ML051990512

Improved Technical Specifications, Volume 14, Revision 0, ITS Section 3.9, Refueling Operations

ML051990512
Person / Time
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: ML051990512 (158)

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IMPROVED TECHNICAL SPECIFICATIONS MONTICELLO NUCLEAR GENERATING PLANT VOLUME 14 ITS Section 3.9, Refueling Operations dNMc Commited to Ncer Excellnce

, Volume 14, Rev. 0, Page 1 of 157 ATTACHMENT I VOLUME 14 MONTICELLO IMPROVED TECHNICAL SPECIFICATIONS CONVERSION ITS SECTION 3.9 REFUELING OPERATIONS Revision 0, Volume 14, Rev. 0, Page 1 of 157

, Volume 14, Rev. 0, Page 2 of 157 LIST OF ATTACHMENTS

1.

ITS 3.9.1

2.

ITS 3.9.2

3.

ITS 3.9.3

4.

ITS 3.9.4

5.

ITS 3.9.5

6.

ITS 3.9.6

7.

ITS 3.9.7

8.

ITS 3.9.8

9.

Relocated/Deleted Current Technical Specifications (CTS)

10.

Improved Standard Technical Specifications (ISTS) not adopted in the Monticello ITS, Volume 14, Rev. 0, Page 2 of 157

, Volume 14, Rev. 0, Page 3 of 157 ATTACHMENT I ITS 3.9.1, Refueling Equipment Interlocks, Volume 14, Rev. 0, Page 3 of 157

, Volume 14, Rev. 0, Page 4 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 4 of 157

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, Volume 14, Rev. 0, Page 6 of 157 DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS 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-1 433, 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 3.1 O.A requires the reactor mode switch to be in the refuel position during core alterations and the refueling interlocks to be OPERABLE. ITS LCO 3.9.1 only requires the refueling "equipment" interlocks associated with the reactor mode switch refuel position to be OPERABLE. This changes the CTS by splitting the requirement of the refueling interlocks into two Specifications. All other refueling interlocks with the reactor mode switch in the refuel position are covered in ITS 3.9.2.

The purpose of CTS 3.1O.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. The requirements for the refueling equipment interlocks are retained in ITS 3.9.1 while ITS 3.9.2 includes the requirements for the refuel position one-rod-out interlock. This change is acceptable because all of the refueling interlocks covered in CTS 3.1 0.A have been incorporated in either ITS 3.9.1 or ITS 3.9.2. Therefore, this change is considered a presentation preference change only and, as such, is considered an administrative change.

A.3 CTS 4.10.A requires refueling interlocks to be functionally tested. ITS SR 3.9.1.1 requires the same test on the required refueling equipment interlock inputs and provides a list of equipment interlocks. This changes the CTS by providing a specific list of refueling equipment interlocks.

The purpose of CTS 3.10.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. ITS SR 3.9.1.1 provides a specific list of refueling equipment interlocks that are applicable when the reactor mode switch is in the refuel position. This change is acceptable because the proposed list of refueling equipment interlocks in ITS SR 3.9.1.1 is consistent with the intent of CTS 4.1O.A. Therefore, this change is considered a presentation preference change only and, as such, is considered an administrative change.

A.4 CTS 3.10.A states that the refueling interlocks are required to be operable "except as specified in specification 3.10.E." The ITS does not include this sentence. This changes the CTS by deleting this cross-reference to another Specification.

This change is acceptable because the requirements have not changed. The CTS reference to Specification 3.10.E is an "information only" statement that neither adds, eliminates, nor modifies requirements. The ITS convention is to not include these types of statements. Any changes to the requirements of CTS 3.10.E will be discussed in the Discussion of Changes for ITS 3.10.2 and Monticello Page 1 of 5, Volume 14, Rev. 0, Page 6 of 157

, Volume 14, Rev. 0, Page 7 of 157 DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS ITS 3.10.6. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M.1 CTS 4.1O.A requires the refueling interlocks (in this case the refueling equipment interlocks) to be functional tested "prior to any fuel handling, with the head off the reactor vessel" and at "weekly intervals thereafter." However, it does not state how soon "prior to" starting the above evolutions. ITS SR 3.9.1.1 requires a similar verification every 7 days. This changes the CTS by eliminating the specific requirement to functionally test the refueling equipment interlocks "prior to any fuel handling, with the head off the reactor vessel," and replaces it with a requirement to perform the test 7 days prior to any fuel handling.

The purpose of CTS 4.10.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. This change is acceptable because ITS SR 3.0.1 states "SRs shall be met during the MODES or other specified conditions in the Applicability for the individual LCOs, unless otherwise stated in the SR." In addition, ITS SR 3.0.4 states "Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency." Therefore, under the ITS, the Surveillances must be met within the 7 day Surveillance Frequency prior to the initiation of in-vessel fuel movements. For CTS 4.1 0.A, the periodic Surveillance Frequency for verifying the refueling interlocks are OPERABLE is acceptable during the MODE of Applicability, and is also acceptable during the period prior to entering the MODE of Applicability, since the specific time prior to starting the above evolutions is not specified. This change is designated as more restrictive since it now requires the test "7 days" prior to any fuel handling, with the head off the reactor vessel.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 2 - Relaxation of Applicability) CTS 3.1 O.A requires the reactor mode switch to be in the refuel position during "core alterations" and the refueling interlocks to be OPERABLE. ITS LCO 3.9.1 requires the refueling equipment interlocks associated with the reactor mode switch refuel position to be OPERABLE during "in-vessel fuel movement with equipment associated with the interlocks when the reactor mode switch is in the refuel position." This changes Monticello Page 2 of 5, Volume 14, Rev. 0, Page 7 of 157

, Volume 14, Rev. 0, Page 8 of 157 DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS the CTS by only requiring the refuel equipment interlocks to be OPERABLE during certain CORE ALTERATIONS (i.e., during in-vessel fuel movement).

The purpose of CTS 3.1O.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. This change is acceptable because the requirements continue to ensure that the structures, systems, and components are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. Currently all the refueling interlocks are required during all types of CORE ALTERATIONS. This change only requires the refueling "equipment" interlocks to be OPERABLE during in-vessel fuel movement with equipment associated with the interlocks. Therefore, these refueling interlocks (those listed in ITS SR 3.9.1.1) will only be required to be OPERABLE during in-vessel fuel movement when the associated equipment is being used. That is, if the refuel platform is being used for in-vessel fuel movement then the "service platform hoist, fuel loaded" interlock is not required to be OPERABLE. In addition, if the refuel platform is in use and the fuel grapple is being used, the "refuel platform fuel grapple, fuel loaded" and "refuel platform fuel grapple fully retracted position" must be OPERABLE as well as the all-rods-in interlock. When the service platform hoist is being used, the "all-rods-in" and the "service platform hoist, fuel load" interlocks are required to be OPERABLE.

This change is acceptable because the only time these interlocks provide a safety function is when they are being used for the specific operation. This change is designated as less restrictive because the LCO requirements are applicable in fewer operating conditions than in the CTS.

L.2 (Category 1 - Relaxation of LCO Requirement) CTS 3.1 O.A requires the reactor mode switch to be "locked" in the refuel position. ITS 3.9.1 is applicable when the reactor mode switch is in the refuel position. This changes the CTS by deleting the requirement to lock the reactor mode switch when in the refuel position.

The purpose of CTS 3.1 O.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. This change is acceptable because the LCO requirements continue to ensure that the refuel equipment interlocks are maintained consistent with the safety analyses. This change deletes the requirement to maintain the reactor mode switch "locked" in the refuel position.

However, the requirement to maintain the reactor mode switch in the refuel position is maintained by the Applicability statement in ITS 3.9.1 (i.e., during in-vessel fuel movement with equipment associated with the interlocks when the reactor mode switch is in the refuel position). The position of the reactor mode switch, average reactor coolant temperature, and the status of the reactor vessel head closure bolts defines the MODE in which the unit is operating. The MODES are defined in ITS Table 1.1-1. The reactor mode switch position is only changed under strict administrative controls to ensure compliance with the Technical Specification's at all times. Before changing the position, unit operators will ensure the requirements of the applicable LCOs are met prior to changing the position of the reactor mode switch. Therefore, the requirement that the reactor mode switch be "locked" in the refuel position during in-vessel fuel movements has been deleted and is acceptable because the position of the reactor mode switch is controlled under strict administrative controls. In addition, ITS 3.9.2 will require the reactor mode switch to be locked when in the refuel position when a Monticello Page 3 of 5, Volume 14, Rev. 0, Page 8 of 157

, Volume 14, Rev. 0, Page 9 of 157 DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS control rod is withdrawn. This change is designated as less restrictive because less stringent LCO requirements are being applied in the ITS than were applied in the CTS.

L.3 (Category 4 - Relaxation of Required Action) CTS 3.1 O.A does not provide specific Actions for when the refueling equipment Interlocks are inoperable.

However, since the refueling interlocks must be OPERABLE during CORE ALTERATIONS, this implies that CORE ALTERATIONS must be suspended if the refueling interlocks are inoperable. ITS 3.9.1 ACTION A covers the condition when one or more required refueling equipment interlocks are inoperable and requires either the immediate suspension of in-vessel fuel movement with equipment associated with the inoperable interlock(s) or the insertion of a control rod withdrawal block and a verification that all control rods are fully inserted. This changes the CTS by providing specific Actions for when a refueling equipment interlock is not met.

The purpose of CTS 3.10.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. 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 continuing the refueling activities while providing time to repair inoperable features. CTS 3.10.A does not provide specific actions for when the refueling interlocks are not met but since the interlocks are required to be met during CORE ALTERATIONS, the unit would immediately stop these activities if the interlocks are inoperable. ITS 3.9.1 Required Action A.1 essentially is the same requirement since it requires the immediate suspension of in-vessel fuel movement with equipment associated with the inoperable interlock(s). However, ITS 3.9.1 Required Actions A.2.1 and A.2.2 provide an option to insert a control rod withdrawal block and to verify that all control rods are fully inserted. This option is acceptable because it helps to ensure the purpose of the refueling equipment interlocks is accomplished; the core does not achieve criticality during refueling. Therefore, it is acceptable to continue with the in-vessel fuel movements since the proposed alternative Required Actions will ensure all control rods are completely inserted and will remain in the inserted position even if control rod withdrawal is attempted. 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.4 (Category 5 - Deletion of Surveillance Requirement) CTS 4.1 O.A requires the refueling interlocks (in this case the refueling equipment interlocks) to be functional tested "following any repair work associated with the interlocks." ITS SR 3.9.1.1 does not require this verification "following any repair work associated with the interlocks." This changes the CTS by eliminating the requirement to functionally test the refueling equipment interlocks "following any repair work associated with the interlocks."

The purpose of CTS 4.1 O.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. 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 continues to be tested in a manner and at a Frequency Monticello Page 4 of 5, Volume 14, Rev. 0, Page 9 of 157

, Volume 14, Rev. 0, Page 10 of 157 DISCUSSION OF CHANGES ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS necessary to give confidence that the equipment can perform its assumed safety function. Any time the OPERABILITY of a system or component has been affected by repair, maintenance, modification, or replacement of a component, post-maintenance testing is required to demonstrate the OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under ITS SR 3.0.1. In addition, the requirements of 10 CFR 50, Appendix B, Section Xi (Test Control) provide adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria.

Compliance with 10 CFR 50, Appendix B, is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS.

Monticello Page 5 of 5, Volume 14, Rev. 0, Page 10 of 157

, Volume 14, Rev. 0, Page 11 of 157 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 11 of 157

, Volume 14, Rev. 0, Page 12 of 157 Refueling Equipment Interlocks 3.9.1 CTS 3.9 REFUELING OPERATIONS 3.10A 3.9.1 Refueling Equipment Interlocks 3.1oA LCO 3.9.1 The refueling equipment interlocks shall be OPERABLE.

[associated with the reactor (mode switch refuel position 0D 3.10A APPLICABILITY:

During in-vessel fuel movement with equipment associated with the interlocks, ltwhen the reactor mode switch (is in the refuel position 0

DOC L.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Suspend in-vessel fuel Immediately refueling equipment movement with equipment interlocks inoperable.

associated with the inoperable interlock(s).

OR A.2.1 Insert a control rod Immediately withdrawal block.

AND A.2.2 Verify all control rods are Immediately fully inserted.

BWR/4 STS 3.9.1-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 12 of 157

, Volume 14, Rev. 0, Page 13 of 157 Refueling Equipment Interlocks 3.9.1

.t1 IF2/II I ARIIF DfhI0 IIF2FMNT.~rr V LILLJ'.INI.JIu m

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SURVEILLANCE FREQUENCY 4.10A SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each of the following required refueling equipment interlock inputs:

a.

All-rods-in

b. Refuel platform position

c.

Refuel platformMfuel grappled fuel loade Md.

Refuel platform fuel grapple fully retracted positions fe. Refuel platform frame mounted hoist, fuel loaded/

zf.

Refuel platform monorail mounted hoist, fuel loaded T and

[g. Service platform hoist, fuel loaded.]

7 days I

-0 0

0D 0

0 I

BWR/4 STS 3.9.1-2, Volume 14, Rev. 0, Page 13 of 157 Rev. 3.0, 03/31/04

, Volume 14, Rev. 0, Page 14 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS

1. The current wording of ISTS 3.9.1 and the associated Applicability could imply that all the refueling equipment interlocks are required at all times during in-vessel fuel movement. The CTS only requires the interlocks associated with the refuel position, not those associated with other positions of the reactor mode switch, and only when the reactor mode switch is in the refuel position, not when it is in the shutdown position. Therefore, to avoid confusion, the LCO and Applicability have been modified to specifically state that the refueling interlocks are those associated with the refuel position, and that it is applicable when the reactor mode switch is in the refuel position. In addition, this change is consistent with the most recently approved BWR ITS conversions (i.e., FitzPatrick, LaSalle Units 1 and 2, Quad Cities Units I and 2, and Dresden Units 2 and 3).

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

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

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 14 of 157

, Volume 14, Rev. 0, Page 15 of 157 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 15 of 157

, Volume 14, Rev. 0, Page 16 of 157 Refueling Equipment Interlocks B 3.9.1 B 3.9 REFUELING OPERATIONS B 3.9.1 Refueling Equipment Interlocks BASES BACKGROUND Refueling equipment interlocks restrict the operation of the refueling equipment or the withdrawal of control rods to reinforce unit procedures that prevent the reactor from achieving criticality during refueling. The refueling interlock circuitry senses the conditions of the refueling equipment and the control rods. Depending on the sensed conditions, interlocks are actuated to prevent the operation of the refueling equipment or the withdrawal of control rods.

