View source for 2CAN030901, License Amendment Request, Technical Specification Change to Modify the Mode of Applicability for Emergency Feedwater Actuation Signal

{{Adams
| number = ML090630163
| issue date = 03/02/2009
| title = License Amendment Request, Technical Specification Change to Modify the Mode of Applicability for Emergency Feedwater Actuation Signal
| author name = Mitchell T
| author affiliation = Entergy Operations, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000368
| license number = NPF-006
| contact person = 
| case reference number = 2CAN030901
| document type = Letter type:, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specifications
| page count = 18
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Entergy Operations, Inc.
1448 S.R. 333 Russellville, AR 72802 Tel 479-858-3110 Timothy G. Mitchell Vice President, Operations Arkansas Nuclear One 2CAN030901 March 2, 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

==SUBJECT:==
License Amendment Request Technical Specification Change To Modify the Mode of Applicability for Emergency Feedwater Actuation Signal Arkansas Nuclear One, Unit 2 Docket No. 50-368 License No. NPF-6

==Dear Sir or Madam:==

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment for Arkansas Nuclear One, Unit 2 (ANO-2). The proposed change will modify Technical Specification (TS) 3.3.1.1, Reactor Protective Instrumentation, and TS 3.3.2.1, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, specifically Table 3.3-1, Table 4.3-1, and Table 3.3-3 (respectively) to adopt a Mode of Applicability for the Logarithmic (Log) Power Level High, Pressurizer Pressure Low, Steam Generator (SG)
Pressure Low, and the SG Differential Pressure and Level Low functions to be consistent with the improved Standard TSs (STS) of NUREG 1432, Revision 3.1, for Combustion Engineering Plants.
ANO-2 is a Combustion Engineering commercial nuclear facility and operates a pressurized water reactor (PWR). The unit is designed to provide primary heat removal during normal shutdown, anticipated transients, and accident events via two SGs. Makeup to these SGs below operational Mode 1 is normally provided by the normal feedwater system via an Auxiliary Feedwater (AFW) pump, or by the Emergency Feedwater (EFW) system following an accident. The makeup systems are designed to support a plant cooldown to Mode 4 conditions (< 300 &deg;F) where the Shutdown Cooling (SDC) system is placed in service. In accordance with this design basis, TS 3.7.1.2, Emergency Feedwater System, requires EFW to be operable in Modes 1, 2, and 3. This is the original Mode of Applicability for this TS since initial plant startup in the mid-1970s.
During plant shutdown for the Spring 2008 refueling outage, ANO-2 operators noted that the TS Mode of Applicability associated with automatic actuation and operation of the EFW systems required operable EFW-related actuation channels in Modes 1, 2, 3, and 4. Requiring operable EFW-related actuation channels in Mode 4 was found to be unnecessary, since the EFW components expected to be actuated were not required to be operable in Mode 4. This

2CAN030901 Page 2 of 3 TS Mode of Applicability has also been unchanged since initial plant startup in the mid-1970s.
Based on this discrepancy, Entergy proposes to revise the Mode of Applicability for the EFAS related functions in TS 3.3.3.2, Table 3.3-3, to be equivalent to that required by the EFW TS 3.1.7.2 and the plant design basis.
Based on the above, a further review of TS instrumentation requirements was performed.
With regard to TS 3.3.1.1, the review identified that the Mode of Applicability for reactor trip on Log Power Level High, Pressurizer Pressure Low, and SG Pressure Low deviated from the STS. Therefore, Entergy proposes to revise the Mode of Applicability for these requirements to be consistent with the STS. The relevant history and justification for change associated with these functions is included in Attachment 1.
In addition to the above, the associated Table 3.3-3 Action notes are revised. These notes were reviewed in light of the proposed Mode of Applicability change described above. During this review, it was discovered that some Action notes associated with equipment required to be operable in Modes 1, 2, and 3 required the operator to place the plant in Mode 5 if an inoperable component was not restored to an operable status within the allotted time period.
In such cases, the Action note should only require exiting the Mode of Applicability, or in other words, place the plant in Mode 4. Those Action notes that are associated with components where some are required in Modes 1, 2, and 3 while others are required in Modes 1, 2, 3, and 4 were revised to simply state that to exit the MODE(s) of Applicability within the currently specified time frame.
The associated TS Bases, controlled in accordance with the TS Bases Control Program of TS 6.5.14, will also be revised accordingly upon approval of this request.
The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that the changes involve no significant hazards consideration. The bases for these determinations are included in the attached submittal.
The proposed change does not include any new commitments.
Entergy requests approval of the proposed amendment by April 1, 2010. Once approved, the amendment shall be implemented within 90 days.
If you have any questions or require additional information, please contact David Bice at 479-858-5338.
I declare under penalty of perjury that the foregoing is true and correct. Executed on March 2, 2009.
Sincerely, TGM/dbb Attachments:

