ML021230234

Revision J JAFNPP B 2.0-5

SUMMARY

OF CHANGES TO ITS SECTION 3.0 - REVISION J Source of Change Summary of Change Affected Pages Consistency issue The ITS Bases LCO 3.0.3 words "Specification 3.0.3" have Specification 3.0 been changed to "LCO 3.0.3" to be consistent with its usage throughout this section.

NUREG Bases markup p B 3.0-4 Retyped ITS Bases p B 3.0-4 Consistency issue A statement in the Bases for ITS 3.0.4 has been deleted Specification 3.0 since it was in conflict with the philosophy of the ITS.

Specifically, the statement in the Bases said that even NUREG Bases markup p Insert if an SR was precluded form being performed during a Page B 3.0-10 certain Mode, it could be credited if the SR were unplanned.

Since the Bases cannot change the Technical JFD TA2 (JFDs p 2 of 3)

Specification requirements (which in this case it does),

this statement has been deleted.

Retyped ITS Bases p B 3.0-12 Editorial change A portion of the A DOC should have been deleted in the Specification 3.0 Revision D submittal.

Amendment 262 approved the change discussed in the second paragraph of the DOC, so it is DOC A12 (DOCs p 5 of 9) now in the CTS and is not added as part of the ITS submittal. Therefore, this second paragraph has been deleted.

TSTF-358 The changes approved in TSTF-358, concerning SR 3.0.3, Specification 3.0 have been made.

DOCs L3 and L4 (DOCs p 7 of

9. 8 of 9, and 9 of 9)

NSHCs L3 and L4 (NSHCs p 4 of 7, 5 of 7, 6 of 7. and 6 of 7)

NUREG ITS markup p 3.0-4 JFD TA4 (JFDs p 1 of 2)

NUREG Bases markup p B 3.0 13 and Insert Page B 3.0-13 Bases JFD TAB (Bases JFDs p 2 of 3)

Retyped ITS p 3.0-4 Retyped ITS Bases p B 3.0-15 and B 3.0-16 Page 1

DISCUSSION OF CHANGES ITS 3.0 - LCO AND SR APPLICABILITY ADMINISTRATIVE CHANGES Al In the conversion of the James A. FitzPatrick Nuclear Power Plant (JAFNPP) Current Technical Specification (CTS) to the proposed plant specific Improved Technical Specifications (ITS) certain wording preferences or conventions are adopted which do not result in technical changes.

Editorial changes, reformatting, and revised numbering are adopted to make the ITS consistent with the conventions in NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4,"

Revision 1 (i.e., Improved Standard Technical Specifications (ISTS)).

A2 CTS 3.0.A states that the LCOs and Actions shall be applicable during the OPERATIONAL CONDITIONS (modes) specified for each specification.

ITS LCO 3.0.1 replaces the CTS phrase "Limiting Conditions for Operation...shall be applicable..." with the phrase "LCOs shall be met..."

This change is made to be consistent with the format of other LCO 3.0 Specifications and with the concept of an LCO being met.

In addition ITS LCO 3.0.1 identifies specific exceptions to other LCO Applicabilities thus eliminating any interpretations that may be required, and avoiding any confusion.

These changes constitute editorial rewording, and presentation preferences consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, and are administrative.

A3 CTS 3.0.B states that the LCO is complied with if the Actions are completed (within the specified time interval) or if the LCO is restored prior to the time interval expiring.

ITS LCO 3.0.2 rewords the current requirement to be consistent with the format of other LCO 3.0 Specifications.

In addition ITS LCO 3.0.2 identifies specific exceptions to other LCO Applicabilities thus eliminating any interpretations that may be required, and avoiding any confusion.

These changes constitute editorial rewording, and presentation preferences consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, and are administrative.

A4 A phrase has been added to CTS LCO 3.0.C for clarity.

ITS LCO 3.0.3 includes the phrase "LCO 3.0.3 is only applicable in MODES 1, 2, and 3."

This phrase has been added since CTS provides no guidance in this area.

No further ACTIONS would be required to be performed if the plant were already in MODE 4 or 5 since CTS LCO 3.0.C only requires the plant to be placed in MODE 4.

This change constitutes editorial rewording, and presentation preferences consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, and is administrative.

Page 1 of 9 Revision J JAFNPP

DISCUSSION OF CHANGES ITS 3.0 -

LCO AND SR APPLICABILITY A5 Two CTS Surveillance Requirements, 4.0.A and 4.0.C have been combined to form ITS SR 3.0.1.

ITS SR 3.0.1 rewords the current requirements to be consistent with the format of other LCO 3.0 Specifications.

ITS SR 3.0.1 also adds clarifying words specifying that "failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO."

CTS implies that failure to meet the Surveillance means failure to meet the LCO, however ITS SR 3.0.1 provides this information in a clearer manner.

This change constitutes editorial rewording, and presentation preferences consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, and are administrative.

A6 CTS 4.0.B allows the Surveillance Frequency to be extended by 25% each Surveillance interval.

ITS SR 3.0.2 rewords the current requirement to be consistent with the format of other LCO 3.0 Specifications.

ITS 3.0.2 also adds the sentence "Exceptions to this Specification are stated in the individual Specifications," to acknowledge the explicit use of exceptions in various Surveillances.

The basic application of the 25% extension to routine Surveillances is maintained.

These changes constitute editorial rewording, and presentation preferences consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, and are administrative.

A7 When it is determined that a Surveillance Requirement has not been performed, CTS 4.0.C provides allowances for delay into the ACTIONS requirements for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for those specifications which include out of service times of less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

This allowance has been modified as described in Li.

CTS 4.0.C has been revised to explicitly state the required ACTIONS if the Surveillance is not performed within the delay period or if the Surveillance is performed within the delay period but it is not met.

The second paragraph of ITS SR 3.0.3 requires the LCO to be immediately declared not met, and the applicable Condition(s) to be entered if the Surveillance is not performed within the delay period.

The third paragraph requires these same actions when the Surveillance is performed within the delay period but is not met.

Since the actions are implied in CTS 4.0.C, this change is considered administrative.

This change is consistent with NUREG-1433, Revision 1.

Page 2 of 9 Revision J JAFNPP

DISCUSSION OF CHANGES ITS 3.0 - LCO AND SR APPLICABILITY A8 CTS 3.0.D does not permit entry into a MODE or other specified condition when an LCO is not met and the associated ACTION requires a shutdown if they are not met within a specified time interval.

Exceptions to these requirements are stated in the individual specifications.

ITS LCO 3.0.4 rewords the current requirement to be consistent with the format of other LCO 3.0 Specifications.

In addition, ITS LCO 3.0.4 states that the LCO is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3.

A review of the current and proposed Specifications has been performed to determine the affects of this allowance on the current and proposed Specifications.

The review has determined that this change does not provide any additional allowances to change MODES beyond those that currently exist, except where justified in individual Specifications (as described in the individual Specification's Discussion of Changes).

These changes constitute editorial rewording, and presentation preferences consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, and are administrative.

A9 CTS 3.0.F states that this LCO is an exception to LCO 3.0.B (ITS 3.0.2).

ITS LCO 3.0.5 includes these requirements and also adds clarifying words specifying that the exception to LCO 3.0.2 is for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

This clarification eliminates any interpretations that may be required, and avoids any confusion.

These changes constitute editorial rewording, and presentation preferences consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, and are administrative.

A1O ITS LCO 3.0.6 is added to provide guidance regarding the appropriate actions to be taken when a single inoperability (e.g., a support system) also results in the inoperability of one or more related systems (e.g.,

supported system(s)).

The existing Technical Specifications and various NRC guidance documents have not provided a consistent approach to the combined support/supported inoperability.

Guidance provided in the June 13, 1979, NRC memorandum from Brian K. Grimes (Assistant Director for Engineering and Projects) to Samuel E. Bryan (Assistant Director for Field Coordination)

Page 3 of 9 Revision J JAFNPP

DISCUSSION OF CHANGES ITS 3.0 -

LCO AND SR APPLICABILITY indicates an intent/interpretation consistent with the proposed LCO 3.0.6 without the necessity of also requiring the additional actions of a Safety Function Determination Program.

That is, only the inoperable support system actions need be taken.

Guidance provided by the NRC in their April 10, 1980, letter to all Licensees regarding the definition of Operability and the impact of a support system on the remainder of the Technical Specifications, indicates a similar philosophy of not taking actions for the inoperable supported equipment.

However, in this case, additional actions similar to the proposed Safety Function Determination Program actions, were addressed and required.

Generic Letter 91-18 and a literal reading of the existing Standard Technical Specifications provide the interpretation that failure to perform a required function, even as a result of an inoperable Technical Specification support system, requires all associated actions be taken.

Certain existing specifications contain actions such as: Declare the supported system inoperable and take the Actions of its specification.

In many cases the supported system would already be considered inoperable.

The implication of this presentation is that the actions of the inoperable supported system would not have been taken without the specific action to do so.

Considering the history of disagreement and misunderstandings in this area, the ISTS were developed with Industry input and approval of the NRC to include ITS LCO 3.0.6.

Since its function is to clarify existing ambiguities and maintain actions within the realm of previous interpretations, this new provision, consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, is deemed to be administrative in nature.

All A requirement has been added to CTS 3.0.C (ITS LCO 3.0.3) which requires entry into LCO 3.0.3 when directed by the associated ACTIONS.

This requirement is not included in the CTS since no specification explicitly directs entry into CTS 3.0.C.

Since the ITS also uses this method of entry into LCO 3.0.3 this statement must be included.

Changes to Specifications to explicitly require direct entry into LCO 3.0.3 (e.g.,

ITS 3.5.1) in the ITS if certain conditions are not met, are discussed in the Discussion of Changes for the specific Specification.

Therefore, Page 4 of 9 Revision J JAFNPP

DISCUSSION OF CHANGES ITS 3.0 -

LCO AND SR APPLICABILITY ADMINISTRATIVE CHANGES All (continued) this change constitutes a presentation preference consistent to NUREG 1433, Revision 1 and is considered administrative.

A12 CTS 4.0.D does not permit entry into a MODE or other specified condition when an LCO's Surveillances have not been met within the applicable interval or as otherwise stated.

ITS SR 3.0.4 rewords the current requirement to be consistent with the format of other LCO 3.0 Specifications.

In addition, ITS SR 3.0.4 states that the SR is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3.

This phrase has been added since CTS 4.0.D provides no guidance in this area.

The change eliminates the restrictions of the SR when in MODES 4 or 5.

Specific restrictions on MODE changes or Required Actions are included in the individual LCOs and discussed in the appropriate DOCs.

These changes constitute editorial rewording, and presentation preferences consistent with the BWR/4 ISTS, NUREG-1433, Revision 1, and are administrative.

Page 5 of 9 JAFNPP Revision J

DISCUSSION OF CHANGES ITS 3.0 -

LCO AND SR APPLICABILITY A13 Not Used.

TECHNICAL CHANGES

- MORE RESTRICTIVE M1 CTS 3.0.C requires the unit be placed in COLD SHUTDOWN (MODE 4) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the LCO or action requirements cannot be satisfied because of circumstances in excess of those addressed in the Specifications.

ITS LCO 3.0.3 requires that the plant take action within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to initiate the shutdown, be in MODE 2 in 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , be in MODE 3 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months , and be in MODE 4 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months (Li).

This change requires the plant to perform the shutdown in a controlled manner which will reduce the chances for a plant transient which could challenge safety systems.

Since this change requires the plant to take action within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and to be at interim conditions, MODE 2 in 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and MODE 3 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months , this change imposes additional time restraints on operations and therefore, is more restrictive.

The times are consistent with NUREG-1433, Revision

1.

This change has no adverse impact on safety.

M2 CTS 4.0.B does not address Frequencies specified as once.

ITS SR 3.0.2 includes the phrase "For Frequencies specified as "once," the above interval extension does not apply."

This is because the interval extension concept is based on scheduling flexibility for repetitive performance and these Surveillances are not repetitive in nature and essentially have no interval as measured from the previous performance.

This change precludes the ability to extend these performances consistent with NUREG-1433, Revision 1.

Since, CTS 4.0.B can be interpreted to apply the extension to all Surveillances, stating that the extension does not apply imposes additional requirements on operations and therefore, is more restrictive.

This change has no adverse impact on safety.

TECHNICAL CHANGES

. LESS RESTRICTIVE (GENERIC)

None Page 6 of 9 Revision J JAFNPP

DISCUSSION OF CHANGES ITS 3.0 - LCO AND SR APPLICABILITY TECHNICAL CHANGES

- LESS RESTRICTIVE (SPECIFIC)

Li CTS 3.0.C requires the unit to be placed in COLD SHUTDOWN (MODE 4) within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if the LCO or action requirements cannot be satisfied because of circumstances in excess of those addressed in the Specification.

ITS LCO 3.0.3 allows 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months to be in MODE 4 which includes the requirements (MU) to initiate the shutdown within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , be in MODE 2 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , and be in MODE 3 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months .

This change is considered less restrictive since the time to get to MODE 4 has increased by 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months (37 versus 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ).

This change is acceptable since the compensating actions added in accordance with M1 and this extended time to reach MODE 4 will ensure a more continuous reduction in power and reactor coolant temperature which is within the specified maximum cooldown rate and within the capabilities of the plant.

This reduces thermal stresses on components of the Reactor Coolant System and also reduces the chances for a plant transient which could challenge safety systems.

This change is consistent with NUREG-1433, Revision 1.

L2 CTS 4.0.B has had the following sentence added, "If a Completion Time requires periodic performance on a "once per..." basis, the above Frequency extension applies to each performance after the initial performance." ITS SR 3.0.2 includes this statement which provides the consistency in scheduling flexibility for all performances of periodic requirements, whether they are Surveillances or Required Actions.

The intent remains to perform the activity, on the average, once during each specified interval.

This change is consistent with NUREG-1433, Revision 1.

L3 When it is determined that a Surveillance was not performed, CTS 4.0.C allows ACTION requirements to be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit completion of the Surveillance if the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ITS SR 3.0.3 continues to allow a delay, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Surveillance Frequency, whichever is greater.

This change is less restrictive since the delay will now apply to any Surveillance instead of those specifications with ACTION requirements of less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The change is also less restrictive since it will allow a delay of greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , but this portion of the change is described in DOC L4.

The current dependance to the ACTION allowable outage time is considered not to be necessary since the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

Page 7 of 9 Revision J JAFNPP

DISCUSSION OF CHANGES ITS 3.0 -

LCO AND SR APPLICABILITY TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

(continued)

L4 CTS 4.0.C allows ACTION requirements to be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit completion of the Surveillance if the allowable outage times of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ITS SR 3.0.3 allows an increase in the delay time of up to the limit of the specified Surveillance Frequency.

However, a delay of greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is only allowed provided a risk evaluation is performed and the risk impact of delaying the Surveillance is managed.

The proposed change will not allow equipment known to be inoperable to be considered Operable until the missed Surveillance is performed.

If it is known that the missed Surveillance could not be met, SR 3.0.3 requires the affected LCO to be declared not met and the appropriate Condition(s) be entered.

In addition, the Bases for SR 3.0.3 state that the use of the delay period is not intended to be used as an operational convenience to extend Surveillance intervals, but only for the performance of missed Surveillance(s).

This change also includes Bases details on how to implement the new allowance.

The Bases provide guidance for Surveillance Frequencies that are not based on time intervals, but are based on specified plant conditions, operating conditions, or requirements of regulations.

In addition, the Bases state that the licensee is expected to perform any missed Surveillance at the first reasonable opportunity, taking into account appropriate considerations, such as impact on plant risk and analysis assumptions, consideration of plant conditions, planning, availability of personnel, and the time required to perform the Surveillance.

The Bases further states that the risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, Regulatory Guide (RG) 1.182, and that the missed Surveillance should be treated as an emergent condition as discussed in RG 1.182.

The degree of depth and rigor of the evaluation should be commensurate with the importance of the component.

Missed Surveillances for important components should be analyzed quantitatively.

If the results of the risk evaluation determine that the risk increase is significant, the evaluation should be used to determine the safest course of action.

All missed Surveillances will be placed in the licensee's Corrective Action Program.

Page 8 of 9 JAFNPP Revision J

DISCUSSION OF CHANGES ITS 3.0 -

LCO AND SR APPLICABILITY TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

L4 (continued)

Therefore, since the most probable result of performing a missed Surveillance is the verification of conformance with the requirements, and the risk of extending the performance of the missed Surveillance greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is controlled as discussed above, this change is considered acceptable.

The change is also consistent with the guidance provided in the Federal Register on September 28, 2001 (66 FR 49714),

which provided the notice of availability for licensees to incorporate TSTF-358, as modified, through the Consolidated Line Item Improvement Process.

TECHNICAL CHANGES RELOCATIONS None Page 9 of 9 Revision J JAFNPP

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0 - LCO AND SR APPLICABILITY TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

Li CHANGE The Licensee has evaluated the proposed Technical Specification change identified as "Technical Changes - Less Restrictive" and has determined that it does not involve a significant hazards consideration.

This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.

The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below.

1.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

This change allows a more gradual plant shutdown path than allowed by CTS 3.0.C.

Currently the plant has to be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ITS LCO 3.0.3 requires the plant to initiate action within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the plant in Mode 2 (Startup/Hot Standby) within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , Mode 3 (Hot Shutdown) within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months (MU) and Mode 4 (Cold Shutdown) within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

The overall time to Cold Shutdown is increased by 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months by the proposed change.

The proposed changes will require the shutdown to proceed in a more orderly and controlled manner.

This reduces thermal stresses on components of the reactor coolant system and the potential for a plant transient that could challenge safety systems under conditions to which this Specification applies.

This relaxation is also acceptable based on the small probability of an event requiring the inoperable Technical Specification structures, systems and components (SSCs) to function or variables to be maintained and the desire to minimize transients.

LCO 3.0.3 is only entered if the Action and Completion Time are not met and no other condition applies or if the condition of the plant is not specifically addressed by the associated actions.

It is the intent of the Technical Specifications to provide action provisions, where possible, to avoid the use of LCO 3.0.3 and subsequent plant shutdown.

The proposed changes to the overall shutdown Completion Times have no impact on any analyzed event.

The change will not allow continuous operation when SSCs are inoperable or parameter limits are not met.

In addition, the consequences of an event occurring during the proposed shutdown Completion Times are the same as the consequences of an event occurring during the existing Completion Times.

The proposed change to extend the time required to reach MODE 4 is less restrictive than present provisions; however, ITS LCO 3.0.3 will provide a more orderly plant shutdown sequence without involving a significant increase in the probability or consequences of an accident previously evaluated.

Page 1 of 7 Revision J JAFNPP

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0 -

LCO AND SR APPLICABILITY TECHNICAL CHANGES LESS RESTRICTIVE (SPECIFIC)

L1 CHANGE (contined)

2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change will not alter the plant configuration (no new or different type of equipment will be installed or removed) nor will the operation of the plant change.

The change still ensures the plant is placed in a specified Mode or condition in a timely manner.

Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does this change involve a significant reduction in a margin of safety?

The relaxation in the time allowed to reach MODE 4 in accordance with proposed LCO 3.0.3 represents a relaxation over the provisions in CTS 3.0.C.

However, this relaxation is acceptable based on the small probability of an event requiring the inoperable Technical Specification components to function or variables to be maintained and the desire to minimize transients.

