ML032670629

License Amendment Request to Various Technical Specification Associated with Replacement of Part Length Control Element Assemblies (Ceas)

ML032670629
Person / Time
Site:	Palo Verde
Issue date:	09/17/2003
From:	Mauldin D Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-04999-CDM/TNW/JAP
Download: ML032670629 (113)

Text

{{#Wiki_filter:David Mauldin 10 CFR 50.90 Vice President Mail Station 7605 Palo Verde Nuclear Nuclear Engineering TEL (623) 393-5553 P.O. Box 52034 Generating Station and Support FAX (623) 393-6077 Phoenix, AZ 85072-2034 1 02-04999-CDM/TNW/JAP September 17, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-37 Washington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2 and 3 Docket Nos. STN 50-52815291530 License Amendment Request to Various Technical Specifications Associated with Replacement of Part Length Control Element Assemblies (CEAs)

Dear Sirs:

In accordance with 10 CFR 50.90, Arizona Public Service Company (APS) hereby requests an amendment to Facility Operating License Nos. NPF-41, NPF-51, and NPF-74 for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3. This License Amendment Request (LAR) revises the following sections of the Technical Specifications: Table of Contents 1.1 "Definitions" 3.1.5 "Control Element Assembly (CEA) Alignment" 3.1.8 "Part Length Control Element Assembly (CEA) Insertion Limits" 3.1.9 "Special Test Exception (STE) - Shutdown Margin (SDM)" 3.1.10 "Special Test Exception (STE) - MODES 1 and 2" 3.1.11 "Special Test Exception (STE) - Reactivity Coefficient Testing" 3.3.3 "Control Element Assembly Calculators (CEACs)" 4.2.2 "Design Features - Control Element Assemblies" 5.6.5 "Reporting Requirements - Core Operating Limits Report (COLR)" This LAR is necessary to support the replacement of Part Length Control Element Assemblies (PLCEAs) with a new design control element assembly described as Part Strenath Control Element Assembly (PSCEA). Additionally, TS 3.1.5 - "Control Element Assembly (CEA) Alignment," Condition B will be modified to prevent a potential unwarranted plant shutdown condition from occurring. A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak
• Diablo Canyon
• Palo Verde
• South Texas Project 0 Wolf Creek

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk License Amendment Request to Various Technical Specifications Associated with Replacement of Part Length Control Element Assemblies (CEAs) Page 2 The enclosure to this LAR provides a description and assessment of the proposed change. Attachment 1 provides the existing TS pages marked up to show the proposed changes. Attachment 2 provides the revised (retyped) TS pages. Additionally, the marked up and retyped Technical Specification pages for LCO 3.3.3 contained in Attachment 1 and 2 are the associated pages for a pending change with the NRC for the approval of the replacement of Core Protection Calculator Systems (CPCS), submitted on 11/07/02 (102-04864-CDMITNW/DWG - Request for Amendment to Technical Specifications: 3.2.4, Departure From Nucleate Boiling Ratio (DNBR), 3.3.1, Reactor Protective System (RPS) Instrumentation - Operating, 3.3.3, Control Element Assembly Calculators (CEACs)). Attachment 3 provides the existing TS Bases pages marked up to show the proposed changes (for information only). Once the implementation of PSCEAs has been completed in all 3 PVNGS Units, APS will submit another LAR to remove from the Technical Specifications references to the PLCEAs. In accordance with the PVNGS Quality Assurance Program, the Plant Review Board and the Offsite Safety Review Committee have reviewed and concurred with this proposed amendment. By copy of this letter, this submittal is being forwarded to the Arizona Radiation Regulatory Agency (ARRA) pursuant to IOCFR 50.91 (b)(1). The anticipated use of the Part Strength Control Element Assemblies (PSCEAs) is scheduled for Unit 1, refueling outage 11 (UIR11). UIR11 is currently scheduled to start April 3, 2004. Approval of this amendment application is requested by February 17, 2004. APS requests to implement the proposed amendment within 60 days of its issuance. No commitments are being made to the NRC by this letter. Should you have any questions, please contact Thomas N. Weber at (623) 393-5764. Sincerely, CDMITNWIJAP

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk License Amendment Request to Various Technical Specifications Associated with Replacement of Part Length Control Element Assemblies (CEAs) Page 3

Enclosures:

Notarized Affidavit APS' evaluation of proposed changes Attachments:

1. Markup of Technical Specification Pages

2. Retyped Technical Specification Pages

3. Associated Changes to Technical Specification Bases (for information only) cc:

Regional Administrator, NRC Region IV M. B. Fields N. L. Salgado A. V. Godwin

STATE OF ARIZONA ) ) ss. COUNTY OF MARICOPA ) 1, David Mauldin, represent that I am Vice President Nuclear Engineering and Support, Arizona Public Service Company (APS), that the foregoing document has been signed by me on behalf of APS with full authority to do so, and that to the best of my knowledge and belief, the statements made therein are true and correct. David Milauldin Sworn To Before Me This/ 1.~ Day or ~X a

kaD, 2003.

I - I H 6Ia4~ OFFICIAL UEAL Susie Lynn Ergish NOTARY PUBLIC-STE of ARIZONA MCOA COUNY WY CMa EXPNMES as U. 2007 _ _ I,, Notaryoiblic () Notary Commission Stamp

Enclosure APS' Evaluation of Proposed LAR Proposed Amendment for Replacement of PLCEAs with PSCEAs

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

Enclosure

1.0 DESCRIPTION

This license amendment request (LAR) will amend Operating Licenses NPF-41, NPF-51, and NPF-74 for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, respectively. The proposed changes would revise sections of the Technical Specifications (TS) to support replacement of the part length control element assemblies (PLCEAs) with a new design that contains neutron absorber over the entire control section of each control element assembly (CEA) finger. The replacements are referred to as part strength control element assemblies (PSCEAs). The current PLCEAs have been in use since the start of operation of each PVNGS unit and are planned to be replaced before reaching 15 effective full power years (EFPYs). This design life requires the replacement of PLCEAs in Units 1 & 3 to be at the end of Cycle 11 and at the end of Cycle 12 for Unit 2. The expected installations of the part strength CEAs (PSCEAs) are planned to occur during upcoming refueling outages as listed below: Unit 1, Refueling Outage 11 - Spring 2004 Unit 3, Refueling Outage 11 - Fall 2004 Unit 2, Refueling Outage 12 - Spring 2005 The proposed changes associated with this LAR are mainly changing the wording from upart length" to "part length or part strength" control element assemblies (CEAs). Along with this change will be the addition of the part strength CEAs description to Section 4.2.2 of the Technical Specifications. Even though there will be no changes or modifications to full length CEAs, for consistency and for ease of reading, the wording for 'full length" CEAs will be changed to "full strength" CEAs. Additionally, TS 3.1.5 -"Control Element Assembly (CEA) Alignment," Condition B, will be modified to eliminate a potential condition which could cause an unwarranted plant shutdown. This condition will be modified such that when more than one CEA in a group has only one operable position indication, a plant shutdown will not be required.

2.0 PROPOSED CHANGE

S The following changes describe the modification to the wording and description associated with part lenath and part strenath CEAs, along with modifying wording for full lenath to full strenath CEAs. In the sections of the Technical Specifications that currently list "part length CEAs," this will be changed to "part length or part strength CEAs". The intent of this change is to accommodate the 1

Enclosure implementation of part strength CEAs during different time frames between the three Palo Verde units. In the "Table of Contents" on page "i", TS 3.1.8 currently is listed as: "Part Length CEA Insertion Limits" The 'Table of Contents". na-ie Y". for TS 3.1.8 will be changed to read: "Part Length! tq"j~ hy CEA Insertion Limits In the "Definitions" section of TS on page 1.1-4, for "Kn.", the definition currently reads: "K,1.1is the K effective calculated by considering the actual CEA configuration and assuming that the fully or partially inserted full-length CEA of highest worth is fully withdrawn." This definition for 'Kn-1" will be changed to read: "Kn is the K effective calculated by considering the actual CEA configuration and assuming that the fully or partially inserted full rgtj CEA of highest worth is fully withdrawn." In the "Definitions" section of TS on page 1.1-6, for "Shutdown Margin (SDM)," the definition currently reads: "SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All full length CEAs (shutdown and regulating) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn. With any full length CEAs not capable of being fully inserted, the withdrawn reactivity worth of these CEAs must be accounted for in the determination of SDM and 2

Enclosure

b. There is no change in part length CEA position."

This definition for "SDM" will be changed to read: "SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All full Etrent j CEAs (shutdown and regulating) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn. With any full grerigt"t CEAs not capable of being fully inserted, the withdrawn reactivity worth of these CEAs must be accounted for in the determination of SDM and

b. There is no change in part length thy a

CEA position." TS Limiting Condition For Operation (LCO) 3.1.5 "Control Element Assembly (CEA) Alignment," currently reads: "All full length CEAs shall be OPERABLE, and all full and part length CEAs shall be aligned to within 6.6 inches (indicated position) of all other CEAs in their respective groups." LCO 3.1.5 will be chanced to read: "All full 13j;eh7g CEAs shall be OPERABLE, and all full H gt and part lengthhkrt.rejje j CEAs shall be aligned to within 6.6 inches (indicated position) of all other CEAs in their respective groups.- TS LCO 3.1.5, Condition C, currently reads: "Required Action and associated Completion Time of Condition A or B not met OR 3

Enclosure One or more full length CEAs untrippable." LCO 3.1.5. Condition C. will be changed to read: "Required Action and associated Completion Time of Condition A or B not met OR One or more full ftredng CEAs untrippable."