2CAN030901 Page 3 of 3
: 1. Analysis of Proposed Technical Specification Change
: 2. Proposed Technical Specification Changes (mark-up) cc:   Mr. Elmo E. Collins Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One P. O. Box 310 London, AR 72847 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS O-7 D1 Washington, DC 20555-0001 Mr. Bernard R. Bevill Director Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Attachment 1 2CAN030901 Analysis of Proposed Technical Specification Change to 2CAN030901 Page 1 of 8

==1.0     DESCRIPTION==

This letter is a request to amend Operating License NPF-6 for Arkansas Nuclear One, Unit 2 (ANO-2).
The proposed change will modify Technical Specification (TS) 3.3.1.1, Reactor Protective Instrumentation, and TS 3.3.2.1, Engineered Safety Feature Actuation System (ESFAS)
Instrumentation, specifically Table 3.3-1, Table 4.3-1, and Table 3.3-3 to adopt a Mode of Applicability for the Logarithmic (Log) Power Level High, Pressurizer Pressure Low, Steam Generator (SG) Pressure Low, and the SG Differential Pressure and Level Low functions to be consistent with the improved Standard TSs (STS) of NUREG 1432, Revision 3.1, for Combustion Engineering Plants.
In addition, Action notes associated with TS Table 3.3-3 are revised to match the correct Modes of Applicability as stated in the table.

==2.0     PROPOSED CHANGE==

The Pressurizer Pressure Low and SG Pressure Low reactor trip functions Mode of Applicability in TS 3.3.1.1, Tables 3.3-1 and 4.3-1 (Functions 5 and 7) are revised from Modes 1, 2, 3*, 4*, 5*
to Modes 1, 2 (* indicates that the reactor Trip Circuit Breakers (TCBs) are closed and capable of Control Element Assembly (CEA) withdrawal).
The Log Power Level High trip function Mode of Applicability in TS 3.3.1.1, Tables 3.3-1 and 4.3-1 (Functions 3.b and 3, respectively) is revised from Modes 3, 4, 5 to Modes 3*, 4*, 5*
(* indicates that the TCBs are closed and capable of CEA withdrawal).
The Emergency Feedwater Actuation Signal (EFAS) SG Differential Pressure and Level Low function Mode of Applicability in TS 3.3.2.1, Table 3.3-3, along with its supporting Manual Trip Pushbuttons, Matrix Logic, Initiation Logic, and Automatic Actuation Logic requirements (Functions 8.a, 8.b, 8.c, 8.d.1, 8.d.2, and 8.e) are revised from Modes 1, 2, 3, 4 to Modes 1, 2, 3. In addition, Action notes associated with TS Table 3.3-3 are revised to match the correct Modes of Applicability as stated in the table.
The associated TS Bases, controlled in accordance with the TS Bases Control Program of TS 6.5.14, will also be revised accordingly upon approval of this request.