LCO 3.0.3 is only entered if the Action and Completion Time are not met and no other condition applies or if the condition of the plant is not specifically addressed by the associated ACTIONS.

It is the intent of the Technical Specifications to provide action provisions, where possible, to avoid the use of LCO 3.0.3 and subsequent plant shutdown.

This change will not affect a margin of safety because it has no impact on the safety analysis assumptions.

The shutdown Completion Times specified in CTS 3.0.C or in ITS LCO 3.0.3 are not assumed in any analyzed accidents.

This proposed change and the compensatory actions added in accordance with M1 (to initiate action within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the plant in MODE 2 in 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and MODE 3 in 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months ) will enhance plant safety by requiring a more orderly plant shutdown while still requiring the plant to reach MODE 4 (Cold Shutdown) within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months of present provisions.

Therefore, the change will not involve a significant reduction in a margin of safety.

Page 2 of 7 Revision J JAFNPP

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0 -

LCO AND SR APPLICABILITY TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

L2 CHANGE New York Power Authority has evaluated the proposed Technical Specification change identified as "Technical Changes - Less Restrictive" and has determined that it does not involve a significant hazards consideration.

This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.

The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below.

1.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The application of the 25% extension to Required Action Completion Times which have a specified frequency on a periodic "once per" basis has been determined to not significantly degrade the reliability that results from performing the surveillance at a specified frequency.

As stated in Generic Letter 87-09, "The vast majority of surveillances do in fact demonstrate that systems or components are operable."

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The possibility of a new or different kind of accident from any accident previously evaluated is not created because the proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant.

3.

Does this change involve a significant reduction in a margin of safety?

The application of the 25% extension to Required Action Completion Times which have a specified frequency on a periodic "once per" basis has been determined to not significantly degrade the reliability that results from performing the surveillance at a specified frequency.

As stated in Generic Letter 87-09, "The vast majority of surveillances do in fact demonstrate that systems or components are operable."

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Page 3 of 7 JAFNPP Revision J

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0 -

LCO AND SR APPLICABILITY TECHNICAL CHANGES

- LESS RESTRICTIVE (SPECIFIC)

L3 CHANGE The Licensee has evaluated the proposed Technical Specification change ld_

identified as "Technical Changes - Less Restrictive" and has determined that it does not involve a significant hazards consideration.

This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.

The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below.

1.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

When it is determined that a Surveillance was not performed, CTS 4.0.C allows ACTION requirements to be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit completion of the Surveillance if the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ITS SR 3.0.3 continues to allow a delay, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater.

Changes to the times permitted to perform a Surveillance is not considered as an initiator of any design basis accident.

Therefore, this change does not significantly increase the probability of an accident previously analyzed.

The most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

The added time allowance is not considered to cause the component or subsystem to become inoperable or parameter to drift out of compliance.

Therefore, the consequences of an event occurring during this extended time period will be bounded by the current allowances.

Therefore, this change does not significantly increase the consequences of an accident previously evaluated.

2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The possibility of a new or different kind of accident from any accident previously evaluated is not created because the proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant.

3.

Does this change involve a significant reduction in a margin of safety?

When it is determined that a Surveillance was not performed, CTS 4.0.C allows ACTION requirements to be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit completion of the Surveillance if the allowable outage time limits of the ACTION requirements are less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ITS SR 3.0.3 continues to allow a delay, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to Page 4 of 7 Revision J JAFNPP

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0 -

LCO AND SR APPLICABILITY TECHNICAL CHANGES

- LESS RESTRICTIVE (SPECIFIC)

L3 CHANGE (continued) the limit of the specified Frequency, whichever is greater.

The most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

The added time allowance is not considered to cause the component or subsystem to become inoperable or a parameter to drift out of compliance.

Therefore, the consequences of an event occurring during this extended time period will be bounded by the current allowances.

Therefore, this change does not result in a significant reduction in a margin of safety.

Page 5 of 7 JAFNPP Revision J

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0 -

LCO AND SR APPLICABILITY TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

L4 CHANGE The Licensee has evaluated the proposed Technical Specification change identified as "Technical Changes - Less Restrictive" and has determined that it does not involve a significant hazards consideration.

This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.

The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below.

1.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change relaxes the time allowed to perform a missed Surveillance.

The time between Surveillances is not an initiator of any accident previously evaluated.

Consequently, the probability of an accident previously evaluated is not significantly increased.

The equipment being tested is still required to be Operable and capable of performing the accident mitigation functions assumed in the accident analysis.

As a result, the consequences of any accident previously evaluated are not significantly affected.

Any reduction in confidence that a standby system might fail to perform its safety function due to a missed Surveillance is small and would not, in the absence of other unrelated failures, lead to an increase in consequences beyond those estimated by existing analyses.

The addition of a requirement to assess and manage the risk introduced by the missed Surveillance will further minimize possible concerns.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or a change in the methods governing normal plant operation.

A missed Surveillance will not, in and of itself, introduce new failure modes or effects and any increased chance that a standby system might fail to perform its safety function due to a missed Surveillance would not, in the absence of other unrelated failures, lead to an accident beyond those previously evaluated.

The addition of a requirement to assess and manage the risk introduced by the missed Surveillance will further minimize possible concerns.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

Page 6 of 7 Revision J JAFNPP

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS 3.0 -

LCO AND SR APPLICABILITY TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

L4 CHANGE (continued)

3.

Does this change involve a significant reduction in a margin of safety?

The extended time allowed to perform a missed Surveillance does not result in a significant reduction in the margin of safety.

As supported by the historical data, the likely outcome of any Surveillance is verification that the LCO is met.

Failure to perform a Surveillance within the prescribed Frequency does not cause equipment to become inoperable.

The only effect of the additional time allowed to perform a missed Surveillance on the margin of safety is the extension of the time until inoperable equipment is discovered to be inoperable by the missed Surveillance.

However, given the rare occurrence of inoperable equipment, and the rare occurrence of a missed Surveillance, a missed Surveillance on inoperable equipment would be very unlikely.

This must be balanced against the real risk of manipulating the plant equipment or condition to perform the missed Surveillance.

In addition, parallel trains and alternate equipment are typically available to perform the safety function of the equipment not tested.

Thus, there is confidence that the equipment can perform its assumed safety function.

Therefore, this change does not involve a significant reduction in a margin of safety.

Page 7 of 7 Revision J JAFNPP

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.

Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO.

Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.

Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous 7performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per..

." basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is'IT. Iis delay(

period is permitted to allow performance of the Surveillance.f If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be BWR/4 STS Rev 1, 04/07/95 REVISIONZ IT P]

rvo0

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS: 3.0 - LCO AND SR APPLICABILITY RETENTION OF EXISTING REQUIREMENT (CLB)

None PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)

PAl Editorial changes have been made for enhanced clarity or to correct a grammatical/typographical error.

PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)

None DIFFERENCE BASED ON AN APPROVED TRAVELER (TA)

TA1 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 104, Revision 0, have been incorporated into the revised Improved Technical Specifications.

TA2 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 166, Revision 0, have been incorporated into the revised Improved Technical Specifications.

TA3 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 208, Revision 0, have been incorporated into the revised Improved Technical Specifications.

TA4 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 358, Revision 6, have been incorporated into the revised Improved Technical Specifications.

DIFFERENCE BASED ON A SUBMITTED, BUT PENDING TRAVELER (TP)

None Page 1 of 2 I

Revision J JAFNPP

LCO Applicability B 3.0 BASES LCO 3.0.3 A uý3 shutdown required in accordance with LCO 3.0.3 may be (continued) terminated and LCO 3.0.3 exited if any of the following occurs:

a.

The LCO is now met.

b.

A Condition exists for which the Required Actions have now been performed.

c.

ACTIONS exist that do not have expired Completion Times.

These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited.

-The time limits of

~

c* f 03a ow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for t

t M

to be in MODE 4 when a shutdown is required during

ýMOD o era ion e

is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies.

If a lower MODE is reached in less time than allowed, however, the total allowable time to reach MODE 4, or other applicable MODE, is not reduced.

For example, if MODE 2 is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , then the time allowed for reaching MODE 3 is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months , because the total time for reaching MODE 3 is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months .

Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, and 3, LCO 3.0.3 provides actions for Conditions not covered in other Specifications.

The requirements of LCO 3.0.3 do not apply in MODES 4 and 5 cause t

(

is already in the most restrictive Condition required by LCO 3.0.3.

The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the

/

ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Exceptionsto LCO 3.0.3 are provided in instances where requiring ashutdown, in accordance with LCO 3.0.3, would not rovide ropiate remedial measures for the associate condition o the An example of this is in LCO 3.7A, Spent Fuel Storage Pool Water Level.'

LCO 3.7.

has an pplicability of 'During movement of irradiated fuel "T3RT0 (continued)

BWR/4 STS B 3.0-4 Rev 1, 04/07/95 REVISIONO

INSERT B 3.0-4 Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR.

In this case, the unplanned event may be credited as fulfilling the performance of the SR.

Insert Page B 3.0-10 Revision J

SR Applicability B 3.0 BASES SR 3.0.3 (continued) period of up to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months sorp 0o te imit of'a e specified Frequency, whichever is applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

Chis delay period provides adequate time to complete CPO Surveillances that have been missed.

This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

)

The basis for this delay period includes consideration of f

conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the

)required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

W n a Surveillance with a Freq ncy based not on time it als, but upon specified uni conditions or operational situ ions, is discovered not to h ye been performed when specif d, SR 3.0.3 allows the full elay period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months "to perfo the Surveillance.

-A2 SR 3.0.3 al provides a time limit for mpletion of Surveillances hat become applicable as a onsequence of DE d b Required Actiot 1_

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence.

Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.

If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside-the specified limits and the Completion Times of the Required Acti6ns for the applicable LCO Conditions begin immediately upon expiration of the delay period.

If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the (continued)

BWR/4 STS B 3.0-13 Rev 1, 04/07/95 REVISION,.V J-

Pinsert SR 3.0.3-A When a ueillce with a Frequency based not on time intervals, but upon specified*

conditions, operating conditions, or requirements of regulations e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance.

However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.

SR 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

SInsert SR 3.0.3-B While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity.

The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and /

impact on any analysis assumptions, in addition to(govc-onditions, planning, Y1 availability of personnel, and the time required to perform the Surveillance.

This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance,(R@ý7Regulatory Guide 1.182, Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants.'s This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown.

The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide.

The risk evaluation may use quantitative, qualitative, or blended methods.

The degree of depth and rigor of the evaluation should be commensurate with the importance of the component.

Missed Surveillances for important components should be analyzed quantitatively.

If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action.

All missed Surveillances will be placed in the licensee's Corrective Action Program.

Insert Page B 3.0-13 Revision J

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.0 - LCO AND SR APPLICABILITY DIFFERENCE BASED ON AN APPROVED TRAVELER (TA)

TA1 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 71, Revision 2, have been incorporated into the revised Improved Technical Specifications.

TA2 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 8, Revision 2, have been incorporated into the revised Improved Technical Specifications.

However, TSTF-8 adds a clarification to the Bases of SR 3.0.1 that allows credit to be taken for unplanned events that satisfy Surveillances.

This clarification also states that this allowance also includes those SRs whose performance is precluded in a given MODE or other specified condition.

This portion of the TSTF has not been adopted.

As documented in the Part 9900 of the NRC Inspection Manual, Technical Guidance - Licensee Technical Specifications Interpretations, and in the ITS Bases Control Program (ITS 5.5.11), neither the Technical Specification Bases nor Licensee generated interpretations can be used to change the Technical Specification requirements.

Thus, if the Technical Specifications preclude performance of an SR in certain MODES (as is the case for some SRs in ITS Section 3.8), the Bases cannot change the Technical Specifications requirement and allow the SR to be credited for being performed in the restricted MODES, even if the performance is unplanned.

Therefore, only the first part of the TSTF-8 change to SR 3.0.1 has been adopted.

TA3 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 52, Revision 3, have been incorporated into the revised Improved Technical Specifications.

TA4 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 104, Revision 0, have been incorporated into the revised Improved Technical Specifications.

TA5 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 122, Revision 0, have been incorporated into the revised Improved Technical Specifications.

TA6 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 165, Revision 0, have been incorporated into the revised Improved Technical Specifications.

TA7 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 273, Revision 2, have been incorporated into the revised Improved Technical Specifications.

TA8 The changes presented in Technical Specification Task Force (TSTF)

Technical Specification Change Traveler number 358, Revision 6, have been incorporated into the revised Improved Technical Specifications.

However, minor changes have been made for consistency.

Page 2 of 3 JAFNPP Revision J

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.0 - LCO AND SR APPLICABILITY DIFFERENCE BASED ON A SUBMITTED, BUT PENDING TRAVELER (TP)

None DIFFERENCE FOR ANY REASON OTHER THAN THE ABOVE (X)

Xl The paragraph in ITS LCO 3.0.4 has been moved consistent with change package BWR-26, C.1.

This change was incorrectly inserted in the wrong position when NUREG-1433, Revision 1 was promulgated.

X2 ITS LCO 3.0.6 Insert B 3.0-2, provided in TSTF-71, Rev. 1, has been revised to include the sentence originally included in TSTF-71, Rev. 0.

The exclusion of this sentence was identified as a typographical error and is addressed by TSTF-71, Rev.

2.

X3 ITS SR 3.0.1 has been revised to reflect the value of Ž 800 psig consistent with ITS 3.1.4 (M6).

X4 The Bases for LCO 3.0.4 and SR 3.0.4 has been revised to reflect the possibility to enter MODE 2 from MODE 5 instead of from just MODES 3 or

4.

The plant can have the Reactor Mode Switch in Refuel and complete the tensioning of all reactor vessel head closure bolts and based on Table 1.1-1 the plant will be immediately in MODE 2 without passing into MODE 3 or 4.

Therefore, this modification simply corrects the Bases to reflect all possible ways of entering MODE 2.

Page 3 of 3 JAFNPP Revision J

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR)

APPLICABILITY SR 3.0.1 SR 3.0.2 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR.

Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO.

Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.

Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per.

" basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater.

This delay period is permitted to allow performance of the Surveillance.

A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

(continued)

Amendment (Rev. J) fi L SR 3.0.3 JAFNPP 3.0-4

SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 (continued)

When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

Entry into a MODE or other specified condition in the Applicability of an LCO shall not be made unless the LCO's Surveillances have been met within their specified Frequency.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with Actions or that are part of a shutdown of the plant.

SR 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, and 3.

Amendment (Rev.

J)

SR 3.0.4 JAFNPP 3.0-5

LCO Applicability B 3.0 BASES LCO 3.0.3 of the following occurs:

(continued)

a.

The LCO is now met.

b.

A Condition exists for which the Required Actions have now been performed.

c.

ACTIONS exist that do not have expired Completion Times.

These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited.

The time limits of LCO 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant to be in MODE 4 when a shutdown is required during MODE 1 operation.

If the plant is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies.

If a lower MODE is reached in less time than allowed, however, the total allowable time to reach MODE 4, or other applicable MODE, is not reduced.

For example, if MODE 2 is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , then the time allowed for reaching MODE 3 is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months , because the total time for reaching MODE 3 is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months .

Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, and 3, LCO 3.0.3 provides actions for Conditions not covered in other Specifications.

The requirements of LCO 3.0.3 do not apply in MODES 4 and 5 because the plant is already in the most restrictive Condition required by LCO 3.0.3.

The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Exceptions to LCO 3.0.3 are provided in instances where requiring a plant shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the plant.

An example of this is in LCO 3.7.7, "Spent Fuel Storage Pool Water Level."

LCO 3.7.7 has an Applicability of "During movement of irradiated fuel assemblies in the spent fuel storage pool."

Therefore, this LCO can be applicable in any or all MODES.

If the LCO and the Required Actions of LCO 3.7.7 are not met while in (continued)

Revision J JAFNPP B 3.0-4

SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications in Sections 3.1 through 3.10 and apply at all times, unless otherwise stated.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs.

This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits.

Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.

Systems and components are assumed to be OPERABLE when the associated SRs have been met.

Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

a.

The systems or components are known to be inoperable, although still meeting the SRs; or

b.

The requirements of the Surveillance(s) are known to be not met between required Surveillance performances.

Surveillances do not have to be performed when the plant is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified.

The SRs associated with a Special Operations LCO are only applicable when the Special Operations LCO is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR.

In this case, the unplanned event may be credited as fulfilling the performance of the SR.

(continued)

SRs JAFNPP B 3.0-12 Revision J

SR Applicability B 3.0 BASES (continued)

SR 3.0.3 I 2ý Revision J SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency.

A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

This delay period provides adequate time to complete Surveillances that have been missed.

This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of plant conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

When a Surveillance with a Frequency based not on time intervals, but upon specified plant conditions, operating conditions, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance.

However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.

SR 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence.

Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals.

While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity.

The determination of the first reasonable opportunity should (continued)

I d)

JAFNPP B 3.0-15

SR Applicability B 3.0 BASES SR 3.0.3 (continued) include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to plant conditions, planning, availability of personnel, and the time required to perform the Surveillance.

This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants."

This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown.

The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide.

The risk evaluation may use quantitative, qualitative, or blended methods.

The degree of depth and rigor of the evaluation should be commensurate with the importance of the component.

Missed Surveillances for important components should be analyzed quantitatively.

If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action.

All missed Surveillances will be placed in the licensee's Corrective Action Program.

If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period.

If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.

(continued)

Revision J B 3.0-16 SR 3.0.4 JAFNPP

SR Applicability B 3.0 BASES SR 3.0.4 This Specification ensures that system and component (continued)

OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the plant.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified condition change.

When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed per SR 3.0.1, which states that Surveillances do not have to be performed on inoperable equipment.

When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed.

Therefore, failing to perform the Surveillance(s) within the specified Frequency, on equipment that is inoperable, does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability.

However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes.

The provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS.

In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any plant shutdown.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary.

The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both.

This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance.

A Surveillance that could not be performed until after entering the LCO Applicability (continued)

Revision J JAFNPP B 3.0-17

SR Applicability B 3.0 BASES SR 3.0.4 (continued) would have its Frequency specified such that it is not "due" until the specific conditions needed are met.

Alternately, the Surveillance may be stated in the form of a Note as not required (to be met or performed) until a particular event, condition, or time has been reached.

Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

SR 3.0.4 is only applicable when entering MODE 3 from MODE 4, MODE 2 from MODE 3, 4, or 5, or MODE 1 from MODE 2.

Furthermore, SR 3.0.4 is applicable when entering any other specified condition in the Applicability only while operating in MODE 1, 2, or 3.

The requirements of SR 3.0.4 do not apply in MODES 4 and 5, or in other specified conditions of the Applicability (unless in MODE 1, 2, or 3) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Revision J JAFNPP B 3.0-18

SUMMARY

OF CHANGES TO ITS SECTION 3.1 -

REVISION J Source of Change Summary of Change Affected Pages Retyped ITS typographical Minor typographical errors in the retyped ITS have been Specification 3.1.4 errors corrected to be consistent with the NUREG ITS markup.

(The scram time to notch position 46 has been changed Retyped ITS p 3.1-14 from 0.441 to 0.44: a comma has been added to the forth line of SR 3.1.7.6; and the wording of SR 3.1.7.11 has Specification 3.1.7 been modified.)

Retyped ITS p 3.1-22 NUREG ITS markup error A minor NUREG markup error has been corrected to be Specification 3.1.3 consistent with the retyped ITS.