• TS Surveillance Requirement (SR) 3.1.5.1, currently reads:

Verify the indicated position of each full and part length CEA is within 6.6 inches of all other CEAs in its group." SR 3.1.5.1 will be chanced to read: Verify the indicated position of each full HF icj and part lengthr ~ CEA is within 6.6 inches of all other CEAs in its group." TS SR 3.1.5.3, currently reads: 'Verify full length CEA freedom of movement (trippability) by moving each individual full length CEA that is not fully inserted in the core at least 5 inches." SR 3.1.5.3 will be changed to read: "Verify full Phi CEA freedom of movement (trippability) by moving each individual full I 5rh tf CEA that is not fully inserted in the core at least 5 inches.' TS SR 3.1.5.5, currently reads: Verify each full length CEA drop time

• 4.0 seconds."

SR 3.1.5.5 will be changed to read: 'Verify each full CEA drop time

• 4.0 seconds."

4

Enclosure The title for LCO 3.1.8, currently reads: "Part Length Control Element Assembly (CEA) Insertion Limits" The title for LCO 3.1.8 will be changed to read: "Part LengthffPT n Control Element Assembly (CEA) Insertion Limits" TS LCO 3.1.8, currently reads: "The part length CEA groups shall be limited to the insertion limits specified in the COLR." LCO 3.1.8 will be changed to read: "The part length CEA groups shall be limited to the insertion limits specified in the COLR." TS LC 03.1.8, Condition A and Required Action A.1, currently read: "Condition A. Part length CEA groups inserted beyond the transient insertion limit. Required Action A.1 Restore part length CEA groups to within the limit." LCO 3.1.8. Condition A and Required Action A.1. will be chanoed to read: "Condition A. Part length H as i jM CEA groups inserted beyond the transient insertion limit. Required Action A.1 Restore part lengthEMtypa !`Eijib CEA groups to within the limit." TS LCO 3.1.8, Condition B and Required Action B.1, currently read: "Condition B. Part length CEA groups inserted between the long term steady state insertion limit and the transient insertion limit for intervals 2 7 effective full power days (EFPD) per 30 EFPD or Ž 14 EFPD per 365 EFPD interval. 5

Enclosure Required Action B.1 Restore part length CEA groups to within the long term steady state insertion limit." LCO 3.1.8. Condition B and Required Action B.1, will be changed to read: "Condition B. Required Action B.1 Part length~o~ -{hJtj CEA groups inserted between the long term steady state insertion limit and the transient insertion limit for intervals 2 7 effective full power days (EFPD) per 30 EFPD or 2 14 EFPD per 365 EFPD interval. Restore part length h CEA groups to within the long term steady state insertion limit." TS SR 3.1.8.1, currently reads: "Verify part length CEA group position." SR 3.1.8.1 will be chanaed to read: "Verify part lengthnf3 CEA group position." TS LCO 3.1.9, Condition A, currently reads: 'Any full length CEA not fully inserted and less than the required shutdown reactivity available for trip insertion. OR All full length CEAs inserted and the reactor subcritical by less than the above required shutdown reactivity equivalent." LCO 3.1.9, Condition A will be changed to read: "Any full Hitieng CEA not fully inserted and less than the required shutdown reactivity available for trip insertion. OR 6

Enclosure All full n CEAs inserted and the reactor subcritical by less than the above required shutdown reactivity equivalent." TS SR 3.1.9.2, currently reads: uVerify each full length CEA not fully inserted is capable of full insertion when tripped from at least the 50% withdrawn position." SR 3.1.9.2 will be chanced to read: 'Verify each full ~ CEA not fully inserted is capable of full insertion when tripped from at least the 50% withdrawn position." TS SR 3.1.9.3, currently reads: Verify that with all full length CEAs fully inserted, the reactor is subcritical within the acceptance criteria." SR 3.1.9.3 will be changed to read: 'Verify that with all full trengt CEAs fully inserted, the reactor is subcritical within the acceptance criteria." TS LCO 3.1.10, currently reads, in part: "During performance of PHYSICS TESTS, the requirements of: LCO 3.1.4, LCO 3.1.5, LCO 3.1.6, LCO 3.1.7, LCO 3.1.8, "Moderator Temperature Coefficient (MTC)"; "Control Element Assembly (CEA) Alignment"; "Shutdown Control Element Assembly (CEA) Insertion Limits"; "Regulating Control Element Assembly (CEA) Insertion Limits"; "Part Length CEA Insertion Limits"; LCO 3.1.10 will be chanaed to read: "During performance of PHYSICS TESTS, the requirements of: LCO 3.1.4, LCO 3.1.5, "Moderator Temperature Coefficient (MTC)"; "Control Element Assembly (CEA) Alignment"; 7

Enclosure LCO 3.1.6, LCO 3.1.7, LCO 3.1.8, "Shutdown Control Element Assembly (CEA) Insertion Limits"; "Regulating Control Element Assembly (CEA) Insertion Limits"; "Part Lengthyor art trent CEA Insertion Limits"..." TS LCO 3.1.11, currently reads, in part: 'During performance of PHYSICS TESTS, the requirements of: LCO 3.1.7, LCO 3.1.8, "Regulating Control Element Assembly (CEA) Insertion Limits"; "Part Length Control Element Assembly (CEA) Insertion Limits;" and..." LCO 3.1.11 will be changed to read: "During performance of PHYSICS TESTS, the requirements of: LCO 3.1.7, LCO 3.1.8, "Regulating Control Element Assembly (CEA) Insertion Limits"; "Part LengthELrIPa T STer th Control Element Assembly (CEA) Insertion Limits;" and..." TS LCO 3.3.3, Required Action B.2, currently reads: "Verify all full length and part length control element assembly (CEA) groups are fully withdrawn and maintained fully withdrawn, except during Surveillance testing pursuant to SR 3.1.5.3 or for control, when CEA group #5 may be inserted to a maximum of 127.5 inches withdrawn." LCO 3.3.3. Required Action B.2. will be chanaed to read: 'Verify all full AGUE and part lengthor tI@gm control element assembly (CEA) groups are fully withdrawn and maintained fully withdrawn, except during Surveillance testing pursuant to SR 3.1.5.3 or for control, when CEA group #5 may be inserted to a maximum of 127.5 inches withdrawn." 8

Enclosure TS Section 4.2.2, "Design Features - Control Element Assemblies," currently reads: "The reactor core shall contain 76 full length and 13 part length control element assemblies (CEAs). The control material shall be boron carbide with Inconel Alloy 625 used as a wear absorber over a portion of the part length control element assemblies as approved by the NRC." TS 4.2.2 will be changed to read: "The reactor core shall contain 76 full tg and e 13 part length E3` control element assemblies (CEAs). The control section iribh-Jl_ ~ shall be boron carbide with Inconel Alloy 625 cladding. Lor units that-nave part length sts, t e control sectjon shal be hconel lloy 625-,in.tnhi ower halt(,followed byperFortd stainiless_ Pteel tubing.-over theriint40%, and boron cabid~.p1ets wth nc X clad WerA-'i_ Alas dtio control ec or units that pave part strength CEAs, the contro section s alb !.rconl All02 i one Alt62 Iad Lnq

• TS 5.6.5.a.7. - 'Core Operating Limits Report (COLR)," currently reads:

"Part Length CEA Insertion Limits for Specification 3.1.8." TS 5.6.5.a.7. will be chanaed to read: "Part Length~ rfiSit'r-Ft CEA Insertion Limits for Specification 3.1.8." TS 5.6.5.b.3. - uCore Operating Limits Report (COLR)," currently reads, in part: "Safety Evaluation Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR system 80,..... ...; 3.1.8, Part Length CEA Insertion Limits and 3.2.3, Azimuthal Power Tilt - Tq]." 9

Enclosure TS 5.6.5.b.3. will be changed to read (in Dart): "Safety Evaluation Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR system 80,..... ... ; 3.1.8, Part Length~g_ CEA Insertion Limits and 3.2.3, Azimuthal Power Tilt - Tq]." TS 5.6.5.b.12. - "Core Operating Limits Report (COLR)," currently reads, in part: Technical Manual for the CENTS code....... ; 3.1.8, Part Length CEA Insertion Limits and 3.2.3, Azimuthal Power Tilt - Tq].' TS 5.6.5.b.12. will be changed to read (in Dart): "Technical Manual for the CENTS code....... ; 3.1.8, Part Lengthr Ei`t Str_ tI CEA Insertion Limits and 3.2.3, Azimuthal Power Tilt - Tq]." In addition to the above changes associated with the replacement of the part length CEAs, the following change associated with TS LCO 3.1.5 is being made: TS 3.1.5, Condition B, currently reads: "Only one CEA position indicator channel OPERABLE for one CEA per CEA Group." TS 3.1.5, Condition B will be changed to read: 'Only one CEA position indicator channel OPERABLE for one EM

3.0 BACKGROUND

The purpose of the control element assemblies (CEAs) are for reactivity control during operation and shutdown of the reactor. The shutdown and regulating groups are made up of 4-and 12-finger full length CEAs. The regulating CEA groups may be used to compensate for changes in reactivity associated with routine power level changes and to compensate for minor variations in moderator temperature and boron concentration changes during operation at power, and to dampen axial xenon oscillations. Thirteen part length CEAs are provided in the design to help control the core power distribution. This function includes the suppression of xenon-induced axial power oscillations. 10