==3.0     BACKGROUND==

The Emergency Feedwater (EFW) system is designed to provide means of supplying water to the intact SG(s) following a postulated main steam line rupture or loss of main feedwater (MFW) event to remove reactor decay heat and provide for cooldown of the Reactor Coolant System (RCS) to within the temperature and pressure at which the shutdown cooling (SDC) system can be placed in operation. The system employs two safety related pumps (turbine driven and motor driven) with two independent feedwater trains, each capable of supplying either of the two SGs, and a non-safety related pump (see next paragraph) whose discharge line is interconnected to each of the safety-related EFW pump's discharge line and to the MFW discharge line. The EFW system is designed so the non-safety auxiliary feedwater (AFW) pump may be used to supply water to the SGs during non-emergency conditions to avoid challenging to 2CAN030901 Page 2 of 8 the safety related EFW pumps. The AFW pump is capable of supplying sufficient water to the SGs for heat load of approximately 4% of full plant power at maximum SG pressure.
The AFW pump is designed to supply condensate to the SGs during plant startup, hot standby, hot shutdown, and normal cooldown and operates in conjunction with the main steam system to maintain SG level.
During an emergency condition, the safety-related EFW pumps are designed to automatically supply water to the SGs upon the initiation of an EFAS or a Diverse Emergency Feedwater Actuation Signal (DEFAS). In addition, in the unlikely failure of both safety related EFW pumps during an emergency condition, the AFW pump can be manually actuated to supply water to the SGs.
When an EFAS is received, both EFW pumps start and provide water to the intact SGs as required by level control. Simultaneously, all valves in the discharge lines will open. However, if a SG isolation is required, as in the case of a postulated main steam line break (MSLB), the EFAS will open only the valves leading to the intact SG and close the valves leading to the isolated SG. A combination of measured variables (level and pressure) for each steam generator are used to determine which SG(s) are intact. The valves to a SG that is not intact will not receive an EFAS and will therefore be closed. Should one of these valves receive a spurious signal to open to a SG which is not intact, the other valve in series will be closed precluding feeding of the SG. This allows long term secondary cooling following the event.
However, the MSLB is an overcooling event that does not challenge the minimum capacity of the EFW pumps. Given that a significant amount of energy will be extracted from the primary during the event, the post event demand on the EFW system will be less than that from either the loss of MFW or the MFW line break events.
The EFW system instrumentation and controls are designed for operation during all phases of plant operation. The system is designed to be operated manually, typically utilizing the AFW pump during normal transients such as startup, shutdown, and hot standby, and automatically utilizing the EFW pumps during emergency situations such as steam line rupture, loss of normal feed, and plant blackout.
The EFAS performs the following functions:
A. Starts the EFW pumps; B. Determines whether a SG is intact; C. Opens the EFW valves to the intact SG; and, D. Prevents a high level condition in the intact SG(s) by closing the EFW valves when the water level is re-established above the low level trip setpoint.
The EFAS is initiated to SG A either by a low SG level coincident with no low pressure trip present on the SG or by a low SG level coincident with a differential pressure between the two SGs with the higher pressure in SG A (an identical EFAS is generated for SG B). Following initial actuation when the level is re-established above the low level actuation point, or when the proper pressure conditions are no longer met, the valve close logic, closes the valves. When the conditions for valve opening are again met, or when the operator takes action, the valves are reopened. The pumps remain running whenever an EFAS signal occurs for either SG.
to 2CAN030901 Page 3 of 8 With regard to reactor trip, the SG Pressure Low and Pressurizer Pressure Low both act to trip the reactor upon a MSLB. During a MSLB, SG pressure will decrease and SG steam flow will increase, resulting in a cooldown of the RCS. Pressurizer pressure will drop as a result of the cooldown. In most cases, both low pressure setpoints will be reached to cause a reactor trip during a MSLB. The Pressurizer Pressure Low trip also protects against a Loss of Coolant Accident (LOCA). The Pressurizer Pressure Low trip is also accompanied by a Departure from Nucleate Boiling (DNBR) trip generated by the Core Protection Calculators (CPCs).
The Logarithmic (Log) Power Level trip aids in protecting against inadvertent reactor startup due to an uncontrolled withdrawal of Control Element Assemblies (CEAs) or an uncontrolled boron dilution event.
The EFW and AFW pumps are further described in the ANO-2 Safety Analysis Report (SAR)
Section 10.4.9. The EFAS, including the determination of the intact SG, is described in ANO-2 SAR Section 3.7.1.1.11.8. RPS and ESF related instrumentation is further discussed in ANO-2 SAR Chapter 7.

