve through its complete range of motion (closed and open) ensures that the valve will function properly during a scram. The 92 day Frequency is based on operating experience and takes into account the level of redundancy in the system design. BWR/4 STS B 3.1.8-3 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 219 of 231

Attachment 1, Volume 6, Rev. 0, Page 220 of 231 SDV Vent and Drain Valves B 3.1.8 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.1.8.3 SR 3.1.8.3 is an integrated test of the SDV vent and drain valves to verify total system performance. After receipt of a simulated or actual scram 3 signal, the closure of the SDV vent and drain valves is verified. The / closure time of seconds after receipt of a scram signal is based on the 0 bounding leakage case evaluated in the accident analysis (Ref. E (3 Similarly, after receipt of a simulated or actual scram reset signal, the Control Rod opening of the SDV vent and drain valves is verified. The LOGIC OPERAILrrY, i SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.j#overlap this Surveillance to provide complete, 2 testing of the assumed safety function. The onT Frequency is 24 based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance 24 when performed at thefmonth Frequency; therefore, the Frequency i) was concluded to be acceptable from a reliability standpoint. REFERENCES TSAR, Section [4.2.31. 0 0D

2. 10 CFR 100. -

3. NUREG-0803, "Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping," August 1981.

BWR/4 STS B 3.1.84 Rev. 3.0, 03/31/04 Attachment 1, Volume 6, Rev. 0, Page 220 of 231

Attachment 1, Volume 6, Rev. 0, Page 221 of 231 JUSTIFICATION FOR DEVIATIONS ITS 3.1.8 BASES, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES

1. Changes are made (additions, deletions, and/or changes) to the ISTS Bases, which reflect the plant specific nomenclature, number, reference, system description, analysis, or licensing basis description.

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

3. Typographical/grammatical error corrected.

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

5. The brackets are removed and the proper plant specific information/value is provided.

Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 221 of 231

Attachment 1, Volume 6, Rev. 0, Page 222 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1,Volume 6, Rev. 0, Page 222 of 231

Attachment 1, Volume 6, Rev. 0, Page 223 of 231 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.1.8, SCRAM DISCHARGE VOLUME (SDV) VENT AND DRAIN VALVES There are no specific NSHC discussions for this Specification. Monticello Page 1 of I Attachment 1, Volume 6, Rev. .0,Page 223 of 231

Attachment 1, Volume 6, Rev. 0, Page 224 of 231 ATTACHMENT 9 RelocatedlDeleted Current Technical Specifications Attachment 1, Volume 6, Rev. 0, Page 224 of 231

Attachment 1, Volume 6, Rev. 0, Page 225 of 231 CTS 3/4.3.B.2i Control Rod Drive Housing Support System Attachment 1, Volume 6, Rev. 0, Page 225 of 231

Attachment 1, Volume 6, Rev. 0, Page 226 of 231 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs) Attachment 1, Volume 6, Rev. 0, Page 226 of 231

c C CTS 3/4.3.B C 3.0 UMING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS I-(b) when the rod Is withdrawn the first time subsequent to each refueling outage, observe discemibte response of the nuclear See ITS 3.1.3 } 0) W

0) Instrunentation. However, for initial rods when ,-

C, response is not discernible, subsequent 2 CD exercising of these rods after the reactor Is critical shall be performed to observe nudesw (D instrumentation response. 2 3

-4

2. The control rod dive housing support system shall be Ip 2. The control rod drive housing support system shall be 0 pace during re .corpower operation and when the Inspected after re 6mbly and the results of the at 0

reactor coo a system Is pressurized above Inspection rec 2 3 atmosphe'pressure with fuel In the reactor ye CD unless all parable control rods are fully Insert ard CD 0s Snedfi fion -A1 Is met. IN I I

;U                                                                               3.(a) To consider the rod worth minimizer operable, the (D        3.(a)Control rod withdrawal sequences shall be established                                                                                                       CD so that the maximum calculated reactivity that could be                 following steps must be performed:

added by dropout of any increment of any one control (I) The control rod withdrawal sequence for the rod 0 blade will not make the core more than 1.3% Ak worth minimizer computer shall be verified as supercrtIcal. corred. See ITS 3.32.1} t 2) Ni 0) (XI The rod worth minimizer computer on-ine diagnostic to to test shall be successfully completed. CD in Proper annunciation of the selection error of at least one out-of-sequence control rod In each fully Ki C; Inserted group shall be verified. 0 0 K) { See ITS 3.1.6 } 3.3/4.3 79 119181 Amendment No. 0 Page 1 of 1

Attachment 1, Volume 6, Rev. 0, Page 228 of 231 DISCUSSION OF CHANGES CTS 3/4.3.B.2, CONTROL ROD DRIVE HOUSING SUPPORT SYSTEM ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 1 - Relaxation of LCO Requirement) CTS 314.3.B.2 requires the control rod drive housing support system to be in place during reactor power operation and when the reactor coolant system is pressurized above atmospheric pressure with fuel in the reactor vessel, unless all operable control rods are fully inserted and Specification 3.3.A.1 is met. CTS 4.3.B.2 requires the control rod drive housing support system to be inspected after reassembly and the results of the inspection recorded. ITS 3.1 does not include the requirements for the control rod drive housing support system. This changes the CTS by deleting the explicit control rod drive housing support system requirements from the Technical Specifications. The purpose of CTS 3/4.3.B.2 is to ensure that the control rod drive housing support system is operable when control rods are withdrawn from the core. This change is acceptable because the LCO requirements continue to ensure that the structures, systems, and components are maintained consistent with the safety analyses and licensing basis. The CTS 3/4.3.B.2 requirement for the control rod drive housing support to be in place is included in the OPERABILITY requirements for control r6ds. Plant configuration management provides adequate controls to assure the control rod drive housing support is in place. The current Technical Specifications require the control rod drive housing support system to be inspected after reassembly and the results of the inspection recorded. This current Technical Specifications requirement verifies that the control rod drive housing support is in place for reactor operation in MODES 1, 2, and 3. Post-maintenance inspections conducted through plant configuration management control have the same function as the current Technical Specifications requirement. Since work is not normally performed on the control rod drive housing support at power, and checks on its installation are not made at power there is no current requirement to verify control rod drive housing support installation in power operating conditions. Therefore, the deletion of this current Monticello Page 1 of 2 Attachment 1, Volume 6, Rev. 0, Page 228 of 231

Attachment 1, Volume 6, Rev. 0, Page 229 of 231 DISCUSSION OF CHANGES CTS 314.3.B.2, CONTROL ROD DRIVE HOUSING SUPPORT SYSTEM Technical Specifications is acceptable based on use of plant configuration management control to ensure proper control rod drive housing support system installation. This change is designated as a less restrictive change because a requirement is being removed from the Technical Specifications. Monticello Page 2 of 2 Attachment 1, Volume 6, Rev. 0, Page 229 of 231

a Attachment 1, Volume 6, Rev. 0, Page 230 of 231 Specific No Significant Hazards Considerations (NSHCs) Attachment 1, Volume 6, Rev. 0, Page 230 of 231

Attachment 1, Volume 6, Rev. 0, Page 231 of 231 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 314.3.B1.2, CONTROL ROD DRIVE HOUSING SUPPORT SYSTEM There are no specific NSHC discussions for this Specification. Monticello Page 1 of 1 Attachment 1, Volume 6, Rev. 0, Page 231 of 231}}