Enclosure A new design for the original equipment part length CEAs was developed by Combustion Engineering for use in the System-80 designed reactors. The originally supplied part length CEAs (PLCEAs) control section design consists of solid Inconel 625 over the bottom 50% of their length, a stainless steel tube open to the reactor coolant over the next 40%, and a sealed chamber containing 73% (theoretical density) boron carbide (B4C) pellets in the top 10%. A holddown spring, similar to the spring in the full length rods, maintains the orientation of the B4C. The new design maintains the same external dimensions as the original design, but with changes to the construction and internal components of the CEA finger. The new CEA is composed of an Inconel 625 tube filled with Inconel 625 slugs throughout the full length of the active region of the finger (nominally 150 inches). This new design is called part strength CEAs (PSCEAs). The perforated tube in the upper 40% section and sealed chamber of B4C pellets at the top of the original PLCEA design is not present in the new PSCEA design. The PLCEAs are currently planned to be replaced with PSCEAs in each PVNGS reactor beginning with Unit 1, during Ul R11 (Spring 2004). There are 13 PLCEAs currently installed in each reactor. PVNGS intends to replace the 13 existing PLCEAs in each unit's reactor with PSCEAs, which are functionally equivalent except for the amount and geometry of neutron absorber inserted into the core. The original design of the PLCEAs, which has Inconel 625 and B4C used as neutron absorber located in the bottom 50% and top 10%, respectively, introduces an incident of moderate frequency which is addressed in the PVNGS Updated Final Safety Analysis Report (UFSAR) Section 7.2.2.1.1.C. A specific safety analysis that comes from this category of single PLCEA misoperation is the malpositioning of a PLCEA between 51% and 90% inserted into the core, resulting in flux peaking in the top of the core. The new design contains neutron absorber (Inconel 625) within 100% of the control section for each CEA finger, which will eliminate the possibility of this incident of moderate frequency. Therefore, this specific analysis will no longer be applicable to the PVNGS licensing bases. The design of the full length CEAs is not changing, but their name will change to "full strength CEAs" (FSCEAs) so that terminology for CEAs will be consistent. TS 3.1.6. Condition B. Modification This change involves a revision to Condition 'B' of Technical Specification 3.1.5 such that it will apply if one or more CEA(s) have only one operable position indication channel. Currently, Condition 'B' of TS 3.1.5 applies if one CEA per group has only one operable position indication channel. 11

Enclosure The change is necessary to provide a Technical Specification Condition that applies to the situation in which more than one CEA in a group has only one operable position indicator. This is being accomplished by simply rewording the existing Condition 'B' to expand the scope of applicability from "one CEA per CEA group", to "one or more CEA(s)." The current alternative to not revising Condition 'B' as proposed is to require entry into TS LCO 3.0.3 when more than one CEA per group has only one operable position indicator. However the requirements of LCO 3.0.3 (i.e.. plant shutdown) are not appropriate for this situation. This position is based on the following:

1) The situation under review is when more than one CEA per group has only one operable position indication channel. Even in this situation, all CEAs still have at least one operable position indication. Entry into LCO 3.0.3 should not be required for situations involving only a loss of redundancy while maintaining operability of the required feature on one train/channel.

2) The existing PVNGS TS Bases for Condition 'B' of Tech Spec 3.1.5 does not address any limitation of applicability to only one CEA per group. The Bases justify continued operation with only one operable position indication channel provided that within 6 hours, either; 1) at least 2 channels are restored to operable status or 2) the affected group is positioned fully withdrawn or fully inserted (while maintaining compliance with the insertion limits).

4.0 TECHNICAL ANALYSES Each reactor at Palo Verde contains 89 control element assemblies (CEAs). Seventy-six of the CEAs are referred to as full length CEAs (FLCEAs) and contain boron carbide (B 4C) neutron absorber pellets, which span the range of the entire height of the fuel core when the CEA is fully inserted. Forty-eight of the 76 FLCEAs are comprised of twelve poison fingers each and the remaining 28 FLCEAs each have four poison fingers. The remaining 13 CEAs are part length CEAs (PLCEAs) each with four fingers which contain B4C neutron absorber sections in only the top 10% of their control section and solid Inconel 625 in the bottom 50%. The remaining 40% of the finger control section region between the B4C and Inconel 625 is made of a perforated stainless steel tube, which allows the presence of primary coolant to act as a neutron moderator. The PLCEAs were intended to provide control of axial power distribution, particularly in the event of axial xenon oscillations. The CEAs are described in the PVNGS UFSAR Sections 4.2 and 4.3. The FLCEAs are categorized by their intended function, i.e., shutdown or regulating. During reactor startup and operation, the shutdown FLCEAs are fully withdrawn followed by the withdrawal of the PLCEAS, before the regulating 12

Enclosure FLCEAs can be withdrawn for controlling the approach to criticality. Only the regulating FLCEAs are then used in a predetermined sequence for reactivity control. A series of testing, prior to plant operation, is performed to assure that each FLCEA will function as expected. Such testing includes a drop test to confirm that each FLCEA safely reaches 90% insertion in less than or equal to four seconds. Another test involves measurement of the reactivity worth of each FLCEA during startup testing for each reload cycle in order to verify the expected design values. Even though the PLCEAs drop times are tested, they are not credited for shutdown margin (SDM) and their drop times and reactivity worth are not considered for accident mitigation in the safety analyses. During plant startup and operation, changes in core reactivity are used to increase or decrease reactor power and can be accomplished using the regulating FLCEA groups to initiate changes in reactivity associated with the desired change in power level. The group sequence and overlap limits for regulating FLCEAs are specified in the Core Operating Limits Report (COLR) for each fuel cycle. Insertion of the PLCEAs during operation is restricted based on the reactor power level. The maximum insertion at or below 50% power is 50% insertion, which drops to 25% insertion for power levels above 50%. All CEA groups (full and part length) are dropped into the core to ensure a rapid shutdown of the reactor following a manual trip or an automatic reactor trip signal from the plant protection system. Although the PLCEAs are released for insertion along with the FLCEAs following a reactor trip signal, the reactivity insertion of the PLCEAs is not credited in the safety analyses. The PLCEAs are used only to adjust the neutron flux distribution within the reactor core during normal operations. The replacements for the PLCEAs are referred to as part strength CEAs (PSCEAs). The PSCEAs also use Inconel 625 as a neutron absorber, but unlike the PLCEAs, Inconel 625 is used over the entire active length (approximately 150 inches) of each finger. The physical dimensions of the PSCEA fingers is essentially the same as used for the 4-finger FLCEAs since the design of the finger cladding and assembly structure (with the exception of the poison being used) will be the same. The FLCEAs will be referred to as full strength CEAs (FSCEAs) for consistency. There are no changes being performed to the full length CEAs. 13

Enclosure A comparison of the significant design characteristic differences are summarized below: PLCEAs vs. PSCEAs Design Criteria Design Criteria PLCEAs PSCEAs Comment Center 61.5 inches of PLCEAs is made of Clad Material Stainless Steel (SS) InconeI625 0.25 inch diameter perforations. Includes the Inconel, Clad Outer Diameter 0.816 + 0.002 in. 0.816 + 0.002 in. 304 SS, and B4C regions of the PLCEAs. The gap between the PSCEA clad and slug corresponds to <2% of the Inconel mass in Clad Inner Diameter 0.746 + 0.002 in. 0.746 + 0.002 in. the solid region of a PLCEA. PLCEA inner diameter is for B4C region at top of each finger. The lower 75 in. of PLCEA fingers is solid Inconel Slug Diameter N/A 0.737 + 0.001 in. Inconel with the same outer diameter as PSCEA fingers. No more than four 0.500 in. Inconel slugs 7.450 + 0.020 in. may be used in any Inconel Slug Length N/A and + one finger at the top of N/A and~~~050 +062i. the stack. These 0.500+/- 0.02 in. slugs are used to adjust the total poison stack length. 75 in. of Inconel in the The bottom nose cap Total Neutron Absorber bottom of each finger 149.000 + 0.005 in. of PSCEAs adds Length and 16 in. of B4C at the of Inconel 0.875 in. to the height top of each finger of Inconel. This section connects Total Length of the the control rod section Control Rod Top 73.625 + 0.005in. 73.625 + 0.005in. containing neutron Assembly absorber to the CEA spider. 14

Enclosure PLCEAs vs. PSCEAs Design Criteria (cont.) Total Finger Length Fingers are the same (including top 244.625 + 0.005in. 244.625 + 0.005in. length for both assembly) designs. Total CEA length is Total CEA Length 252.969 + 0.005in. 252.969 + 0.005in. the same for both designs. The weight of a 12-finger FSCEA (211.9 Estimated Total CEA 116.8 lbs 141.1 lbs lbs from same ref.) is Weight much more limiting than a PSCEA, which has only 4 fingers. The design of the outer geometries of the PLCEAs and PSCEAs are very similar. The principal design differences between the PLCEA and PSCEA are associated with the cladding and the neutron absorber materials used throughout each finger. As mentioned previously, the PLCEAs are comprised of solid Inconel 625, hollow 304 SS perforated tubing, and B4C pellets with Inconel cladding, within 50%, 40%, and 10% of the absorber volume, respectively. The neutron absorber in PSCEA is made up entirely of Inconel 625 slugs with a clad gap of 0.009 inches. The two designs are geometrically very similar and contain essentially the same amount of neutron absorber in the lower 50% of each finger. This region also corresponds to the limiting PLCEA power dependent insertion limit (PDIL) which will be applied to the PSCEAs. Each PSCEA contains substantially more Inconel 625 resulting in a weight increase. This weight increase is still within the capability and design of the control element assembly design and its associated control mechanism design. This resulting weight is still much less than the weight involved with a 12-finger FSCEA, but it is comparable to that of a 4-finger full length CEA. Therefore, operation of the CEA drive mechanism system with the PSCEAs installed will not be adversely affected. The principal design function of the PLCEAs is to control axial power distribution. However, the current part length design can cause undesirable flux redistribution if inserted past the PDIL due to the lack of a neutron absorber in 40% of the upper region of each PLCEA finger. The design of the PSCEAs contains Inconel slugs over the entire control section of each CEA finger. As a result, the accident event of concern regarding the PLCEAs does not apply to the new design of the PSCEAs. This occurs due to the fact that the neutron absorber is present throughout the entire control section of each CEA finger and this will not promote the undesired neutron flux shift to the upper region of the core when inserted past 50%. The PDILs established for the PLCEAs minimize undesirable axial power redistributions since the maximum allowed insertion of 50% corresponds to the lower region containing Inconel as a neutron absorber. This same PDIL will be conservatively maintained for the PSCEAs. 15

Enclosure As mentioned above, the PSCEA design will eliminate an accident scenario from PVNGS licensing bases. This event analysis involves the insertion of a PLCEA past the PDIL which results in an axial shift in power due to a portion of the upper region of the PLCEAs which does not have a neutron absorber. This condition will not occur with the PSCEAs because they are filled with neutron absorber over 100% of the control section of the CEA. Additionally, the following constraints will be maintained for the PSCEAs:

1.