==4.0     TECHNICAL ANALYSIS==

ANO-2 is designed to provide primary heat removal during normal shutdown, anticipated transients, and accident events via two SGs. Makeup to these SGs below operational Mode 1 is normally provided by the non-safety feedwater system via an AFW pump, or by the EFW system following an accident or other transient that causes a low SG level setpoint to be reached. The makeup systems are designed to support a plant cooldown to Mode 4 conditions
(< 300 &deg;F), where the Shutdown Cooling (SDC) system can be placed in service. In accordance with this design basis, TS 3.7.1.2, Emergency Feedwater System, requires EFW to be operable in Modes 1, 2, and 3. This is the original Mode of Applicability for this TS since initial plant startup in the mid-1970s.
During plant shutdown for the Spring 2008 refueling outage, ANO-2 operators noted that the TS Mode of Applicability associated with automatic actuation and operation of the EFW systems required operable EFW-related actuation channels in Modes 1, 2, 3, and 4. Requiring operable EFW-related actuation channels in Mode 4 was found to be unnecessary, since the EFW components expected to be actuated were not required to be operable in Mode 4. This TS Mode of Applicability has also been unchanged since initial plant startup in the mid-1970s.
Based on this discrepancy, Entergy proposes to revise the Mode of Applicability for the EFAS related functions in TS 3.3.3.2, Table 3.3-3, to be equivalent to that required by the EFW TS 3.1.7.2 and the plant design basis.
The EFW system is designed to provide makeup to the SGs following a feedwater line break (FWLB) or MSLB. The Background Section above describes EFW response and function when either of these events occur in Modes 1, 2, or 3. In Mode 4, the SDC system is placed in operation and used for decay heat removal purposes. EFW may be maintained available in this Mode as a backup, but is not required by plant design or by TSs.
SG temperature and pressure is relatively low during Mode 4 operation (temperature < 300 &deg;F and pressure < 70 psia). Because SDC is available, a FWLB will not prevent a means of continued primary system cooling in Mode 4, even with EFW unavailable. A Mode 4 MSLB would cause a cooldown of the primary system until the affected SG blows dry. At this point, to 2CAN030901 Page 4 of 8 SDC again will be the primary means to maintain post cooldown temperature of the primary system. This design illustrates that EFW makeup is not required in Mode 4.
Functional Unit 8.b and 8.c act to determine which SG is intact following a MSLB and permit makeup to the intact SG. Because a MSLB results in a cooldown of the primary system, continuing to feed a MSLB-affected SG would enhance the cooldown. SG isolation is initiated by a low SG pressure signal. This same signal blocks the feeding of the SGs from any source.
As stated previously, feeding of the intact SG is not required in Mode 4 since SDC is the primary source for decay heat removal. If SDC was lost and EFW is unavailable, feedwater may remain available via the AFW pump or the non-safety Condensate pump contained within the MFW system. Although the AFW and MFW sources are lost on a loss of offsite power event, ANO has an Alternate AC Diesel Generator (station blackout rule) that can be quickly started and connected to non-vital electrical buses to restore power to AFW or MFW components as needed. In addition, procedures are available that provide step-by-step guidance for restoring AFW/MFW capability following an MSLB, if needed.
The original ANO-2 TSs contained a Mode of Applicability for the EFAS SG Pressure Differential Pressure and Level Low functions of Modes 1, 2, 3, and 4. Likewise, original and current EFW TSs only require the EFW system to be operable in Modes 1, 2, and 3 (see TS 3.7.1.2). ANO-2 TS Amendment 159 added similar criteria (as described above) and Mode of Applicability for the Matrix Logic, Initiation, Logic, and Automatic Actuation Logic for the EFAS SG Pressure Differential Pressure and Level Low functions. Based on the information provided, Entergy believes it is acceptable and necessary to change the applicability of the parameters associated with this function from Modes 1, 2, 3, and 4, to Modes 1, 2, and 3, consistent with the STS and design basis of the unit.