(A period has been added to Required Action A.1.)

NUREG ITS markup p 3.1-7 Retyped ITS Bases Minor typographical errors in the retyped ITS Bases have Specification 3.1.1 typographical errors been corrected to be consistent with the NUREG Bases markup.

(The words "shuffling within the reactor Retyped ITS Bases p B 3.1-5 pressure vessel," and the word "or" has been changed to and B 3.1-6 "and" in SR 3.1.1.1; a comma has been added to a sentence in the ITS 3.1.3 Applicable Safety Analyses Specification 3.1.3 section: the words "THERMAL Power" have been changed to "THERMAL POWER" in SR 3.1.3.2 and SR 3.1.3.3; the Retyped ITS Bases p B 3.1-14 apostrophe has been deleted from the ITS 3.1.4 Reference and B 3.1-20 6 word "Owners'"; the comma has been moved inside the end quote mark in ITS 3.1.5 Applicable Safety Analyses Specification 3.1.4 section; and the term "BWR16" has been changed to "BWR/6" in ITS 3.1.6 Reference 3.)

Retyped ITS Bases p B 3.1-28 Specification 3.1.5 Retyped ITS Bases p B 3.1-29 Specification 3.1.6 Retyped ITS Bases p B 3.1-38 NUREG Bases markup errors Minor NUREG markup errors have been corrected to be Specification 3.1.1 consistent with the retyped ITS Bases.

(The word "or" has been deleted from the SR 3.1.1.1 section; the proper NUREG Bases markup p B 3.1-5 punctuation marks have been added to various References and B 3.1-6 in ITS 3.1.1, 3.1.2, 3.1.3. and 3.1.4; and the words "of the" have been deleted from the ITS 3.1.4 LCO section.)

Specification 3.1.2 NUREG Bases markup p B 3.1 12 Specification 3.1.3 NUREG Bases markup p B 3.1 21 Specification 3.1.4 NUREG Bases markup p B 3.1 23 and B 3.1-28 Page 1

SUMMARY

OF CHANGES TO ITS SECTION 3.1 - REVISION J Source of Change Summary of Change Affected Pages Typographical errors Minor typographical errors have been corrected in the Specification 3.1.3 NUREG Bases markups and retyped ITS Bases.

(A comma has been deleted from the ITS 3.1.3 Background section: and NUREG Bases markup p B 3.1 the name "Lianes" has been change to "Lianas" in ITS 13 3.1.6 Reference 2.)

Retyped ITS Bases p B 3.1-13 Specification 3.1.6 NUREG Bases markup p B 3.1 37 Retyped ITS Bases p B 3.1-38 Consistency issue Minor consistency issue corrections have been made. (The Specification 3.1.1 revision and date of certain Bases References, which are also identified in the COLR, have been deleted and a NUREG Bases markup p B 3.1-6 statement that the Revision is specified in the COLR has and B 3.1-7 been added.

This is consistent with Specification 5.6.5.)

Retyped ITS Bases p B 3.1-7 Specification 3.1.4 NUREG Bases markup p B 3.1 28 Retyped ITS Bases p B 3.1-28 Specification 3.1.6 NUREG Bases markup p B 3.1 37 Retyped ITS Bases p B 3.1-38 Consistency issue The word "solution" in the second Frequency of SR Specification 3.1.7 3.1.7.9 has been changed to "piping" for consistency with the Bases.

SR 3.1.7.9 requires a verification that DOC M2 (DOCs p 1 of 7) the heat traced piping between the storage tank and the pump suction is unblocked.

Thus, the Frequency should NUREG markup p 3.1-22 be keyed to when the piping temperature is restored, not to when the storage tank temperature is restored.

The JFD PAl (JFDs p 1 of 2)

NUREG Bases states it is the piping temperature that is of concern.

This change is also consistent with the Retyped ITS p 3.1-22 most recently approved BWR ITS conversions (LaSalle and NMP2 ITS).

Consistency issues The word "boron" in SR 3.1.7.5, including the second Specification 3.1.7 Frequency, has been changed to "sodium pentaborate" for consistency with the similar words in SR 3.1.7.1, SR NUREG ITS markup p 3.1-21 3.1.7.2. SR 3.1.7.10, and SR 3.1.7.11.

Also, the Bases for SR 3.1.7.9 has been modified to reflect the current JFD PA2 (JFDs p 1 of 2) manner in which the Surveillance is being performed (pumping from the storage tank and recirculating it back NUREG Bases markup p B 3.1 to the storage tank).

45 Retyped ITS p 3.1-21 Retyped ITS Bases p B 3.1-44 Page 2

SUMMARY

OF CHANGES TO ITS SECTION 3.1 - REVISION J Page 3 Source of Change Summary of Change Affected Pages Consistency issues For consistency and clarity throughout the ITS, each of Specification 3.1.8 the two Scram Discharge Volumes has been described as having a header and an instrument volume.

Also, the NUREG Bases markup p 3.1-47 Bases for SR 3.1.8.3, describing why the Frequency is and B 3.1-51 acceptable, has been modified to be consistent with the NUREG (i.e., the words "and analysis" have been deleted Retyped ITS Bases p B 3.1-46 and the word "usually" has been added).

and B 3.1-50 Editorial The proper References/Reference number has been Specification 3.1.1 provided.

NUREG Bases markup p B 3.1 5, B 3.1-6, B 3.1-7 Retyped ITS Bases p B 3.1-6 and B 3.1-7 Specification 3.1.4 NUREG Bases markup p B 3.1 28 Retyped ITS Bases p B 3.1-34 Specification 3.1.6 NUREG Bases markup p B 3.1 37 Retyped ITS Bases p B 3.1-38 Editorial The description of the plant specific reduced notch Specification 3.1.6 worth procedure (RNWP) has been deleted from the Applicable Safety Analyses section, since the RNWP is a NUREG Bases markup p B 3.1 subset of the BPWS.

Thus the description is not 35, Insert Page B 3.1-35 necessary.

A Reference has been renumbered due to this (deleted), and Insert Page B deletion.

3.1-38 JFD CLB1 (deleted) (JFDs p 1 of 2)

Retyped ITS Bases p B 3.1-35 and B 3.1-38

SUMMARY

OF CHANGES TO ITS SECTION 3.1 -

REVISION J Page 4 Source of Change Summary of Change Affected Pages Technical change The second Frequency for SR 3.1.2.1 has been changed Specification 3.1.2 from "Each full power month..." to "1000 MWD/T" to be consistent with the NUREG wording.

An appropriate DOC CTS markup p 1 of 1 and NSHC has been provided.

DOC L3 (DOCs p 4 of 4)

NSHC L3 (NSHCs p 5 of 5)

NUREG ITS markup p 3.1-6 JFD CLB1 (deleted) (JFDs p 1 of 1)

NUREG Bases markup p B 3.1 12 Bases JFD CLB1 (deleted)

(Bases JFDs p 1 of 1)

Retyped ITS p 3.1-6 Retyped ITS Bases p B 3.1-12 Technical change The proper normal water pressure of the CRD accumulators Specification 3.1.5 has been provided.

A Reference concerning this pressure has also been added.

NUREG Bases markup p B 3.1 33 JFD DB1 (JFDs p I of 1)

Retyped ITS Bases p B 3.1-33 Technical change The Discussion of Changes have been modified to properly Specification 3.1.7 justify deleting the requirement to replace all (i.e.,

both) SLC System explosive primer assemblies every 24 CTS markup p 2 of 5 months.

The NUREG and ITS only require firing and replacing one of the two primer assemblies every 24 DOCs M5.

LAI, and L5 (DOCs p months.

3 of 7 and 7 of 7)

NUREG Bases markup p B 3.1 45 Retyped ITS Bases p B 3.1-43

SDM B 3.1.1 6D© ACTIONS E.1. E.2. E.3, E.4, and E.5 (continued)

Tke a*,*rjVkW eoAvos assumed to~

isolate~to mitigate radioactivity releases.-*

C.

-S his may be performed as an administrative check, by 0ý to -f d por

,y#j wexamining logs or other information, to determine if the components are out of service for maintenance or other reasons.

It is not necessary to perform the Surveillances as needed to demonstrate the OPERABILITY of the components.

-4&c

.o4-*,(k b-4If, however, any required component is inoperable, then it

s~oh v ctc~.A".

must be restored to OPERABLE status.

In this case, SRs may need to be performed to restore the component to OPERABLE d,(

status.

Action must continue until all required components l'<-S

(

'e OPERABLE.

irk>">,

C1,1 SURVEILLANCE SR 3.1.1J..1 REQUIREMENTS SAdequate SDM must be to ensure that the reactor can be made subcritica T o iny initial operating con io.

Adequate SDM is demonstratedby testin before or urin the first startup after fuel movemen contro rod es61esse on ro rod replacement refers o cupling

4.f.-"Ldrand removal of a control rod from a core location, and Ssubsequent replacement with a new control rod or a control o

,,M6; rod from another core location.

Since core reactivity will 5F Ivary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also

/J44*

iatlP account for changes in core reactivity during the cycle.

Therefore, to obtain the Sod, the initial measured valueOp-)

incrmust be rased by an adder, OR, which is the difference

[,

1-"

A between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core 4/,reactivity. If the value of R is negative tha is BOC is the most reactive point in the cycle) o correc ion o -lei

~L ighest rth control rod,' dditional argin

.0% Ak/k)

The SDM may.be.demonstrated during an inequence control I

rod withdrawal n

i hcW the highest worth control rod is analytically determinedJi ourn oca cra ca (continued)

Rev 1, 04/07/95 fp A RAIZ174Z BWR/4 STS B 3.1-5

SiM B 3.1.1 BASES SURVEILLANCE REQUIREMENTS SR 3.1.1.1 (continued)

Local critical tests require the withdrawal of out of sequence control rods.

This testing would therefore require bypassing of the rod worth minimizer to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LCO 3.10.7, "Control Rod Testing-Operating".*--

Lejo 4 /0i'P~

T3*

JA' The Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.

ýDuring MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals.

An evaluation of each in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling.

This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern.

For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence.

These bounding analyses include additional margins to the associated uncertainties.

Spiral offload/reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle.

Removing fuel from the core will always result in an increase in SDO.

(at.?f)

(

L U

N..

REFERENCES 1.

~10 Crr r

tt$, A I.,uh A, fBDCZG.>

~AR, Section RE 44.,a-~i~i..

NEDE-24011-P-A e

Gner~alElectric Stanpdard Application foreacto Fuel *Supple~ment for United States6' Section'S.2.2.3.1,

'ffSAR, Section 5*. v ioFSo0.

G7(j(

(continued)

Rev 1, 04/07/95 BWR/4 STS B 3.1-6

1DM B 3.1.1 BASES

-SAR, Sectiq [4.3.2.4.1].)

1,IEt24011-P-A

'*General E*ectric Standard Application for Reactor Fuel,t Section 3.2.4.1, (4,

-'k%og sr&4ir Rev 1, 04/07/95 BWR/4 STSB B 3.1-7

SDM B 3.1.1 BASES ACTIONS E.1, E.2, E.3, E.4, and E.5 (continued) least one SGT subsystem is OPERABLE; and secondary containment isolation capability is available in each associated secondary containment penetration flow path not isolated that is assumed to isolate to mitigate radioactivity releases (i.e., at least one secondary containment isolation valve and associated instrumentation are OPERABLE, or acceptable administrative controls assure isolation capability.

These administrative controls consist of stationing a dedicated operator, who is in continuous communication with the control room, at the controls of the isolation device.

In this way, the penetration can be rapidly isolated when a need for secondary containment isolation is indicated).

This may be performed as an administrative check, by examining logs or other information, to determine if the components are out of service for maintenance or other reasons.

It is not necessary to perform the Surveillances as needed to demonstrate the OPERABILITY of the components.

If,

however, any required component is inoperable, then it must be restored to OPERABLE status.

In this case, SRs may need to be performed to restore the component to OPERABLE status.

Action must continue until all required components are OPERABLE.

SURVEILLANCE SR 3.1.1.1 REQUIREMENTS Adequate SDM must be verified to ensure that the reactor can be made subcritical from any initial operating condition with the highest reactivity worth control rod fully withdrawn and all other control rods fully inserted.

This can be accomplished by a test (by withdrawing control rods),

an evaluation, or a combination of the two.

Adequate SDM is demonstrated by testing before or during the first startup after fuel movement, shuffling within the reactor pressure vessel, or control rod replacement.

Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location.

Since core reactivity will vary during the cycle as a function of fuel depletion and poison burnup, the beginning of cycle (BOC) test must also account for changes in core reactivity during the cycle.

Therefore, to obtain the SDM, the initial measured value must be increased by an adder, "R",

which is (continued)

Revision J JAFNPP B 3.1-5

SDM B 3.1.1 BASES SURVEILLANCE SR 3.1.1.1 (continued)

REQUIREMENTS the difference between the calculated value of maximum core reactivity during the operating cycle and the calculated BOC core reactivity.

If the value of R is negative (that is, BOC is the most reactive point in the cycle), no correction to the BOC measured value is required (Ref. 6).

The SDM may be demonstrated during an in-sequence control rod withdrawal or during local criticals. In both cases, the highest worth control rod is analytically determined.

Local critical tests require the withdrawal of out of sequence control rods.

This testing would therefore require bypassing of the rod worth minimizer to allow the out of sequence withdrawal, and therefore additional requirements must be met (see LCO 3.10.7, "Control Rod Testing-Operating" and LCO 3.10.8, "SHUTDOWN MARGIN Test - Refueling").

The Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reaching criticality is allowed to provide a reasonable amount of time to perform the required calculations and have appropriate verification.

During MODES 3 and 4, analytical calculation of SDM may be used to assure the requirements of SR 3.1.1.1 are met.

During MODE 5, adequate SDM is required to ensure that the reactor does not reach criticality during control rod withdrawals.

An evaluation of each in-vessel fuel movement during fuel loading (including shuffling fuel within the core) is required to ensure adequate SDM is maintained during refueling.

This evaluation ensures that the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern.

For example, bounding analyses that demonstrate adequate SDM for the most reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel movement sequence.

These bounding analyses include additional margins to the associated uncertainties.

Spiral offload/reload sequences inherently satisfy the SR, provided the fuel assemblies are reloaded in the same configuration analyzed for the new cycle.

Removing fuel from the core will always result in an increase in SDM.

(continued)

Revision J JAFNPP B 3.1-6

SDM B 3.1.1 BASES (continued)

REFERENCES

1.

UFSAR, Section 16.6.

2.

UFSAR, Section 14.6.1.2.

3.

NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, Supplement for United States, Section S.2.2.3.1, (Revision specified in the COLR).

4.

UFSAR, Section 14.5.4.3.

5.

6.

NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, Section 3.2.4.1, (Revision specified in the COLR).

7.

UFSAR, Section 14.5.4.4.

Revision J JAFNPP B 3.1-7

3.: 1-..27 JAFNPP Ol Reactivity Anomalies T

the difference between the actual During caitical rod configuration and the xpected configuration during shall not exceed 1 percent

k. If this limit Ist.i expectdd donliguratkol shuAdown shall be D above cannot be met, and the reactor shau be I 5<

Amendment No. 155 97

/9 q /

REVISION,/r e4,-f; -Aj

3. 1. ?L 2 It

'Pt. /

q'O a CMI 4M U1

DISCUSSION OF CHANGES ITS: 3.1.2 -

REACTIVITY ANOMALIES TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

(continued)

L2 The CTS 3.3.E requirement to be in a cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if CTS 3.3.D cannot be met, is being deleted.

This deletion is acceptable since ITS 3.1.2 ACTION B (M2) requirement to be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , if the Required Action and associated Completion Time are not met, has been added, which places the plant in a Condition outside the ITS 3.1.2 (CTS 3.3.D) Applicability.

In MODE 3 all control rods are fully inserted and therefore the reactor is in the least reactive state, where measuring core reactivity is not necessary, a continuation to cold shutdown (MODE 4) will not reduce core reactivity and therefore also is not necessary.

In addition, if the reactivity anomaly specification is not met and if ITS 3.1.1 SHUTDOWN MARGIN (SDM) cannot be met, entry into the appropriate ITS 3.1.1 ACTION is required.

In MODE 3, primary and secondary containment OPERABILITY is required, therefore adequate protection exists if a reactivity anomaly were to occur.

The requirements of ITS 3.1.2 provide adequate protection and therefore is considered acceptable.

This change is consistent with NUREG-1433, Revision 1.

L3 CTS 4.3.D requires the shutdown margin to be verified within limits every full power month (approximately 708 MWD/T).

In ITS SR 3.1.2.1, the Frequency has been changed to every "1000 MWD/T during operations in MODE 1" after the first performance (the first performance requirement is described in DOC M3).

Both Frequencies consider the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity.

The proposed change is consistent with the BWR Standard Technical Specification, NUREG-1433, Revision 1.

TECHNICAL CHANGES - RELOCATIONS None Page 4 of 4 JAFNPP Revision J

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS: 3.1.2 - REACTIVITY ANOMALIES TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

Li CHANGE New York Power Authority has evaluated the proposed Technical Specification change identified as "Technical Changes

- Less Restrictive" and has determined that it does not involve a significant hazards consideration.

This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.

The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below.

1.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change would allow 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to evaluate and determine the cause of any reactivity anomalies prior to requiring a plant shutdown.

Such a reactivity anomaly is not considered an initiator of any accident previously evaluated and therefore would not affect their probability.

Substantial margin exists in the analyses which predict core reactivity and in those which analyze the accidents.

In addition, adequate shutdown margin is demonstrated by test during plant startup after in vessel fuel movement or control rod replacement and is followed by a reactivity anomaly test within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of reaching equilibrium conditions at greater than 75% rated thermal power.

Since the first reactivity anomaly test is typically performed within a few days following the shutdown margin demonstration the reactivity difference between the measured and predicted rod density is expected to be small.

Based on experience, the reactivity differences determined by periodic performance of reactivity anomaly tests are also expected to be small, slow developing and insignificant with respect to the probability or consequences of accidents previously evaluated.

Further, the consequences of an event occurring during the proposed 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time are the same as the consequences of an event occurring under the current Actions.

Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation.

The proposed change will only provide a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time to restore the core reactivity difference to within limits before requiring a plant shutdown.

Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

JAFNPP Page 1 of 5 Revision A

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS: 3.1.2 - REACTIVITY ANOMALIES TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

Li CHANGE

3.

Does this change involve a significant reduction in a margin of safety?

The proposed 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time to restore core reactivity difference to within limits prior to requiring a plant shutdown is acceptable based on the small probability of an event occurring during this time period.

Further, reactivity anomaly conditions develop slowly so there will not be a substantial change in the anomaly during the longer allowed interval before plant shutdown.

Any minor decrease in the margin of safety during the additional time is offset by minimization of the potential for plant transients which may occur while shutting down the plant.

Therefore, this change does not involve a significant reduction in a margin of safety.

Page 2 of 5 JAFNPP Revision A

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS: 3.1.2 - REACTIVITY ANOMALIES TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

L2 CHANGE New York Power Authority has evaluated the proposed Technical Specification change identified as "Technical Changes - Less Restrictive" and has determined that it does not involve a significant hazards consideration.

This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.

The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below.

1.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change deletes the requirement to be in a cold condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the Reactivity Anomalies Specification of CTS 3.3.D is not met.

Placing the plant in a cold condition does not place the plant in a less reactive condition.