PSCEAs will be in the same location as the existing PLCEAs with no change in subgroup assignments.

2.

The PSCEAs will consist of four axially uniform fingers constructed of materials that have the same nuclear properties as the active region (lowest 50%) of the current PLCEA design. In particular, the bounding reactivity worth per inch of insertion in the active region is not significantly different.

3.

The Power Dependent Insertion Limit (PDIL) for the PSCEAs will be the same as the current PDIL for the PLCEAs, which limits insertion to less than the length of the current active region (50% insertion). The name change from "part length CEA" to "part strength CEA" is the principle change being made to the affected Technical Specifications. Although this change is principally an editorial change, the name change also reflects the physical and geometrical changes associated with the replacement CEAs. This name change effectively represents the function of these replacement CEAs in comparison to that of the "full strength" CEAs. The following discusses each specific Technical Specification change: TS Section 1.1, "Definitions" The definition for uShutdown Margin (SDM)" currently includes a discussion of how the full length CEAs are involved in the determination of SDM. The proposed change will replace ufull length" with "full strength". Since there are no changes involving the design or operation of the existing full length CEAs, this change is strictly editorial. The definition also states that the SDM is accurately assessed by restricting the movement of the part length CEAs with insertion of the FSCEAs. This criterion shall remain applicable to the replacement PSCEAs that have reactivity worths essentially the same as the existing PLCEAs at or above their PDlLs. The definition for Kn-1 also refers to the ufull length" CEAs with regard to determining the value of K-effective. Referring to 'full strength" CEAs represents an editorial change with no technical impact since the design of the existing FLCEAs will not change. 16

Enclosure TS Section 3.1.5, "Control Element Assembly (CEA) Alignment" This section refers to both the FLCEAs and PLCEAs. The terminology for FLCEAs and PLCEAs shall be changed to FSCEAs and PSCEAs or PLCEAs with no technical impact. LCO 3.1.5.C is related to untrippable FLCEAs and will also be revised. The PLCEAs are required to be aligned within 6.6 inches of all other CEAs in their respective groups. This requirement will remain unchanged for the PSCEAs. The event of primary concern has been the misalignment of the FLCEAs or PLCEAs. The existing Technical Specification Basis describes the PLCEA drop and PLCEA subgroup drop events as resulting in changes to the core power distribution, departure from nucleate boiling ratio (DNBR), and fuel centerline temperature which could result in a reactor trip. The design of the PLCEAs introduces a slightly different response than the FLCEAs as a result of the flux redistribution toward the top of the core due to 40% of the upper control section of each finger containing no neutron absorber. Replacing the PLCEAs with the PSCEAs will eliminate the flux redistribution resulting from these events. In addition, the design of the new PSCEAs is similar to the FLCEAs except for a weaker neutron absorber, which effectively prevents the PSCEAs from being more limiting than the FLCEAs for any accident scenario currently analyzed in the UFSAR. The FLCEA drop event remains the bounding event. The changes to this TS LCO, Condition, and Surveillance Requirements (SRs) will only consist of name changes from 'full length CEAs" to "full strength CEAs" and "part length CEAs" to "part length or part strength CEAs". TS Section 3.1.8, "Part Length Control Element Assembly (CEA) Insertion Limits" The insertion limits developed for the PLCEAs represent initial assumptions used in the existing safety analysis for CEA misoperation events. They are intended to prevent undesired neutron flux redistribution toward the top of the core. The associated LCO refers to the COLR for the explicit PLCEA insertion limits. The maximum designated insertion is 50%, which corresponds to the solid Inconel region of the PLCEAs. The limitations for insertion between the long term (steady-state) and transient insertion limits provided in Fig. 3.1.8-1 of the COLR remain applicable due to the similarity in design between the PLCEAs and PSCEAs. However, the core response following insertion of the PSCEAs beyond the PDIL will not be as undesirable as it would be with the PLCEAs. The existing safety analysis does not credit any neutron absorber in the upper 50% of the PLCEAs, which can result in a core power increase or undesirable flux redistribution. This concern is not applicable to the PSCEAs since neutron absorber is present throughout the entire active region of each finger, which would prevent the core response exhibited by the PLCEAs. The effect of the PSCEAs exceeding the PDIL would be similar to the FLCEAs in that the fingers in both CEAs contain neutron poison throughout the entire control section of the CEA. The current limit for returning the FLCEAs and PLCEAs to within the PDIL 17

Enclosure is two hours. Although the flux redistribution resulting from the PLCEAs can be different from the PSCEAs due to the neutron absorber distribution, the insertion of PSCEAs result in a similar flux redistribution as that of the FLCEAs, although it is not as strong. Therefore, the two-hour limit remains conservative and is still applicable to the PSCEAs. The changes to this TS LCO, Conditions, Required Actions, and SR will only consist of the name change from "part length CEAs" to "part length or part strength CEAs". TS Section 3.1.9, "Special Test Exception (STE) - Shutdown Margin (SDM)" This section addresses suspending FLCEA insertion requirements, to assure SDM, during approved physics tests. The insertion requirements are not being changed and only the CEA terminology will be changed with no technical impact. The changes to this TS Condition and SRs will only consist of the name change from "full length CEAs" to "full strength CEAs". TS Section 3.1.10, "Special Test Exception (STE) - MODES I and 2" This section refers to LCO 3.1.8, "Part Length Control Element Assembly (CEA) Insertion Limits" which may be suspended during physics tests. As discussed above, LCO 3.1.8 will be renamed by replacing 'part length" CEAs with "part length or part strength" to reflect the new design. The new design of the PSCEAs does not introduce any new technical or operational considerations and no changes to the insertion limits are required. Suspending these limits during testing will not introduce any new concerns. The neutron absorber in the PSCEAs is located throughout the entire control section of each finger and would not result in a positive addition to the reactivity of the upper core region, which could cause an undesired axial flux redistribution. Therefore, suspending the insertion limits of the PSCEAs as currently identified for the PLCEAs in LCO 3.1.10 will have no impact on safe operation. The change to this TS LCO will only consist of the name change from "Part Length CEA" to "Part Length or Part Strength CEA". TS Section 3.1.11, "Special Test Exception (STE) - Reactivity Coefficient Testing" This TS section refers to suspending LCO 3.1.8, "Part Length Control Element Assembly (CEA) Insertion Limits" for reactivity coefficient testing. As discussed above, LCO 3.1.8 will be renamed by replacing "part length" with "part length or part strength" to reflect the new design. LCO 3.1.11 refers to the COLR for the PLCEA positioning requirements. Suspending these limits during testing will not introduce any new considerations since the design of the PSCEAs which use the neutron absorber throughout the entire control section of each finger effectively eliminates the concern associated with the axial flux redistribution to the top of 18

Enclosure the core. Therefore, suspending the insertion limits of the PSCEAs as currently identified for the PLCEAs in LCO 3.1.1 1 will have no impact on safe operation. The change to this TS LCO will only consist of the name change from "Part Length CEA" to "Part Length or Part Strength CEA". TS Section 3.3.3, "Control Element Assembly Calculators (CEACs)" The CEACs are used by the Core Protection Calculator System (CPCS) to assure the position of the CEAs in each subgroup is within acceptable limits. LCO 3.3.3 requires that full length and part length CEAs be fully withdrawn in the event of certain conditions for CEAC(s) inoperability. The operation of the new PSCEAs is equivalent to the current PLCEAs (above the PDILs). In addition, the PSCEAs will not functionally impact operation of the CEACs since the same CEA extension shafts, control element drive mechanisms (CEDMs), and rod position indicators are used and will continue to provide position indication for the CEACs. This section also refers to the "full length" CEAs but the change has no technical impact on their design or operation. However, they will be renamed to "full strength CEAs" to be consistent with the new naming convention. The DNBR-Low trip will provide protection against core damage in the event of PLCEA subgroup drop based on the expected impact on core conditions resulting from no neutron absorber in 40% of the upper control section of each PLCEA finger. Dropping a PLCEA subgroup can result in an increase in core power, in the upper region of the core, due to the lack of neutron absorber in the top half of each finger which results in a core flux redistribution to the top of the core. However, the design of the PSCEAs with neutron absorber covering 100% of the control section will not cause a similar shift in core flux redistribution if accidentally lowered or dropped within the core. Accident events applicable to the PSCEAs (e.g., dropped CEAs) are bounded by the existing safety analyses for the FSCEAs. Therefore, the requirements specified in LCO 3.3.3 will not be impacted by the design of the PSCEAs. The change to this TS Required Action will only consist of the name changes from "full length CEAs" to "full strength CEAs" and "part length CEAs" to "part length or part strength CEAs". TS Section 4.2.2, "Control Element Assemblies" This section provides a summary description of the CEAs used at PVNGS. This section will be revised to provide a description of the PSCEAs along with maintaining a description of the PLCEAs to accommodate staggered installation of PSCEAs in each Unit. In addition, the name used for the FLCEAs will be changed from "full length CEAs" to "full strength CEAs". 19