In addition to the above, the associated Table 3.3-3 Action notes are revised. These notes were reviewed in light of the proposed Mode of Applicability change described above. During this review, it was discovered that some Action notes associated with equipment required to be operable in Modes 1, 2, and 3 required the operator to place the plant in Mode 5 if an inoperable component was not restored to an operable status within the allotted time period. In such cases, the Action note should only require exiting the Mode of Applicability, or in other words, place the plant in Mode 4. Those Action notes that are associated with components where some are required in Modes 1, 2, and 3 while others are required in Modes 1, 2, 3, and 4 were revised to simply state that to exit the MODE(s) of Applicability within the currently specified time frame. This is consistent with TS Limiting Condition for Operation (LCO) 3.0.1 which states:
Limiting Conditions for Operation and ACTION requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for each specification.
The overall justification for this change is provided by the original (and current) plant design basis. In addition, the above discussion illustrates significant defense-in-depth with regard to primary system cooling sources. The proposed change resolves a discrepancy between the TS required Mode of Applicability of the EFW system and the instrumentation actuation channels that automatically initiate and control the EFW system. The proposed Mode of Applicability for the EFAS channels is also consistent with the STS.
to 2CAN030901 Page 5 of 8 With regard to reactor trip functions, the Mode of Applicability for the Log Power Level High, Pressurizer Pressure Low, and SG Pressure Low trips are modified to be consistent with the STS.
With regard to the Pressurizer Pressure Low and SG Pressure Low trip functions, the original ANO-2 TSs contained a Mode of Applicability consistent with the STS. However, in Amendment 24 to the ANO-2 TS (June 19, 1981), the Mode of Applicability was changed to include conditions when TCBs were closed in order to support the Cycle 2 core reload analysis for a postulated MSLB accident. Amendment 24 changed the Mode of Applicability from Modes 1, 2 to Modes 1, 2 * (the
* relating to TCB position). In ANO-2 TS Amendment 159 (May 10, 1995),
the Mode of Applicability was changed to Modes 1, 2, 3*, 4*, 5* since there was confusion with regard to what Modes 1, 2
* implied. The latter amendment was considered an editorial change and no additional justification was provided.
Entergy has verified that the addition of the
* is no longer required based on administrative and TS controls that prevent an unintended approach to criticality in Modes 3, 4, and 5. Other trip functions protect against inadvertent criticality from a subcritical condition. Note also that procedures require declaration of Mode 2 conditions prior to withdrawal of Regulator Group 3.
This means that Regulating Groups, 3, 4, 5, 6, and P are fully inserted in Mode 3 or below.
A reduction in RCS temperature following a MSLB in Mode 3 will not introduce enough positive reactivity to bring the reactor critical given the number of CEAs that remain inserted until Mode 2 is entered during an intentional reactor startup. Even if criticality were achieved, the Log Power Level High reactor trip (occurs at 10-4% power) will ensure safety limits are not exceeded.
Because the reactor is already shutdown in Modes 3, 4, 5, the production of a reactor trip signal initiated by Pressurizer Pressure Low or SG Pressure Low functions is unnecessary. In addition, the STS only require this instrumentation to be operable in Modes 1 and 2 with respect to RPS functions. This proposed change affects both Table 3.3-1 relating to operability and Table 4.3-1 relating to surveillance testing of these instruments.
With regard to the Log Power Level High trip, this function is designed to prevent exceeding safety limits should an inadvertent criticality occur due to an uncontrolled CEA withdrawal event.
This event cannot reasonably occur when TCBs are opened since the CEAs have no means of receiving electrical power to their drive systems in this condition. Therefore, consistent with the STS, the Mode of Applicability for this function is changed from Modes 3, 4, 5 to Modes 3*, 4*,
5* (* indicates TCBs are closed and CEAs are capable of withdrawal). As in the trip functions above, this results in a change to both Table 3.3-1 relating to operability and Table 4.3-1 relating to surveillance testing of these instruments.
The proposed changes support consistency with the STS and resolve variations among individual TSs and/or plant design basis. No physical changes are being made to the plant as a result of the aforementioned changes. Reactor trip and reactor mitigating functions will continue to operate as designed, when needed, in order to meet the current safety analysis requirements and to protect the safety of the general public. Therefore, Entergy believes the proposed changes are both warranted and acceptable.
to 2CAN030901 Page 6 of 8