The reactor core is more reactive at colder temperatures, therefore the requirement to be in a cold condition does not decrease significance of the reactivity anomaly.

The new requirement (M2) will be to be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (ITS 3.1.2 Required Action B.1).

The proposed action is considered sufficient when Reactivity Anomalies is not met.

The requirement to be in a cold condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if Reactivity Anomalies is not met is not considered in the initiation of any accident.

Therefore, this change does not significantly increase the probability of any accident previously evaluated.

The proposed ACTION exits the Applicability of the LCO and limits core reactivity.

In Mode 3 the primary and secondary containment are required to be OPERABLE to limit the consequences of any design bases accident.

Thus, the consequences of an accident will not be increased as a result of this change.

Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change will not involve a physical alteration of the plant (no new or different type of equipment will be installed) or changes in methods governing normal plant operation.

The proposed change limits core reactivity and other Specifications will provide additional requirements to ensure sufficient components are OPERABLE to limit any radioactivity release if an event were to occur.

Therefore, this change will not create the possibility of a new or different type of accident from any accident previously analyzed.

Page 3 of 5 JAFNPP Revision A

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS: 3.1.2 - REACTIVITY ANOMALIES TECHNICAL CHANGES

- LESS RESTRICTIVE (SPECIFIC)

L2 CHANGE

3.

Does this change involve a significant reduction in a margin of safety?

The proposed change deletes the requirement to be in a cold condition in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the Reactivity Anomalies Specification of CTS 3.3.D is not met.

Placing the plant in a cold condition does not place the plant in a less reactive condition.

The reactor core is more reactive at colder temperatures, therefore the requirement to be in a cold condition does not decrease significance of the reactivity anomaly.

The new requirement (M2) will be to be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (ITS 3.1.2 Required Action B.1).

The proposed ACTIONs are considered sufficient when Reactivity Anomalies is not met.

The proposed action limits core reactivity and exits the Applicability of the LCO.

In Mode 3 the primary and secondary containment are required to be Operable to limit the consequences of any design bases accident.

Thus, the consequences of an accident will not be increased as a result of this change.

Deleting this requirement to be in a cold condition when Reactivity Anomalies is not met will effectively decrease the core reactivity.

This change will not impact any safety analysis assumptions.

As such, no question of safety is involved.

Therefore, this change does not involve a significant reduction in a margin of safety.

Page 4 of 5 JAFNPP Revision A

NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS: 3.1.2 - REACTIVITY ANOMALIES TECHNICAL CHANGES

- LESS RESTRICTIVE (SPECIFIC)

L3 CHANGE The Licensee has evaluated the proposed Technical Specification change identified as "Technical Changes - Less Restrictive" and has determined that it does not involve a significant hazards consideration.

This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92.

The bases for the determination that the proposed change does not involve a significant hazards consideration are discussed below.

1.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed change revises the Surveillance Frequency for the verification of the reactivity difference between the measured rod density and the predicted rod density.

The proposed change continues to provide assurance that plant operation is maintained within the assumptions of the DBA and transient analysis.

The proposed change in Frequency does not involve the operation of, or change to, any equipment assumed to be an initiator for any analyzed accidents.

Therefore, this change will not significantly increase the probability or consequences of an accident previously evaluated.

2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed change does not involve a physical modification to the plant and does not introduce a new mode of operation.

Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does this change involve a significant reduction in a margin of safety?

The extension in the surveillance test interval is insignificant given that the proposed Frequency considers the relatively slow change in core reactivity with exposure, and operating experience related to variations in core reactivity.

The proposed change does not impact the ability of the equipment to maintain the plant within acceptable limits.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Page 5 of 5 Revision J JAFNPP

Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2.1 Verify core reactivity edifferenceg between Once within the I i

rod density and the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after predicted rod densityg is within +/- 1%,Ak/k.

reaching equilibrium conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 1000 MWD/T thereafter during operations in MODE 1 Rev 1, 04/07/95 BWR/4 STS L4. 3. Dý 3.1-6 ACIV ýIekT

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS: 3.1.2 - REACTIVITY ANOMALIES RETENTION OF EXISTING REQUIREMENT (CLB)

None PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)

PAl Changes have been made (additions, deletions and/or changes to the NUREG) to reflect the plant specific system/structure/component nomenclature, equipment identification or description.

PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)

None DIFFERENCE BASED ON AN APPROVED TRAVELER (TA)

None DIFFERENCE BASED ON A SUBMITTED, BUT PENDING TRAVELER (TP)

None DIFFERENCE FOR ANY REASON OTHER THAN THE ABOVE MX)

None Page 1 of 1 Revision J JAFNPP

Reactivity Anomalies B3.S.

BASES SURVEILLANCE REQUIREMENTS SR 3.1.2.1 (continued) rod from another core location.

Also, core reactivity changes during the cycle.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the gonstore and predicted rod density can be made.

For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes) at k 75% RTP have been obtained.

The 1000 MWD/T Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity.

This comparison requires the core to be operating at power levels which minimize the uncertainties and measurement errors, in order to obtain meaningful results.

Therefore, the comparison is only done when in MODE 1.rA Rev 1, 04/07/95

[ nx BWR/4 STS B 3.1-12 AQ4 (410A)-J'-

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.1.2 - REACTIVITY ANOMALIES RETENTION OF EXISTING REQUIREMENT (CLB)

None 1%

PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)

PAl Editorial changes have been made for enhanced clarity or to correct a grammatical/typographical error.

PA2 Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific system/structure/component nomenclature, equipment identification or description.

PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)

DB1 JAFNPP was designed and under construction prior to the promulgation of Appendix A to 10 CFR 50 - General Design Criteria for Nuclear Power Plants.

The JAFNPP Construction Permit was issued on May 20, 1970.

The proposed General Design Criteria (GDC) were published in the Federal Register on July 11, 1967 (32 FR 10213) and became effective on February 20, 1971 (32 FR 3256).

UFSAR, Section 16.6 - Conformance to AEC Design Criteria, describes the JAFNPP current licensing basis with regard to the GDC.

ISTS statements concerning the GDC are modified in the ITS to reference UFSAR, Section 16.6.

The brackets have been removed from the Reference and the proper plant specific reference included.

DB2 ITS 3.1.2 has been revised to reflect the specific JAFNPP reference requirements of, UFSAR, Chapter 14.

DIFFERENCE BASED ON AN APPROVED TRAVELER (TA)

None DIFFERENCE BASED ON A SUBMITTED, BUT PENDING TRAVELER (TP)

None DIFFERENCE FOR ANY REASON OTHER THAN THE ABOVE MX)

X1 NUREG-1433, Revision 1, Bases references to "the NRC Policy Statement" has been replaced with 10 CFR 50.36(c)(2)(ii), in accordance with 60 FR 36953 effective August 18, 1995.

Page 1 of 1 JAFNPP Revision J

Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1.2.1 Verify core reactivity difference between the measured rod density and the predicted rod density is within +/- 1% Ak/k.

1*

FREQUENCY 1*

Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching equilibrium conditions following startup after fuel movement within the reactor pressure vessel or control rod replacement AND 1000 MWD/T thereafter during operations in MODE 1 Amendment (Rev.

J) 3.1-6 JAFNPP

Reactivity Anomalies B 3.1.2 BASES SURVEILLANCE SR 3.1.2.1 (continued)

REQUIREMENTS another core location.

Also, core reactivity changes during the cycle.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the measured and predicted rod density can be made.

For the purposes of this SR, the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes) at k 75% RTP have been obtained.

The 1000 MWD/T Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating experience related to variations in core reactivity.

This comparison requires the core to be operating at power levels which minimize the uncertainties and measurement errors, in order to obtain meaningful results.

Therefore, the comparison is only done when in MODE 1.

The tests performed at this Frequency also use base data obtained during the first test of the specific cycle.

REFERENCES

1.

UFSAR, Section 16.6.

2.

UFSAR, Chapter 14.

3.

Revision J JAFNPP B 3.1-12

Control Rod OPERABILITY 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Control Rod OPERABILITY 33.A

[?A 1 LCO 3.1.3 (All APPLICABILITY:

ACTIONS Each control rod shall be OPERABLE.

MODES 1 and 2.

NOTE-Separate Condition entry is allowed for each control rod.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One withdrawn control rod stuck.

[,41]

--- -----N O T E - - - - - - - - -

Rod worth minimizer (RWM) may be bypassed as allowed by LCO 3.3.2.1, "Control Rod Block Instrumentation," if required, to allow continued operation.

Disarm the associated control rod drive (CRD).

AND 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (continued)

.K*p=r=t^

3.1-7 REVISION CA 1]

týý FrA fe %

41']

P Al --e C' tP

Control Rod OPERABILITY B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod OPERABILITY BASES BACKGROUND Control rods are components of th ij&ptro Qi~~iiive (CRD)

System, which is the primary react'l*ty co 'AoFystem for the reactor. In conjunction with the Reactor Protection SysemtheCRD System provides the means for the reliable a6?0V'At 6.A ontrol of reactivity changes to ensure under conditions of C nomloeain including s, that specified acceptable fuel design limits are not exceeded. In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to-limit the potential amount and rate of reactivity increase caused by a malfunction in the CR0 frA e %I System. The CRD System is designed to-satisfy the.

(SR

~requirements, and_29__

)

The CR0 System consists of 137 locking piston Mrrrral an CR(s and a hydraulic control unit for LAI i~iI~ft~TttSII.The locking piston type CR011' is a Idouble aacting hyraulic piston, which uses condensate water as the operating fluid. Accumulators provide additional energy for scram. An index tube and piston, coupled to the control rod, are locked at fixed increments by a collet mechanism. The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion.

This Specification, along with LCO 3.1.4, "Control Rod Scram

-Times," and LCO 3.1.5, wControl Rod Scram Accumulators,"

ensure that the performance of the control rods in the event of a Design Basis Accident (DBA) or transient meets the assumptions used in the safety analyses of References.

APPLICABLE SAFETY ANALYSES evalu ions inv ving *con s re prese ted in)

Refe ences 2.

and 4 The contro ro ss provide the r primary means fdor rapid reactivity control (reactor scram),

for maintaining the reactor subcritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CR0 System.

(continued)

B 3.1-13 REVISION Pr

)

BWR SJý:S:!D

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE SR 3.1.3.5 (continued)

REQUIREMENTS affect coupling.

This includes control rods inserted one notch and then returned to the "full out" position during the performance of SR 3.1.3.2.

This Frequency is acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved and operating experience related to uncoupling events.

REFERENCES

1.

5n i-,WAen cix2-D28C 2.ýAAR, Section.2 M AR, Section 4.~~(

XQ. S~.3~ (C.)&JC2 5.

NEDO-21231, #Banked Position Withdrawal Sequence,'K,-.

Section 7.2, January 1977.

BWR/4 STS B 3.1-21 Rev 1, 04/07/95 REVISIONS/" --

Control Rod OPERABILITY B 3.1.3 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.3 Control Rod OPERABILITY BASES BACKGROUND APPLICABLE SAFETY ANALYSES Control rods are components of the Control Rod Drive (CRD)

System, which is the primary reactivity control system for the reactor.

In conjunction with the Reactor Protection System, the CRD System provides the means for the reliable control of reactivity changes to ensure under conditions of normal operation, including abnormal operational transients, that specified acceptable fuel design limits are not exceeded.

In addition, the control rods provide the capability to hold the reactor core subcritical under all conditions and to limit the potential amount and rate of reactivity increase caused by a malfunction in the CRD System.

The CRD System is designed to satisfy the requirements specified in Reference 1.

The CRD System consists of 137 locking piston CRDs and a hydraulic control unit for each CRD.

The locking piston type CRD is a double acting hydraulic piston, which uses condensate water as the operating fluid.

Accumulators provide additional energy for scram.

An index tube and piston, coupled to the control rod, are locked at fixed increments by a collet mechanism.

The collet fingers engage notches in the index tube to prevent unintentional withdrawal of the control rod, but without restricting insertion.

This Specification, along with LCO 3.1.4, "Control Rod Scram Times," and LCO 3.1.5, "Control Rod Scram Accumulators,"

ensure that the performance of the control rods in the event of a Design Basis Accident (DBA) or transient meets the assumptions used in the safety analyses of References 2 and 3.

The control rods provide the primary means for rapid reactivity control (reactor scram), for maintaining the reactor subcritical and for limiting the potential effects of reactivity insertion events caused by malfunctions in the CRD System.

(continued)

Revision J B 3.1-13 JAFNPP

Control Rod OPERABILITY B 3.1.3 BASES APPLICABLE SAFETY ANALYSES (continued)

The capability to insert the control rods provides assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated (Refs. 2 and 3).

Since the SDM ensures the reactor will be subcritical with the highest worth control rod withdrawn (assumed single failure), the additional failure of a second control rod to insert, if required, could invalidate the demonstrated SDM and potentially limit the ability of the CRD System to hold the reactor subcritical.

If the control rod is stuck at an inserted position and becomes decoupled from the CRD, a control rod drop accident (CRDA) can possibly occur.

Therefore, the requirement that all control rods be OPERABLE ensures the CRD System can perform its intended function.

The control rods also protect the fuel from damage which could result in release of radioactivity.

The limits protected are the MCPR Safety Limit (SL) (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)"),

the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)",

and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"),

and the fuel damage limit (see Bases for LCO 3.1.6, "Rod Pattern Control")

during reactivity insertion events.

The negative reactivity insertion (scram) provided by the CRD System provides the analytical basis for determination of plant thermal limits and provides protection against fuel damage limits during a CRDA.

The Bases for LCO 3.1.4, LCO 3.1.5, and LCO 3.1.6 discuss in more detail how the SLs are protected by the CRD System.

Control rod OPERABILITY satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).

The OPERABILITY of an individual control rod is based on a combination of factors, primarily, the scram insertion times, the control rod coupling integrity, and the ability to determine the control rod position.

Accumulator OPERABILITY is addressed by LCO 3.1.5.

The associated scram accumulator status for a control rod only affects the scram insertion times; therefore, an inoperable accumulator does not immediately require declaring a control rod inoperable.

Although not all control rods are required to be OPERABLE to (continued)

LCO Revision J B 3.1-14 JAFNPP

Control Rod OPERABILITY B 3.1.3 BASES SURVEILLANCE SR 3.1.3.2 and SR 3.1.3.3 (continued)

REQUIREMENTS performance and the ease of performing notch testing for fully withdrawn control rods.

Partially withdrawn control rods are tested at a 31 day Frequency, based on the potential power reduction required to allow the control rod movement and considering the large testing sample of SR 3.1.3.2.

Furthermore, the 31 day Frequency takes into account operating experience related to changes in CRD performance.

At any time, if a control rod is immovable, a determination of the control rods OPERABILITY must be made and appropriate action taken.

These SRs are modified by Notes that allows 7 days and 31 days respectively, after withdrawal of the control rod and increasing power to above the LPSP of the RWM, to perform the Surveillance.

This acknowledges that the control rod must be first withdrawn and THERMAL POWER must increase to above the LPSP before performance of the Surveillance, and therefore the Notes avoid potential conflicts with SR 3.0.3 and SR 3.0.4.

SR 3.1.3.4 Verifying that the scram time for each control rod to notch position 04 is

7 seconds provides reasonable assurance that the control rod will insert when required during a DBA or transient, thereby completing its shutdown function.

This SR is performed in conjunction with the control rod scram time testing of SR 3.1.4.1, SR 3.1.4.2, SR 3.1.4.3, and SR 3.1.4.4.

The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation," and the functional testing of SDV vent and drain valves in LCO 3.1.8, "Scram Discharge Volume (SDV)

Vent and Drain Valves," overlap this Surveillance to provide complete testing of the assumed safety function.

The associated Frequencies are acceptable, considering the more frequent testing performed to demonstrate other aspects of control rod OPERABILITY and operating experience, which shows scram times do not significantly change over an operating cycle.

(continued)

Revision J B 3.1-20 JAFNPP

Control Rod Scram Times B 3.1.4 BASES APPLICABLE The scram function of the CRD System protects the MCPR SAFETY ANALYSES Safety Limit (SL)

(see Bases for SL 2.1.1, "Reactor Core (continued)

SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)")

and the 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)*),

which ensure that no fuel damage will occur if these limits are not exceeded.

Above 800 psig, the scram function is designed to insert negative reactivity at a rate fast enough to prevent the actual MCPR from becoming less than the MCPR SL, during the analyzed limiting power

, rnsient, Below 800 psig, the scram function is a'sumed to

(-er dd*oIn the control rod drop accident (Ref.

an,

jfo l also provides protection against violating fuel damage limits during reactivity insertion accidents (see ( D.

Bases for LCO 3.1.6, "Rod Pattern Control").

For the reactor vessel overpressure protection analysis, the scram function, along with the safety/relief valves, ensure that the peak vessel pressure is maintained within the applicable ASME Code limits.

Control rod scram times satisfy Criterion 3 of he RC Control Wo or6 Re. !!6 The scram times specified in Table 3.1.4-16Iý ý

\\-(?Ai' acco an in CO are required to ensure th-a-t te scram1....

reactivity assumed in the DBA and transient analysis is met (Ref. 6).

To account for single failures and "slow" scramming control rods, the scram times specified in Table 3.1.4-1 are faster than those assumed in the design basis ansis. The scram times have a mar in that allows rj, UP 0 ppro7 ate 7%

control rods e..,

137 7 %

rf0o ave scram times exceeding e specifiedýý )

limits (i.e., "slow" control rods) assuming a single stuck control rod (as allowed by LCO 3.1.3, "Control Rod T

OPERABILITY")

and an additional control rod failing to scram per the single failure criterion.

The scram times are specified as a function of reactor steam dome pressure to account for the pressure dependence of the scram times.

The scram times are specified relative to measurements based on reed switch positions, which provide the control rod position indication.

The reed switch closes ("pickup") when the index tube passes a specific location and then opens

("dropout") as the index tube travels upward.

Verification of the specified scram times in Table 3.1.4-1 is accomplished (continued)

BWR/4 STS B 3.1-23 REVISION,0'*

)

LCO Rev 1, 04/07/95

Control Rod Scram Times B 3.1.4 BWR/4 STS B 3.1-28 Rev 1, 04/07/95 REVISIONO

Control Rod Scram Times 3.1.4 Table 3.1.4-1 (page 1 of 1)

Control Rod Scram Times

NOTES ------------.----------------------

1.

OPERABLE control rods with scram times not within the limits of this Table are considered "slow."

2.

Enter applicable Conditions and Required Actions of LCO 3.1.3, "Control Rod OPERABILITY," for control rods with scram times > 7 seconds to notch position 04.

These control rods are inoperable, in accordance with SR 3.1.3.4, and are not considered "slow."

SCRAM TIMES(a)(b) (seconds) when REACTOR STEAM DOME NOTCH POSITION PRESSURE t 800 psig 46 0.44 36 1.08 26 1.83 06 3.35 (a)

Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids at time zero.

(b)

Scram times as a function of reactor steam dome pressure, when

< 800 psig, are within established limits.

Amendment (Rev. J)

JAFNPP 3-1-14

Control Rod Scram Times B 3.1.4 BASES REFERENCES (continued)

4.

NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, Supplement for United States, Section S.2.2.3.1, (Revision specified in the COLR).

5.

6.

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

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

7.

Technical Requirements Manual.

Revision J I L6; JAFNPP R 3.1-28

Control Rod Scram Accumulators B 3.1.5 BASES ACTIONS D.