Enclosure TS Section 5.6.5, "Core Operating Limits Report (COLR)" This section identifies the core operating limits required to be identified in the COLR along with their technical basis (i.e., Technical Specification referenced topicals). Item 5.6.5.a.7 of the TS identifies the insertion limits of the part length CEAs to be included in the COLR. This section will be reworded to specify the "Part Length or Part Strength CEA Insertion Limits for Specification 3.1.8". The same technical information provided will apply to the new PSCEAs. Item 5.6.5.b.3 of the TS identifies the reference for the analytical methodology used for specifying limiting data to be included in the COLR. This item includes reference to TS Section 3.1.8 relating to the PLCEAs. The operating limitations for the replacement PSCEAs (i.e., the PDILs) will not change and their effective reactivity worth, when inserted up to the limits of the PDIL, is essentially the same as the current PLCEAs. Therefore, the same analytical methodology will apply to the proposed change for TS 3.1.8 to address the PSCEAs. This section will be reworded to specify the "Part Length or Part Strength CEA Insertion Limits for Specification 3.1.8". Item 5.6.5.b.12 of the TS refers to the technical basis documentation for the computer code CENTS as being applicable to TS 3.1.8 for the part length insertion limits. CENTS is used for transient accident analysis required in support of the plants operating license. Since the physical design characteristics of the new PSCEAs are similar to the PLCEAs, the analytical modeling of the PSCEAs can be implemented into the CENTS based analyses, which currently model the PLCEAs. Consequently, the methodology referring to the CENTS code can apply to the proposed change for TS 3.1.8 to address the PSCEAs. This section will be reworded to specify the "Part Length or Part Strength CEA Insertion Limits for Specification 3.1.8". Summary of Changes for Part Length CEA Replacements The design of the PLCEAs utilizes a solid region of Inconel in the bottom 50% and B4C in the top 10% of each finger with no neutron absorber located in the middle region. The design of the PSCEAs contains solid Inconel slugs inside an Inconel tube throughout the entire control section of each finger. The outer geometry of the PLCEA fingers is similar to the PSCEAs and the gap between the slugs and the cladding in the PSCEAs is very small. As a result, the effective neutron absorption of the PSCEAs is equivalent to the solid region of the PLCEA fingers. Other general design issues (e.g., weight difference, vibrational difference, seismic, assembly specifications, etc...) have been evaluated and determined to be within analyses parameters. With the installation of the 20

Enclosure PSCEAs, a part strength rod drop will not cause the addition of positive reactivity from any initial position in the core because the PSCEAs are entirely made of neutron absorber. The design of the "spider" assembly, which holds the CEA fingers, is unchanged. The PSCEA fingers are heavier than the PLCEAs. Current analyses have been evaluated and are bounding for this additional weight. Consequently, the only design difference, which presents any technical significance, is the extension of the neutron absorber region from the lower 50% to 100% of the control section of each finger. The new neutron absorber distribution extending through the entire control section for the PSCEAs will result in similar and less severe core power and neutron flux distributions following anticipated operational occurrences (AOOs) for that of FLCEAs. The failure mode associated with aging for the PSCEA fingers is different than that of the FSCEAs (i.e., B4C pellet swelling causing clad cracking). This is primarily due to the same material (Inconel) being used for both the cladding and neutron absorber slugs within the cladding of the PSCEAs. Due to the neutron absorber slugs and the cladding being made of the same material, no significant strain on the clad which could cause cracking, is expected from swelling of the neutron absorber slugs due to neutron irradiation. All mechanical design aspects of the PSCEA meet applicable mechanical design criteria. Principal results and conclusions of this evaluation are summarized below: Topic Objective Method Results Conclusion Seismic / Confirm stresses meet Evaluated existing Design Allowables Acceptable LOCA design criteria. analyses and are satisfied for all accounted for design conditions. changes. Threaded Confirm joint preload Evaluated existing Preload stresses Acceptable Joints induced stresses are analyses and are within allowable acceptably low, and accounted for design values. Preloads that preload is changes. exceed operating sufficient to keep loads even with connections tight. relaxation Consider relaxation. considered. Fatigue Confirm Utilization Evaluated existing All factors Acceptable Factor for areas analyses and essentially zero, susceptible to fatigue accounted for design including effects of are less than the changes. heavier weight and maximum criterion. potentially longer operating time. Control Rod Confirm PSCEA Evaluated existing Confirmed that all Acceptable Stress meets all acceptance analyses and design allowables criteria with added accounted for added are satisfied at all weight. weight. conditions. 21

Enclosure Topic Objective Method Results Conclusion Clad Welds Confirm PSCEA Accounted for added Confirmed that all Acceptable meets all acceptance weight and added design allowables criteria with added vent holes in top are satisfied at all weight. assembly. conditions. CEA Scram Show displacement Results of existing Required Drop Times vs. time behavior. analysis for other displacement vs. Acceptable System 80 Units that time is met by the use PSCEAs are PSCEA design. applicable. Spider Confirm stresses are Added weight and All design Acceptable Structure less than design referred to existing allowables are allowables under all analysis of Spider satisfied with wide conditions. structural integrity to margins. make assessment. Spider Confirm arresting Existing CEA Spring absorbs Acceptable Spring spring is sufficient to SCRAM analysis for remaining energy absorb energy of other System 80 (after dashpot falling PSCEA without Units that use deceleration) hard impact. PSCEAs is without impacting applicable. on Upper Guide Structure. Plenum Confirm criteria for Previous analysis of Meets appropriate Acceptable Spring stack preloads are other PSCEA shipping and met. designs is applicable. handling requirement and also BOL hot operational requirement. Never reaches solid height Collapse Confirm minimum Previous analysis is Margin against Acceptable Resistance required margin applicable. collapse at max. against collapse ovality is greater considering ovality. than the required margin. Clad Strain Typically for CEAs, No IASCC limits are The projected time Acceptable (CEA Life) define and identified for to reach a wear communicate the life PSCEAs; however, limit criterion is limiting parameters. an active wear provided. mechanism was considered. Heating and Confirm that the Existing analyses for Existing analyses Acceptable Cooling Available guide tube future FSCEAs that bound the PSCEA (T&H) flow is greater than will contain AgInCd, for the same plant the Required flow to which has a much conditions. No bulk suppress bulk boiling higher heating rate, boiling. in the annulus. are bounding. 22

Enclosure Topic Objective Method Results Conclusion RSGs & Confirm that RSG and Documentation on Recent fuel and Acceptable Power Power Uprate do not record demonstrates CEA evaluations Uprate adversely affect the no impact on CEAs. show no impact on PSCEAs. CEAs due to RSGs and Power Uprate. This result is considered applicable to PSCEAs as well. TS 3.1.5. Condition B. Modification This LAR is also modifying the words associated with LCO 3.1.5, Condition B. This Condition is currently written to address what actions are required when one CEA, per CEA group, has only one operable position indicator available. As currently written, if more than one CEA, per CEA group, had only one operable CEA position indicator, the required action would be to enter LCO 3.0.3. Entering LCO 3.0.3 would require a plant shutdown in a very short period of time. Entry into LCO 3.0.3 should not be required for situations involving only a loss of redundancy while maintaining operability of the required feature on one train/channel. There are three position indication channels for each individual CEA:

1) Reed Switch Position Transmitter (RSPT) #1

2) Reed Switch Position Transmitter (RSPT) #2

3) Pulse Counter indication Although it is possible to have independent malfunctions that affect 2 indicator channels for more than one CEA, the most likely cause would be a loss of either the Channel 'C' or Channel 'D' 120 VAC Vital Instrument Bus. If either the Channel 'C' or 'D' Vital Instrument Bus is de-energized, both RSPT #2, and Pulse Counter indication channels are lost on multiple CEAs such that more than one CEA per group would have only one operable indication channel (RSPT #1).