==5.0     REGULATORY ANALYSIS==

5.1     Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.
General Design Criteria (GDC) 13, 20, 21, 22, 23, and 24 describe requirements associated with automatic actuation systems for reactor protection. GDC 34 describes requirements associated with primary system heat removal methods that may include secondary heat removal capability. The proposed change does not alter physical system design or challenge any of the aforementioned GDCs or other regulatory requirements. Based on the considerations discussed in Section 4.0 above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will continue to be conducted in accordance with the site licensing basis, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.
In conclusion, Entergy has determined that the proposed change does not require any exemptions or relief from regulatory requirements, other than the TS, and does not affect conformance with any GDC differently than described in the Safety Analysis Report (SAR).
5.2     No Significant Hazards Consideration The proposed change will modify Technical Specification (TS) 3.3.1.1, Reactor Protective Instrumentation, and TS 3.3.2.1, Engineered Safety Feature Actuation System (ESFAS)
Instrumentation, specifically Table 3.3-1, Table 4.3-1, and Table 3.3-3 (respectively) to adopt a Mode of Applicability for the Logarithmic (Log) Power Level High, Pressurizer Pressure Low, Steam Generator (SG) Pressure Low, and the SG Differential Pressure and Level Low functions to be consistent with the improved Standard TSs (STS) of NUREG 1432, Rev. 3.1, for Combustion Engineering Plants. In addition, Action notes associated with TS Table 3.3-3 are revised to match the correct Modes of Applicability as stated in the table.
Entergy Operations, Inc. has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
: 1.      Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change acts to reconcile a difference between Emergency Feedwater (EFW) TS 3.7.1.2 and Table 3.3-3 of TS 3.3.3.2, or differences between the current ANO-2 TSs and the STS in relation to Reactor Protective System (RPS) or ESFAS functions. The TS 3.7.1.2 Mode of Applicability for EFW is based on plant design basis.
Revising the associated actuation instrumentation Mode of Applicability to match that of TS 3.7.1.2 will continue to ensure that automatic actuation of the EFW system will occur during any Mode 1, 2, or 3 event that results in a Steam Generator (SG) actuation setpoint being reached. The change is not associated with any accident precursor or to 2CAN030901 Page 7 of 8 initiator. EFW will continue to be automatically actuated and capable of a supporting plant cooldown through to Mode 4, where the Shutdown Cooling (SDC) system may be placed in service for decay heat removal purposes. Upon a loss of SDC, EFW may be manually initiated (if available) or a back-up source of SG makeup can be placed in service, such as the non-safety Auxiliary Feedwater (AFW) pump or other non-safety Main Feedwater (MFW) system pumps. These non-safety pumps can be powered from the onsite Alternate AC Diesel Generator should a loss of offsite power event occur.
Changes to the Modes of Applicability for the Log Power Level High, Pressurizer Pressure Low, and SG Pressure Low reactor trip functions do not involve physical plant changes or changes to the current safety analysis. These functions will continue to provide their respective protective feature in the operational modes consistent with the design basis and STS. None of these functions are associated with accident precursors.
Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not result in any plant modifications or change in the way the plant is designed to function. The proposed change is not associated with any accident precursor or initiator.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
EFW will continue to be automatically actuated and capable of supporting a plant cooldown to Mode 4, where the Shutdown Cooling (SDC) system may be placed in service for decay heat removal purposes. Upon a loss of SDC, EFW may be manually initiated (if available) or a back-up source of SG makeup can be placed in service, such as the non-safety Auxiliary Feedwater (AFW) pump or other non-safety Main Feedwater (MFW) system pumps. These non-safety pumps can be powered from the onsite Alternate AC Diesel Generator should a loss of offsite power event occur.
Changes to the Modes of Applicability for the Log Power Level High, Pressurizer Pressure Low, and SG Pressure Low reactor trip functions do not involve physical plant changes or changes to the current safety analysis. These functions will continue to provide their respective protective feature in the operational modes consistent with the design basis and STS.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
to 2CAN030901 Page 8 of 8 Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.
5.3     Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6.0     PRECEDENCE The current TSs are a result of a plant specific issue relating to core reload activities in the early 1980s. Because the changes are plant specific, no generic application is evident. In general, the changes act to realign the affected ANO-2 TSs to be consistent with the Standard TSs of NUREG-1432, Rev. 3.1.