(continued) ensures that all insertable control rods are inserted and that the reactor is in a condition that does not require the active function (i.e., scram) of the control rods.

This Required Action is modified by a Note stating that the action is not applicable if all control rods associated with the Inoperable scram accumulators are fully inserted, since the function of the control rods has been performed.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force.

The primary indicator of accumulator OPERABILITY is the accumulator pressure.

A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator i considered inoperable.

The minimum accumulato essure of 9 0 5i is well below the expected pressure of ilOOpsig (Ref.

Declaring the accumulator Inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur.

The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.

NKF)

REFERENCES

2.

SAR, Section Rev 1, 04/07/95 BWR/4 STS B 3.1]-33

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.1.5 - CONTROL ROD SCRAM ACCUMULATORS RETENTION OF EXISTING REQUIREMENT (CLB)

None PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)

PAl Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific nomenclature.

PA2 Typographical/grammatical error corrected.

PA3 The Bases have been revised for enhanced clarity.

PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)

DB1 The word "approximately" has been added for clarity.

[

DB2 Changes have been made to reflect the plant specific references.

In addition, the brackets have been removed from the References and the plant specific References have been included.

DIFFERENCE BASED ON AN APPROVED TRAVELER (TA)

None DIFFERENCE BASED ON A SUBMITTED, BUT PENDING TRAVELER (TP)

None DIFFERENCE FOR ANY REASON OTHER THAN THE ABOVE (X)

Xl NUREG-1433, Revision 1, Bases reference to "the NRC Policy Statement" has been replaced with 10 CFR 50.36(c)(2)(ii), in accordance with 60 FR 36953 effective August 18, 1995.

Page 1 of 1 JAFNPP Revision J

Control Rod Scram Accumulators B 3.1.5 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.5 Control Rod Scram Accumulators BASES BACKGROUND APPLICABLE SAFETY ANALYSES The control rod scram accumulators are part of the Control Rod Drive (CRD)

System and are provided to ensure that the control rods scram under varying reactor conditions.

The control rod scram accumulators store sufficient energy to fully insert a control rod at any reactor vessel pressure.

The accumulator is a hydraulic cylinder with a free floating piston.

The piston separates the water used to scram the control rods from the nitrogen, which provides the required energy.

The scram accumulators are necessary to scram the control rods within the required insertion times of LCO 3.1.4, "Control Rod Scram Times."

The Design Basis Accident (DBA) and transient analyses assume that all of the control rods scram at a specified insertion rate (Refs. 1 and 2).

OPERABILITY of each individual control rod scram accumulator, along with LCO 3.1.3, "Control Rod OPERABILITY,"

and LCO 3.1.4, ensures that the scram reactivity assumed in the DBA and transient analyses can be met.

The existence of an inoperable accumulator may invalidate prior scram time measurements for the associated control rod.

The scram function of the CRD System, and therefore the OPERABILITY of the accumulators, protects the MCPR Safety Limit (see Bases for SL 2.1.1, "Reactor Core SLs," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)")

and 1% cladding plastic strain fuel design limit (see Bases for LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR),"

and LCO 3.2.3, "LINEAR HEAT GENERATION RATE (LHGR)"),

which ensure that no fuel damage will occur if these limits are not exceeded (see Bases for LCO 3.1.4).

In addition, the scram function at low reactor vessel pressure (i.e., startup conditions) provides protection against violating fuel design limits during reactivity insertion accidents (see Bases for LCO 3.1.6, "Rod Pattern Control").

Control rod scram accumulators satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 3).

(conti nued)

Revision 3 JAFNPP B 3.1-29

Control Rod Scram Accumulators B 3.1.5 BASES (continued)

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS SR 3.1.5.1 requires that the accumulator pressure be checked every 7 days to ensure adequate accumulator pressure exists to provide sufficient scram force.

The primary indicator of accumulator OPERABILITY is the accumulator pressure.

A minimum accumulator pressure is specified, below which the capability of the accumulator to perform its intended function becomes degraded and the accumulator is considered inoperable.

The minimum accumulator pressure of 940 psig is well below the expected pressure of approximately 1100 psig (Ref. 4).

Declaring the accumulator inoperable when the minimum pressure is not maintained ensures that significant degradation in scram times does not occur.

The 7 day Frequency has been shown to be acceptable through operating experience and takes into account indications available in the control room.

REFERENCES

1.

UFSAR, Section 14.6.

2.

UFSAR, Section 14.5.

3.

4.

GEK-9582C, "Hydraulic Control Unit," December 1987.

Revision J JAFNPP B 3.1-33

Rod Pattern Control B 3.1.6 APPLICABLE SAFETY ANALYSES (continued)

Control od atterns analyzed in Reference I follow the banked position withdrawal sequence (BPWS).

The BPWS is applicable from the condition of all control rods fully inserted to.O RTP (Ref. 2).

For the BPWS, the control rods are req ired to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12).

The banked positions are established to minimize

)the maximum incremental control rod worth without being overly restrictive during normal plant operation.

Generic analysis of the BPWS (Ref. 1) has demonstrated that the 3ZBU Pca Igm

!!Luel jlimit ill not be violated during a

-CRDA while following

-e--PWSH of operation.

The generic BPWS analysis (Refo$) also evaluates the effect of fully inserted, inoperable control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, inoperable control rods.

Rod atten control sati sf ies Cri teri on 3 o W7R lic LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS.

This LCO only applies to OPERABLE control rods.

For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

APPLICABILITY In MODES I and 2, when THERMAL POWER is :g((

RTP, th CRDA is a Design Basis Accident and, there or, compliance with the assumptions of he safety analysis is required.

When THERMAL POWER is >

RTP, there is no credible <

control rod configuration tiat results in a control rod worth that could exceed the 280 cal/gm fuel 4#ga imit during a CRDA (Ref. 2).

In MODES 3, 4, and 5, since the reactor is shut down and only a single control rod can be withdrawn from a core cell containing fuel assemblies, adequate SDM ensures that the consequences of a CRDA are acceptable, since the reactor will remain subcritical with a single control rod withdrawn.

(continued)

BWR/4 STS B 3.1-35 Rev 1, 04/07/95 REVISION 0 JI BASES 1I BASES

Rod Pattern Control B 3.1.6 BASES ACTIONS B.1 and B-2 control rods has less impact on control rod worth than withdrawals have.

Required Action B.1 is modified by a Note which allows the RWM to be bypassed to allow the affected control rods to be returned to their correct position.

LCO 3.3.2.1 requires verification of control rod movement by When nine or more OPERABLE control rods are not in compliance with BPWS, the reactor mode switch must be placed in the shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

With the mode switch in shutdown, the reactor is shut down, and as such, does not meet the applicability requirements of this LCO.

The allowed Completion Time of I hour is reasonable to allow insertion of control rods to restore compliance, and is appropriate relative to the low probability of a CRDA occurring with the control rods out of sequence.

SURVEILLANCE REQUIREMENTS 0y-3 BWR/4 STS (continued)

The control rod pattern is verified to be in compliance with the BPWS at a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency to ensure the assumptions of the CRDA analyses are met.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency was developed considering that the primary check on compliance with the BPWS is performed by the RWM (LCO 3.3.2.1), which provides control rod blocks to enforce the required sequence an arequired to be OPERABLE when operating at B 3.1-37 Rev 1, 04/07/95 REVISION Z'"

i Ju I y I Naty. T-

1% r I-eci ee't %

r r

apr NUREG-0979, Section 4.2.1.3.2, April AC4 NUREG-0800,ýSection 15.4.9, Revision 2, July 1981.

r4j R'Yo;f fl'T'b't bTý TAA( h-est, Ativw, e

41,eA 9,C

ýhe-8A)e M) r1r---Z Des,ý,L (A'4 sueplew-ft (continued)

Insert Ref 1

6.

NEDO-10527, Rod Drop Accident Analysis For Large BWRs, March 1972.

7.

NEDO-10527, Supplement 1, Rod Drop Accident Analysis For Large Boiling Water Reactors, Addendum No.

1, Multiple Enrichment Cores With Axial Gadolinium, July 1972.

8.

NEDO-10527, Supplement 2, Rod Drop Accident Analysis For Large Boiling Water Reactors, Addendum No. 2, Exposed Cores, January 1973.

Insert Ref 2

12.

Insert Page B 3.1-38 Revision J

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.1.6 -

ROD PATTERN CONTROL RETENTION OF EXISTING REQUIREMENT (CLB)

None 1(t PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)

PAl Changes have been made consistent with the Specification.

PA2 Changes have been made to correct a typographical error.

PA3 Not used PA4 Not used PA5 The quotations used in the Bases References have been removed.

The Writer's Guide does not require the use of quotations.

PA6 Changes have been made to be consistent with the plant specific terminology.

PA7 The title for the Bases References have been included for clarity.

PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)

DB1 Changes have been made (additions, deletions and/or changes to the NUREG) to reflect the plant specific design references.

References have been renumbered, where required.

DB2 Existing Reference 2 is actually an attachment to another document.

The actual reference has been revised to reflect this other document in order to facilitate location of the references in the future.

DB3 The brackets have been removed and the proper plant specific value included.

Page 1 of 2 Revision J JAFNPP

Rod Pattern Control B 3.1.6 BASES APPLICABLE number of fuel rods would reach a fuel enthalpy of SAFETY ANALYSES 170 cal/gm, which is the enthalpy limit for eventual (continued) cladding perforation.

Control rod patterns analyzed in Reference 1 follow the banked position withdrawal sequence (BPWS).

The BPWS is applicable from the condition of all control rods fully inserted to 10% RTP (Ref. 2).

For the BPWS, the control rods are required to be moved in groups, with all control rods assigned to a specific group required to be within specified banked positions (e.g., between notches 08 and 12).

The banked positions are established to minimize the maximum incremental control rod worth without being overly restrictive during normal plant operation.

Generic analysis of the BPWS (Ref. 1) has demonstrated that the 280 cal/gm fuel energy deposition limit will not be violated during a CRDA while following the BPWS mode of operation.

The generic BPWS analysis (Ref. 11) also evaluates the effect of fully inserted, inoperable control rods not in compliance with the sequence, to allow a limited number (i.e., eight) and distribution of fully inserted, inoperable control rods.

Rod pattern control satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 12).

LCO Compliance with the prescribed control rod sequences minimizes the potential consequences of a CRDA by limiting the initial conditions to those consistent with the BPWS.

This LCO only applies to OPERABLE control rods.

For inoperable control rods required to be inserted, separate requirements are specified in LCO 3.1.3, "Control Rod OPERABILITY," consistent with the allowances for inoperable control rods in the BPWS.

(continued)

Revision J JAFNPP B 3.1-35

Rod Pattern Control B 3.1.6 BASES (continued)

REFERENCES

1.

NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, Supplement for United States, Section S.2.2.3.1, (Revision specified in the COLR).

2.

Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),

Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, BWROG-8644, August 15, 1986.

3.

NUREG-0979, Safety Evaluation Report Related to the Final Design Approval of the GESSAR II, BWR/6 Nuclear Island Design (and Supplements 1 through 5),

Section 4.2.1.3.2, April 1983.

4.

NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, Section 15.4.9, Spectrum of Rod Drop Accidents (BWR),

Revision 2, July 1981.

5.

6.

NEDO-10527, Rod Drop Accident Analysis For Large BWRs, March 1972.

7.

NEDO-10527, Supplement 1, Rod Drop Accident Analysis For Large Boiling Water Reactors, Addendum No.

1, Multiple Enrichment Cores With Axial Gadolinium, July 1972.

8.

NEDO-10527, Supplement 2, Rod Drop Accident Analysis For Large Boiling Water Reactors, Addendum No. 2, Exposed Cores, January 1973.

9.

NEDO-21778-A, Transient Pressure Rises Affecting Fracture Toughness Requirements For Boiling Water Reactors, December 1978.

10.

ASME, Boiler and Pressure Vessel Code,Section III, 1965 Edition, Addenda Winter of 1966.

11.

NEDO-21231, Banked Position Withdrawal Sequence, January 1977.

12.

Revision J JAFNPP B 3.1-38

~per-C-r6-3 (7

JAFNPP 5g the expo ive valv solut in th eirculat pah ci

]Exploie one On ee prmer 5La u,isse.oluie a(fcod in heame

/*

"*'~

FA xp

e. one pac o

v e

erif firo er uc-n

)~ ~ -"'\\*as/mblies t

ofh same bI~tch iqt7f S*h* ~explosie/ie:/

Sým!lerglize-jwater shall be injected into the reactor vessel to v

3s exPt ex osive vIves) ot ch ked the r circula on tes ar L*

"6.

Test that the setting 9/ the system I

pressure relief valve is between 1,400 and 1,490 psig.I "Disassemble Inspect one explosive I

\\

valve so tha it can be established S[

that the val a is n4t clogged. Both

*wih Io~era~e Coz~ont*

valves shall be inspected within B.

O v e l i t i i

t h I n a a r i n-

.^

-wo n s t i n t e r v a l s...F From and after the date that a redundant component is made

0.

0 eraton with In er 0om on:ts or found to be inoperable, Specification 3.4.A shall be L/

considered fulfilled, and continued operation permitted, rWhen a compon nt becomes inoperabl its reduni provided that:

component sh be verified to be operble immed!

thereafter.

1.

The component is returned to an operable condition within 7 days.

Amendment No. 38, 134, 148, 232, 241 106 REVISION V I

DISCUSSION OF CHANGES ITS: 3.1.7 - STANDBY LIQUID CONTROL (SLC)

SYSTEM ADMINISTRATIVE CHANGES Al In the conversion of the James A. FitzPatrick Nuclear Power Plant (JAFNPP) Current Technical Specification (CTS) to the proposed plant specific Improved Technical Specifications (ITS) certain wording preferences or conventions are adopted which do not result in technical changes.

Editorial changes, reformatting, and revised numbering are adopted to make the ITS consistent with the conventions in NUREG-1433, "Standard Technical Specifications, General Electric Plants, BWR/4",

Revision 1 (i.e., Improved Standard Technical Specifications (ISTS)).

A2 CTS 3.4.C states that the solution temperature including the pump suction piping temperature has to be maintained above the temperature limits.

CTS 4.4.C.2 requires that the solution temperature be checked at least once per day.

The ITS has two separate surveillances (SR 3.1.7.2 and SR 3.1.7.3) which require that the temperature of the sodium pentaborate solution (SR 3.1.7.2) and the temperature of the pump suction piping (SR 3.1.7.3) be verified every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Since the pump suction piping temperature has always been a requirement for SLC OPERABILITY, having a separate SR to verify this temperature is considered an administrative change.

This is consistent with NUREG 1433, Revision 1.

TECHNICAL CHANGES - MORE RESTRICTIVE M1 CTS 4.4.A.1 requires the verification that all valves (manual, power operated, or automatic) in the system flowpath that is not locked, sealed or otherwise secured in position is in the correct position.

There are no power operated or automatic valves in the system except for the explosive valves.

This Surveillance is included as ITS SR 3.1.7.6 for all manual valves, and a new requirement has been added to verify the continuity of each explosive charge (ITS SR 3.1.7.4).

Since the ITS is more explicit on the method of verification for the explosive valve this change is considered more restrictive on plant 'operation.

This change is consistent with NUREG-1433, Revision 1.

M2 CTS 4.4.A.3 requires verification that heat traced piping between the SLC storage tank and the pump suction is unblocked (by manually initiating the system, except explosive valves, and pump boron solution from the SLC storage tank through the recirculation path) once every 24 months.

ITS SR 3.1.7.9 requires verification that heat traced piping between the SLC storage tank and the pump suction is unblocked once per 24 months and "Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after piping temperature is restored within the limits of Figure 3.1.7-2."

The addition of this second Surveillance Frequency represents a more restrictive change necessary to ensure the piping is unblocked after conditions have existed with the Page 1 of 7 JAFNPP Revision J

DISCUSSION OF CHANGES ITS: 3.1.7 -

SYSTEM TECHNICAL CHANGES MORE RESTRICTIVE M2 (continued) potential for causing the piping to become blocked due to precipitation of boron from solution.

M3 CTS 4.4.C.1 has requirements for checking the concentration of sodium pentaborate in the SLC Tank after certain events which could affect boron concentration occur (adding water to tank, adding boron to tank, or if temperature of solution in tank drops below the temperature limit).

The CTS does not specify any time requirement for performing these checks.

ITS (SR 3.1.7.5) adds a time limit of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months into the requirement to check sodium pentaborate concentration after additions to the SLC Tank are made (water or boron).

This ensures that the concentration is checked on a timely basis after additions to the tank are made rather than the current open ended specification.

SR 3.1.7.5 also adds a second time requirement to check the concentration within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within limits.

This checks for the amount of boron that may have precipitated out of solution.

The addition of new requirements reflects a more restrictive change necessary to ensure SLC System Operability is adequately maintained.

M4 CTS 4.4.C.4 requires that the enrichment of the Boron-lO (in the SLC tank) be checked once per 24 months, but the CTS contains no requirement for checking the Boron-lO enrichment of sodium pentaborate being added to the tank.

ITS SR 3.1.7.10 requires that a Boron-lO enrichment verification be done prior to adding sodium pentaborate to the tank.

Since the enrichment of a batch/lot of sodium pentaborate will not change with time, a single isotopic test of any given batch/lot can suffice as the required analysis for any number of mixings and additions from that batch/lot.

For sodium pentaborate supplied and purchased under controls assuring appropriate 10CFR50, Appendix B and ANSI N45.2 compliance, the required analysis may be satisfied by certified vendor analytical test results.

While this is consistent with current practice, this SR is considered more restrictive in that the requirement is not expressly stated in the CTS.

Once the Boron-10 is in the SLC tank the enrichment of the solution will not change. ITS SR 3.1.7.5 requires that the concentration of the boron solution in the SLC tank be verified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the boron addition.

ITS SRs 3.1.7.1, 3.1.7.2 and 3.1.7.3 verify proper boron solution volume and temperature.

ITS SR 3.1.7.11 (retained from CTS 4.4.C.4) verifies the enrichment of boron in the SLC tank every 24 months.

These verifications, in addition to proposed ITS SR 3.1.7.10, help maintain the required quantity of B-10 in the tank.

ITS SR 3.1.7.10 is Page 2 of 7 JAFNPP Revision J

DISCUSSION OF CHANGES ITS: 3.1.7 - STANDBY LIQUID CONTROL (SLC)

SYSTEM TECHNICAL CHANGES - MORE RESTRICTIVE M4 (continued) considered more restrictive but provides assurance that SLC System Operability is adequately maintained.

This change is consistent with NUREG-1433, Revision 1.

M5 CTS 4.4.A.4 requires the firing of SLC primer assemblies prior to being installed into the plant, but does not require firing of the primer assemblies during the flow testing of CTS 4.4.A.5.

ITS SR 3.1.7.8 will require the firing of an installed primer assembly as part of the SLC flow test to ensure the primer assembly opens the associated SLC valve.

This requirement is considered more restrictive but provides assurance that SLC System Operability is adequately maintained. This change is consistent with NUREG-1433, Revision 1.