I FINSERT 1~

g0 GDC 26 of 10 CF pt ppendix A, require ta ne of the two requiredl d

independent reatvy ontrol systems be Iaa o oding the reactor l k

core subcritical une od conios (e.1/Thcotrol rods, when Ad fully inserted, serve as the system capable of maintaining the reactor 0

subcritical in cold conditions during all fuel movement activities and accidents.

chanof instrumentationkprovided to sense the position of the refueling platformthe loading o te re ue inp arm ue ra eand the full insertion of all control rods. Additionally, inputs are provided for-j the loading of the refueling platform frame mounted hoist, the loading of the refueling platform monorail mounted hoist, the full retraction of the fuel grapple, and the loading.of the service platform hoist. With the reactor mode switch in the u

n o refuefjfposition, the indicated conditions are combined in logic circuits to determine if all restrictions on refueling equipment operations and control rod insertion are satisfied.

A control rod not at its full-in position interrupts power to the refueling equipment and prevents operating the equipment over the reactor core when loaded with a fuel assembly.

Conversely, the refueling equipment located over the core and loaded with fuel inserts a control rod withdrawal l block in the Contro d DriveSystem to prevent withdrawing a control The refueling platform has two mechanical switches that open before the platform or any of its hoists are physically located over the reactor vessel.

All refueling hoists have switches that open when the hoists are loaded with fuel.

BWR/4 STS B 3.9.1-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 16 of 157

, Volume 14, Rev. 0, Page 17 of 157 B 3.9.1 0

INSERT I USAR, Section 1.2.2 (Ref. 1), requires the reactor core to be 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.

Insert Page B 3.9.1-1, Volume 14, Rev. 0, Page 17 of 157

, Volume 14, Rev. 0, Page 18 of 157 Refueling Equipment Interlocks B 3.9.1 BASES BACKGROUND (continued)

-The refueling interlocks use these indications to prevent operation of the refueling equipment with fuel loaded over the core whenever any control rod is withdrawn, or to prevent control rod withdrawal whenever fuel loaded refueling equipment is over the core (Ref. 2).

The hoist switches open at a load lighter than the weight of a single fuel assembly in water.

APPLICABLE

$he refueling interlocks arelexplicitl assumed in th FSAR anasefor SAFETY Proper fthe control rod removal error dung refueling R. 3] and the fuel ANALYSES operation ofassembly insertion error during refuelin(R

.4.

41hdse analysesl fe-is-ievaluate the conspquenc-es of control rod withdraws dur ng refueling andl Ialso fuel assemj~y insertion with a control rod withdrawn.1 A prompt niticality will not resul reactivity excursion during refueling could potentially result in fuel failure Wth adequate SDM with subsequent release of radioactive material to the environment.

refueling Interlocks.

Criticality and, therefore, subsequent prompt reactivity excursions are prevented during the insertion of fuel, provided all control rods are fully inserted during the fuel insertion. The refueling interlocks accomplish this by preventing loading of fuel into the core with any control rod withdrawn or by preventing withdrawal of a rod from the core during fuel loading.

The refueling platform location switches activate at a point outside of the reactor core suchjtha~t con~sidering switch hysteresn maiumll platform mome toward the core at the timezw power loss with a fuel assembly loaded and a control rod withdrawn, the fuel is not over the core.

Refueling equipment interlocks satisfy Criterion 3 of ssociated withthreaor 10 CFR 50.36(c)(2)(ii).

e switch refuel position LCO To prevent criticality during refueling, the refueling interlocksensure that fuel assemblies are not loaded with any control rod withdrawn.

To prevent these conditions from developing, the all-rods-in, the refueling platform position, the refueling platform fuel grapple fuel loaded, the refueling platform trolley frame mounted hoist fuel loaded, the refueling platform monorail mounted hoist fuel loaded, the refueling platform fuel when the grapple fully retracted position, and the service platform hoist fuel loaded reactor mode inputs are required to be OPERABL These inputs are combined in switch Is In the refuel ogic circuits, which provide refueling equipment or control rod blocks to position prevent operations that could result in criticality during refueling operations.

BWR/4 STS B 3.9.1-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 18 of 157

, Volume 14, Rev. 0, Page 19 of 157 Refueling Equipment Interlocks B 3.9.1 BASES APPLICABILITY In MODE 5, a prompt reactivity excursion could cause fuel damage and subsequent release of radioactive material to the environment. The refueling equipment interlocks protect against prompt reactivity excursions during MODE 5. The interlocks are required to be OPERABLE during in-vessel fuel movement with refueling equipment associated with the interlocks.

0 In MODES 1, 2, 3, and 4, the reactor pressure vessel head is on, and CORE ALTERATIONS are not possible. Therefore, the refueling interlocks are not required to be OPERABLE in these MODES.

ACTIONS A.1, A.2.1, and A.2.2 With one or more of the required refueling equipment interlocks inoperable, the unit must be placed in a condition in which the LCO does NR 3

not appl. Therefore, Required Action A.1 requires that in-vessel fuel movement with the affected refueling equipments be immediately suspended. This action ensures that operations are not performed with equipment that would potentially not be blocked from unacceptable operations (e.g., loading fuel into a cell with a control rod withdrawn).

Suspension of in-vessel fuel movement shall not preclude completion of movement of a component to a safe position.

0 Alternatively, Required Actions A.2.1 and A.2.2 require a control rod withdrawal block to be insertec6 and all control rods to be subsequently (i

verified to be fully inserted. Required Action A.2.1 ensures no control rods can be withdrawn, because a block to control rod withdrawal is in INSERT 4 place. The withdrawal block utilized must ensure that if rod withdrawal is /

requested, the rod will not respond (i.e., it will remain inserted).;Required 2

o2 Operformed after placing the rod withdrawal block in effect, and provides a verification that all control rods are fully inserted. This verification that all control rods are fully inserted is in addition to the periodic verifications required by SR 3.9.3.1.,s (i

Ike Required Action A.1, Re uired Actions A.2.1 and A.2.2 ensure unacceptable operations are bI (e.g., loading fuel into a cell with the control rod withdrawn).

rhbe SURVEILLANCE SR 3.9.1.1 REQUIREMENTS Performance of a CHANNEL FUNCTIONAL TEST demonstrates each required refueling equipment interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of the 0

BWR/4 STS B 3.9.1-3 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 19 of 157

, Volume 14, Rev. 0, Page 20 of 157 B 3.9.1 0

INSERT 2 when the reactor mode switch is in the refuel position. The Interlocks are not required when the reactor mode switch is in the shutdown position since a control rod block (LCO 3.3.2.1, "Control Rod block Instrumentation") ensures control rod withdrawal cannot occur simultaneously with in-vessel fuel movements CD INSERT 3 or is not necessary. This can be performed by ensuring fuel assemblies are not moved in the reactor vessel or by ensuring that the control rods are inserted and cannot be withdrawn IN2 INSERT 4 This action can be accomplished rod withdrawal.

by inserting an electrical or hydraulic block to control Insert Page B 3.9.1-3, Volume 14, Rev. 0, Page 20 of 157

, Volume 14, Rev. 0, Page 21 of 157 Refueling Equipment Interlocks B 3.9.1 BASES SURVEILLANCE REQUIREMENTS (continued) other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested.

The 7 day Frequency is based on engineering judgment and is considered adequate in view of other indications of refueling interlocks and their associated input status that are available to unit operations personnel.

REFERENCES

1.

0 AR 0,>

A,, Section G

0O 00(D

3.

FSA9 /ection [11 13].

4-SA, Sectior[5.1. 4]l

{D BWR/4 STS B 3.9.1-4 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 21 of 157

, Volume 14, Rev. 0,. Page 22 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.1 BASES, REFUEL EQUIPMENT INTERLOCKS

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

4. Grammatical/typographical errors corrected.

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

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 22 of 157

, Volume 14, Rev. 0, Page 23 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 23 of 157

, Volume 14, Rev. 0, Page 24 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.1, REFUELING EQUIPMENT INTERLOCKS There are no specific NSHC discussions for this Specification.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 24 of 157

, Volume 14, Rev. 0, Page 25 of 157 ATTACHMENT 2 ITS 3.9.2, Refuel Position One-Rod-Out Interlock, Volume 14, Rev. 0, Page 25 of 157

, Volume 14, Rev. 0, Page 26 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 26 of 157

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, Volume 14, Rev. 0, Page 28 of 157 DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK 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).

This change is designated as administrative because it does not result in technical changes to the CTS.

A.2 CTS 3.10.A requires the reactor mode switch to be in the refuel position during core alterations and the refueling interlocks to be OPERABLE. ITS LCO 3.9.2 only requires the refueling "position one-rod-out" interlock to be OPERABLE.

This changes the CTS by splitting the requirement of the refueling interlocks into two Specifications. All other refueling interlocks with the reactor mode switch in the refuel position are covered in ITS 3.9.1.

The purpose of CTS 3.10.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. The requirements for the refueling equipment interlocks are retained in ITS 3.9.1 while ITS 3.9.2 includes the requirements for the refuel position one-rod-out interlock. This change is acceptable because all of the refueling interlocks covered in CTS 3.10.A have been incorporated in either ITS 3.9.1 or ITS 3.9.2. Therefore, this change is considered a presentation preference change only and, as such, is considered an administrative change. This change is designated as more restrictive, because it adds a new Surveillance Requirement to the CTS.

A.3 CTS 3.1 O.A states that the refueling interlocks are required to be operable "except as specified in specification 3.1 0.E." The ITS does not include this sentence. This changes the CTS by deleting this cross-reference to another Specification.

This change is acceptable because the requirements have not changed. The CTS reference to Specification 3.10.E is an "information only" statement that neither adds, eliminates, or modifies requirements. The ITS convention is to not include these types of statements. Any changes to the requirements of CTS 3.10.E will be discussed in the Discussion of Changes for ITS 3.10.2 and ITS 3.10.6. This change is designated as administrative because it does not result in a technical change to the CTS.

MORE RESTRICTIVE CHANGES M.1 CTS 3.1 0.A does not provide specific Actions for when the refueling equipment interlocks are inoperable. However, since the interlock must be OPERABLE during CORE ALTERATIONS, this implies that CORE ALTERATIONS must be suspended if the interlock is inoperable. ITS 3.9.2 ACTION A covers the condition when the refuel position one-rod-out interlock is inoperable and it requires the immediate suspension of control rod withdrawal and the immediate initiation of action to fully insert all insertable control rods in core cells containing Monticello Page 1 of 5, Volume 14, Rev. 0, Page 28 of 157

, Volume 14, Rev. 0, Page 29 of 157 DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK one or more fuel assemblies. This changes the CTS by adding specific Actions for when the refuel position one-rod-out interlock is not met.

The purpose of CTS 3.10.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. This change adds ITS 3.9.2 ACTION A that covers the condition when the refuel position one-rod-out interlock is inoperable and it requires the suspension of control rod withdrawal and the initiation of action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. The addition of the specific requirement to suspend control rod withdrawal is considered administrative since CTS 3.1 0.A is applicable during CORE ALTERATIONS. When the CORE ALTERATIONS are suspended the LCO is not required to be met since the unit will be outside the applicability of the Specification. However, the additional requirement to initiate action to fully insert all insertable control rods in core cells containing one or more fuel assemblies is considered more restrictive because the CTS does not explicitly require the control rods to be inserted since the suspension of the operations will place the unit outside the applicability of the Specification. This change is acceptable because with the refueling position one-rod-out interlock inoperable the refuel interlocks may not be capable of preventing more than one control rod from being withdrawn. This condition may lead to criticality. Control rod withdrawal must be immediately suspended, and action must be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies to assure the core does not become critical. Action must continue until all such control rods are fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted. This changes the CTS by adding specific Actions for when the refuel position one-rod-out interlock is not met.

M.2 CTS 4.10.A requires the refueling interlocks (in this case the refuel position one-rod-out interlock) to be functional tested "prior to any fuel handling, with the head off the reactor vessel" and at "weekly intervals thereafter." However, it does not state how soon "prior to" starting the above evolutions. ITS SR 3.9.2.2 requires a similar verification every 7 days. This changes the CTS by eliminating the specific requirement to functionally test the refuel position one-rod-out interlock "prior to any fuel handling, with the head off the reactor vessel," and replaces it with a requirement to perform the test 7 days prior to any fuel handling.

The purpose of CTS 4.10.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. This change is acceptable because ITS SR 3.0.1 states "SRs shall be met during the MODES or other specified conditions in the Applicability for the individual LCOs, unless otherwise stated in the SR." In addition, ITS SR 3.0.4 states "Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency." Therefore, under the ITS, the Surveillances must be met within the 7 day Surveillance Frequency prior to the initiation of in-vessel fuel movements. For CTS 4.1 0.A, the periodic Surveillance Frequency for verifying the refuel position one-rod-out interlock is OPERABLE is acceptable during the MODE of Applicability, and is also acceptable during the period prior to entering the MODE of Applicability, since the specific time prior to starting the above evolutions is not specified. This Monticello Page 2 of 5, Volume 14, Rev. 0, Page 29 of 157

, Volume 14, Rev. 0, Page 30 of 157 DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK change is designated as more restrictive since it now requires the test "7 days" prior to any fuel handling, with the head off the reactor vessel.

M.3 CTS 3.1 O.A requires the reactor mode switch to be locked in the refuel position, however there is no Surveillance Requirement to verify that it is locked in the refuel position. ITS SR 3.9.1.2 requires verification every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the reactor mode switch is locked in the refuel position. This changes the CTS by adding this new Surveillance.

The purpose of CTS 3.1Q.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. Proper functioning of the refuel position one-rod-out interlock requires the reactor mode switch to be in refuel.

During control rod withdrawal in MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. This change is acceptable because it provides an additional level of assurance that the refuel position one-rod-out interlock will be OPERABLE when required. By 'locking" the reactor mode switch in the proper position (i.e.,

removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation. This change is designated as more restrictive, because it adds a new Surveillance Requirement to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.1 (Category 2 - Relaxation of Applicability) CTS 3.1 O.A requires the reactor mode switch to be in the refuel position during "core alterations" and the refueling interlocks to be OPERABLE. ITS LCO 3.9.2 requires the refuel position one-rod-out interlock to be OPERABLE in MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn. This changes the CTS by only requiring the refuel position one-rod-out interlock to be OPERABLE during certain CORE ALTERATIONS (in MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn).