23

Enclosure The table below provides the effects on CEA indication channels due to loss of each Vital Instrument Bus. Vital Bus Number of CEAs with lost indication De-energized RSPT #1 RSPT #2 Pulse Counter PNA-D25 22 Not Affected Not Affected (Channel A (Channel B) 67 Not Affected Not Affected P(N eC-D2 7 Not Affected 67 67 P(ND-nDe 28 Not Affected 22 22 As shown above, a loss of Channel 'C' or Channel 'D' results in a loss of CEA position indication that is beyond the scope currently addressed by Condition 'B' of Tech Spec 3.1.5, since more than one CEA per group will have only one operable position indication channel. This problem is more significant for Unit 1 since unlike Units 2 and 3, there is no automatic transfer of the power source for the Vital Instrument Busses. Units 2 and 3 have static transfer switches which will maintain the Vital Instrument Bus energized on a loss of the normal (inverter) power supply by automatically transferring to the backup (voltage regulator) power supply. In Unit 1, there is no static transfer switch so the Vital Instrument Bus will be de-energized on a loss of the associated inverter. In addition, on a planned transfer between the inverter and the voltage regulator, the Vital Instrument Bus must be de-energized (for Unit 1 only) prior to powering from the alternate source. Upon loss of any of the above Vital Instrument Buses, entry into Abnormal Operating Procedure 40AO-9ZZ1 3, 'Loss of Class Instrument or Control Power', is warranted. The applicable section directs declaring CEAC #1 inoperable for loss of either Channel 'A' or Channel 'B'; CEAC #2 inoperable for loss of either Channel 'C' or Channel 'D'. Condition 'A' of Tech Spec 3.3.3 [Control Element Assembly Calculators (CEAC)] will be entered and 40ST-92Z23, CEA Position Data Log, will be performed every 4 hours to verify the indicated position of each full and part length or part strength CEA is within 6.6 inches of all other CEAs in its group. This action is performed to comply with the 'Required ActionlCompletion Time' of LCO 3.3.3. In addition, the loss of any Vital Instrument Bus requires entry into Condition 'B' of TS 3.8.9 (Distribution Systems - Operating) which provides 2 hours to restore the bus operable. In the event the bus cannot be restored to operable, then the unit must be in Mode 3 within 6 hours. Loss of two Vital Instrument Buses (Condition E) will require entry into LCO 3.0.3. 24

Enclosure Current TS Bases states that, "At least two of the following three CEA position indicator channels shall be OPERABLE for each CEA." Additionally the Bases states, "If only one CEA position indicator channel is OPERABLE, continued operation in MODES 1 and 2 may continue, provided, within 6 hours, at least two position indicator channels are returned to OPERABLE status; or within 6 hours and once per 12 hours, verify that the CEA group with the inoperable position indicators are either fully withdrawn or fully inserted while maintaining the insertion limits of LCO 3.1.6, LCO 3.1.7 and LCO 3.1.8." Current analyses already assumes that more than one CEA in a subgroup could have only one position indicator OPERABLE. This change will still require at least one position indication channel be available for each CEA. The intent of this change is not to permit operation with less than 2 operable CEA position indication channels, per CEA. The operability requirements for CEA position indication will remain unchanged (at least 2 position indication channels for each CEA). This change is needed solely to address the lack of any existing Condition/Required Actions for situations in which more than one CEA per group has only one operable position indication channel. With no applicable Condition/Required Action, LCO 3.0.3 is required. However, providing a 6-hour completion time to restore the CEA indication is preferable to entering LCO 3.0.3 which would require shutdown to Mode 3 within 7 hours (and require a considerable amount of CEA manipulations during the power reduction). Also, the only credible single failure that would result in more than one CEA per group having only one operable position indication channel is the failure of Channel 'C' or Channel 'D', as discussed above. However in this case, the most limiting Tech Spec requirement would not be for CEA position indication. Required Action B.1 of Tech Spec 3.8.9 provides a 2-hour completion time to restore a de-energized Vital Instrument Bus. Thus, the 6-hour completion time associated with the proposed Condition 'B' of Tech Spec 3.1.5 would not be available for use if the vital instrument bus (Channel 'C' or Channel 'D') was not restored within 2 hours. When the vital instrument bus is restored, then the CEA position indication would also be restored. These time constraints serve as a limit to unit operation with only 1 CEA position indication for one or more CEA(s).

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Arizona Public Service Company (APS) 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: This license amendment request (LAR) is to amend Operating Licenses NPF-41, NPF-51, and NPF-74 for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 25

Enclosure 2, and 3, respectively. The proposed changes would revise sections of the Technical Specifications (TS) to support replacement of the part length control element assemblies (PLCEAs) with a new design that contains neutron absorber over the entire control section of the CEA. The replacements are referred to as part strength control element assemblies (PSCEAs). The proposed changes associated with this LAR are mainly changing the wording from "part length" to "part length or part strength" control element assemblies (CEAs) in several sections of TS. Included with this change will be the addition of the part strength CEAs description to Section 4.2.2 of the Technical Specifications. Even though there will be no changes or modifications to full length CEAs, for consistency and for ease of reading, the wording for "full length" CEAs will be changed to "full strength" CEAs. Additionally, TS 3.1.5 - "Control Element Assembly (CEA) Alignment," Condition B, will be modified to eliminate a potential condition which could cause an unwarranted plant shutdown.

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response

No. The physical difference between the 4-finger full strength control element assemblies (FSCEAs) and the PSCEAs involves using Inconel rather than B4C (boron carbide) over 100% of the active control section of each CEA finger. In addition, the PSCEAs use Inconel tubing to encase solid Inconel slugs, which cover the entire control section of the control element assembly (CEA). The current PLCEAs (also have only 4-fingers) use solid Inconel rods for only the lower half of each finger and B4C pellets in the top 15 inches (10%) of the control section of the CEA. Although failure of the solid Inconel region due to neutron fluence would be less likely than a typical clad design, the differences in swelling between the Inconel slugs encased by Inconel clad for the PSCEAs will be minor and result in a minimal impact on clad integrity. With the exception of the neutron absorber, the cladding design used for the PSCEAs is similar to the cladding of the full strength CEAs (FSCEAs). The geometry, cladding materials, and the spider assembly that supports the CEA fingers are essentially the same for the 4-finger FSCEAs and the PSCEAs. The principal difference results from the Inconel slugs contained in the PSCEAs being heavier than the B4C pellets used in the FSCEAs. Even though the weight of a 4-finger PSCEA is greater than the weight of a 4-finger PLCEA or a 4-finger FSCEA, this weight difference is bounded by the 12-finger FSCEAs which are operated by the same CEA drive mechanism system. The PSCEAs use Inconel as a neutron absorber in the entire control section of each CEA finger and will be operationally used the same way as the PLCEAs. In particular, the insertion restraints that are defined by the power dependent insertion limits (PDILs) for the PLCEAs will remain the same for the PSCEAs. This existing requirement will not result in any significant operational impact on 26

Enclosure the PSCEAs since the solid Inconel cylinder in the bottom 50% (operating range of the PDILs) of the PLCEAs has essentially the same reactivity worth as that of the PSCEAs. In addition, renaming the full length CEAs and part length CEAs to full strength CEAs and part strength CEAs, respectively, and providing definition for the PSCEAs will not impact the safe operation of the plant. The terminology will be appropriately changed in any related document, equipment tag, or indication on a control panel. The PLCEAs are not credited in the accident analyses for accident mitigation. The PSCEA design eliminates an accident scenario involving the insertion of a PLCEA past the PDIL, which results in an axial shift in power due to the upper region of the PLCEAs which has no neutron absorber. This condition will not occur with the PSCEAs because they are filled with neutron absorber over 100% of the control section of each finger. Concerning TS Limiting Condition for Operation (LCO) 3.1.5, Condition B, proposed change; there are three position indicator channels available for each CEA. Current TS Bases state that, "At least two of the following three CEA position indicator channels shall be OPERABLE for each CEA." Additionally the TS Bases states, "If only one CEA position indicator channel is OPERABLE, continued operation in MODES 1 and 2 may continue, provided, within 6 hours, at least two position indicator channels are returned to OPERABLE status; or within 6 hours and once per 12 hours, verify that the CEA group with the inoperable position indicators are either fully withdrawn or fully inserted while maintaining the insertion limits of LCO 3.1.6, LCO 3.1.7 and LCO 3.1.8." The TS Bases make no restriction or condition limiting only one CEA within a subgroup to having only one CEA position indication channel. Current analyses already assume that more than one CEA in a subgroup could have only one position indicator OPERABLE. Modifying the wording for Condition B, of LCO 3.1.5, will not affect the likelihood or consequences of a CEA drop, slip, ejection, or misalignment. This change will still require at least one position indication channel be available for each CEA. Consequently, the proposed change does not involve a significant increase in the probability or consequences of an accident.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response

No. The proposed changes do not introduce any new mode of plant operation and the PSCEAs, like the PLCEAs, are not relied upon for accident mitigation. The PSCEAs will be operated in exactly the same manner in which the PLCEAs are operated. The existing operating restrictions for the PLCEAs will apply to the 27

Enclosure PSCEAs. In particular, the power dependent insertion limit (PDIL) restrictions identified in the Core Operating Limits Report (COLR) will remain the same for the PSCEAs. The PSCEA design uses Inconel over the entire control section of each CEA finger, which will prevent the potential undesired flux redistribution currently associated with the misoperation of PLCEAs. Therefore, the analysis associated with the undesired flux redistribution misoperation for the PLCEAs will be eliminated from PVNGS safety analyses. PSCEA misoperation events are bounded by the existing PLCEA and FSCEA misoperation safety analyses. In addition, renaming (within the Technical Specifications) the "full length CEAs" and "part length CEAs" to "full strength CEAs" and "part length or part strength CEAs," respectively, and providing a definition for the PSCEAs will not impact the safe operation of the plant. The terminology will be appropriately changed in any related document, equipment tag, or indication on a control panel. Concerning TS LCO 3.1.5, Condition B proposed change, CEA position indication channels have no control function and provide input to the CEA Calculators (CEACs) and Core Protection Calculators (CPCs) for generation of a penalty factor. This change will still require at least one position indication channel be available for each CEA. Allowing Condition 'B' of LCO 3.1.5 to apply to more than one CEA per group does not create the possibility of a different type of malfunction than previously evaluated in the UFSAR. Therefore, the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in the margin of safety?