TECHNICAL CHANGES

- LESS RESTRICTIVE (GENERIC)

LA1 The details of method of performing CTS 4.4.A.2 to verify flow by recirculating demineralized water to the test tank; the details in CTS 4.4.A.3, to demonstrate all piping between the SLC storage tank and the pump suction is unblocked (by manually initiating the system, except the explosive valves and pump solution in the recirculation path); the details in CTS 4.4.A.5 to verify flow through the SLC subsystem into the reactor pressure vessel (to test that the valves except explosive valves not checked by the recirculation test are not clogged); and the details in CTS 4.4.A.4 to explode one of three primer assemblies manufactured in same batch to verify proper function and (as modified by DOC L5) that the replacement primer assembly is from a batch previously tested are proposed to be relocated to the Bases.

These details are not necessary to ensure that the SLC System is maintained Operable.

The requirements of ITS 3.1.7 and SRs 3.1.7.7, 3.1.7.8, and 3.1.7.9 are adequate to ensure the capability to provide flow through each SLC subsystem to the test tank and into the reactor pressure vessel, to ensure the piping between the SLC storage tank and the pump suction is unblocked, and to ensure SLC System Operability.

Therefore, the relocated details are not necessary to be in the ITS to provide adequate protection of the public health and safety.

Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications.

LA2 The testing requirements of CTS 4.4.A.6 (to verify the proper operation and setpoints of the relief valves) and CTS 4.4.A.7 (to disassemble and inspect one explosive valve) are proposed to be relocated to the Inservice Testing (IST) Program.

These testing requirements demonstrate the SLC System relief valves and explosive valves are OPERABLE.

Revision J JAFNPP Page 3 of 7

DISCUSSION OF CHANGES ITS: 3.1.7 - STANDBY LIQUID CONTROL (SLC)

SYSTEM TECHNICAL CHANGES

- LESS RESTRICTIVE (GENERIC)

LA2 (continued)

However, the IST Program, required by 10 CFR 50.55a, provides requirements for the testing of all ASME Code Class 1, 2, and 3 valves in accordance with Section XI of the ASME Code.

ITS Section 5.5.7 provides controls over the IST Program.

These controls are adequate to ensure the required testing to demonstrate Operability is performed.

Therefore, the relocated requirements are not necessary to be in the ITS to provide adequate protection of the public health and safety.

Changes to the relocated requirements in the IST Program will be controlled by the provisions of 10 CFR 50.59.

LA3 CTS 3.4.C contains detailed information concerning the boron solution for the SLC storage tank, and what support components and variables are required to assure SLC OPERABILITY is maintained.

The ITS relocates this detailed information to the Bases for Specification 3.1.7.

The requirements of ITS 3.1.7 including the LCO, ACTIONS and Surveillances are adequate to ensure SLC System OPERABILITY.

Therefore, the relocated details are not necessary to be in the ITS to provide adequate protection of the public health and safety.

Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications.

LA4 The detail in CTS Figure 3.4-2 that the saturation temperature of enriched sodium pentaborate solution curve includes a 10OF margin is proposed to be relocated to the Bases.

The requirements in ITS SR 3.1.7.2 to verify the temperature of sodium pentaborate solution is within the limits of Figure 3.1.7-2 (CTS Figure 3.4-2) and Figure 3.1.7 2 (Sodium Pentaborate Solution Temperature Versus Concentration Requirements curve) are adequate to ensure the proper evaluation is performed and therefore help ensure SLC System OPERABILITY.

Therefore, the relocated details are not necessary to be in the ITS to provide adequate protection of the public health and safety.

Changes to the Bases will be controlled by the provisions of the proposed Bases Control Program described in Chapter 5 of the Technical Specifications.

LB1 The requirements in CTS 4.4.C.3 to calibrate the temperature and level elements is proposed to be relocated to the Technical Requirements Manual (TRM).

These temperature and level indications do not necessarily relate directly to SLC System OPERABILITY.

In general NUREG-1433, Revision 1, does not specify requirements for equipment which only provide indication to support OPERABILITY of a system or component.

Control of the availability of, and necessary compensatory activities if not available, for indications, monitoring instruments, and alarms are addressed by plant operational procedures and policies.

Therefore, the SLCs temperature and level instrument surveillances are Page 4 of 7 JAFNPP Revision J

DISCUSSION OF CHANGES ITS: 3.1.7 - STANDBY LIQUID CONTROL (SLC)

SYSTEM TECHNICAL CHANGES - LESS RESTRICTIVE (GENERIC)

LB1 (continued) removed from the Technical Specifications and relocated to the TRM.

The requirements in ITS 3.1.7 including the LCO, ACTIONS and Surveillances are adequate to ensure the SLC System is Operable.

Therefore, the relocated requirements are not necessary to be in the ITS to provide adequate protection of the public health and safety.

At ITS implementation, the relocated requirement will be incorporated by reference into the UFSAR.

As such changes to the relocated requirements in the TRM will be controlled by the provisions of 10 CFR 50.59.

TECHNICAL CHANGES

- LESS RESTRICTIVE (SPECIFIC)

Li CTS requires that the Standby Liquid Control System be Operable during a period when fuel is in the reactor and prior to startup from cold condition.

This System need not be Operable when the reactor is in the cold condition, control rods are fully inserted and CTS 3.3.A (Reactivity Limitations) is met.

The ITS 3.1.7 Applicability is MODES 1 and 2.

The current Applicability corresponds to MODES 1, 2 and may even imply MODES 3, 4 and 5 with any control rod withdrawn.

This change is less restrictive since the new Applicability does not include MODES 3, 4, and 5.

The SLC system is not needed during Hot or Cold Shutdown (MODES 3 or 4) since control rods can only be withdrawn in accordance with Section 3.10, "Special Operations," and adequate SDM prevents criticality under these conditions.

While in the refueling MODE, the SLC System is not needed because only a single control rod can be withdrawn and adequate SDM prevents criticality when under these conditions.

L2 CTS 3.4.C includes an action to restore certain components (e.g., tank heaters) or variables (e.g., sodium pentaborate volume-concentration and temperature requirements) within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or take action to be in hot shutdown in the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All the components or variables discussed in CTS 3.4.C will cause both subsystems of the SLC System to be inoperable.

However, the list is not all inclusive of the possible events which could lead to both subsystems being inoperable.

ITS 3.1.7, ACTION B is being added to allow the entire SLC System (e.g., both pumps) to be inoperable for any reason up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to requiring a plant shutdown.

The 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides time to restore minor problems (e.g., some pump inoperabilities) prior to requiring a plant shutdown.

The 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is considered acceptable since the time is short, the SLC System is not the primary method of shutting down the plant, reduces the possibility of plant shutdown transients and the probability of an ATWS event is very small.

Page 5 of 7 JAFNPP Revision J

DISCUSSION OF CHANGES ITS: 3.1.7 -

SYSTEM TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

L3 CTS 4.4.B requires that when a SLC subsystem or component becomes inoperable, the redundant subsystem or component be verified to be OPERABLE immediately and daily thereafter.

ITS 3.1.7 does not have this cross system check.

This change will allow credit to be taken for normal periodic Surveillances as a verification of OPERABILITY and availability of the remaining SLC subsystem.

The periodic Frequencies specified to verify OPERABILITY of the remaining SLC subsystem has been shown to be adequate to ensure equipment OPERABILITY.

As stated in NRC Generic Letter 87-09, "It is overly conservative to assume that systems or components are inoperable when a surveillance requirement has not been performed.

The opposite is in fact the case; the vast majority of surveillances demonstrate the systems or components in fact are operable."

Therefore, reliance on the specified Surveillance intervals does not result in a reduced level of confidence concerning the equipment availability.

The ITS and current BWR operating philosophy accept the philosophy of system OPERABILITY based on satisfactory performance of monthly, quarterly, refueling interval, post-maintenance or other specified performance tests without requiring additional testing when another system is inoperable (except for diesel generator testing, which is not being changed).

L4 CTS 4.4.A.1 requires that each SLC subsystem "valve (manual, power operated, or automatic) in the system flow path that is not locked, sealed or otherwise secured in position, is in the correct position" once per 31 days.

ITS SR 3.1.7.6 requires that "each SLC subsystem manual valve (see M1) in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position or can be aligned to the correct position" every 31 days.

The proposed change permits the SLC subsystem to be considered OPERABLE as long as the valves can be manually realigned to their correct position.

The Bases stipulates that this realignment must be capable of being done from the control room, or locally by a dedicated operator at the valve control.

The SLC System is a manually initiated system.

Therefore allowing the system to be considered OPERABLE whenever the system valves can be correctly aligned does not reduce the level of safety and is considered acceptable.

This change is consistent with NUREG-1433, Revision 1.

Page 6 of 7 JAFNPP Revision J

DISCUSSION OF CHANGES ITS: 3.1.7 - STANDBY LIQUID CONTROL (SLC)

SYSTEM TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)

L5 CTS 4.4.A.5 requires that every 24 months demineralized water be injected into the reactor vessel to test that valves (except explosive valves) not checked by the recirculation test (CTS 4.4.A.3) are not clogged.

This test involves testing entire subsystems; including portions common to both subsystems as well as non-common portions.

As such, testing either subsystem can satisfy the necessary testing for the common portions of both subsystems.

To accomplish this, ITS SR 3.1.7.8 requires the verification of flow through one SLC subsystem from the pump into reactor pressure vessel every 24 months on a STAGGERED TEST BASIS (i.e., such that the subsystems use for the test are alternated each 24 months).

Since the CTS could be inferred to require testing both subsystems each 24 months, this change is a relaxation in the frequency of testing an individual subsystem (i.e., on the Staggered Test Basis), and is classified as a less restrictive change.

In addition, CTS 4.4.A.4 requires replacing both primer assemblies every 24 months.

ITS SR 3.1.7.8 will only require replacing one primer assembly every 24 months; the one that has been fired during performance of the SR.

Over the course of two surveillance intervals (48 months),

both primer assemblies will be required to be replaced.

Testing of the non-common portions, which are also the subject of the relaxed testing frequency, are appropriately surveilled by other ITS SRs:

specifically, each pump is tested per the IST Program as required by SR 3.1.7.7, the continuity of each explosive charge is verified every 31 days in accordance with SR 3.1.7.4, the temperature of pump suction piping is verified within limits every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with SR 3.1.7.3, and proper manual valve position is verified every 31 days as required by SR 3.1.7.6.

These surveillance tests, which are performed more frequently than the proposed surveillance interval of SR 3.1.7.8, provide assurance that unacceptable conditions associated with the SLC System will be detected in a timely manner.

This change is also consistent with NUREG-1433, Revision 1.

TECHNICAL CHANGES -

RELOCATIONS None Page 7 of 7 JAFNPP Revision J

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pentaborate solution is ithin the limits of Figure 3.1.7-1 j

1

+1.....

~LLL~iJ 3.1.7.3 Verify temperature of sodium pentaborate olution is wi in the limits of igure 3.1.7-Verify temperature of pump suction pip is within the limits ofj*igure 3.1.7-:

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

/

SR 3.1.7.4 Verify continuity of explosive charge.

31 days SR 3.1.7.5 Verify,the cooncentration solution is(vithin the 1 F i g u r e J.. -. -

31 days AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water o is added to solution AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of igure (continued)

Rev 1, 04/07/95 REVISION T"

BWR/4 STS 3.1-21 C. Y, 0, C, ZI

)

j

SLC System 3.1.7 SR 3.1.7.6 ye d

an a valv inthefIow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the

-correct position.

SR 3.1.7.7 Verify each pump develops a flow rate

>4+42-gpm at a discharge pressure

~p~sig.

Verify flow through one SLC subsystem from pump into reactor pressure vessel.

sodium pentaborate enrichment is

atom percent B-10.

31 days LI Lk CSTS B;1/4 STS Rev 1, 04/07/95 REVISION,I, SURVEII 3.1.7.9 Verify all heat traced piping between storage tank and pump suction is unblocked.

Oil Dý'k

[& HM,*,,*IIR 3.1. 7.8

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS: 3.1.7 -

SYSTEM RETENTION OF EXISTING REQUIREMENT (CLB)

CLB1 The brackets have been removed and the Frequency "In accordance with the Inservice Testing Program" retained consistent with the current licensing basis in CTS 4.4.A.2.

CLB2 The brackets have been removed and a 24 month Frequency included in SR 3.1.7.8 and SR 3.1.7.9 (first frequency) consistent with CTS 4.4.A.4 and CTS 4.4.A.3, respectively.

In addition, the brackets have been removed from SR 3.1.7.9 and the second frequency has been added in accordance with M2.

CLB3 The requirements of CTS 4.4.C.4 for verifying Boron-lO enrichment of the sodium pentaborate solution in the SLC tank on a 24 month Frequency are retained as SR 3.1.7.11.

Although the provisions of SR 3.1.7.10 are adequate to ensure proper Boron-lO enrichment, periodic verification of the SLC solution enrichment is a good practice providing added assurance that the proper Boron-lO enrichment is maintained.

PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT (PA)

PAl The second Frequency for ISTS SR 3.1.7.9 (ITS SR 3.1.7.9) is being changed from being based on solution temperature to piping temperature.

The SR requires a verification that all heat traced piping is unblocked.

A change in solution temperature in the tank does not necessarily have an impact on the piping temperature, as long as the piping heat trace circuit is functioning properly.

The intent of the second Frequency is to ensure that, if the heat tracing is inoperable such that piping temperature falls below the limits of Figure 3.1.7-2, after the heat tracing is restored to OPERABLE status and the piping temperature is within the limits of Figure 3.1.7-2 the piping is still unblocked.

This is supported by the ISTS Bases description for this second Frequency, which describes the requirement as required to be performed after "piping" temperature is restored.

PA2 The proper JAFNPP nomenclature has been used.

This is also consistent with SR 3.1.7.1 and SR 3.1.7.2.

PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)

DB1 NUREG-1433 ACTION A, is not applicable to JAFNPP and, has been deleted.

JAFNPP requires the same concentration of boron in solution to meet the original licensing basis of the SLC System (cold shutdown) as it does for the ATWS rule (10 CFR 50.62).

Therefore low boron concentration Page 1 of 2 Revision J JAFNPP

JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, ITS: 3.1.7 -

REVISION 1 SYSTEM PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)

DB1 (continued) would result in both SLC subsystems being inoperable.

The remaining Conditions and Required Actions have been renumbered or revised to reflect this deletion.

DB2 The brackets have been removed and the proper limits included.

DB3 The brackets have been removed and the information deleted since the system does not include any power operated or automatic valves other than the explosive valves.

DB4 NUREG Figures 3.1.7-1 and Figure 3.1.7-2 have been modified in accordance with the current requirements.

DB5 The brackets have been removed and SR 3.1.7.10 retained in accordance with CTS 4.4.C.4 and DOC M4.

DIFFERENCE BASED ON AN APPROVED TRAVELER (TA)

None DIFFERENCE BASED ON A SUBMITTED, BUT PENDING TRAVELER (TP)

None DIFFERENCE FOR ANY REASON OTHER THAN THE ABOVE (X)

None Page 2 of 2 Revision J JAFNPP

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.7 (continued) pvmp

(,Oý-Op REQUIREMENTS requirements, the rate of negativ reactivity insertion from the SLC System will adequatelyi ompensate for the positive reactivity effects encountered during power recducti on, Scooldown of the moderator, and xenon decay.

This test, confirms dner b t n

h

.*~sa quvJB3f3 and is 1 -~7-

\\

indicative of overall performance.

Such inservice G

ff confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

The Frequency of this Surveillance is rdance with the Inservice Testing Progran tA.,

SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution'storage tank to the RPV, including the firing of an explosive valv The replacement CflIa for the explosive valveshall be from the same manufactured batch as the one fired or from another batch L

/

that has been certified by having one of that batch successfully fired.

The pump and explosive valve tested should be alternated such that both corplete flow paths are q

tested everyQ months at alternating Wmonth intervals.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV.

An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV.

The ont requency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating xperience has shown these component usually pass the J.* *

/,

Surveillance when performed at the

'month Frequency; e*e-Sk s rtherefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Demonstrating that all heat traced piping between the boron solution storage tank and the suction inlet to the injection anltd-pfoi pumps is unblocked ensures that there is a functioning flow f;*

path for injecting the sodium pentaborate solution.

An acceptable method for verifying that the suction piping is unblocked is~to pump from the storage tanko e

tnk.

BR4SSB 3.1-45 ifýRev

, 04/07/95 REVISIONY' *"

SLC System 3.1.7 SURVE I LLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Verify available volume of sodium 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months pentaborate solution is within the limits of Figure 3.1.7-1.

SR 3.1.7.2 Verify temperature of sodium pentaborate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months solution is within the limits of Figure 3.1.7-2.

SR 3.1.7.3 Verify temperature of pump suction piping 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is within the limits of Figure 3.1.7-2.

SR 3.1.7.4 Verify continuity of explosive charge.

31 days SR 3.1.7.5 Verify the concentration of sodium 31 days pentaborate in solution is within the limits of Figure 3.1.7-1.

AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after water or sodium pentaborate is added to sol ution AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the limits of Figure 3.1.7-2 (continued)

Amendment (Rev.

J)

I L-JAFNPP 3.1-21

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.6 Verify each SLC subsystem manual valve in 31 days the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.

SR 3.1.7.7 Verify each pump develops a flow rate In accordance

Ž 50 gpm at a discharge pressure with the 2 1275 psig.

Inservice Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem 24 months on a from pump into reactor pressure vessel.

STAGGERED TEST BASIS SR 3.1.7.9 Verify all heat traced piping between 24 months storage tank and pump suction is unblocked.

AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after piping temperature is restored within the limits of Figure 3.1.7-2 SR 3.1.7.10 Verify sodium pentaborate enrichment is Prior to S34.7 atom percent B-1O.

addition to SLC tank SR 3.1.7.11 Verify sodium pentaborate enrichment in 24 months solution in the SLC tank is 2 34.7 atom percent B-10.

Amendment (Rev.

J)

JAFNPP 3.1-22

SLC System B 3.1.7 BASES SURVEILLANCE REQUIREMENTS (continued)

SR 3.1.7.5 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron in the storage tank is maintained per Figure 3.1.7-1.

SR 3.1.7.5 must be performed anytime boron or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.

SR 3.1.7.5 must also be performed anytime the temperature is restored to within the limits of Figure 3.1.7-2, to ensure that no significant boron precipitation occurred.

The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.

SR 3.1.7.7 Demonstrating that each SLC System pump develops a flow rate

Ž 50 gpm at a discharge pressure ; 1275 psig by recirculating demineralized water to the test tank ensures that pump performance has not degraded during the surveillance interval.

This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay.

This test confirms pump and motor capability and is indicative of overall performance.

Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance.

The Frequency of this Surveillance is in accordance with the Inservice Testing Program.

SR 3.1.7.8 and SR 3.1.7.9 These Surveillances ensure that there is a functioning flow path from the boron solution storage tank to the RPV, including the firing of an explosive valve primer assembly.

The replacement primer assembly for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired.

The pump and explosive (continued)

Revision J B 3.1-43 JAFNPP

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.8 and SR 3.1.7.9 (continued)

REQUIREMENTS valve pathway tested should be alternated such that both complete flow paths are tested every 48 months at alternating 24 month intervals.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV.

An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Demonstrating that all heat traced piping between the boron solution storage tank and the suction inlet to the injection pumps is unblocked ensures that there is a functioning flow path for injecting the sodium pentaborate solution.

An acceptable method for verifying that the suction piping is unblocked is to manually initiate the system, except the explosive valves, and pump from the storage tank and recirculating it back to the storage tank.

Upon completion of this verification, the pump suction piping must be flushed with demineralized water to ensure piping between the storage tank and pump suction is unblocked.