The purpose of CTS 3.10.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. This change is acceptable because the requirements continue to ensure that the structures, systems, and components are maintained in the MODES and other specified conditions assumed in the safety analyses and licensing basis. Currently all the refueling interlocks are required during all types of CORE ALTERATIONS. This change Monticello Page 3 of 5, Volume 14, Rev. 0, Page 30 of 157

, Volume 14, Rev. 0, Page 31 of 157 DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK only requires the refuel position one-rod-out interlock to be OPERABLE during MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn. Therefore, this interlock will only be required to be OPERABLE during MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn. This change is acceptable because the only time this interlock provides a safety function is during MODE 5 with the reactor mode switch in the refuel position and any 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 7-Relaxation Of Surveillance Frequency, Non-24 Month Type Change) CTS 4.1 0.A requires the refueling interlocks (in this case the refuel position one-rod-out interlock) to be functionally tested every 7 days. ITS SR 3.9.2.2 includes a Note that states the Surveillance is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn. This changes the CTS by allowing the test to be delayed up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn.

The purpose of CTS 4.1 0.A is to.ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. To properly perform a CHANNEL FUNCTIONAL TEST of the one-rod-out interlock without the use of jumpers, a control rod must be withdrawn. However, ITS SR 3.0.1 requires a Surveillance to be met within the specified Frequency while in the applicable MODE or condition.

This essentially ensures that the Applicability of the LCO is not entered with the Surveillance not current. If this specific Surveillance Requirement were not performed within the specified Frequency prior to entering the applicable MODE and condition, then as soon as the applicable MODE and condition are entered, this would result in the LCO not being met. The ACTION for ITS 3.9.2 requires immediate action to be taken to exit the Applicability of the LCO. Therefore, an allowance in ITS SR 3.9.2.2 is provided to enter the LCO's Applicability for a short time (1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) to provide adequate time to perform the required Surveillance.

The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time is considered acceptable because of the procedural controls on control rod withdrawals and indications in the control room to alert the operator of controls rods not fully inserted. This change is designated as less restrictive because Surveillances will be performed less frequently under the ITS than under the CTS.

L.3 (Category 5 - Deletion of Surveillance Requirement) CTS 4.10.A requires the refueling interlocks (in this case the refuel one-rod-out interlock) to be functionally tested "following any repair work associated with the interlocks." ITS SR 3.9.2.2 does not require this verification 'following any repair work associated with the interlocks." This changes the CTS by eliminating the requirement to functionally test the refueling equipment interlocks "following any repair work associated with the interlocks."

The purpose of CTS 4.10.A is to ensure the refueling interlocks are OPERABLE to help prevent criticality during refueling. 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 continues to be tested In a manner and at a Frequency necessary to give confidence that the equipment can perform its assumed safety function. Any time the OPERABILITY of a system or component has been Monticello Page 4 of 5, Volume 14, Rev. 0, Page 31 of 157

, Volume 14, Rev. 0, Page 32 of 157 DISCUSSION OF CHANGES ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK affected by repair, maintenance, modification, or replacement of a component, post-maintenance testing is required to demonstrate the OPERABILITY of the system or component. This is described in the Bases for ITS SR 3.0.1 and required under ITS SR 3.0.1. In addition, the requirements of 10 CFR 50, Appendix B, Section Xi (Test Control) provide adequate controls for test programs to ensure that testing incorporates applicable acceptance criteria.

Compliance with 10 CFR 50, Appendix B, is required under the unit operating license. As a result, post-maintenance testing will continue to be performed and an explicit requirement in the Technical Specifications is not necessary. This change is designated as less restrictive because Surveillances which are required in the CTS will not be required in the ITS.

Monticello Page 5 of 5, Volume 14, Rev. 0, Page 32 of 157

, Volume 14, Rev. 0, Page 33 of 157 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 33 of 157

, Volume 14, Rev. 0, Page 34 of 157 Refuel Position One-Rod-Out Interlock 3.9.2 OTrs 3.9 REFUELING OPERATIONS 3.10A 3.9.2 Refuel Position One-Rod-Out Interlock 3.10A LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.

3.10A APPLICABILITY:

MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.

ACTIONS DOC M.1 3.10.f DOC M.2 CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one-rod-A.1 Suspend control rod Immediately out interlock inoperable.

withdrawal.

AND A.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY A,SR 3.9.2.1 Verify reactor mode switch locked ingefuel position.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 0D 4.10A SR 3.9.2.2 Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn.

Perform CHANNEL FUNCTIONAL TEST.

7 days BWR14 STS 3.9.2-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 34 of 157

, Volume 14, Rev. 0, Page 35 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK

1. Editorial change made to be consistent with the LCO.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 35 of 157

, Volume 14, Rev. 0, Page 36 of 157 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 36 of 157

, Volume 14, Rev. 0, Page 37 of 157 Refuel Position One-Rod-Out Interlock B 3.9.2 B 3.9 REFUELING OPERATIONS B 3.9.2 Refuel Position One-Rod-Out Interlock BASES BACKGROUND The refuel position one-rod-out interlock restricts the movement of control rods to reinforce unit procedures that prevent the reactor from becoming critical during refueling operations. During refueling operations, no more than one control rod is permitted to be withdrawn.

INET1 GDC 26 of 1 0 CFR 5p, Appendix A, requires thatoeof the two requiredlAi independent reactivyty control systems be capal ofhlg the reactor_lx core subcritical under cold conditions (Ref.

rods serve as the system capable of maintaining the reactor subcritical in cold conditions.

The refuel position one-rod-out interlock prevents the selection of a second control rod for movement when any other control rod is not fully inserted (Ref. 2). It is a logic circuit that has redundant channels. It uses the all-rods-in signal (from the control rod full-in position indicators discussed in LCO 3.9.4, "Control Rod Position Indication") and a rod selection signal (from the Reactor Manual Control System).

contrl rod withdrawal t[error during refueling This Specification ensures that the performance of the refuel position one-rod-out interlock in the event of arDesign Basis ccident meets the Iassumptions used in tWe safety an-alysis of Refere ce 3!

APPLICABLE

1. che refuemjpfosition one-rod-out interlock is explicitly assume in the SAFETY Proper0 ana sisfor the control rod withdrawal error during refuelingt ANALYSES~prto OlRf

.l his ana ysis evaluates the consequences/of controlErod 1 3 0 t 3 iil aldurn rfeIn prompt reactivity excursion during ne contrl d refueling could potentially result in fuel failure with subsequent release of refueling will not result radioactive material to the environment.

in criticality.

The refuel position one-rod-out interlock and adequate SDM (LCO 3.1.1,

.SHUTDOWN MARGIN (SDM)rprevent criticality by preventing m withdrawal of more than one control rod. With one control rod withdrawn, the core will remain subcritical, thereby preventing any prompt critical excursion.

The refuel position one-rod-out interlock satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

BWR/4 STS B 3.9.2-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 37 of 157

, Volume 14, Rev. 0, Page 38 of 157 B 3.9.2 0

INSERT 1 USAR, Section 1.2.2 (Ref. 1), requires the reactor core to be 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.

Insert Page B 3.9.2-1, Volume 14, Rev. 0, Page 38 of 157

, Volume 14, Rev. 0, Page 39 of 157 Refuel Position One-Rod-Out Interlock B 3.9.2 BASES LCO To prevent criticality during MODE 5, the refuel position one-rod-out interlock ensures no more than one control rod may be withdrawn. Both channels of the refuel position one-rod-out interlock are required to be OPERABLE and the reactor mode switch must be locked in the refuel position to support the OPERABILITY of these channels.

APPLICABILITY In MODE 5, with the reactor mode switch in the refuel position, the OPERABLE refuel position one-rod-out interlock provides protection against prompt reactivity excursions.

I irorProtection Syvstem (RPS) ntnumentati6n' In MODES 1, 2, 3, and 4, the refuel position one-rod-out interlock is not required to be OPERABLE and is bypassed. In MODES 1 and 2, the

• Contro Rod Reactor Protection System (LCO 3.3.1.1t and the control rods IOPERABILITLCO 3.1.t provide mitigation of potential reactivity excursions. In In MODES 3 n Ath the reactor mode switch in the shutdown position, a contro rod block (LCO 3.3.2.

ensures all control rods are inserted J thereby preventing criticality dns.

CIockj 0

ACTIONS A.1 and A.2 Withlone or flanneis otlthe refueEjg position one-rod-out interlock D (ID inoperable, the refueling interlocks may not be capable of preventing more than one control rod from being withdrawn. This condition may lead to criticality.

Control rod withdrawal must be immediately suspended, and action must be immediately initiated to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Action must continue until all such control rods are fully inserted. Control rods in core cells containing no fuel assemblies do not affect the reactivity of the core and, therefore, do not have to be inserted.

SURVEILLANCE REQUIREMENTS SR 3.9.2.1 Proper functioning of the refuefWg position one-rod-out interlock requires the reactor mode switch to be in Fefuel. During control rod withdrawal in(3)

MODE 5, improper positioning of the reactor mode switch could, in some instances, allow improper bypassing of required interlocks. Therefore, this Surveillance imposes an additional level of assurance that the refuel position one-rod-out interlock will be OPERABLE when required.

By "locking" the reactor mode switch in the proper position (i.e., removing the reactor mode switch key from the console while the reactor mode switch is positioned in refuel), an additional administrative control is in place to preclude operator errors from resulting in unanalyzed operation.

0 0

BWR/4 STS B 3.9.2-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 39 of 157

, Volume 14, Rev. 0, Page 40 of 157 Refuel Position One-Rod-Out Interlock B 3.9.2 BASES SURVEILLANCE REQUIREMENTS (continued)

The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other administrative controls utilized during refueling operations to ensure safe operation.

SR 3.9.2.2 Performance of a CHANNEL FUNCTIONAL TEST on each channel demonstrates the associated refuel position one-rod-out interlock will function properly when a simulated or actual signal indicative of a required condition is injected into the logic. A successful test of the required contact(s) of a channel relay may be performed by the verification of the change of state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay.

This is acceptable because all of the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specifications tests at least once per refueling interval with applicable extensions. The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps so that the entire channel is tested. The 7 day Frequency is considered adequate because of demonstrated circuit reliability, procedural controls on control rod withdrawals, and visual and audible indications available in the control room to alert the operator to control rods not fully inserted. To perform the required testing, the applicable condition must be entered (i.e., a control rod must be withdrawn from its full-in position). Therefore, SR 3.9.2.2 has been modified by a Note that states the CHANNEL FUNCTIONAL TEST is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn.

REFERENCES

1. IO0 CFR 5O, 1x A, GDC2 U

Er-ip 0D 0 0 1

3.

YSAR, Section [1.4.1.1]4 BWR/4 STS B 3.9.2-3 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 40 of 157

, Volume 14, Rev. 0, Page 41 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.2 BASES, REFUEL POSITION ONE-ROD-OUT INTERLOCK

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 the Writer's Guide or similar statements in other places in the Bases.

3. Typographical/grammatical error corrected.

4. Changes have been made to be consistent with the requirements in the Specification.

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

J Monticello Page 1 of 1, Volume 14, Rev. 0, Page 41 of 157

, Volume 14, Rev. 0, Page 42 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 42 of 157

, Volume 14, Rev. 0, Page 43 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.2, REFUEL POSITION ONE-ROD-OUT INTERLOCK There are no specific NSHC discussions for this Specification.

Monticello Page 1 of I, Volume 14, Rev. 0, Page 43 of 157

, Volume 14, Rev. 0, Page 44 of 157 ATTACHMENT 3 ITS 3.9.3, Control Rod Position, Volume 14, Rev. 0, Page 44 of 157

, Volume 14, Rev. 0, Page 45 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 45 of 157

, Volume 14, Rev. 0, Page 46 of 157 ITS 3.9.3

-4 i Add proposed ITS 3.9.3 Page 1 of I, Volume 14, Rev. 0, Page 46 of 157

, Volume 14, Rev. 0, Page 47 of 157 DISCUSSION OF CHANGES ITS 3.9.3, CONTROL ROD POSITION ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not specify any requirements for control rod position when loading fuel assemblies into the core. ITS LCO 3.9.3, "Control Rod Position," requires all control rods to be fully inserted. An appropriate ACTION and a Surveillance Requirement are also provided. This changes the CTS by incorporating the requirements of ITS 3.9.3.

Control rods provide one of the two required independent reactivity control systems capable of maintaining the reactor subcritical under cold conditions. To minimize the probability of an inadvertent criticality during refueling, all control rods must be fully inserted during applicable refueling conditions. This change is acceptable since the requirement for Control Rod Position satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). This change is designated as more restrictive because it adds new requirements to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1, Volume 14, Rev. 0, Page 47 of 157

, Volume 14, Rev. 0, Page 48 of 157 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 48 of 157

, Volume 14, Rev. 0, Page 49 of 157 Control Rod Position 3.9.3 J

CTS 3.9 REFUELING OPERATIONS 3.9.3 Control Rod Position Doc M.1 LCO 3.9.3 All control rods shall be fully inserted.

APPLICABILITY:

ACTIONS When loading fuel assemblies into the core.

CONDITION 1

REQUIRED ACTION COMPLETION TIME DOC M.1 A. One or more control rods not fully inserted.

A.1 Suspend loading fuel assemblies into the core.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify all control rods are fully inserted.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> DOC M.1 BWR/4 STS 3.9.3-1, Volume 14, Rev. 0, Page 49 of 157 Rev. 3.0, 03/31/04

, Volume 14, Rev. 0- Page 50 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3, CONTROL ROD POSITION There are no deviations from NUREG-1433, Rev. 3 for this Specification.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 50 of 157

, Volume 14, Rev. 0, Page 51 of 157 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 51 of 157

, Volume 14, Rev. 0, Page 52 of 157 Control Rod Position B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Control Rod Position BASES BACKGROUND Control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the Control Rod Drive System.

EquipmentDung refuelinr ment of control rods is limited by the refueling interlocks (LCO 3.9. 1and LCO 3.9.Z) or the control rod block with the

[,Refuel Posito One

[Rod-Out Interloci reactor mode switch in the shutdown position (LCO 3.3.2.1).

EVE}-GDC 26of 1 zFR 50, Appendix A, reurs that one of th euredl lnl:E.pendef reactivity control sytsb aal fhofi h ecorl core su crtical under cold conYos(e.1) /Tecoto rods serve as the system capable of maintaining the reactor subcritical in cold conditions.