Response

No. The design of the PSCEAs is very similar to the FSCEAs except for the neutron absorber within each finger of a PSCEA. The PSCEAs do not have as strong of a neutron absorber (Inconel) as that which is contained in the FSCEAs (B4C). There is a weight difference which results from the Inconel slugs contained in the PSCEAs being heavier than the B4C pellets used in the FSCEAs. Even though the weight of the 4-finger PSCEAs is greater than the weight of the 4-finger PLCEAs, the CEA drive mechanism and support components shall operate within their design bases. Therefore, the PSCEAs can be considered adequate for safety-related applications. Consequently, the differences in design between the current PLCEAs and the PSCEAs do not adversely impact safe operation. The PLCEAs are not relied upon for shutdown margin or accident mitigation and no new requirements will apply to the PSCEAs. However, the design of the PSCEAs is effectively eliminating the concern associated with the insertion of the PLCEAs past the PDILs which could result in an undesirable shift in neutron flux 28

Enclosure to the top of the core due to the region within the PLCEAs that do not have neutron absorber. The PSCEAs have neutron absorber throughout their entire control section, which prevents a neutron flux shift to the top of the core if inserted past the PDIL, when compared to that of the PLCEAs. In addition, renaming the "full length CEAs" and "part length CEAs" to "full strength CEAs" and "part length or part strength CEAs," respectively, and providing definition for the PSCEAs will not impact the safe operation of the plant. The terminology will be appropriately changed in any related document, equipment tag, or indication on a control panel. Concerning TS LCO 3.1.5, Condition B, proposed change, the current licensing bases already considers having more than one CEA in a CEA group with only one available position indication. The TS Bases for LCO 3.1.5, Condition B states that, "At least two of the following three CEA position indicator channels shall be OPERABLE for each CEA." Additionally the Bases states, "If only one CEA position indicator channel is OPERABLE, continued operation in MODES 1 and 2 may continue, provided, within 6 hours, at least two position indicator channels are returned to OPERABLE status; or within 6 hours and once per 12 hours, verify that the CEA group with the inoperable position indicators are either fully withdrawn or fully inserted while maintaining the insertion limits of LCO 3.1.6, LCO 3.1.7 and LCO 3.1.8." The TS Bases make no restriction or condition limiting only one CEA within a subgroup, to having only one CEA position indication channel OPERABLE. Therefore, modifying the wording for LCO 3.1.5, Condition B, does not involve a significant reduction in the margin of safety since loss of indication to more than one CEA is already considered in the licensing bases. Therefore, the proposed changes do not involve a significant reduction in the margin of safety. Based on the above, APS concludes that the activities associated with 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. 5.2 Applicable Regulatory RequirementslCriteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met. In the application for a license to operate a facility, 10 CFR 50.34(b)(6)(ii) requires that the following shall be part of the Updated Final Safety Analysis Report (UFSAR): "Managerial and administrative controls to be used to assure safe operation. Appendix B, "Quality Assurance Criteria for 29

Enclosure Nuclear Power Plants and Fuel Reprocessing Plants," sets forth the requirements for such controls for nuclear power plants and fuel reprocessing plants. The information on the controls to be used for a nuclear power plant or a fuel reprocessing plant shall include a discussion of how the applicable requirements of Appendix B will be satisfied." In accordance with 10 CFR 50 Appendix B, a Quality Assurance Program, as outlined in Chapter 17.2 of the Palo Verde UFSAR, is utilized by APS for designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, and modifying activities that affect the safety-related functions of structures, systems, and components. As stated in the PVNGS Equipment Qualification Program, "The design, specification and procurement of new, replacement, or reworked equipment and parts shall consider the specific requirements necessary to maintain the continued qualification of installed equipment and environmental performance requirements of any 'new" equipment." Also, it states, "The qualification of new equipment and designs shall be verified prior to their installation in the plant." In accordance with the Palo Verde Quality Assurance Program, the qualification requirements involving the PSCEAs such as suitability, functionality, environmental, seismic, electromagnetic and radio interference, human factors, software life cycle failure mode analysis, defense in depth and diversity analysis, and TMI action items were evaluated to ensure that the PSCEAs meet or exceed the original PLCEA requirements. 10 CFR Appendix A to Part 50, "General Design Criteria for Nuclear Power Plants," related to the design and operability requirements of the CEAs has been assessed to assure that the PSCEAs will satisfy regulatory design requirements. The criteria associated with the CEAs are summarized below. Criterion 10 - Reactor design - The principle difference associated with the PSCEA is the total mass and distribution of neutron absorber. However, PSCEAs are not subject to the potential anticipated operational occurrences (AOOs) currently seen by the PLCEAs due to the uniform distribution of the neutron absorber over the entire control section of each CEA finger for the PSCEAs. The only other significant difference is the weight of a PSCEA which is greater that a PLCEA. However, this difference has been analyzed for, as has the performance capability of the 30

Enclosure CEA drive mechanisms, and found to be within design capabilities and design analyses. Criterion 12 - Suppression of reactor power oscillations - Axial power oscillations are controlled using the PLCEAs andlor FLCEAs. The PSCEAs will be equally effective since their reactivity worth within the PDILs is essentially the same. The ability to reliably detect and suppress power oscillations is unaffected by the proposed changes. Criterion 13 - Instrumentation and control - The existing systems and components used for monitoring and control of CEA positions are unaffected by the proposed changes and will be equally effective and relied upon for the control of the PSCEAs. The change for LCO 3.1.5, Condition B only addresses a more appropriate action to be taken given that the number of operable CEA position indications are less than that which is required for more than one CEA within a subgroup. Criterion 26 - Reactivity control system redundancy and capability - The operational reactivity control characteristic of the PSCEAs is nearly identical to the PLCEAs. Redundancy and capability for the PSCEAs to control reactivity is not impacted and remains bounded by maintaining the operational restrictions required by the PDILs. Criterion 27 - Combined reactivity control systems capability - The current design of the Reactor Control System includes a more than adequate capability for reactivity control using only the FLCEAs. As a result, neither the PLCEAs nor the PSCEAs are considered for shutdown margin and are not relied upon for accident mitigation. The design of the PSCEAs will not introduce any new effect which could impact the performance of the FLCEAs. Therefore, the reactivity control systems remain capable of reliably controlling reactivity changes to assure that under postulated accident conditions and with appropriate margin for stuck rods, the capability to cool the core is maintained. Criterion 28 - Reactivity limits - Operation of each reactor may require partial insertion of the PSCEAs in order to reconfigure the neutron flux distribution within the core. The ability of the PSCEAs to control reactivity for this effect is not impacted and remains bounded by maintaining the operational restrictions required by the PDILs. The PLCEAs and PSCEAs have nearly identical reactivity worth above the PDILs. Once inserted past the PDILs, the PSCEAs will add more negative reactivity than that of the PLCEAs when fully inserted into the core. Therefore, this proposed change would not cause a change in the amount or rate of reactivity increase different than what is already assumed in accident analyses. Criterion 29 - Protection against anticipated operational occurrences - The PLCEAs are not relied upon for accident mitigation and provide no safety function; however, insertion of the PLCEAs past the PDIL could result in 31

Enclosure an event which is qualified as an AOO. A potential problem results due to 40% of the upper control section of each finger containing no neutron absorber. Significant insertion past the PDIL could result in undesirable core power redistribution. Since the design of the PSCEAs provides neutron absorber through the entire control section of each CEA finger, violation of the PDIL will be bounded by the AQOs involving the FLCEAs, which use a more reactive neutron absorber than present in the PSCEAs. Additionally, changing the name of the FLCEAs to FSCEAs does not affect safety function. The requirements for Limiting Conditions for Operation (LCO) and Surveillance Requirements (SRs) to be included in the Technical Specifications (TS) are found in 10 CFR 50.36. As stated previously, the replacement PSCEAs (as they will be used with the existing PDILs) are functionally equivalent to the existing PLCEAs. Similarly, the proposed TS revisions are written to meet the same intent as the previous. Therefore, the lowest functional capability or performance levels of equipment required for safe operation of the facility will be retained in the proposed amendment. Likewise, Surveillance Requirements in the proposed amendment will continue to assure that the necessary quality of systems and components are maintained that facility operation will be within safety limits, and that limiting conditions for operation will be met.

6.0 ENVIRONMENTAL CONSIDERATION

S Arizona Public Service Company has evaluated the proposed changes and has determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amount of effluent that may be released offsite, or (iii) a significant increase in the individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), an environmental assessment of the proposed change is not required. 32

ATTACHMENT I MARKUP OF TECHNICAL SPECIFICATION PAGES NOTE: The attached marked up Technical Specification pages for LCO 3.3.3 (only) are the associated pages for a pending change with the NRC for the approval of the replacement of Core Protection Calculator Systems (CPCS), submitted on 11107102 (102-04864-CDMtTNWIDWG - Request for Amendment to Technical Specifications: 3.2.4, Departure From Nucleate Boiling Ratio (DNBR), 3.3.1, Reactor Protective System (RPS) Instrumentation - Operating, 3.3.3, Control Element Assembly Calculators (CEACs)) /-

PALO VERDE NUCLEAR GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.2 SL Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) -- Reactor Trip Breakers Open 3.1.2 SHUTDOWN MARGIN (SDM) -- Reactor Trip Breakers Closed 3.1.3 Reactivity Balance 3.1.4 Moderator Temperature Coefficient (MTC) 3.1.5 Control Element Assembly (CEA) Alignment 3.1.6 Shutdown CEA Insertion Limits 3.1.7 Regulating CEA Insertion Limits 3.1.8 Part Length or Part Strength)CEA Insertion Limits 3.1.9 Special Test Exception (STE) -- SHUTDOWN MARGIN (SDM) 3.1.10 STE -- MODES 1 and 2 3.1.11 STE -- Reactivity Coefficient Testing 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Linear Heat Rate (LHR) 3.2.2 Planar Radial Peaking Factors (Fxy) 3.2.3 Azimuthal Power Tilt (Tq) 3.2.4 Departure From Nucleate Boiling Ratio (DNBR) 3.2.5 Axial Shape Index (ASI) PAL VED UNT 1=, = AMENDMENT.NO. _~ PALO VERDE UNITS 1,2.3 i AMENDMENT NO. +i-
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Definitions 1.1 1.1 Definitions (continued) ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e.. the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential. overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. Knlis the K effective calculated by considering the actual CEA configuration and assuminrg Ah-at-the fully or partially inserted full ke thotrEgNt CEA of highest worth is fully withdrawn. LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff),

that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System.