The 24 month Frequency is acceptable since there is a low probability that the subject piping will be blocked due to precipitation of the boron from solution in the heat traced piping.

This is especially true in light of the temperature verification of this piping required by SR 3.1.7.3.

However, if, in performing SR 3.1.7.3, it is determined that the temperature of this piping has fallen below the specified minimum, SR 3.1.7.9 must be performed once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the piping temperature is restored to within the limits of Figure 3.1.7-2.

SR 3.1.7.10 Enriched sodium pentaborate solution is made by mixing granular, enriched sodium pentaborate with water.

Isotopic (continued)

Revision J B 3.1-44 JAFNPP

SDV Vent and Drain Valves B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Scram Discharge Volume (SDV)

Vent and Drain Valves BASES BACKGROUND The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram.

During a scram, the SDV vent and drain valves close to contain reactor water.

The SDV is a volumi of header piping that connects to each hydraulic control

,uitj (HCU1 and drain to an instrument volume.

There arc two SDVis-headerd* an&

instrument vol ume)V,7e~ach

-T9 r*rcigarone half of the control rod drive (CRD) discharges.

-instrument volumej

_conn

-cME adrain two valves in serie a......

"header is ected to a vent line o valves series or a total of four ven valves.

The he er piping "is sized to receive and contain the water discharged bý the CRDs during a scram. The design and functions of the SOV are described in Reference 1.

APPLICABLE SAFETY ANALYSES The Design Basis Accident and transient analyses assume all of the control rods are capable of scramming.

The acceptance criteria for the SDV vent and drain valves are that they operate automatically to:

a.

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

b.

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

Isolation of the SDV can also be accomplished by manual closure of the SDV valves.

Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves.

For a bounding leakage case, the offsite doses are well within the limits of 10 CFR 100 (Ref. 2), and adequate core cooling is maintained (Ref. 3).

The SDV vent and drain valves allow continuous drainage of the SDV during normal plant operation (continued)

B 3.1-47 ev 440

/71'e (continued)

SDV Vent and Drain Valves B 3.1.8 BASES I

SURVEILLANCE SR 3.1.8.3 (continued)

REQU IREM ENTS R Eseconds after receipt of a scram signal is based on the bounding leakage case evaluated in the accident analysis (Regf,).

Similarly, after receipt of a simulated or actual sc*-sram reset signal, the opening of the SDV vent and drain

i valves is verified.

The LOGIC SYSTEM FUNCTIONAL TEST in

-- ) LCO 3.3.1:1 and the scram time testing of control rods inn LCO 3.1.3 overlap this Surveillance to provide complete testing of the assumed safety function.

The

%*1 Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance Lz.

were performed with the reactor at power.

Operating experience' has shown these components. usually pass the Surveillance when performed at the Wx-Im i-tFrequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES I. QSAR, Section

2.

3.

NUREG-0803, 6Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping,0 August 1981.

BWR/4 STS B 3.1-51 Rev 1, 04/07/95

SDV Vent and Drain Valves B 3.1.8 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.8 Scram Discharge Volume (SDV)

Vent and Drain Valves BASES BACKGROUND APPLICABLE SAFETY ANALYSES The SDV vent and drain valves are normally open and discharge any accumulated water in the SDV to ensure that sufficient volume is available at all times to allow a complete scram.

During a scram, the SDV vent and drain valves close to contain reactor water.

The SDV is a volume of header piping that connects to each hydraulic control unit (HCU) and drains into an instrument volume.

There are two SDVs (each SDV consisting of a header and an instrument volume), each receiving approximately one half of the control rod drive (CRD) discharges.

Each instrument volume has a drain line each having two valves in series for a total of four drain valves.

Each header is connected to a separate vent line each having two valves in series for a total of four vent valves.

The header piping is sized to receive and contain all the water discharged by the CRDs during a scram.

The design and functions of the SDV are described in Reference 1.

The Design Basis Accident and transient analyses assume all of the control rods are capable of scramming.

The acceptance criteria for the SDV vent and drain valves are that they operate automatically to:

a.

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

b.

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

Isolation of the SDV can also be accomplished by manual closure of the SDV valves.

Additionally, the discharge of reactor coolant to the SDV can be terminated by scram reset or closure of the HCU manual isolation valves.

For a bounding leakage case, the offsite doses are well within the limits of 10 CFR 100 (Ref. 2), and adequate core cooling is maintained (Ref. 3).

The SDV vent and drain valves allow (continued)

Revision J JAFNPP B 3.1-46

SDV Vent and Drain Valves B 3.1.8 BASES SURVEILLANCE REQUIREMENTS SR 3.1.8.3 (continued)

(Ref. 3).

Similarly, after receipt of a simulated or actual scram reset signal, the opening of the SDV vent and drain valves is verified.

The LOGIC SYSTEM FUNCTIONAL TEST in LCO 3.3.1.1 and the scram time testing of control rods in LCO 3.1.3 overlap this Surveillance to provide complete testing of the assumed safety function.

The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when performed at the 24 month Frequency; therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES

1.

UFSAR, Section 3.5.5.2.

2.

3.

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

4.

Revision J I ni-JAFNPP B 3.1-50

SUMMARY

OF CHANGES TO ITS SECTION 3.2 - REVISION J Source of Change Summary of Change Affected Pages Retyped ITS typographical A minor typographical error in the retyped ITS has been Specification 3.2.4 error corrected. (The acronym "(FRTP)"

has been deleted, consistent with the NUREG markup.)

Retyped ITS p 3.2-5 Retyped Bases Minor typographical errors in the retyped ITS have been Specification 3.2.1 typographical errors corrected. (The words "Design Basis" have been changed to "design basis" in the ITS 3.2.1 Background section; Retyped Bases p B 3.2-1 and the words "of 0.84" has been deleted from the ITS 3.2.1 B 3.2-2 Applicable Safety Analyses section: a comma has been moved inside the end quote mark in the ITS 3.2.4 Specification 3.2.4 Background section: and the comma has been moved inside the end quote mark in two places in ITS 3.2.4 Applicable Retyped Bases p B 3.2-13 and Safety Analyses section.)

B 3.2-14 NUREG Bases markup errors Minor NUREG markup errors have been corrected to be Specification 3.2.2 consistent with the retyped ITS Bases.

(The word "Safety" has been added to ITS 3.2.2 Reference 8: the NUREG Bases markup p B 3.2 proper Reference number has been added to the ITS 3.2.4 10 Background section; and the words "Allowable value" have been changed to "Allowable Value" in the ITS 3.2.4 LCO Specification 3.2.4 Section.)

NUREG Bases markup p B 3.2 15 and B 3.2-17 Typographical error The dash has been removed from between the words "Flow" Specification 3.2.4 and "Biased" in the Background section.

NUREG Bases markup p B 3.2 14 Retyped ITS Bases p B 3.2-13 Consistency issue Minor consistency issue corrections have been made. (The Specification 3.2.1 revision and date of certain Bases References, which are also identified in the COLR, have been deleted and a NUREG Bases markup p B 3.2-4 statement that the Revision is specified in the COLR has been added.

This is consistent with Specification Retyped ITS Bases p B 3.2-4 5.6.5.)

Specification 3.2.2 NUREG Bases markup p B 3.2-9 Retyped ITS Bases p B 3.2-8 Specification 3.2.4 NUREG Bases markup p Insert Page B 3.2-19 Retyped ITS Bases p B 3.2-18 Consistency issue The word "or" has been deleted from LCO 3.2.4.a. and the Specification 3.2.4 word "Function" has been added after the words "(Flow Biased)" in all places that where the term "APRM Neutron NUREG markup p 3.2-5 and Flux - High (Flow Biased) Allowable Value" is used.

3.2-6 Retyped ITS p 3.2-5 and 3.2 6

Page 1

SUMMARY

OF CHANGES TO ITS SECTION 3.2 - REVISION J Page 2 Source of Change Summary of Change Affected Pages Editorial The proper References have been provided for ITS 3.2.1 Specification 3.2.1 and ITS 3.2.2 Bases.

NUREG Bases markup p B 3.2 2, B 3.2-4. and B 3.2-5 Retyped Bases p B 3.2-2 and B 3.2-4 Specification 3.2.2 NUREG Bases markup p B 3.2 6, B 3.2-7, and B 3.2-10 Retyped Bases p B 3.2-5, B 3.2-6, and B 3.2-9

APLHGR B 3.2.1 BASES APPLICABLE which turbine stop valve closure and turbine contra valve\\

SAFETY ANALYSES fast closure scr trips are bypassed, both high d low (continued) core flow MAPFA limits are provided for operat' n at power levels between 5% RTP and the previously menti ned bypass power level.

Te exposure dependent APLHGR limi s are reduced by MAPF Cp and MAPFACf at various operati g conditions to nsure that all fuel design crite a are met for normal Olation and AO0s.

A complete d_*isfsinoth anysis cod is-provided in Reference 9.

LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46.

The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly.

The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor.

A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated

p~l~d

o

l For single recirculation 1opoeaiont*

A

+maximum limit is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one 40 L0 P*;

recirculation loop available, resulting in a more severe OP P-P,,ki0 T.,cladding heatup, during a LOCA.

The APLHGR satisfies Criterion 2 of LCO The APLHGR limits sdeifie in the COLR are the result of the fuel design, DBAand transient analyses.

For two (w

c t

recirculation loops perating, the limit is determined am Oý a oultiplying T. smaller o the NAP Cp and PFA art S*,,,,*

l(

J imestheexmburede-n.d nt A-DING limit~d With anl one*

Vý#-

OLrecirculation loop in operation, in conformance with the requirements of LCO 3.4.1, "RecirculationLoops Operating,"

4Kt4P (continued)

S 3-R BWR/4 STS B 3.2-2 Rev 1, 04/07/95

APLHGR B 3.2.1 BASES ACTIONS L.1 (continued) operating experience, to reduce THERMAL POWER to < 25% RTP in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS REFERENCES SR 3.2.1.1

-APLHGRs are required to be initially calculated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is k 25% RTP and then every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

They are compared to the specified limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on in-n anf recognition of the slowness of changes in power distribution during normal operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER Ž 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

1.

NED6-24011-P-A GM for Reactor Fuel*

2.

)FSAR, Chapter FSAR, Chapter J61--,

4.)FSAR, Chapter 3

6.

l.pa secilc Io e-is.

7.

A6AEt A, V

Fe69

&dff freJ vuetq (Je4,f(continued)

BWR/4 STS B 3.2-4 Rev 1, 04/07/95 1,J3'b Z LIZ &-13, 1

C i?*%'( iljý vk.ý-trm I R2 e a.c 'f,,

L -. &,

( L 1 a r ago.

/197 0 C- - 3 2-j-p r" ej A, r-,' f 1 He-rt^ --c C " 444 "r-e

APLHGR B 3.2.1 BASES REFERENCES NEDO-254, Qualificatio of the One-Dime onal Cor-e (continued) transi nt Model for Boili g Water Reactors

10. -P1lt specific loss coolant accidet analysis L

AP 10 C-Fe Ia II Rev 1, 04/07/95 BWR/4 STS B 3.2-5

&ri4ý

-T

APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location.

Limits on the APLHGR are specified to ensure that the fuel design limits identified in Reference 1 are not exceeded during abnormal operational transients and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified in 10 CFR 50.46.

APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the fuel design limits are presented in References 1 and 2.

The analytical methods and assumptions used in evaluating Design Basis Accidents MDBAs),

abnormal operational transients, and normal operation that determine the APLHGR limits are presented in References 1, 2, 3, 4, 5, 6, 7, and 8.

Fuel design evaluations are performed to demonstrate that the 1% limit on the fuel cladding plastic strain and other fuel design limits described in Reference 1 are not exceeded during abnormal operational transients for operation with LHGRs up to the operating limit LHGR.

APLHGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the fuel assembly.

APLHGR limits are developed as a function of exposure to ensure adherence to fuel design limits during the limiting abnormal operational transients (Refs. 5, 6, and 7).

LOCA analyses are then performed to ensure that the above determined APLHGR limits are adequate to meet the PCT and maximum oxidation limits of 10 CFR 50.46.

The analysis is performed using calculational models that are consistent with the requirements of 10 CFR 50, Appendix K. A complete discussion of the analysis code is provided in Reference 8.

The PCT following a postulated LOCA is a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an (continued)

Revision J JAFNPP B 3.2-1

APLHGR B 3.2.1 BASES APPLICABLE SAFETY ANALYSES (continued)

LCO APPLICABILITY assembly.

The APLHGR limits specified are equivalent to the LHGR of the highest powered fuel rod assumed in the LOCA analysis divided by its local peaking factor.

A conservative multiplier is applied to the LHGR assumed in the LOCA analysis to account for the uncertainty associated with the measurement of the APLHGR.

For single recirculation loop operation, a conservative multiplier is applied to the exposure dependent APLHGR limits for two loop operation (Refs.

5 and 7).

This maximum limit is due to the conservative analysis assumption of an earlier departure from nucleate boiling with one recirculation loop available, resulting in a more severe cladding heatup during a LOCA.

The APLHGR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii)

(Ref. 9).

The APLHGR limits specified in the COLR are the result of the fuel design, and DBA and transient analyses.

For two recirculation loops operating, the limit is determined for each lattice type as a function of average planar exposure and is approved by the NRC.

With only one recirculation loop in operation, in conformance with the requirements of LCO 3.4.1, "Recirculation Loops Operating," the limit is determined by multiplying the exposure dependent APLHGR limit by a conservative multiplier determined by a specific single recirculation loop analysis (Ref. 5).

The APLHGR limits are primarily derived from fuel design evaluations and analyses of LOCAs and transients that are assumed to occur at high power levels.

Design calculations and operating experience have shown that as power is reduced, the margin to the required APLHGR limits increases.

This trend continues down to the power range of 5% to 15% RTP when entry into MODE 2 occurs.

When in MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2.

Therefore, at THERMAL POWER levels < 25% RTP, the reactor is operating with substantial margin to the APLHGR limits; thus, this LCO is not required.

(continued)

Revision J B 3.2-2 JAFNPP

APLHGR B 3.2.1 BASES (continued)

REFERENCES

1.

NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, (Revision specified in the COLR).

2.

UFSAR, Chapter 3.

3.

UFSAR, Chapter 6.

4.

UFSAR, Chapter 14.

5.

Supplemental Reload Licensing Report for James A.

FitzPatrick (Revision specified in the COLR).

6.

NEDO-24243, General Electric Boiling Water Reactor Load Line Limit Analysis For James A. FitzPatrick Nuclear Power Plant, February 1980.

7.

NEDC-32016P-1, Power Uprate Safety Analysis For James A. FitzPatrick Nuclear Power Plant, April 1993, including Errata and Addenda Sheet No.

1, dated January 1994.

8.

NEDC-31317P, Revision 2, James A. FitzPatrick Nuclear Power Plant SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis, April 1993.

9.

Revision J JAFNPP B 3.2-4

MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power.

The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.l.Z, The operating limit MCPR is established to ensure that no fuel damage results duri~ng c~~*o-r~aa

,~

r A

Although fuel damage does not necessarily occur if a uel rod actually experienced boiling transition (Ref. 1),

the critical power at which boiling transition is calculated S to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs.

Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,

the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling).

Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE SAFETY ANALYSES The analytical methods and assumptis used in evaluating

the to establish the operatin Tmit MCPR are presented in References 2, 3, 4, 5, 6, 7,: IIdU.

To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR).

The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest change in CPR (ACPR).

When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.

The NCPR operating limits derived from the transient

/*

analysis are dependent on the operating core flow and (continued)

Rev 1, 04/07/95 REVISION)3/*"

Th B 3.2-6 M

BWR T

.- A f A111M

MCPR B 3.2.2 BASES 1(hI fuel design limits duir-ing ti with moderate frequency (Re, CPR limits /redetermined methods wi key physics re, the three imensional BWR s' slow flo 'runout transients dependen) on the maximum coi (RecirculXtion Flow Control )

7ýICPR limits (MCPRp) are detrMinee mainlyb3 obnal transient code (Ref.

).

Due to the the transient response to nitial core flow levels below those at whi h the turbine stol Lnd turbine control valve f st closure scrams uigh and low flow MCPRp op rating limits are,,

ierating between 25% RTP nd the previously

(

LCO

/

,)

The MCPR operating limits specified in the COLR are the q

ali 01 result of the Design Basis Accident (DBA) and transien t

eG(P°7o4 TA /

- Mrs

()

analysis.

The operating limit MCPR ise er ne C

t-he dk Al "k,-Ve 4M k14 P*g l o h

g. a d MC I.

1 i **{

s*o ro RMcPa. OAPPLICABILITY TheM MJKoperating limits are primarily derived from Sr_*

ansienUthat are assumed to occur at high power Slevels. Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small.

Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

Statistical analyses indicate that the 41I nominal value of the initial MCPR expected at 25% RTP is 4a VIU4

> 3.5.

Studies of the variation of limiting transient

behavior have been performed over the range of power and flow conditions.

These studies'encompass the range of ojy

.nt

.rmerew vue important to typically limiting transients.

The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as (continued)

BWR/4 STS B 3.2-7 Rev 1, 04/07/95 REVISION J

If I

MCPR B 3.2.2 BASES SURVEILLANCE REQUIREMENTS SR 3.2.2.1 (continued) recognition of the slowness of changes in power distribution during normal operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER t 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis.

SR 3.2.2.2 determines the value of T, which is a measure of the actual scram speed distribution compared with the assumed distribution.

The SnMCPR operating limit is then determined based on an 4interpolation between the applicable limits for Option A (scram times of LCO 3.1.4,"Control Rod Scram Times") and Option B (realistic scram times) analyses.

The parameterr Tr/

SCY

/

must be determined once within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each set of scram time tests required by SR 3.1.4.,I.*.4 V

F*RR3.1, because the effective scram speed dist'Nbution may change d

th 1

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is acceptable due to the relatively minor changes in T expected during the R E E EN E

-NU E G. 40 11 A (Gen e

r a l

-rý

ý Z

2.

NE1240IZ11-P-A, Ieneral lectric Standard Application Mi

for Reactor Fuels -a apparo-e v rM n.on

3.

T

)FSAR, Chapter V6".

/;)

ý 41v Ai,41 t

If La S.ýFSAR, Chapter itinued)

Rev 1, 04/07/95

REVISION,

!r

' E0 BWR/4 STS B 3.2-9 le-~c 13 71R-14,,

A

-#+.i

(;c*-

,.,s 4

,÷zP-/,

J *t-..

ov

MCPR B 3.2.2 BASES REFERENCES

8.

Pant specific Aver ge Power Range Monitor, Yod Block (continued) onitor and Technic Specification Improve nts ARTS) Program].A NEOO0ý-30130-A.

eady State Nuclear Method,

May 1985.Pr h

10.

NEDO-24154, "Qual fication of the One-Dj ensional Core Transient Model or Boiling Water Reac ors October 1978

=ASES el t4 pr+P4e",*c'(

4he(-k

~

pr; I !5 73J

(

o.iv crz sor0.CczC')

io q' COLP).

pC!

Rev 1, 04/07/95 I

BWR/4 STS B 3.2-10 REVISION. "2 1'

MCPR B 3.2.2 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

BASES BACKGROUND MCPR is a ratio of the fuel assembly power that would result in the onset of boiling transition to the actual fuel assembly power.