0D 0D 0

The refueling interlocks allow a single control rod to be withdrawn at any time unless fuel is being loaded into the core. To preclude loading fuel assemblies into the core with a control rod withdrawn, all control rods must be fully inserted. This prevents the reactor from achieving criticality during refueling operations.

APPLICABLE Prevention and mitigation of prompt reactivity excursions during refueling SAFETY are Provided b the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the ANALYSES SDM (LCO 3.1.15, the intermediate range monitor neutron flux scram

,SHUTDOWN (LCO 3.3.1.,

the average po~er range monitor nephron flux scram I

./

l(LC) 3.3..1)Jand the control rod block instrumentation (LCO 3.3.2.1).

\\_y, '

Rer PS) ovent relies on the proper operation l

A Instrunentaion' The safebi-a rvsis for thel control rod withdrawal error during refueling:

the FSAR (Re.2 assumes thefunc ionind of the refueling interlocks and adequate SDM.7 he nalvsis fithelfuel assembly insertion error(

eliesosall control rodfully inserted. Thus, prior to fuel reload, all control rods must be fully inserted to minimize the probability of a J

inadvertent criticality.

LIeven 0D 0

Control rod position satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO All control rods must be fully inserted during applicable refueling conditions to minimize the probability of an inadvertent criticality during refueling.

BWR14 STS

-B 3.9.3-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 52 of 157

, Volume 14, Rev. 0, Page 53 of 157 B 3.9.3 INSERT I USAR, Section 1.2.2 (Ref. 1), requires the reactor core to be 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.

Insert Page B 3.9.3-1, Volume 14, Rev. 0, Page 53 of 157

, Volume 14, Rev. 0, Page 54 of 157 Control Rod Position B 3.9.3 BASES APPLICABILITY During MODE 5, loading fuel into core cells with control rods withdrawn may result in inadvertent criticality. Therefore, the control rods must be inserted before loading fuel into a core cell. All control rods must be inserted before loading fuel to ensure that a fuel loading error does not result in loading fuel into a core cell with the control rod withdrawn.

In MODES 1, 2, 3, and 4, the reactor pressure vessel head is on, and no fuel loading activities are possible. Therefore, this Specification is not applicable in these MODES.

ACTIONS A.1 With all control rods not fully inserted during the applicable conditions, an inadvertent criticality could occur that is not analyzed in theMSAR. All fuel loading operations must be immediately suspended. Suspension of these activities shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS During refueling, to ensure that the reactor remains subcritical, all control rods must be fully inserted prior to and during fuel loading. Periodic checks of the control rod position ensure this condition is maintained.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Frequency takes into consideration the procedural controls on control rod movement during refueling as well as the redundant functions of the refueling interlocks.

0 0

REFERENCES

1. 110CFR50, ixA, GDC2 2.

3.

7AR Section ASAR, Section

[15 1.13].

[1

.14].

0 BWR/4 STS B 3.9.3-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 54 of 157

, Volume 14, Rev. 0, Page 55 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.3 BASES, CONTROL ROD POSITION

1. Editorial changes made to be consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 3.3.2.

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

3. The APRM neutron flux scram is not required to be OPERABLE while in MODE 5, therefore reference to it has been deleted.

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

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 55 of 157

, Volume 14, Rev. 0, Page 56 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 56 of 157

, Volume 14, Rev. 0, Page 57 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.3, CONTROL ROD POSITION There are no specific NSHC discussions for this Specification.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 57 of 157

, Volume 14, Rev. 0, Page 58 of 157 ATTACHMENT 4 ITS 3.9.4, Control Rod Position Indication, Volume 14, Rev. 0, Page 58 of 157

, Volume 14, Rev. 0, Page 59 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 59 of 157

, Volume 14, Rev. 0, Page 60 of 157 ITS 3.9.4 4 -

AdpoTS3.9.4 Page 1 of I, Volume 14, Rev. 0, Page 60 of 157

, Volume 14, Rev. 0, Page 61 of 157 DISCUSSION OF CHANGES ITS 3.9.4, CONTROL ROD POSITION INDICATION ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not specify any requirements for control rod position indication.

ITS LCO 3.9.4, "Control Rod Position Indication," requires the "full-in" position indication channel for each control rod to be OPERABLE. An appropriate ACTION and a Surveillance Requirement are also provided. This changes the CTS by incorporating the requirements of ITS 3.9.4.

The control rod "full-in" position indication channel provides the required input to the refueling interlocks to ensure that fuel cannot be loaded with any control rod withdrawn and that no more than one control rod can be withdrawn at a time.

This change is acceptable since the requirement for Control Rod Position Indication satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii). This change is designated as more restrictive because it adds new requirements to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1, Volume 14, Rev. 0, Page 61 of 157

, Volume 14, Rev. 0, Page 62 of 157 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 62 of 157

, Volume 14, Rev. 0, Page 63 of 157 Control Rod Position Indication 3.9.4 K-_)

CTS 3.9 REFUELING OPERATIONS 3.9.4 Control Rod Position Indication DOC M.1 LCO 3.9.4 APPLICABILITY:

ACTIONS The control rod 'full-in" position indication channel for each control rod shall be OPERABLE.

MODE 5.

Separate Condition entry is allowed for each re re channel.

0D DOC M.1 CONDITION REQUIRED ACTION COMPLETION TIME A. One or more re re A.1.1 Suspend in vessel fuel Immediately control rod position movement.

indication channels inoperable.

AND A.1.2 Suspend control rod Immediately withdrawal.

AND A.1.3 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.

OR A.2.1 Initiate action to fully insert Immediately the control rod associated with the inoperable position indicator.

AND 0

BWR/4 STS 3.9.4-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 63 of 157

, Volume 14, Rev. 0, Page 64 of 157 Control Rod Position Indication 3.9.4 DOC M.1 DOC M.1 ACTIONS (continued)

I CONDITION REQUIRED ACTION COMPLETION TIME A.2.2 Initiate action to disarm the Immediately control rod drive associated with the fully inserted control rod.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify the reired channel has no "full-in" indication Each time the on each control rod that is not "full-in."

control rod is withdrawn from the "full-in" position 0

BWR/4 STS 3.9.4-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 64 of 157

, Volume 14, Rev. 0, Page 65 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4, CONTROL ROD POSITION INDICATION

1. Since the Monticello design includes only one "full-in" position indicator for each control rod, all "full-in" channels are required. Therefore, the word "required" has been deleted from the ACTIONS Note, Condition A and the Surveillance.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 65 of 157

, Volume 14, Rev. 0, Page 66 of 157 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 66 of 157

, Volume 14, Rev. 0, Page 67 of 157 Control Rod Position Indication B 3.9.4 B 3.9 REFUELING OPERATIONS B 3.9.4 Control Rod Position Indication BASES BACKGROUND

'Refuellng Equipment

[interocks'

, 'Refuel Position One-

[Rodut Interlock' l indi catibo n- -

2E -- i The full-in position indication channel for each control rod provides necessary information to the refueling interlocks to prevent inadvertent criticalities during refueling operations. During refueling, the refueling interlocks (LCO 3.9.1 and LCO 3.9.2) use the full-in position indication J channel to limit the operation of the refueling equipment and the FmINSERT 1 movement of the control rods.iThe absence of the full-in positiorichanne signal for any control rod removes the all-rods-in permissive for the refueling equipment interlocks and prevents fuel loading. Also, this condition causes the refuel position one-rod-out interlock to not allow the withdrawal of any other control rod.

.GDC 26 of 1,VCFR 50, Appendix A, r quires that one of thebiequiredI independon reacti~vity control systems be capable of bkinqg the reactorI core subcrtical under cold conditons (Ref. 1!W The control rods serve as the system capable of maintaining the reactor subcritical in cold conditions.

0 APPLICABLE SAFETY Prevention and mitigation of prompt reactivity excursions during refueling are provided by the refueling interlocks (LCO 3.9.1 and LCO 3.9.2), the SDM (LCO 3.1.1tj, the intermediate range monitor neutron flux scram

'(LCO 3.3.1.1'), and the control rod block instrumentation (LCO 3.3.2.1).

A l

Control Rod Block Instrumentation: r

[The safet sis for thel control rod withdrawal error during refueling e (Ref. 2) assumee functionina of the refueling interlocks and adequate SDM. I or the fuel assembly insertion error

e.

sumes

• all control rods fully inserted The full-in position indication channel is required to be OPERABLE so that the refueling interlocks can ensure that fuel cannot be loaded with any control rod withdrawn and that no more than one control rod can be withdrawn at a time.

3 Control rod position indication satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Eib h

control rod full-in position indication channelmust be OPERABLE to provide the required input to the refueling interlocks. A channel is OPERABLE if it provides correct position indication to the refueling interlock logic.

0D BWR/4 STS B 3.9.4-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 67 of 157

, Volume 14, Rev. 0, Page 68 of 157 0

B 3.9.4 INSERT I Two full-in position indication switches (S51 and S52) provide input to the all-rods-in logic for each control rod. Switch S51 provides full core display beyond full-in (scram) position indication (double dashes - no number) and switch S52 provides full core display normal green full-in position indication. Switch S52 is set slightly beyond switch SOO, which provides the digital "00" full-in position readout (switch SOO does not provide input to the all-rods-in logic and is not considered a full-in channel). When switch S52 is actuated, the color of the full core display "00" readout is changed from amber to green, indicating the control rod is full-in and latched. Switches S51 and S52 are wired in parallel, such that, if either switch indicates full-in, the all-rods-in logic will receive a full-in signal for that control rod. Therefore, each control rod is considered to have only one "full-in" position indication channel.

0 INSERT 2 USAR, Section 1.2.2 (Ref. 1), requires the reactor core to be 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.

Insert Page B 3.9.4-1, Volume 14, Rev. 0, Page 68 of 157

, Volume 14, Rev. 0, Page 69 of 157 Control Rod Position Indication B 3.9.4 BASES APPLICABILITY During MODE 5, the control rods must have OPERABLE full-in position indication channels to ensure the applicable refueling interlocks will be OPERABLE.

In MODES I and 2, requirements for control rod position are specified in LCO 3.1.3, "Control Rod OPERABILITY." In MODES 3 and 4, with the reactor mode switch in the shutdown position, a control rod block (LCO 3.3.2.1) ensures all control rods are inserted, thereby preventing criticality during shutdown conditions.

ACTIONS A Note has been provided to modify the ACTIONS related to control rod position indication channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable control rod position indication channels provide appropriate compensatory measures for separate inoperable channels. As such, this Note has been Provided, which allows separate Condition entry for each inoperable ed control rod position indication channel.

A.1.J. A.1.2, A.1.3, A.2.14and A.2.2(i With one or more re re full-in position indication channels inoperable, compensating actions must be taken to protect against potential reactivity excursions from fuel assembly insertions or control rod withdrawals. This may be accomplished by immediately suspending in-vessel fuel movement and control rod withdrawal, and immediately initiating action to fully insert all insertable control rods in core cells containing one or more fuel assemblies. Actions must continue until all insertable control rods in core cells containing one or more fuel assemblies are fully inserted.

Suspension of in-vessel fuel movements and control rod withdrawal shall not preclude moving a component to a safe position.

Alternatively, actions must be immediately initiated to fully insert the control rod(s) associated with the inoperable full-in position indicator(s) and disarm the drive(s) to ensure that the control rod is not withdrawn.

RI~

Actions must continue until all associated control rods are fully inserted and drives are disarmed Under these conditions (control rod fully (9

inserted and disarmed), an inoperable full-in channel may be bypassed to allow refueling operations to proceed. An alternate method must be used to ensure the control rod is fully inserted (e.g., use the "00" notch position indication).

BWR/4 STS B 3.9.4-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 69 of 157

, Volume 14, Rev. 0, Page 70 of 157 B 3.9.4 0

INSERT 3 electrically or hydraulically. The control rods can be electrically disarmed by disconnecting power from all four directional control valve solenoids. The control rods can be hydraulically disarmed by closing the drive water and exhaust water isolation valves.

Insert Page B 3.9.4-2, Volume 14, Rev. 0, Page 70 of 157

, Volume 14, Rev. 0, Page 71 of 157 Control Rod Position Indication B 3.9.4 BASES SURVEILLANCE REQUIREMENTS SR 3.9.4.1 The full-in position indication channels provide input to the one-rod-out interlock and other refueling interlocks that require an all-rods-in permissive. The interlocks are actuated when the full-in position indication for any control rod is not present, since this indicates that all rods are not fully inserted. Therefore, testing of the full-in position indication channels is performed to ensure that when a control rod is withdrawn, the full-in position indication is not present. Note that failure to indicate full-in when the control rod is not withdrawn results in conservative actuation of the one-rod-out interlock, and therefore, is not explicitly required to be verified by this SR. The full-in position indication channel is considered inoperable even with the control rod fully inserted, if it would continue to indicate full-in with the control rod withdrawn.

Performing the SR each time a control rod is withdrawn is considered adequate because of the procedural controls on control rod withdrawals and the visual and audible indications available in the control room to alert the operator to control rods not fully inserted.

REFERENCES

1..10CFR590

.JxAGDC SRSColZ 0D

2.

5tSAR, Section [1 1.13].

3.

FSAR, Section [ 5.1.14].

D BWR/4 STS B 3.9.4-3 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 71 of 157

, Volume 14, Rev. 0, Page 72 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.4 BASES, CONTROL ROD POSITION INDICATION

1. Editorial changes made to be consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 3.3.2.

2. Changes have been made to more closely reflect the Specification.

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

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

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

6. Typographical error corrected.

7. Changes have been made to be consistent with similar words in other Specifications.

Monticello Page 1 of I, Volume 14, Rev. 0, Page 72 of 157

, Volume 14, Rev. 0, Page 73 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 73 of 157

, Volume 14, Rev. 0, Page 74 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.4, CONTROL ROD POSITION INDICATION There are no specific NSHC discussions for this Specification.

Monticello Page 1 of I, Volume 14, Rev. 0, Page 74 of 157

, Volume 14, Rev. 0, Page 75 of 157 ATTACHMENT 5 ITS 3.9.5, Control Rod OPERABILITY - Refueling, Volume 14, Rev. 0, Page 75 of 157

, Volume 14, Rev. 0, Page 76 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 76 of 157

C CI ITS 3.9.5 0

ITS ITS 3.0 UMmNG CONDmONS FOR OPERATION 4.f0 SURVEILLANCE REOUIRFMENTS I

Su 0

CD 0

(I)

-Co CD

-4'

-4 Control Sod Accumulators Once check the status in the copM ro of the required Operable accumulator pressurea4 e

i5 > 940 pSi9 IaI G

LCO 3.9.5 Applicability _

LCO 3.9.5 1 ACTION A

{ See ITS LI If a control rod with an Inoperable accumulator is

_ inserted *tull-iWand either its directonal control valves

/

are _e y

or it I hydrau _ctVsolat It I

snall not 0e constoerea to nave an Inoperable accumulator.