Attachment 2 2CAN030901 Proposed Technical Specification Changes (mark-up)

TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION MINIMUM TOTAL NO.         CHANNELS    CHANNELS     APPLICABLE FUNCTIONAL UNIT                   OF CHANNELS          TO TRIP   OPERABLE        MODES          ACTION
: 1. Manual Reactor Trip              2 sets of 2       1 set of 2  2 sets of 2      1,2             5 2 sets of 2       1 set of 2  2 sets of 2    3*,4*,5*          8
: 2. Linear Power Level - High             4                 2           3           1,2            2,3
: 3. Logarithmic Power Level - High
: a. Startup and
* 4              2(a)(d)        3        2,3*,4*,5*        2,3
: b. Shutdown                      4                 0           2          3,4,5            4
: 4. Pressurizer Pressure - High           4                 2           3           1,2            2,3
: 5. Pressurizer Pressure - Low            4                2(b)         3       1,2,3*,4*,5*       2,3
: 6. Containment Pressure - High           4                 2           3           1,2            2,3
: 7. Steam Generator Pressure - Low      4/SG              2/SG        3/SG      1,2,3*,4*,5*       2,3
: 8. Steam Generator Level - Low         4/SG              2/SG        3/SG          1,2            2,3
: 9. Local Power Density - High            4              2(c)(d)        3           1,2            2,3 ARKANSAS - UNIT 2                                 3/4 3-2                     Amendment No. 24,134,159,

TABLE 4.3-1 REACTOR PROTECTION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL         MODES IN WHICH CHANNEL        CHANNEL       FUNCTIONAL        SURVEILLANCE FUNCTIONAL UNIT                      CHECK       CALIBRATION         TEST             REQUIRED
: 1. Manual Reactor Trip               N.A.           N.A.           S/U (1)               N.A.
: 2. Linear Power Level - High          S            D (2,4)         TA (10)                1,2 M (3,4)
Q (4)
: 3. Logarithmic Power Level - High     S             R (4)          TA (10)         1,2,3*,4*,5 and
* S/U (1)
: 4. Pressurizer Pressure - High        S               R            TA (10)                1,2
: 5. Pressurizer Pressure - Low         S               R            TA (10)           1,2,3*,4*,5*
: 6. Containment Pressure - High        S               R            TA (10)                1,2
: 7. Steam Generator Pressure - Low     S               R            TA (10)           1,2,3*,4*,5*
: 8. Steam Generator Level - Low        S               R            TA (10)                1,2
: 9. Local Power Density - High         S            D (2,4)         TA (10)                1,2 R (4,5)          R (6)
: 10. DNBR - Low                         S             S (7),         TA (10)                1,2 D (2,4),         R (6)
M (8),
R (4,5)
: 11. Reactor Protection System Logic   N.A.           N.A.           TA (10)           1,2,3*,4*,5*
: 12. Reactor Trip Breakers             N.A.           N.A.             M               1,2,3*,4*,5*
: 13. Core Protection Calculators        S            D (2,4)        TA (9,10)               1,2 R (4,5)          R (6)
: 14. CEA Calculators                    S               R            TA (10)                1,2 R (6)
ARKANSAS - UNIT 2                          3/4 3-7           Amendment No. 24,39,77,159,186,216,

TABLE 3.3-3 (Continued)
ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.            CHANNELS         CHANNELS     APPLICABLE FUNCTIONAL UNIT                       OF CHANNELS             TO TRIP        OPERABLE       MODES           ACTION
: 7. LOSS OF POWER
: a. 4.16 kv Emergency Bus Undervoltage (Loss of Voltage)     2/Bus                1/Bus            2/Bus         1,2,3            9
: b. 460 volt Emergency Bus Undervoltage (Degraded Voltage)                           1/Bus                1/Bus            1/Bus         1,2,3            9
: 8. EMERGENCY FEEDWATER (EFAS)
: a. Manual (Trip Switches)         2 sets of 2 per        2 sets of 2      2 sets of 2    1,2,3,4            9 S/G               per S/G          per S/G
: b. SG Level and Pressure (A/B) -
Low and P (A/B) - High             4/SG                 2/SG             3/SG        1,2,3,4         10,11
: c. SG Level (A/B) - Low and No S/G Pressure - Low Trip (A/B)       4/SG                 2/SG             3/SG        1,2,3,4         10,11
: d. ESFAS Logic
: 1. Matrix Logic                     6                    1                3         1,2,3,4           12
: 2. Initiation Logic                 4                    2                4         1,2,3,4            9
: e. Automatic Actuation Logic             2                    1                2         1,2,3,4           13 ARKANSAS - UNIT 2                                        3/4 3-13                          Amendment No. 134,159,186,