The MCPR Safety Limit (SL) is set such that 99.9% of the fuel rods avoid boiling transition if the limit is not violated (refer to the Bases for SL 2.1.1.2).

The operating limit MCPR is established to ensure that no fuel damage results during abnormal operational transients.

Although fuel damage does not necessarily occur if a fuel rod actually experienced boiling transition (Ref. 1), the critical power at which boiling transition is calculated to occur has been adopted as a fuel design criterion.

The onset of transition boiling is a phenomenon that is readily detected during the testing of various fuel bundle designs.

Based on these experimental data, correlations have been developed to predict critical bundle power (i.e.,

the bundle power level at the onset of transition boiling) for a given set of plant parameters (e.g., reactor vessel pressure, flow, and subcooling).

Because plant operating conditions and bundle power levels are monitored and determined relatively easily, monitoring the MCPR is a convenient way of ensuring that fuel failures due to inadequate cooling do not occur.

APPLICABLE SAFETY ANALYSES The analytical methods and assumptions used in evaluating the abnormal operational transients to establish the operating limit MCPR are presented in References 2, 3, 4, 5, 6, 7, 8, and 9.

To ensure that the MCPR SL is not exceeded during any transient event that occurs with moderate frequency, limiting transients have been analyzed to determine the largest reduction in critical power ratio (CPR).

The types of transients evaluated are loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient yields the largest change in CPR (ACPR).

When the largest ACPR is added to the MCPR SL, the required operating limit MCPR is obtained.

(continued)

Revision J JAFNPP B 3.2-5

MCPR B 3.2.2 BASES APPLICABLE SAFETY ANALYSES (continued)

LCO The MCPR operating limits derived from the transient analysis are dependent on the operating core flow and core exposure to ensure adherence to fuel design limits during the worst transient that occurs with moderate frequency (Refs. 6, 7, 8, and 9).

A generator load reject without bypass and a feedwater controller transient normally result in the worst case MCPR transients for a given fuel cycle.

During operations at low core flows the MCPR operating limit must be increased by a factor of Kf (specified in the COLR) which is derived from the recirculation flow runout transient and is a function of core flow.

This will ensure the MCPR safety limit is not exceeded during a recirculation flow runout event.

The MCPR satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii)

(Ref. 10).

The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis.

The operating limit MCPR is a function of exposure, control rod scram times and core flow.

The MCPR values for each fuel assembly must remain above the operating limit MCPR.

APPLICABILITY The MCPR operating limits are primarily derived from the analyses of transients that are assumed to occur at high power levels.

Below 25% RTP, the reactor is operating at a minimum recirculation pump speed and the moderator void ratio is small.

Surveillance of thermal limits below 25% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL is not exceeded even if a limiting transient occurs.

Statistical analyses indicate that the nominal value of the initial MCPR expected at 25% RTP is

> 3.5.

Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions.

These studies encompass the range of actual values for key plant parameters important to typically limiting transients.

The results of these studies demonstrate that a margin is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 25% RTP.

This trend is expected to (continued)

Revision J JAFNPP B 3.2-6

MCPR B 3.2.2 BASES SURVEILLANCE SR 3.2.2.1 (continued)

REQUIREMENTS The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER

Ž 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

SR 3.2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis.

SR 3.2.2.2 determines the value of T, which is a measure of the actual scram speed distribution compared with the assumed distribution.

The MCPR operating limit is then determined based on an interpolation between the applicable limits for Option A (scram times of LCO 3.1.4,"Control Rod Scram Times") and Option B (realistic scram times) analyses.

The parameter T must be determined once within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after each set of scram time tests required by SR 3.1.4.1, SR 3.1.4.2, and SR 3.1.4.4 because the effective scram speed distribution may change during the cycle or after maintenance that could affect scram times.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is acceptable due to the relatively minor changes in T expected during the fuel cycle.

REFERENCES

1. NUREG-0562, Fuel Rod Failure as a Consequence of Departure From Nucleate Boiling or Dry Out, June 1979.

2.

NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, (Revision specified in the COLR).

3.

UFSAR, Chapter 3.

4.

UFSAR, Chapter 6.

5.

UFSAR, Chapter 14.

6.

NEDO-24281, FitzPatrick Nuclear Power Plant Single Loop Operation, August 1980.

(continued)

Revision J JAFNPP B 3.2-8

MCPR B 3.2.2 BASES REFERENCES (continued)

7.

NEDO-24243, General Electric Boiling Water Reactor Load Line Limit Analysis For James A. FitzPatrick Nuclear Power Plant, February 1980.

8.

NEDC-32016P-1, Power Uprate Safety Analysis For James A. FitzPatrick Nuclear Power Plant, April 1993, including Errata and Addenda Sheet No.

1, dated January 1994.

9.

Supplemental Reload Licensing Report for James A.

FitzPatrick (Revision specified in the COLR).

10.

Revision J JAFNPP B 3.2-9

APRM Gain and Setpoin' 3.2 POWER DISTRIBUTION LIMITS 3.2.4 Average Power Range Monitor (APRM)

Gain and Setpoinl LCO 3.2.4

a.

MFLPD shall be less than or equal to Fraction of RTP;

b. Each required APRM\\

specified in the COLR shall be made appllicable; or C.

tacn re uirea rKM !

APPLICABILITY:

THERMAL POWER Ž 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months LCO not met.

requirements of the LCO.

B.

Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to < 25% RTP.

Time not met.

3.2-5 REVISION 0 f1e.

@KEEý

ýýFTEý

APRM Gain and Setpoint

(

,3bi SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1

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

Not required to be met if SR 3.2.4.2 is satisfied for LCO 3.2.4 (

b or c requirements.

Verify MFLPD is within limits.

Once within LCZ-¶ 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

Ž 25% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter SR 3.2.4.2 NOTE--------------

Not required to be met if SR 3.2.4.1 is satisfied for LCO3.2.4 1 a

requirements.

324 a or gains are adjusted TPD.

3.2-6 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Rev 1, 04/07/95 REVISION2r',

BWR/4 STS

APRM B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 Average Power Range Monitor (APRM)

Gain and Setpoint$

BASES BACKGROUND The OPERABILITY of the APRtqs and their setpoints is an initial condition of all safety anal ses that assume ro,

-}

insertion upon reactor scram. i. pplIca

~e G~s are IC 10*

/Re~tr

.ein, D1*-Instrumntatimq and Contr~q'

_(GC*,

uproeto S

em Functions," and DC 23,

'Prote tion againt Anticipated Operation Oc u rence (Re

. 1 This LCO is provided to require t e APRM gain or ApRM rTW__eM gse= scrm *int to be adjusted when operating under conditions of excessive power peaking to rmaintain acceptable margin to the fuel cladding integrity Safety Limit (SL) and the fuel cladding 1% plastic strain

/limit.

I The condition of excessive power peaking is determined by the ratio of the actual power peaking to the limiting power peaking at RTP.

This ratio is equal to the ratio of the core limiting MFLPD to the Fraction of RTP (FRTP),

where FRTP is the measured THERMAL POWER divided by the RTP.

Excessive power peaking exists when:

MFLPD > 1, FRTP indicating that MFLPD is not decreasing proportionately to the overall power reduction, or conversely, that power peaking is increasing.

To maintain margins similar to those at RTP conditions, the excessive power peaking is compensated by a gain adjustment on the APRMs or adjustment e APRM1 Q*

ý.

Either of these adjustments has effectively the same result as maintaining MFLPD less than('

or equal to FRTP and thus maintains RTP margins for APLHGR?

40 CP1.NEvfre).

Ftux. -

1 (P/k.

&iuPeJ uVý- 6^

The normally selected APRM Wf positio he scram above the upper bound of the normal power/flow operating region that has been considered in the design of the fuel rods.

Thee flow biased with a slope that approximates the upper flow control line, such that an approximately constant margin is maintainedbetween the flow biased trip level and the upper operating boundary for core flows in excess of about 45% of rated core flow.

In the range of infrequent operations below 45% of rated core flow, (continued)

Try' B 3.2-14 ILA Jr,

APRM Gain and Setpoint4 C-3 BASES (J

BACKGROUND the margin to scr is reduced because of the nonlinear core (continued) flow versus drive flow relationship.

The normally selected supported by the analyses presented i e erences n

, that concentrate on events initiated from rate Wconlitions.

Design experience has shown that minimum Alhkwa&

VaCU i

deviations occur with.in expected margins to operating limits 14CP

, at rated conditions for normal power distributions.

However, at other than rated conditions, control rod patterns can be established that significantly reduce the margin to thermal limits.

Therefore, the Ne 1

'125 APRM4 E

may be reduced during operation eYo w en e combination of THERMAL POWER and MFLPD indicates an CF(ouj %aM Tv4d excessive power peaking distribution,#

he APRM neutron flux sign e

o more closely follow the fuel adding heat flux during power transients.

The APR14 utron flux signal is a measure o the core thermal pow during steady state operation.

During power trans* nts, the APRM signal leads the ual core thermal pow response because of the fuel t al time constant.

Ther ore, on power increase transie s, the APRM signal provi a conservatively high measur f core thermal pow By passing the APRM signal rough an electroni filter with a time constant 1 s than, but approxi tely equal to, that of the fu thermal time const t, an APR3 transient response hat more closely foll ws actual fuel cladding heat ux is obtained, whi a

co ervative margin is maintaine.

The delayed resp e of e filtered APRM signal allow he flow-biasedAP scram.

,evels to be positioned clos to the upper boun of the normal power and flow range without unnecess.

ly causing reactor scrams during sho duration neutro nlux spikes.

These spikes can be cau by insignifica transients such as performance of mai team line valve rveillances or momentary flow incr es of only sever percent.

APPLICABLE

AFFTY ANAL YSES The acceptance criteria for the APRM gain or setpoint adjustments are that acceptable margins (to APLHGR, MCP.)

be maintained to the fuel cladding integrity bL anu tne TUe!

cladding 1% plastic strain limit.

(~

safety analyses (Refs. 2 and-3) concentrate on the ra"ed power condition for which the minimum expected margin to the operating limits (APLHGR MCPR occurs.

r (continued)

Rev 1, 04/07/95 BWR/4 STS

)

A16--XU0- Volve-B 3.2-15 (fed

APRM Gain and Setpoint$ q4_.*

AR4Gi B 3.2.4 BASES LCO (continued)

c.

Increasing APRM gains to causee"thee APR1 to read greater than 10 times MFLPD J

This condition is to account for the reduction in margin to the fuel cladding integrity SL and the fuel cladding 1% plastic strain limit.

MFLPD is the ratio of the limiting LHGR to the LHGR limit for the specific bundle type.

As power is reduced, if the design power distribution is maintained, MFLPD is reduced in proportion to the reduction in power.

However, if power peaking increases above the design value, the MFLPD is not reduced in proportion to the reduction in power.

Under these conditions, the APR1 gain is adjusted upward or t-4r s

AFRR2*ow

____dsr__ina.i~reduce4accoidingly.

When the reactor is operating with peaking less than the(W'"3) design value, it is not necessary to modify the APRM14

-> ias "o

Adjusting APRM gain or -ela*tsi

),equivalent to MFLPD less than or equal to FRTP, as stated in Stthe LCO..

For compliance with LCO Item b (APR1 thj) or Item c (APR14 gain adjustment), only APRMs required to be OPERABLE per LCO 3.3.1.1, *ec s m rs mDen aare required to be adjusted.

In addition, each APRM may be allowed to have its gain or 4--M---1--*

adjusteddindependently of other APRP4s that are having thei\\rV6ouý/)1 gain or s n

adjusted.

APPLICABILITY The MFLPD limit, APRM gain adjustment, and APRM Mlbiild provided to ensure that the fuel cladding integrity SL and the fuel cladding 1% plastic strain limit are not violated during design basis transients.

As discussed in the Bases for LCO 3.2.1 and LCO 3.2.2, sufficient margin to these limits exists below 25% RTP and, therefore, these requirements are only necessary when the reactor is operating at ; 25% RTP.

If the APR14 gain o 01no within limits while the MFLPD has exceede-FRT, the margin to the fuel cladding integrity SL and the fuel cladding 1% plastic strain limit (continued)

Rev 1, 04/07/95

/

)

BWR/4 STS B 3.2-17

O'

( Insert Ref

3.

NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, (Revision specified in the COLR).

4.

NEDO-31960-A, BWR Owners' Group Long Term Stability Solutions Licensing Methodology, June 1991.

5.

NEDO-31960-A, Supplement 1, BWR Owners' Group Long Term Stability Solutions Licensing Methodology, March 1992.

6.

GENE-637-004-0295, Application Of The "Regional Exclusion With Flow-Biased APRM Neutron Flux Scram" Stability Solution (Option I-D) To The James A.

FitzPatrick Nuclear Power Plant, February 1995.

7.

Insert Page B 3.2-19 Revision J

APRM Gain and Setpoint 3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 Average Power Range Monitor (APRM)

Gain and Setpoint

a.

MFLPD shall be less than or equal to Fraction of RTP;

b.

Each required APRM Neutron Flux-High Function Allowable Value specified in made applicable: or (Flow Biased) the COLR shall

c.

Each required APRM gain shall be adjusted as specified in the COLR.

APPLICABILITY:

THERMAL POWER Ž 25% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Satisfy the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months LCO not met.

requirements of the LCO.

B.

Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion to < 25% RTP.

Time not met.

Amendment (Rev.

J)

LCO 3.2.4 be I6\\

JAFNPP 3.2-5

APRM Gain and Setpoint 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 NOTE ------------------

Not required to be met if SR 3.2.4.2 is satisfied for LCO 3.2.4.b or LCO 3.2.4.c requirements.

Verify MFLPD is within limits.

Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after

> 25% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter SR 3.2.4.2

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

Not required to be met if SR 3.2.4.1 is satisfied for LCO 3.2.4.a requirements.

Verify required APRM Neutron Flux-High 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (Flow Biased) Function Allowable Value or ARPM gains are adjusted for the calculated MFLPD.

Amendment (Rev. J)

JAFNPP 3.2-6

APRM Gain and Setpoint B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 Average Power Range Monitor (APRM)

Gain and Setpoint BASES BACKGROUND The OPERABILITY of the APRMs and their setpoints is an initial condition of all safety analyses that assume rod insertion upon reactor scram.

Applicable design criteria is discussed in UFSAR, Section 16.6 (Ref. 1).

This LCO is provided to require the APRM gain or APRM Neutron Flux-High (Flow Biased) Function Allowable Value (LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation," Function 2.b) to be adjusted when operating under conditions of excessive power peaking to maintain acceptable margin to the fuel cladding integrity Safety Limit (SL) and the fuel cladding 1% plastic strain limit.

The condition of excessive power peaking is determined by the ratio of the actual power peaking to the limiting power peaking at RTP.

This ratio is equal to the ratio of the core limiting MFLPD to the Fraction of RTP (FRTP),

where FRTP is the measured THERMAL POWER divided by the RTP.

Excessive power peaking exists when:

MFLPD 1,

FRTP indicating that MFLPD is not decreasing proportionately to the overall power reduction, or conversely, that power peaking is increasing.

To maintain margins similar to those at RTP conditions, the excessive power peaking is compensated by a gain adjustment on the APRMs or adjustment of the APRM Neutron Flux-High (Flow Biased) Function Allowable Value.

Either of these adjustments has effectively the same result as maintaining MFLPD less than or equal to FRTP and thus maintains RTP margins for APLHGR,

MCPR, and LHGR.

The normally selected APRM Neutron Flux-High (Flow Biased)

Function Allowable Value positions the scram above the upper bound of the normal power/flow operating region that has been considered in the design of the fuel rods.

The Allowable Value is flow biased with a slope that approximates the upper flow control line, such that an approximately constant margin is maintained between the flow biased trip level and the upper operating boundary for core (continued)

Revision J JAFNPP B 3-2-13

APRM Gain and Setpoint B 3.2.4 BASES BACKGROUND (continued) flows in excess of about 45% of rated core flow.

In the range of infrequent operations below 45% of rated core flow, the margin to scram is reduced because of the nonlinear core flow versus drive flow relationship.

The normally selected APRM Allowable Value is supported by the analyses presented in References 2 and 3 that concentrate on events initiated from rated conditions.

Design experience has shown that minimum deviations occur within expected margins to operating limits (APLHGR,

MCPR, and LHGR),

at rated conditions for normal power distributions.

However, at other than rated conditions, control rod patterns can be established that significantly reduce the margin to thermal limits.

Therefore, the APRM Neutron Flux-High (Flow Biased) Function Allowable Value may be reduced during operation when the combination of THERMAL POWER and MFLPD indicates an excessive power peaking distribution.

In addition, the APRM Neutron Flux-High (Flow Biased) Function provides protection from reactor thermal hydraulic instability consistent with Boiling Water Reactors Owners' Group Long-Term Solution, Option I-D (Refs. 4, 5 and 6).

APPLICABLE SAFETY ANALYSES The acceptance criteria for the APRM gain or setpoint adjustments are that acceptable margins (to APLHGR,

MCPR, and LHGR) be maintained to the fuel cladding integrity SL and the fuel cladding 1% plastic strain limit.

The safety analyses (Refs. 2 and 3) concentrate on the rated power condition for which the minimum expected margin to the operating limits (APLHGR, MCPR, and LHGR) occurs.

LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR),"

and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),"

and LCO 3.2.3, "Linear Heat Generation Rate (LHGR),"

limit the initial margins to these operating limits at rated conditions so that specified acceptable fuel design limits are met during transients initiated from rated conditions.

At initial power levels less than rated levels, the margin degradation of either the APLHGR, the MCPR, or the LHGR during a transient can be greater than at the rated condition event.

This greater margin degradation during the transient is primarily offset by the larger initial margin to limits at the lower than rated power levels.

However, power distributions can be hypothesized that would result in reduced margins to the pre-transient operating limit.

When combined with the increased severity of certain transients (continued)

Revision J B 3.2-14 JAFNPP

APRM Gain and Setpoint B 3.2.4 BASES SURVEILLANCE SR 3.2.4.1 and SR 3.2.4.2 (continued)

REQUIREMENTS circuitry.

SR 3.2.4.1 and SR 3.2.4.2 have been modified by Notes which clarify that the respective SR does not have to be met if the alternate requirement demonstrated by the other SR is satisfied.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of SR 3.2.4.1 is chosen to coincide with the determination of other thermal limits, specifically those for the APLHGR (LCO 3.2.1) and LHGR (LCO 3.2.3).

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on the recognition of the slowness of changes in power distribution during normal operation.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER 2 25% RTP is achieved is acceptable given the large inherent margin to operating limits at low power levels.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency of SR 3.2.4.2 requires a more frequent verification than if MFLPD is less than or equal to FRTP.

When MFLPD is greater than FRTP, more rapid changes in power distribution are typically expected.

REFERENCES

1.

UFSAR, Section 16.6.

2.

UFSAR, Section 14.5.

3.

NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, (Revision specified in the COLR).

4.

NEDO-31960-A, BWR Owners' Group Long Term Stability Solutions Licensing Methodology, June 1991.

5.

NEDO-31960-A, Supplement 1, BWR Owners' Group Long Term Stability Solutions Licensing Methodology, March 1992.

6.

GENE-637-044-0295, Application Of The "Regional Exclusion With Flow-Biased APRM Neutron Flux Scram" Stability Solution (Option I-D) To The James A.

FitzPatrick Nuclear Power Plant, February 1995.

7.