I

1.

In the Startup or Run Mode, a rod accumulator may be inoperable provided that no other control rod within two control rod cells In any direction has a:

(a) Inoperable accumulator, or (b) Directional control valve electrically disarmed while In a non-fully inserted position.

NSee ITS 3.1.51 I

Add proposed SR 3.9.5.1 M) a)

CD 0

CD (0)

ID 0)

-4'

-4 Add proposed ACTION A for when control rod scram Insertion capability Is not rnet 3.314.3 82 10/26101 Amendment No. 5r1, 1$3,054,-63,4014, 123 Page 1 of 3

(7-C C

ITS 3.9.5 0

3.0 UMITING CONDITIONS FOR OPERATION J4.0 SURVEILLANCE REQUIREMENTS l 2. lnjge Refuel Modea rod accumulaj& may be I

iv 0

CD 3

CD 0

0)

Co CD 4:1 0

to

-9, U'

I Jroioperable pr od:

/

I-f (a) All fuel Is removed from the cell containing the I

essodated control rod, or 1-See ITS 3.10.6}

(b/Tha one -rod refuel interi or the

/ assodated od drhve Is operapfe.

j h

(

M33 0)

El CD 0

0 0)

CD CD

-9

U

.4 CX

. O 3.3/43 882a 4/18189 Amendment No. 63 Page 2 of 3

C C

C ITS 3.9.5 0

03 C,

0 F

03 CD

-4

-4' lu

-4 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 shall volume vent and drain valves shall be operable, be cycled quarterly.

except as specified below.

Once per operating cycle verify the scram discharge

.voume vent arnd drain valves dose within 30 seconds

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 shall be maintained by elther

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

or

b.

Maintaining the Inoperable valve(s), or the associated redundant valve(s), In the closed position. Periodically the Inoperable and the 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 all but one operable control rods (not Including rods removed per specification 3.10.E or Inoperable rods allowed by 3.3A2) shal be fully Insetted within ten hours.

I See ITS 3.1.8}

f..

03 CD

-A 0

0 2

CD 03

-.1

'-N G. RI if

.3 roug not met, an o

rly shutd St be Initiated an ave reactor In I

/

I e cold shutdown ditlon within 2 ours.

i nvAn mvMM IQ In the r%

ceptwhen F(.-
Pt"-Te mode switch Is In t efuel poslidon) 83a 5t1184 Amendment No. 24 Page 3 of 3

, Volume 14, Rev. 0, Page 80 of 157 DISCUSSION OF CHANGES ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING 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 an inoperable control rod is inserted full in, it shall not be considered to have an inoperable accumulator. ITS LCO 3.9.5 states "Each withdrawn control rod shall be OPERABLE." ITS 3.9.5 ACTION A requires action to be initiated immediately to fully insert any inoperable control rods. This changes the CTS by restating the existing control rod OPERABILITY requirement and specifying the implied action required to exit the OPERABILITY requirement.

The purpose of CTS 3.3.D, in part, is to establish the OPERABILITY requirements for control rod scram accumulators and the conditions under which the inoperability may be accepted. Similarly, ITS LCO 3.9.5 requires each withdrawn control rod to be OPERABLE. This requirement includes the control rod accumulators specified by CTS and also includes the control rod insertion capability added by DOC M.1. ITS 3.9.5 ACTION A, to immediately initiate action to fully insert any inoperable control rods, provides definitive actions consistent with the exceptions provided in CTS 3.3.D. This change is designated as administrative because it does not result in technical changes to the CTS.

A.3 CTS 3.3.G.1 states, in part, if Specification 3.3.D is not met the unit must be in cold shutdown in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. ITS LCO 3.9.5 is applicable only in MODE 5. This changes the CTS by deleting the reference to unit shutdown requirements associated with an inoperable control rod accumulator in Refuel Mode.

CTS 3.3.G.1 provides the required actions necessary to ensure the unit has exited the LCO. Since ITS LCO 3.9.5 is only applicable in MODE 5, ITS 3.9.5 ACTION A (DOC A.2) to initiate action to fully insert inoperable withdrawn control rods provides the necessary actions under these conditions. This change is designated as administrative because it does not result in technical changes to the CTS.

A.4 This change to CTS 3.3.G is 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, this change is administrative.

MORE RESTRICTIVE CHANGES M.1 CTS 3.3.D requires each control rod accumulator to be operable. ITS LCO 3.9.5 requires each withdrawn control rod accumulator to be OPERABLE and has Monticello Page 1 of 4, Volume 14, Rev. 0, Page 80 of 157

, Volume 14, Rev. 0, Page 81 of 157 DISCUSSION OF CHANGES ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING added the requirement that each withdrawn control rod must be capable of insertion upon receipt of a scram signal. ITS 3.9.5 ACTION A has been added to provide proper actions when the insertion capability is not met. ITS SR 3.9.5.1 has been added to insert each withdrawn control rod at least one notch every 7 days. This changes the CTS by adding an OPERABILITY requirement for control rod insertion capability and a subsequent Surveillance Requirement to demonstrate this insertion capability.

The purpose of CTS 3.3.D is to ensure control rod accumulators are capable of providing adequate operating pressure to insert the control rods upon receipt of a scram signal. Similarly, ITS LCO 3.9.5 retains the control rod accumulator pressure requirement but also requires the control rods be capable of being automatically inserted upon receipt of a scram signal and ITS 3.9.5 ACTION A provides the proper actions for when this is not met. ITS SR 3.9.5.1 verifies the ability of the withdrawn control rod to be inserted. This change is designated as more restrictive because it adds an OPERABILITY requirement, and ACTION, and a Surveillance Requirement not required in the CTS.

M.2 CTS 4.3.D requires a check of the accumulator pressure alarm located in the control room. ITS SR 3.9.5.2 requires 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, associated with a withdrawn control rod, is OPERABLE. ITS SR 3.9.5.2 includes acceptance criteria for accumulator pressure (2 940 psig) consistent with current Monticello practice, and requires verification that each accumulator associated with a withdrawn control rod meets this pressure criterion. Although this change is consistent with current practice, adding this acceptance criterion and verification requirement in ITS SR 3.9.5.2 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.

M.3 CTS 3.3.D.2 allows a control rod accumulator to be inoperable if the one-rod-out interlock for the associated control rod is operable. ITS 3.9.5 does not provide this allowance. This changes the CTS by deleting a control rod accumulator inoperability allowance.

The purpose of CTS 3.3.D.2 is to provide an allowance for special conditions that would preclude the requirement for a control rod accumulator to be operable.

ITS 3.9.5 does not provide this allowance. The requirements for control rods to be positioned, and therefore OPERABLE, per ITS 3.9.5 are addressed in associated refueling Specifications ITS 3.9.1, "Refueling Equipment Interlocks,"

ITS 3.9.2, "Refuel Position One-Rod-Out Interlock," ITS 3.9.3, "Control Rod Position," and ITS 3.9.4, "Control Rod Position Indication." This change is designated as more restrictive because it removes a control rod accumulator inoperability allowance that appears in the CTS.

Monticello Page 2 of 4, Volume 14, Rev. 0, Page 81 of 157

, Volume 14, Rev. 0, Page 82 of 157 DISCUSSION OF CHANGES ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES L.I (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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. ITS SR 3.9.5.2 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to every 7 days.

The purpose of CTS 4.3.D is to ensure the control rod scram accumulators associated with withdrawn control rods 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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.D 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.D is to ensure each control rod scram accumulator associated with a withdrawn control rod 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 3.9.5.2 requires verification that the accumulator pressure is within the pressure limit for each accumulator associated with a withdrawn control rod. The actual limit has been added as described in DOC M.2. This change deletes the requirement to verify OPERABILITY of the control rod accumulators via the accumulator level alarm in the control room.

The ISTS does 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 Monticello Page 3 of 4, Volume 14, Rev. 0, Page 82 of 157

, Volume 14, Rev. 0, Page 83 of 157 DISCUSSION OF CHANGES ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING policies. The requirement to verify control rod scram accumulator pressure (which does relate directly to accumulator OPERABILITY) is within limits is still maintained in ITS SR 3.9.5.2. 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.

Monticello Page 4 of 4, Volume 14, Rev. 0, Page 83 of 157.

, Volume 14, Rev. 0, Page 84 of 157 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 84 of 157

, Volume 14, Rev. 0, Page 85 of 157 Control Rod OPERABILITY - Refueling 3.9.5 3.9 REFUELING OPERATIONS 3.9.5 Control Rod OPERABILITY - Refueling LCO 3.9.5 Each withdrawn control rod shall be OPERABLE.

3.3.D.

DOC M.1 3.3.D 3.3.D.

DOC M.1 DOC M.1 4.3.D APPLICABILITY:

ACTIONS MODE 5.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more withdrawn A.1 Initiate action to fully insert Immediately control rods inoperable.

inoperable withdrawn control rods.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1

-NOTES Not required to be performed until 7 days after the control rod is withdrawn.

Insert each withdrawn control rod at least one notch.

7 days SR 3.9.5.2 Verify each withdrawn control rod scram 7 days accumulator pressure is Ž[940a psig.

0D BWR/4 STS 3.9.5-1, Volume 14, Rev. 0, Page 85 of 157 Rev. 3.0, 03/31/04

, Volume 14, Rev. 0, Page 86 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING

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

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 86 of 157

, Volume 14, Rev. 0, Page 87 of 157 Improved Standard Technical Specifications (ISTS) Bases Markup I

and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 87 of 157

, Volume 14, Rev. 0, Page 88 of 157 Control Rod OPERABILITY - Refueling B 3.9.5 B 3.9 REFUELING OPERATIONS B 3.9.5 Control Rod OPERABILITY - Refueling BASES BACKGROUND INSERT 1}

Control rods are components of the Control Rod Drive (CRD) System, 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 during refueling operation. In addition, the control rods provide the capability to maintain the reactor subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System.

GDC 26 of 1,YCFR 50, Appendix A, re uires that one of th~,cfequiredI independ t reactivity control systfns becapable of Godina the reactorI core subc~ritical under cold cond ons (Ref 1)./ The CRD System is the system capable of maintaining the reactor subcritical in cold conditions.

0D I, 'Refueling Equipment InterIocks, I APPLICABLE Prevention and mitigation of promDt reactivity excursions during refuelin SAFETY are provided by refueling interlocks (LCO 3.9.1iand LCO 3.9.e, the

'Refuei ANALYSES SDM (LCO 3.14), the intermediate range monitor neutron flux scra One-Rod-

-SHUTDOWN MARGIN (SDMr I (LCO 3.3.14.), and the control rod block instrumentation (LCO 3.3.2. ). out Reactor Protection System Control Rod Block Instrumentation' Inteiloc (RPS) Instrumentation The safety a lyses for the control rod withdrawal error during fueling (Ref. 2) wanthe fuel assembly inserti error (Ref. 3) evalua the consequences of control rod withdr al during refuelin a also fuel assemby insertion with a control od withdrawn. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. Control rod scram provides protection should a prompt reactivity excursion occur.

Control rod OPERABILITY during refueling satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO Each withdrawn control rod must be OPERABLE. The withdrawn control rod is considered OPERABLE if the scram accumulator pressure is 2 E94Ca psig and the control rod is capable of being automatically inserted upon receipt of a scram signal. Inserted control rods have already completed their reactivity control function, and therefore are not required to be OPERABLE.

0 APPLICABILITY During MODE 5, withdrawn control rods must be OPERABLE to ensure that in a scram the control rods will insert and provide the required negative reactivity to maintain the reactor subcritical.

BWR/4 STS B 3.9.5-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 88 of 157

, Volume 14, Rev. 0, Page 89 of 157 B 3.9.5 0

INSERT I USAR, Section 1.2.2 (Ref. 1), requires the reactor core to be 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.

Insert Page B 3.9.5-1, Volume 14, Rev. 0, Page 89 of 157

, Volume 14, Rev. 0, Page 90 of 157 Control Rod OPERABILITY - Refueling B 3.9.5 BASES APPLICABILITY (continued)

For MODES 1 and 2, control rod requirements are found in LCO 3.1.2, "Reactivity Anomalies," LCO 3.1.3, "Control Rod OPERABILITY,"

LCO 3.1.4, "Control Rod Scram Times," and LCO 3.1.5, "Control Rod Scram Accumulators." During 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.

ACTIONS A.1 With one or more withdrawn control rods inoperable, action must be immediately initiated to fully insert the inoperable control rod(s). Inserting the control rod(s) ensures the shutdown and scram capabilities are not adversely affected. Actions must continue until the inoperable control rod(s) is fully inserted.

SURVEILLANCE SR 3.9.5.1 and SR 3.9.5.2 REQUIREMENTS During MODE 5, the OPERABILITY of control rods is primarily required to ensure a withdrawn control rod will automatically insert if a signal requiring a reactor shutdown occurs. Because no explicit analysis exists for automatic shutdown during refueling, the shutdown function is satisfied if the withdrawn control rod is capable of automatic insertion and the associated CRD scram accumulator pressure is 2f9403 psig.

The 7 day Frequency takes into consideration equipment reliability, procedural controls over the scram accumulators, and control room alarms and indicating lights that indicate low accumulator charge pressures.

SR 3.9.5.1 is modified by a Note that allows 7 days after withdrawal of the control rod to perform the Surveillance. This acknowledges that the control rod must first be withdrawn before performance of the Surveillance, and therefore avoids potential conflicts with SR 10Ed Z

SR

.0.4.

REFERENCES 110 CFR 50, ixpe xA, GOC 2 USARSection12.2 W2R SSART SectB 3.952 13]

BWR/4 STS B 3.9.5-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 90 of 157

, Volume 14, Rev. 0, Page 91 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.5 BASES, CONTROL ROD OPERABILITY - REFUELING

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 changes made to be consistent with the Writer's Guide for the Improved Standard Technical Specifications, NEI 01-03, Section 3.3.2.

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

4. Editorial change made for enhanced clarity.

Monticello Page 1 of I, Volume 14, Rev. 0, Page 91 of 157

, Volume 14, Rev. 0, Page 92 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 92 of 157

, Volume 14, Rev. 0, Page 93 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.5, CONTROL ROD OPERABILITY - REFUELING There are no specific NSHC discussions for this Specification.