TABLE 3.3-3 (Continued)
TABLE NOTATION (a) Trip function may be bypassed in this MODE when pressurizer pressure is below 400 psia; bypass shall be automatically removed before pressurizer pressure exceeds 500 psia.
(b) An SIAS signal is first necessary to enable CSAS logic.
(c) Remote manual not provided for RAS. These are local manuals at each ESF auxiliary relay cabinet.
ACTION STATEMENTS ACTION 9 -    With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and exit the MODE(s) of Applicabilityin COLD SHUTDOWN within the following 30 hours.
ACTION 10 - With the number of channels OPERABLE one less than the Total Number of Channels, operation in the applicable MODES may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour.
If the inoperable channel is bypassed for greater than 48 hours, the desirability of maintaining this channel in the bypassed condition shall be reviewed as soon as possible but no later than the next regularly scheduled OSRC meeting in accordance with the Quality Assurance Program Manual (QAPM). The channel shall be returned to OPERABLE status prior to startup following the next COLD SHUTDOWN.
If an inoperable Steam Generator P or RWT Level - Low channel is placed in the tripped condition, remove the inoperable channel from the tripped condition within 48 hours or be in at least HOT STANDBY within the next 6 hours and in HOTCOLD SHUTDOWN within the following 30 hours.
With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below.
Process Measurement Circuit                    Functional Unit Bypassed
: 1. Containment Pressure - NR                   Containment Pressure - High (RPS)
Containment Pressure - High (ESFAS)
Containment Pressure - High-High (ESFAS)
: 2. Steam Generator 1 Pressure                  Steam Generator 1 Pressure - Low Steam Generator 1 P (ESFAS 1)
Steam Generator 2 P (ESFAS 2)
: 3. Steam Generator 2 Pressure                  Steam Generator 2 Pressure - Low Steam Generator 1 P (ESFAS 1)
Steam Generator 2 P (ESFAS 2)
: 4. Steam Generator 1 Level                     Steam Generator 1 Level - Low Steam Generator 1 P (EFAS 1)
: 5. Steam Generator 2 Level                     Steam Generator 2 Level - Low Steam Generator 2 P (EFAS 2)
ARKANSAS - UNIT 2                            3/4 3-14       Amendment No. 134,159,186,195,196, 216,255,

TABLE 3.3-3 (Continued)
TABLE NOTATION ACTION 11 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement, operation in the applicable MODES may continue provided the following conditions are satisfied:
: a. Verify that one of the inoperable channels has been bypassed and place the other inoperable channel in the tripped condition within 1 hour, and
: b. All functional units affected by the bypassed/tripped channel shall also be placed in the bypassed/tripped condition as listed below:
Process Measurement Circuit                   Functional Unit Bypassed/Tripped
: 1. Containment Pressure - NR                  Containment Pressure - High (RPS)
Containment Pressure - High (ESFAS)
Containment Pressure - High-High (ESFAS)
: 2. Steam Generator 1 Pressure                 Steam Generator 1 Pressure - Low Steam Generator 1 P (EFAS 1)
Steam Generator 2 P (EFAS 2)
: 3. Steam Generator 2 Pressure                 Steam Generator 2 Pressure - Low Steam Generator 1 P (EFAS 1)
Steam Generator 2 P (EFAS 2)
: 4. Steam Generator 1 Level                    Steam Generator 1 Level - Low Steam Generator 1 P (EFAS 1)
: 5. Steam Generator 2 Level                    Steam Generator 2 Level - Low Steam Generator 2 P (EFAS 2)
If an inoperable Steam Generator P or RWT Level - Low channel is placed in the tripped condition, remove the inoperable channel from the tripped condition within 48 hours or be in at least HOT STANDBY within the next 6 hours and in HOTCOLD SHUTDOWN within the following 30 hours.
Operation in the applicable MODES may continue until the performance of the next required CHANNEL FUNCTIONAL TEST. Subsequent operation in the applicable MODES may continue if one channel is restored to OPERABLE status and the provisions of ACTION 10 are satisfied.
ACTION 12 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in HOTCOLD SHUTDOWN within the following 30 hours.
ACTION 13 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in exit the MODE(s) of ApplicabilityCOLD SHUTDOWN within the following 30 hours; however, one channel may be bypassed for up to 1 hour for surveillance testing provided the other channel is OPERABLE.
ARKANSAS - UNIT 2                          3/4 3-15              Amendment No. 159,195,216,}}