Monticello Page I of 1, Volume 14, Rev. 0, Page 93 of 157

, Volume 14, Rev. 0, Page 94 of 157 ATTACHMENT 6 ITS 3.9.6, Reactor Pressure Vessel (RPV) Water Level, Volume 14, Rev. 0, Page 94 of 157

, Volume 14, Rev. 0, Page 95 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 95 of 157

, Volume 14, Rev. 0, Page 96 of 157 ITS 3.9.6

=

ITS Page 1 of 1, Volume 14, Rev. 0, Page 96 of 157

, Volume 14, Rev. 0, Page 97 of 157 DISCUSSION OF CHANGES

-ITS 3.9.6, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any requirements for the reactor pressure vessel (RPV) water level to ensure the consequences of design basis refuel accident is maintained within analysis calculations. ITS LCO 3.9.6 requires the RPV water level to be > 21 ft 11 inches above the top of the RPV flange during the movement of irradiated fuel assemblies within the RPV and during movement of new fuel assemblies or handling of control rods within the RPV when irradiated fuel assemblies are seated within the RPV. An appropriate ACTION and a Surveillance Requirement are also provided. This changes the CTS by incorporating the requirements of ITS 3.9.6.

RPV water level is required to ensure the consequences of a design basis refuel accident remain within the bounds of the radiological dose calculations. The change is acceptable since the RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). This change is designated as more restrictive because it adds new requirements to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1, Volume 14, Rev. 0, Page 97 of 157

, Volume 14, Rev. 0, Page 98 of 157 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 98 of 157

, Volume 14, Rev. 0, Page 99 of 157

[RP\\JlIWater Level Irradi d Fuel] C) 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 fReactor Pressure Vessel (RPV)MWater Level - [Irradi Fuel](

CM DOC M.1 LCO 3.9.6 APPLICABILITY:

RRPVfjwater level shall be 2WRabove the top of theRRPV flange].

During movement of irradiated fuel assemblies within theQRP\\4 1

M During movement of new fuel assemblies or handling of control rods within the ILRP\\/V, when irradiated fuel assemblies are seated within J the IRP4. a 0

0 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME DOC A. ERPVjwater level not

• A within limit.

A.1 Suspend movement of fuel assemblies and handling of control rodsMwithin the MRP\\q.

Immediately }0 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY DOC SR 3.9.6.1 MA1 Verify MRPVM water level is 2 aabove the top of the TRPV flangelJ L

n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

}0 BWR/4 STS 3.9.6-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 99 of 157

, Volume 14, Rev. 0, Page 100 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.6, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL

1. The bracketed term "Irradiated Fuel" has been deleted since all fuel assembly movement over the reactor pressure vessel (RPV) will be covered by ITS 3.9.6.

ISTS 3.9.7, "Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods," has not been adopted in the Monticello ITS. Therefore during the movement of irradiated fuel assemblies within the RPV or during the movement of new fuel assemblies or handling of control rods within the RPV when irradiated fuel assemblies are seated within the RPV, the RPV Water Level will be maintained

> 21 ft 11 inches above the top of the RPV flange in accordance with ITS 3.9.6.

The allowances in ISTS 3.9.7 are not necessary since Monticello outages are planned in such a way that all these operations are performed at a high water level.

Although the safety analyses will support the allowances provided in ISTS 3.9.7, the proposed method of operation is conservative.

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

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 100 of 157

, Volume 14, Rev. 0, Page 101 of 157 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 101 of 157

, Volume 14, Rev. 0, Page 102 of 157 RPV Water Levell-Irradiated Fuel B 3.9.6 0D B 3.9 REFUELING OPERATIONS B 3.9.6 Reactor Pressure Vessel (RPV) Water Level - [Irradiad Fuell 0

BASES

[

1 tI1 Inches BACKGROUND The movement of (irraated] fuel assemblies Bfr handlinof control rodsM 1

(2 within the RPV requires a minimum water level ofl above the top of 2 the RPV flange. During refueling, this maintains a sufficient water level in the reactor vessel cavity and spent fuel pool. Sufficient water is necessary to retain iodine fission product activity in the water in the event of a fuel handling accident (Refs. 1 Sufficient iodine activity would]

be retained to limit offsite doses from the accident to 2 o O 11C-wlimU11imits as provided by e

e of Reference J

APPLICA SAFETY ANALYSE f;;i-E lar. NURE&

[(Ref. 5)

BLE During movement ofl[irr

• ated fuel assembliespor handling of control

(

rodsa, the water level in the RPV is an initial condition design parameter 0 ES in the analysis of a fuel handling accident in contatnmeM postulated by Regulatory Guide 1.25 (Ref. 1 A minimum water level of 23 ft (D Requlato ifon C.1.c o Ref. Iyallows a decontamination factor of

[I.g of Ref. 1)to be used in tohe accident 0

analysis for iodine. This relates to the assumption that of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the water. The fuel pellet to cladding gap is raectons assumed to cona 0% of the total fu de-invento e.

Analysis of the fuel handling accident inside goainmen is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to fuel handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and that offsite doses are maintained within allowable limits (Ref. 4).

0O While the worst case assumptions include the dropping of the irradiated fuel assembly being handled onto the reactor core, the possibility exists of the dropped assembly striking the RPV flange and releasing fission products. Therefore, the minimum depth for water coverage to ensure acceptable radiological consequences is specified from the RPV flange.

Since the worst case event results in failed fuel assemblies seated in the core, as well as the dropped assembly, dropping an assembly on the RPV flange will result in reduced releases of fission gases. jBased on this judglment, and the physical dimensions which preclude normal operation with water level 23 feet above the flange, a slight reduction in this water level is acceptable (Ref.,R)M RPV water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

0D BWR/4 STS B 3.9.6-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 102 of 157

, Volume 14, Rev. 0, Page 103 of 157 RPV Water Levell-Irradiated FL B 3.9 Iel 1.6 BASES 1

inches 3 (D 0

0 LCO A minimum water level of above the top of the RPV flange is required to ensure that the radiological consequences of a postulated fuel handling accident are within acceptable limits, as provided by the guidance of Reference P-M

-1 APPLICABILITY LCO 3.9.6 is applicable when moving irra te fuel assemblies Tor (1

handling control rods (i.e., movement with other than the normal control rod driveH within the RPV. The LCO minimizes the possibility of a fuel handling accidentlin comment that is beyond the assumptions of the safety anal

[is irraia ue is not present wi in, ere can be no significant radio ivity release as a result postulated fuel handling accideni equirements for hand of new fuel assemblies or control rods ere water depth to the flange is not of concern) are covered LCO 3.9.7. "RPV Water vel - New Fuel or Control Rods."

Requirements for fuel handling accidents in the spent fuel storage pool are covered by LCO 3.7.8, "Spent Fuel Storage Pool Water Level."

0!

0D 0D

---

REVWER'S NOTE----

LCO 3.9.6 is written to cove new fuel and control rods well as irradiated fuel. If a plant dopts LCO 3.9.7, however he second bracketed portion of t Applicability is adopted iieu of the first bracketed portion, d the LCO name and Re ired Action A.1 modified appropriately.

ACTIONS A.1 1

If the water level is <[

above the top of the RP flange, all operations involving movement o iena ted uetassemblies aand handling of control rodsiwithin the RPV shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of

[irrap.ated] fuel movementland control rod handlinglshall not preclude completion of movement of a component to a safe position.-

SURVEILLANCE REQUIREMENTS SR 3.9.6.1 Verification of a minimum water level o 2 above the top of the RPV flange ensures that the design basis forte postulated fuel handling accident analysis during refueling operations is met. Water at the required level limits the consequences of damaged fuel rods, which are postulated to result from a fuel handling accidentlin cormen l(Ref. 2).

The Frequency of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls on valve positions, which make significant unplanned level changes unlikely.

0D 0D BWR/4 STS B 3.9.6-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 103 of 157

, Volume 14, Rev. 0, Page 104 of 157 RPV Water Levell-Irradiated Fuel (D

B 3.9.6 BASES REFERENCE'S

1. Regulatory Guide 1.25, March 23, 1972.

147o6

3. Regulatory Guide 1.183, July 2000.

2SR etol[15.41] 4

. 1 0 CFR 50.67.(i()

M4. NUREG-0800, Section 15.7.4.

I 4-10100.11.l 0

BWR/4 STS B 3.9.6-3 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 104 of 157

, Volume 14, Rev. 0, Page 105 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.6 BASES, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL

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

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

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

4. This reviewer's type of note has been deleted. This is not meant to be retained in the final version of the plant specific submittal. In the Monticello ITS, ISTS 3.9.7 is not being adopted. Thus, ITS 3.9.6 will cover all the Applicabilities.

Monticello Page 1 of I, Volume 14, Rev. 0, Page 105 of 157

, Volume 14, Rev. 0, Page 106 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 106 of 157

, Volume 14, Rev. 0, Page 107 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.6, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL There are no specific NSHC discussions for this Specification.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 107 of 157

, Volume 14, Rev. 0, Page 108 of 157 ATTACHMENT 7 ITS 3.9.7, Residual Heat Removal (RHR) - High Water Level, Volume 14, Rev. 0, Page 108 of 157

, Volume 14, Rev. 0, Page 109 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 109 of 157

, Volume 14, Rev. 0, Page 110 of 157 ITS 3.9.7 4 -AddprposdTS.9.7 Page 1 of 1, Volume 14, Rev. 0, Page 110 of 157.

, Volume 14, Rev. 0, Page 111 of 157 DISCUSSION OF CHANGES ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any requirements for the Residual Heat Removal (RHR)

Shutdown Cooling System during MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level 2 21 ft 11 inches above the top of the RPV flange. ITS LCO 3.9.7 requires one RHR shutdown cooling subsystem to be OPERABLE and in operation. Appropriate ACTIONS and a Surveillance Requirement are also provided. This changes the CTS by incorporating the requirements of ITS 3.9.7.

Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. The change is acceptable since the RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). This change is designated as more restrictive because it adds new requirements to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1, Volume 14, Rev. 0, Page 111 of 157

, Volume 14, Rev. 0, Page 112 of 157 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 112 of 157

, Volume 14, Rev. 0, Page 113 of 157 RHR - High Water Level 3.9 CTS 3.9 REFUELING OPERATIONS 3.9 Residual Heat Removal (RHR) - High Water Level DOC LCO 3.9.l One RHR shutdown cooling subsystem shall be OPERABLE and in MA1 operation.

0C 0

(

NU]t P-The required RHR shutdown cooling subsystem may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

APPLICABILITY:

MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level 2 above the top of thejRPV flange].

ME 0

DOC M.1 DOC M.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required RHR shutdown A.1 Verify an alternate method 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> cooling subsystem of decay heat removal is inoperable.

available.

AND Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter B. Required Action and B.1 Suspend loading irradiated Immediately associated Completion fuel assemblies into the Time of Condition A not RPV.

met.

AND B.2 Initiate action to restore Immediately TsecondarI containment to OPERABLE status.

AND 0

BWR/4 STS 3.9.8-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 113 of 157

, Volume 14, Rev. 0, Page 114 of 157 RHR - High Water Level ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME DOC M.1 DOC C. No RHR shutdown M.1 cooling subsystem in operation.

B.3 Initiate action to restore one standby gas treatment subsystem to OPERABLE status.

Immediately Immediately AND B.4 Initiate action to restore isolation capability in each required secondarA containment penetration flow path not isolated.

C.1 Verify reactor coolant circulation by an alternate method.

AND C.2 Monitor reactor coolant temperature.

.0 1 hourfrom discovery of no reactor coolant circulation AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter Once per hour SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9-1 Verify one RHR shutdown cooling subsystem is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> operating.

DOC M.1 0D BWRI4 STS 3.9.8-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 114 of 157

, Volume 14, Rev. 0, Page 115 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL

1. ISTS 3.9.8 is renumbered as ITS 3.9.7 since ISTS 3.9.7, "Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods," is not included in the Monticello ITS.

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

Monticello Page 1 of I, Volume 14, Rev. 0, Page 115 of 157

, Volume 14, Rev. 0, Page 116 of 157 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 116 of 157

, Volume 14, Rev. 0, Page 117 of 157 RHR - High Water Leve,/l..

B 3.9 B 3.9 REFUELING OPERATIONS m

B 3.9.rResidual Heat Removal (RHR) - High Water Level 0

described by USAR, Section 10.2.4.2 (Ref. 1) )

BASES

====i BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and sensible heat from the reactor coolant, as require

.. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump Ier discharges the reactor coolant, after it has been cooled by circulation 7 through the respective heat exchangers, to the reactor via recirculation loop lor to the reactor v l:pre ss re coolant lent.

The RHR heat exchangers transfer heat to the RHR Service Water System. The RHR shutdown cooling mode is manually controlled.

l In addition to the RHR subsystems, the volume of water above the reactor pressure vessel (RPV) flange provides a heat sink for decay heat removal.

I APPLI1 SAFEI ANAL)

SABLE With the unit in MODE 5, the RHR stem is not required to mitigate any

-Y events or accidents evaluated in the safety analyses. The RH stem

'SES is required for removing decay heat to maintain the temperature of the reactor coolant.

SdThe RHf $System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

0 LCO Only one RHR shutdown cooling subsystem is required to be OPERABLE i and in oeration in MODE 5 with irradiated fuel in the RPV and the water L

v

((

bove the RPV flange. Only one subsystem is required 43) because the volume of water above the RPV flange provides backup decay heat removal capability.

to be OPERABLE In addition, the necessary portions of the RHRSW System must be capable of providing cooling water to the RHR heat ehanger.

An OPERABLE RHR shutdown cooling subsystem consists of an RHR pump, a heat exchanger, valves, piping instruments, and controls to ensureanOPERABLE flowpathl In MODE 5, e RHR cross tie valve is not required to by closed; thus, the valve may e opened to allow pumps in one loop to di charge through the opposite oop's heat exchanger to make a complete subsystem. I 0-0 BWR/4 STS B 3.9.8-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 117 of 157

, Volume 14, Rev. 0, Page 118 of 157 RHR - High Water Level (i

B 3.9 BASES LCO (continued)

[ the--I Additionally,lRRHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exception for the operating subsystem to be removed from operation every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

0 APPLICABILITY One RHR shutdown cooling subsystem must be OPERABLE and in RPV operation in MODE 5, with irradiated fuel in thelreactor pressure vesse(

Ihand with the water level 23 eoeabove the top of the RPV flange, to (i) provide decay heat removal. RHR$ystem requirements in other MODES m

Lou1"OwTn 1are covered by LCOs in Section 3.4, Reactor Coolant System (RCS Section 3.5, Emorgency Core Cooling Systemf(ECCS) and Reactor Core Isolation aoolinq (RCIC) Svstem; and Soction 3.6, Containment

[Sub S Sstem. RHR hutdown ¢oolin ystem requirements in MODE 5 with irradiated fuel in the reactor pressure vessel and with the water level above the RPV flange are given in LCO 3.9 E3' ACTIONS A 1 f Residual Heat Removal (RHR) -

[Low Water Level.'

I J

With no RHR shutdown cooling subsystem PERABLE, an alternate method of decay heat removal must be est she within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. In this condition, the volume of water above the RPV flange provides adequate capability to remove decay heat from the reactor core. However, the overall reliability is reduced because loss of water level could result in reduced decay heat removal capability. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of the alternate methodW must be reconfirmed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. This will ensure continued heat removal capability. The required cooling capacity of the alternate method should be ensured by verification

-(,)

.(by calculation or demonstration) of Its capability to maintain or reduce temperature.

Fuel Pool Cooling System or the Alternate decay heat removal methods are available to the operators for review and preplanning in the unit[Eoperating procedures(. For example, this may include the use of thd Reactor Water Cleanup System 5 operating with the regenerative heat exchanger bypassedThe method used to remove the decay heat should be the most prudent choice based on unit conditions.

I (heat reject mode). Either or both systems may be used to reject hot water while using a cooler source of water for makeup 0D 0

BWR/4 STS B 3.9.8-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 118 of 157

, Volume 14, Rev. 0, Page 119 of 157 RHR - High Water Level B 3.9.

BASES ACTIONS (continued)

B.1. B.2, B.3, and B.4 If no RHR shutdown cooling subsystem is OPERABLE and an alternate method of decay heat removal is not available in accordance with Required Action A.1, actions shall be taken immediately to suspend operations involving an increase in reactor decay heat load by suspending loading of irradiated fuel assemblies into the RPV.

Additional actions are required to minimize any potential fission product release to the environment. This includes ensuring seconda a

containment is OPERABLE; one standby gas treatment svstem Isl OPERABLE; and secondary containment isolation capa iitie.,

o eX secondary containment isolation valve and associae intuentations are OPERABLE or other acceptable administrative controls to assure islto apbl N

iea~ascaeddeerbo~~oa that i ff I

assumed to be isolated to mitigate radioactiverases.This nay be l perormed as an administrative check, by examining logs or ot e S

information to determine whether 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, a surveillance may need to be performed to restore the component to OPERABLE status.

Actions must continue until all required components are OPERABLE.

C.1 and C.2 If no RHR Shutdown Cooling System is in operation, an alternate method of coolant circulation is required to be established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Completion Time is modified such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is applicable separately for each occurrence involving a loss of coolant circulation.

During the period when the reactor coolant is being circulated by an J alternate method (other than by the required RHR $hutdown tooling

'ystem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.

BWR/4 STS B 3.9.8-3 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 119 of 157

, Volume 14, Rev. 0, Page 120 of 157 B 3.9.7 0

INSERT 1 These administrative controls consist of stationing a dedicated qualified individual, 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.

0 INSERT 2 (ensuring components are OPERABLE)

Insert Page B 3.9.8-3, Volume 14, Rev. 0, Page 120 of 157

, Volume 14, Rev. 0, Page 121 of 157 RHR - High Water Level B 3.9 BASES SURVEILLANCE 0R 3

REQUIREMENTS This Surveillance demonstrates that the RHRisubsystem is in operation Ishutdown E

and circulating reactor coolant.,

rhe required flow rate is determined by the flow rate necessary to provide I Oufficient decay heat removal capabilit. The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is i v o

end audible indications available to the sco I operator for monitoring the R subsystem in the control room.

REFERENCES 3

(E)

BWR/4 STS B 3.9.8-4 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 121 of 157

, Volume 14, Rev. 0, Page 122 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.7 BASES, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL

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

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 the Writer's Guide or similar statements in other places in the Bases.

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

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

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

7. RHR shutdown cooling subsystem requirements, which is what this LCO is governing, are not covered in other MODES in Section 3.5 or 3.6. Therefore, this statement has been deleted.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 122 of 157

, Volume 14, Rev. 0, Page 123 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 123 of 157

, Volume 14, Rev. 0, Page 124 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.7, RESIDUAL HEAT REMOVAL (RHR) - HIGH WATER LEVEL There are no specific NSHC discussions for this Specification.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 124 of 157

, Volume 14, Rev. 0, Page 125 of 157 ATTACHMENT 8 ITS 3.9.8, Residual Heat Removal (RHR) - Low Water Level, Volume 14, Rev. 0, Page 125 of 157

, Volume 14, Rev. 0, Page 126 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 126 of 157

, Volume 14, Rev. 0, Page 127 of 157 ITS 3.9.8 4 -F Add oS3.9.8 Page 1 of 1, Volume 14, Rev. 0, Page 127 of 157

, Volume 14, Rev. 0, Page 128 of 157 DISCUSSION OF CHANGES ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES M.1 The CTS does not have any requirements for the Residual Heat Removal (RHR)

Shutdown Cooling System during MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level < 21 ft 11 inches above the top of the RPV flange. ITS LCO 3.9.8 requires two RHR shutdown cooling subsystems to be OPERABLE, and one RHR shutdown cooling subsystem in operation.

Appropriate ACTIONS and a Surveillance Requirement are also provided. This changes the CTS by incorporating the requirements of ITS 3.9.8.

Decay heat removal by operation of the RHR System in the shutdown cooling mode is not required for mitigation of any event or accident evaluated in the safety analyses. Decay heat removal is, however, an important safety function that must be accomplished or core damage could result. The change is acceptable since the RHR Shutdown Cooling System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). This change is designated as more restrictive because it adds new requirements to the CTS.

RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES None LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1, Volume 14, Rev. 0, Page 128 of 157

, Volume 14, Rev. 0, Page 129 of 157 Improved Standard Technical Specifications (ISTS) Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 129 of 157

, Volume 14, Rev. 0, Page 130 of 157 RHR - Low Water Level 3.9 CTS 3.9 REFUELING OPERATIONS 3.9 Residual Heat Removal (RHR) - Low Water Level DOC LCO 3.9 Two RHR shutdown cooling subsystems shall be OPERABLE, and one MA

.RHR shutdown cooling subsystem shall be in operation.

0D

-NOTI The required operating shutdown cooling subsystem may be removed from operation for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period.

APPLICABILITY:

MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and the water level 2

above the top of the MRPV flangesJl L421 f 11inchesl DOC M.1 DOC M.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or two required A.1 Verify an alternate method 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

/

RHR shutdown cooling of decay heat removal is subsystems inoperable.

available forkR~*--EI AND NOTE inoperable required RHR Separate Condition entry Is allowed for shutdown cooling Once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> each inoperable required RHR subsystem.

thereafter Lshutdown cooling subsystem.suyte.hrafr B. Required Action and B.1 Initiate action to restore Immediately associated Completion secondarya containment to Time of Condition A not OPERABLE status.

met.

AND B.2 Initiate action to restore one Immediately standby gas treatment subsystem to OPERABLE status.

AND 0

0 BWR/4 STS 3.9.9-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 130 of 157

, Volume 14, Rev. 0, Page 131 of 157 RHR - Low Water Level 3.9 (o)

CTS DOC M.1 DOC M.1 DOC M.1 ACTIONS (continued)

REQUIRED ACTION COMPLETION TIME CONDITION B.3 Initiate action to restore Immediately isolation capability in each required asecondarAj containment penetration flow path not isolated.

C. No RHR shutdown C.1 Verify reactor coolant 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery cooling subsystem in circulation by an alternate of no reactor coolant operation.

method.

circulation AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter AND C.2 Monitor reactor coolant Once per hour temperature.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9 M.

Verify one RHR shutdown cooling subsystem is 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> operating.

0 0D BWRI4 STS 3.9.9-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 131 of 157

, Volume 14, Rev. 0, Page 132 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL

1. ISTS 3.9.9 is renumbered as ITS 3.9.8 since ISTS 3.9.7, "Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods," is not included in the Monticello ITS.

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

3. Condition A has been modified by the addition of a Note that allows separate Condition entry for each inoperable required RHR shutdown cooling subsystem.

Currently, the Condition is required to be entered if one or two required RHR shutdown cooling subsystems are inoperable. The Required Actions require the verification of an alternate method of decay heat removal for each inoperable required RHR shutdown cooling subsystem within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. According to ITS 1.3, Completion Times, when one required RHR shutdown cooling subsystem is inoperable, entry into the Condition is required and the Completion Times start upon entry into the Condition. When the second required RHR shutdown cooling subsystem becomes inoperable, a new Condition entry is not allowed; the Completion Times from the initial entry are still applicable. Thus, if the second required RHR shutdown cooling subsystem becomes inoperable more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the first subsystem, no time is provided to verify a second alternate method; the time has already expired. Therefore, the Note to Condition A has been added to allow separate Condition entry for each inoperable required RHR shutdown cooling subsystem. In addition, the Required Action has been modified to be applicable to the associated RHR shutdown cooling subsystem (by changing the word "each" to "the"). In addition, this change is consistent with previously approved BWR ITS conversions (i.e., LaSalle Units 1 and 2, Quad Cities Units 1 and 2, and Dresden Units 2 and 3).

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 132 of 157

, Volume 14, Rev. 0, Page 133 of 157 Improved Standard Technical Specifications (ISTS) Bases Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 133 of 157

, Volume 14, Rev. 0, Page 134 of 157 RHR - Low Water LeveL/

B 3.91O(i B 3.9 REFUELING OPERATIONS B 3.9.9 Residual Heat Removal (RHR) - Low Water Level 0D

[described by USAR.

Section 10.2.4.2 (Ref. 1)

BASES BACKGROUND The purpose of the RHR System in MODE 5 is to remove decay heat and l sensible heat from the reactor coolant, asjrequireeGDC 3. Each of the two shutdown cooling loops of the RHR System can provide the required decay heat removal. Each loop consists of two motor driven pumps, a heat exchanger, and associated piping and valves. Both loops have a common suction from the same recirculation loop. Each pump i

discharges the reactor coolant, after it has been cooled by circulation through the respective heat exchangers, to the reactor via thesiae recirculation loop lor to the reactor viamge~lu r

ooant iwectioffE E gth. The RHR heat exchangers transfer heat to the RHR Service Water7 J System. The RHR shutdown cooling mode is manually controlled.

RJ APPLI1 SAFEI ANAL) tShutdown Coolin CABLE With the unit in MODE 5, the RHR ystem is not required to mitigate any IY events or accidents evaluated in the safety analyses. The RH ysem

'SES is required for removing decay heat to maintain the temperature of the reactor coolant.

The RH ;System satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

LCO In MODE 5 with irradiated fuel in the reactor pressure vessel (RPV) and 21fnches the watr level *j above the reaoressur e sel (RPVqflange (7) both RHR shutdown cooling subsystems must be OPERABLI

[and one RIIR

)

00 shutdown cooling subsystem must be I operation I n addition, the necessary portions of the RHRSW System must be capable of providing cooling water to the RHR heat exchanger.

i An OPERABLE RHR shutdown cooling subsystem consists of an RHR J pump, a heat exchanger, valves, piping, instruments, and controls to ensure an OPERABLE flow path. To meet the LCO, both pumps in one loop or one pump in each of the two loops must be OPERABLE~l MODE 5, the R11 ross tie valve is not requid to be closed; hs th valve may be opnd to allow pumps in one l~o odshretruhte opposite loop' ha exchanger to make a cofnlt usse.

Additionally, each RHR shutdown cooling subsystem is considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required. A Note is provided to allow a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exception for the operating subsystem to be removed from operation every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

BWR/4 STS B 3.9.9-1 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 134 of 157

, Volume 14, Rev. 0, Page 135 of 157 RHR - Low Water Level B 3.9t

')

BASES APPLICABILITY Two RHR shutdown cooling subsystems are required to be OPERABLE, and one must be in operation in MODE 5, with irradiated fuel in the RPV 2and with the water level <

ft above the top of the RPV flange, to 4

provide decay heat removal. RHRyystem requirements in other MODES shtonl subare covered by LCOs in Section 3.4, Reactor Coolant System (RCS, ection 3.5, Em rgency Core Cooling System (FCaS) and Reactor 7

Core Isolation ooling (RCIC) Systems and S ction 3.6. Containment stems.RHRIShutdown

¢oolin ystem requirements in MODE 5 with (E)

Inches irradiated tuel in the RPV and with the water level ft above the RPV lnge are given in LCO 3.9 "Residual Heat Removal (RHR)- High Water Level."

i ACTIONS A.1 With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing the required decay heat removal. However, the overall reliability is reduced.

L erefoiran alternate method of decay heat removal must be provided.

With both required RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of this alternate method(s) must be reconfirmed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter. This will ensure continued heat removal capability. [The required cooling capacity of the alternate method should be ensured by verifcation A

)

_(by calculation or demonstration) of Its capability to maintain or reduce temperature.

2 Alternate decay heat removal methods are available to the operators for F~uel pool review and preplanning in the uniM Operating Procedures. For example, (i

coolingnis may include e use o th Reactor Water Cleanup Systemoperating-Systemorthe with the regenerative heat exchanger bypassed,; The method used to (2) remove decay heat should be the most prudentlchoice based on unit conditions.

{(heat reject mode). Either or both systems may be used to reject Lhot water while using a cooler source of water for makeup B.1. B.2. and B.3 1I INSROI With the required decay heat removal subsystem(s) inoperable and the required alternate method(s) of decay heat removal not available in accordance with Required Action A.1, additional actions are required to minimize any potential fission product release to the environment. This includes ensuring secondary containment is OPERABLE; one standby gas treatment subsystem is OPERABLE; and secondary containment BWR/4 STS B 3.9.9-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 135 of 157

, Volume 14, Rev. 0, Page 136 of 157 0

B 3.9.8 INSERT I Condition A is modified.by a Note allowing separate Condition entry for each inoperable required RHR shutdown cooling subsystem. This is acceptable since the Required Actions for this Condition provide appropriate compensatory actions for each inoperable required RHR shutdown cooling subsystem. Complying with the Required Actions allow for continued operation. A subsequent inoperable required RHR shutdown cooling subsystem is governed by subsequent entry into the Condition and application of the Required Actions.

l Insert Page B 3.9.9-2, Volume 14, Rev. 0, Page 136 of 157

, Volume 14, Rev. 0, Page 137 of 157 RHR - Low Water Level B 3.9SD BASES ACTIONS (continued) ab isolation capa iit i'i.e., one secondary containment isolation valve anl Iassociated i sruentation are OPERABLE or other accetbl Iadministrative controls to assure isolation capabiliyi ech associated at penetratiornot isolatedthat is assumed to be isolated to mitigate raioac iv reThis~may be performed as an administrative check, or other information to determine whether the INSERT 3 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, the surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.

0 C.1 and C.2

[shutdwn cooing j

If no RHRsubsystem is in operation, an alternate method of coolant circulation is required to be established within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Completion Time is modified such that the 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is applicable separately for each occurrence involving a loss of coolant circulation.

During the period when the reactor coolant is being circulated by an

) alternate method (other than by the required RHR $hutdown tooling ity 4stem), the reactor coolant temperature must be periodically monitored to ensure proper functioning of the alternate method. The once per hour Completion Time is deemed appropriate.

0D 0

SURVEILLANCE REQUIREMENTS This Surveillance demonstrates that one RHR shutdown cooling subsystem is in operation and circulating reactor coolant. The required flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability.

0 The Frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient in view of other visual and audible indications available to the operator for monitoring the RHR subsystems in the control room.

0D 0D REFERENCES BWR/4 STS B 3.9.9-3 Rev. 3.0, 03/31104, Volume 14, Rev. 0, Page 137 of 157

, Volume 14, Rev. 0, Page 138 of 157 B 3.9.8 0

INSERT 2 These administrative controls consist of stationing a dedicated qualified individual, 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.

0 INSERT 3 (ensuring components are OPERABLE)

Insert Page B 3.9.9-3, Volume 14, Rev. 0, Page 138 of 157

, Volume 14, Rev. 0, Page 139 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.8 BASES, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL

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

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 the Writer's Guide or similar statements in other places in the Bases.

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

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

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

7. RHR shutdown cooling subsystem requirements, which is what this LCO is governing, are not covered in other MODES in Section 3.5 or 3.6. Therefore, this statement has been deleted.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 139 of 157

, Volume 14, Rev. 0, Page 140 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 140 of 157

, Volume 14, Rev. 0, Page 141 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.9.8, RESIDUAL HEAT REMOVAL (RHR) - LOW WATER LEVEL There are no specific NSHC discussions for this Specification.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 141 of 157

, Volume 14, Rev. 0, Page 142 of 157 ATTACHMENT 9 Relocated/Deleted Current Technical Specifications, Volume 14, Rev. 0, Page 142 of 157

, Volume 14, Rev. 0, Page 143 of 157 CTS 3.10.D, Decay Time, Volume 14, Rev. 0, Page 143 of 157

, Volume 14, Rev. 0, Page 144 of 157 Current Technical Specification (CTS) Markup and Discussion of Changes (DOCs), Volume 14, Rev. 0, Page 144 of 157

C C

C CTS 3.10.D 3.0 LIMITING CONDmONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS C)

W 0

0-0 F

CD

-o t)

CD 0

to CD ul en 0

-9' L"

B. Core Monitoring During core alterations two SRM's shall be operable.

one In and one adjacent to any core quadrant where fuel or control rods are being moved. For an SRM to be considered operable, the following conditions shall be satisfied:

1. The SRM shall be inserted to the normal operating level. (Use of special moveable. dunking type detectors during Initial fuel loading and major core alterations is permissible as long as the detector is connected Into the normal SRM circuit.)

2. The SRM shall have a minimum of 3 CPS with all rods fully Inserted In the core except when both of the following conditions are fulfilled:

a. No more than two fuel assemblies are present In the core quadrant associated with the SRM.

b. While in core. these fuel assemblies are In

-lrotons adlacent to the SRM B. Core Monitoring Prior to making any alterations to the core while mnore than two fuel assemblies are present In any reactor quadrant, the SRM's shall be functionally tested and checked for neutron response. Thereafter, the SRMs will be checked daily for response.

ISee ITS 3.3.1.2}

0) 0 0

CD 2

0 0

o CD

-9 Ut

.Pb CD en 0

IN C. Fuel Storage Pool Water Level Whenever Irradiated fuel Is stored In the fuel storage pool, the pool water level shall be maintained at a level of greater or equal to 33 feet.

C. Fuel Storage Pool Water Level Whenever Irradiated fuel is stored in the fuel storage See ITS 3.7.8 pool the pool level shall be recorded daily.

I

/

D. The/eacor shall beXhutdown for a rplnimum of 24 h6urs prior to movdment of fuel witVn the reactor.

-I 9LA I

3.10/4.10 207 Amendment No. 2v01 23 10126/01 Page 1 of 1

, Volume 14, Rev. 0, Page 146 of 157 DISCUSSION OF CHANGES CTS 3.1 O.D, DECAY TIME ADMINISTRATIVE CHANGES None MORE RESTRICTIVE CHANGES None RELOCATED SPECIFICATIONS None REMOVED DETAIL CHANGES LA.l (Category 6 - Removal of LCO, SR, or other TS requirement to the TRM, USAR, ODCM, OQAP, IST Program, or lIP) CTS LCO 3.1 0.D requires the reactor to be subcritical for a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to movement of fuel within the reactor.

The ITS does not include the requirements for decay time. This changes the CTS by moving the explicit decay time requirements from the Technical Specifications to the Technical Requirements Manual (TRM).

The removal of these details from the Technical Specifications is acceptable because this type of information is not necessary to provide adequate protection of public health and safety. The purpose of CTS LCO 3.1 0.D to ensure that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products in the irradiated fuel consistent with the assumptions used in the fuel handling accident analysis. Although CTS LCO 3.10.D satisfies Criterion 2 of the Technical Specifications Selection Criteria in 10 CFR 50.36 (c)(2)(ii) (for the radioactive decay assumptions in the fuel handling accident), the requirements for decay time following subcriticality will always be met for a refueling outage because of the operations required prior to moving Irradiated fuel in the reactor vessel (e.g., reactor cooldown, containment entry, removal of vessel head, removal of vessel internals, etc.). Also, this change is acceptable because the removed information will be adequately controlled in the TRM. The TRM is incorporated by reference into the USAR and any changes to the TRM are made under 10 CFR 50.59, which ensures changes are properly evaluated. This change is designated as a less restrictive removal of detail change because a requirement is being removed from the Technical Specifications.

LESS RESTRICTIVE CHANGES None Monticello Page 1 of 1, Volume 14, Rev. 0, Page 146 of 157

, Volume 14, Rev. 0, Page 147 of 157 Specific No Significant Hazards Considerations (NSHCs), Volume 14, Rev. 0, Page 147 of 157

, Volume 14, Rev. 0, Page 148 of 157 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS CTS 3.10.D, DECAY TIME There are no specific NSHC discussions for this Specification.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 148 of 157

, Volume 14, Rev. 0, Page 149 of 157 ATTACHMENT 10 Improved Standard Technical Specifications (ISTS) not adopted in the Monticello ITS, Volume 14, Rev. 0, Page 149 of 157

, Volume 14, Rev. 0, Page 150 of 157 ISTS 3.9.7, Reactor Pressure Vessel (RPV) Water Level - New Fuel or Control Rods, Volume 14, Rev. 0, Page 150 of 157

, Volume 14, Rev. 0, Page 151 of 157 ISTS 3.9.7 Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 151 of 157

, Volume 14, Rev. 0, Page 152 of 157

[RPV] Water Level - [Ne Fuel or Control Rods]

3.9.7 3.9 REFUELI G OPERATIONS 3.9.7

[R actor Pressure Vessel (RP

] Water Level - [New Fuel or Control Rods]

LCO 3.9.7

[RPV] water level sh 11 be 2 [23] ft above the top o irradiated fuel assemblies seated ithin the [RPV].

APPLICABILI During movement o new fuel assemblies or hand ing of control rods within the [RP

, when irradiated fuel assem lies are seated within the [RPV].

ACTIONS C NDITION EQUIRED ACTION COMPLETION TIME A. [RPV]

ater level not A.1 Suspend movement of fuel Immediately within imit.

assemblies and handling of control rods within the

[RPV].

SURVEIL NCE REQUIREMENTS l_l SURVEIL NCE FREQUENCY SR 3.9..1 Verify [RPV] wate level is 2 [23] ft above the to of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> irradiated fuel ass mblies seated within the [R

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, Volume 14, Rev. 0, Page 153 of 157

.:JUSTIFICATION FOR DEVIATIONS ISTS 3.9.7, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL - NEW FUEL OR CONTROL RODS 1.

ISTS 3.9.7 has not been adopted since it is not applicable to Monticello.

Therefore, during the movement of irradiated fuel assemblies within the Reactor Pressure Vessel (RPV) or during the movement of new fuel assemblies or handling of control rods within the RPV when irradiated fuel assemblies are seated within the RPV, the RPV Water Level will be maintained 21 ft 11 inches above the top of the RPV flange in accordance with ITS 3.9.6, "Reactor Pressure Vessel (RPV) Water Level." The allowances in ISTS 3.9.7 are not necessary since Monticello outages are planned in such a way that all these operations are performed at a high water level. Although the safety analyses will support the allowances provided in ISTS 3.9.7, the proposed method of operation is conservative.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 153 of 157

, Volume 14, Rev. 0, Page 154 of 157 ISTS 3.9.7 Bases Markup and Justification for Deviations (JFDs), Volume 14, Rev. 0, Page 154 of 157

, Volume 14, Rev. 0, Page 155 of 157 RPV Water Level - Ne Fuel or Control Rods B 3.9.7 B 3.9 REFUE ING OPERATIONS B 3.9.7 React r Pressure Vessel (RPV) W ter Level - New Fuel or Cont I Rods BASES BACKGROU D The movement of n (N fuel assemblies or handlin of control rods within the RPV when fuel semblies seated within the r actor vessel are irradiated requires a minimum water level of [23]

above the top of irradiated fuel asse blies seated within the RPV. During refueling, this maintains a sufficie t water level above the irradi ted fuel. Sufficient water is necessary retain iodine fission produc activity in the water in the event of a fuel ndling accident (Refs. 1 and 2). Sufficient iodine activity would be re ained to limit offsite doses fro the accident to s 25%

of 10 CFR 100 limit,as provided by the guidanc of Reference 3.

APPLICABL SAFETY ANALYSES During movement f new fuel assemblies or han ling of control rods over irradiated fuel ass mblies, the water level in the PV is an initial condition design p rameter in the analysis of a fel handling accident in containment postu ated by Regulatory Guide 1. 5 (Ref. 1). A minimum water level of [23]

(Regulatory Position C.1.c f Ref. 1) allows a decontamination ctor of 100 (Regulatory Posi on C.1.g of Ref. 1) to be used in the accide t analysis for iodine. This re ates to the assumption that 99% of the to al iodine released from the p nlet to cladding gap of all the dropped fuel ssembly rods is retained by t e water. The fuel pellet to cladding gap is assumed to contain 10% of t e total fuel rod iodine inventory (Ref. 1)

Analysis of the fu I handling accident inside co tainment is described in Reference 2. Wi h a minimum water level of [ 3] ft and a minimum decay time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> rior to fuel handling, the ana sis and test programs demonstrate tha the iodine release due to a p stulated fuel handling accident is adeq ately captured by the water nd that offsite doses are maintained withi allowable limits (Ref. 4).

The related ass mptions include the worst ca e dropping of an irradiated fuel assembly o to the reactor core loaded wih irradiated fuel assemblies..

RPV water leve satisfies Criterion 2 of 1 0 CF 50.36(c)(2)(ii).

-o0 LCO A minimum wa r level of [23] ft above the to of irradiated fuel assemblies se ted within the RPV flange is r quired to ensure that the radiological co sequences of a postulated fu I handling accident are within accepta le limits, as provided by the uidance of Reference 3.

BWRI4ISTS I

B 3.9.7-1 I

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, Volume 14, Rev. 0, Page 156 of 157 P

RPV Water Level - New uel or Control Rods e

le B 3.9.7 BASES APPLICABILI LCO 3.9.7 is applicabl when moving new fuel asse blies or handling control rods (i.e., mov ment with other than the no al control rod drive) over irradiated fuel as emblies seated within the R

. The LCO minimizes the possibil ty of a fuel handling accident n containment that is beyond the assumpti s of the safety analysis. If ir adiated fuel is not present within the RP, there can be no significant adioactivity release as a result of a postul ted fuel handling accident.

equirements for fuel handling accidents in he spent fuel storage pool ar covered by LCO 3.7.8, "Spent F I Storage Pool Water Level.' Requirements for handling irradiated fuI over the RPV are covered y LCO 3.9.6,

"[Reactor Pressure ssel (RPV)] Water Level - [Ir adiated Fuel]."

ACTIONS A.1 If the water level is

[23] ft above the top of irradi ted fuel assemblies seated within the R V, all operations involving m vement of new fuel assemblies and han ling of control rods within th RPV shall be suspended immedi ely to ensure that a fuel han ling accident cannot occur. The suspen ion of fuel movement and co trol rod handling shall not preclude compl tion of movement of a comp nent to a safe position.

-o SURVEILLA CE REQUIREM NTS SR 3.9.7.1 Verification of a mi mum water level of [23] ft a ve the top of irradiated fuel assemblies se ted within the RPV ensures t at the design basis for the postulated fuel handling accident analysis d ring refueling operations is met. Water at t e required level limits the co equences of damaged fuel rods, which ar postulated to result from a f el handling accident in containment (Ref.

).

The Frequency of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is based on engine ring judgment and is considered adequ te in view of the large volum of water and the normal procedural contros on valve positions, which ke significant unplanned level changes unl kely.

l ES

1. Regulatory uide 1.25, March 23, 1972.

2.

FSAR, Secti n [1 5.1.41].

3.

NUREG-08

, Section 15.7.4.

4.

10CFR10.11.

B 3.9.7-2 Rev. 3.0, 03/31/04, Volume 14, Rev. 0, Page 156 of 157

, Volume 14, Rev. 0, Page 157 of 157 JUSTIFICATION FOR DEVIATIONS ITS 3.9.7 BASES, REACTOR PRESSURE VESSEL (RPV) WATER LEVEL-NEW FUEL OR CONTROL RODS

1. Changes have been made to reflect those changes made t6 the Specification.

Monticello Page 1 of 1, Volume 14, Rev. 0, Page 157 of 157