(continued) PALO VERDE UNITS 1.2.3 1.1-4 AMENDMENT NO. 1441, +8&

Definitions 1.1 1.1 Definitions (continued) RATED THERMAL POWER (RTP) REACTOR PROTECTIVE SYSTEM (RPS) RESPONSE TIME SHUTDOWN MARGIN (SDM) RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3876 MWt. The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All full lenigth (strength)CEAs (shutdown and regulating) are fully inserted-exZept-for the single CEA of highest reactivity worth, which is assume e

to be fully withdrawn. With any full teint strength CEAs not capable of being fully inserted, the withdrawn reactivity worth of these CEAs must be accounted for in the determination of SDM and

b. There is position.

no change in part length or part strength CEA (continued) PALO VERDE UNITS 1.2.3 1.1-6 AMENDMENT NO. 1J, 4

&

CEA Alignment 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Element Assembly (CEA) Alignment LCO 3.1.5 APPLICABILITY: All full all be OPERABLE. and all full and part length or ar Strength CEAs shall be aligned to within 6.6 inches in icate position) of all other CEAs in their respective groups. MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more CEAs A.1 Reduce THERMAL POWER 1 hour trippable and in accordance with misaligned from its the limits in the group by > 6.6 inches COLR. and < 9.9 inches. AND OR A.2 Restore CEA 2 hours One CEA trippable and alignment. misaligned from its group by > 9.9 inches. (continued) PALO VERDE UNITS 1.2.3 3.1.5-1 AMENDMENT NO. +ii-L

CEA Alignment 3.1.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Only one CEA position B.1 Restore at least two 6 hours indicator channel position indicator OPERABLE for one channels to OPERABLE pep -GA CGreo. status. OR For more CEA(s). OR B.2 Verify the CEA 6 hours Group(s) with the inoperable position AND indicators are fully withdrawn or fully Once per 12 inserted while hours maintaining the thereafter. insertion limits of LCO 3.1.6. LCO 3.1.7 and LCO 3.1.8. C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A or B not met OR One or m~reful length-strength CEAs untrippaDTe. D. Two or more CEAs D.1 Open the reactor trip Immediately trip able and breakers. misaligned from their group by > 9.9 inches. PALO VERDE UNITS 1.2.3 3.1.5-2 AMENDMENT NO. 44-jL

CEA Alignment 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 hif the indicated positiornof eachfull 12 hours strength and part length or part strength LIA is within 6.6 inches OT all other s i its group. SR 3.1.5.2 Verify that, for each CEA. its OPERABLE CEA 12 hours position indicator channels indicate within 5.2 inches of each other. SR 3.1.5.3 Verify full +emgth CEA freedom of 92 days movement (trippabiliTUt7ycY).i g each individual full length strength CEA that is not fully inserted in tfecgoreffat least 5 in hes. SR 3.1.5.4 Perform a CHANNEL FUNCTIONAL TEST of each 18 months reed switch position transmitter channel. SR 3.1.5.5 Verify each full +en& CEA drop Prior to time < 4.0 seconds. reactor criticality, after each removal of the reactor head PALO VERDE UNITS 1.2,3 3.1.5-3 AMENDMENT NO. 44 Part Length or Part Strength CEA hnsertion .ifif nj 3T.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 Part Length or Part Strength Control Element Assembly (CEA) Insertion Limits LCO 3.1.8 APPLICABILITY: The part lengtho pat tength) CEA groups shall be limited to the insertion limits specified in the COLR. MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION ICOMPLETION TIME A. EPertlength fart (strength CEA groups Tinseted beyond the transient insertion lim A.1 ERelstre paLtlength & ,part strength)CEA group to within the limit. 2 hours it. OR A. 2 Reduce THERMAL POWER to less than or equal to that fraction of RTP specified in the COLR. 2 hours 4. B. Part lengthTEA groups inserted between the long term steady state insertion limit and the transient insertion limit for intervals 2 7 effective full power days (EFPD) per 30 EFPD or 2 14 EFPD per 365 EFPD interval. Restore part length -;CEA groups to within the long term steady state insertion limit. 2 hours (continued) PALO VERDE UNITS 1.2.3 3.1.8-1 AMENDMENT NO. +1-
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Part Length CEA Insertion Limits N 3.1.8 lor Part Strength I ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Verify part lengt CEA group position. 12 hours lor part strength PALO VERDE UNITS 1,2.3 3.1.8-2 AMENDMENT NO. 444

STE-SDM 3.1.9 3.1 REACTIVITY CONTROL SYSTEMS 3.1.9 Special Test Exception (STE) - SHUTDOWN MARGIN (SDM) LCO 3.1.9 During performance of PHYSICS TESTS. the requirements of: LCO 3.1.2. LCO 3.1.6, LCO 3.1.7 "SHUTDOWN MARGIN (SOM)-Reactor Trip Breakers Closed"; "Shutdown Control Element Assembly (CEA) Insertion Limits", and "Regulating Control Element Assembly (CEA) Insertion Limits" may be suspended for measurement of CEA worth, provided shutdown reactivity equivalent to at least the highest estimated CEA worth (of those CEAs actually withdrawn) is available for trip insertion or the reactor is subcritical by at least the reactivity equivalent of the highest CEA worth. APPLICABILITY: MODES 2 and 3 during PHYSICS TESTS. NOTE---------------------------- Operation in MODE 3 shall be limited to 6 consecutive hours. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any full l A A.1 Initiate boration to 15 minutes not fully inser and restore required less than the requ dshutdown reactivity. shutdown reactivity available for trip insertion. OR s All full EAs inserted and the reactor subcritical by less than the above required shutdown reactivity equivalent. PALO VERDE UNITS 1.2,3 3.1.9-1 AMENDMENT NO. +1-
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STE-SDM 3.1.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.9.1 Verify that the position of each CEA not 2 hours fully inserted is within the acceptance criteria for available negative reactivity addition. SR 3.1.9.2 Verify each full CEA not fully Within 7 days inserted is capabl f full insertion when prior to tripped from at ISst the 50% withdrawn reducing SDM position. requirements to less than the limits of LCO 3.1.2 SR 3.1.9.3 E T------------------- Only required to be p rmed in Mode 3. Verify that with all full CEAs fully 2 hours inserted, the reactor is subcritical within the acceptance criteria. PALO VERDE UNITS 1.2,3 3.1.9-2 AMENDMENT NO. 44 STE - MODES 1 and 2 3.1.10 3.1 REACTIVITY CONTROL SYSTEMS 3.1.10 Special Test Exceptions (STE) - MODES 1 and 2 LCO 3.1.10 During performance of PHYSICS TESTS, the requirements of: LCO 3.1.4. LCO 3.1.5. LCO 3.1.6. LCO 3.1.7, "Moderator Temperature Coefficient (MTC)"; "Control Element Assembly (CEA) Alignment"; "Shutdown Control Element Assembly (CEA) Insertion Limits"; Strntht "Regulating Control Eleme y (CEA) Insertion Lim t "~ "Part Length STA Insertion Limits"; "Planar Radial Peaking Factors (Fxy)"; "AZIMUTHAL POWER TILT (Tq)"; "AXIAL SHAPE INDEX (ASI)"; and "Control Element Assembly Calculators (CEACs)" LCO LCO LCO LCO LCO 3.1.8. 3.2.2. 3.2.3. 3.2.5. 3.3.3. may be suspended. provided THERMAL POWER is restricted to the test power plateau, which shall not exceed 85% RTP. APPLICABILITY: MODES 1 and 2 during PHYSICS TESTS. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Test power plateau A.1 Reduce THERMAL POWER 15 minutes exceeded. to less than or equal to the test power plateau. B. Required Action and B.1 Suspend PHYSICS 1 hour associated Completion TESTS. Time not met. PALO VERDE UNITS 1,2.3 3.1.10-1 AMENDMENT NO. +1-7

STE - Reactivity Coefficient Testing 3.1.11 3.1 REACTIVITY CONTROL SYSTEMS 3.1.11 Special Test Exceptions (STE) - Reactivity Coefficient Testing LCO 3.1.11 During performance of PHYSICS TESTS, the requirements of: LCO 3.1.7. "Regulating Control Element Assemblr Insertion Limits" LCO 3.1.8. "Part Length rol Element Assembly (CEA) Insertion Limits;" and LCO 3.4.1. "RCS Pressure, Temperature and Flow limits" (LCO 3.4.1.b. RCS Cold Leg Temperature only) may be suspended, provided LHR and DNBR do not exceed the limits in the COLR. APPLICABILITY: MODE 1 with Thermal Power > 20% RTP during PHYSICS TESTS. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LHR or DNBR outside A.1 Reduce THERMAL POWER 15 minutes the limits specified to restore LHR and in the COLR. DNBR to within limits. B. Required Action and B.1 Sus end PHYSICS 1 hour associated Completion TESTS. Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.11.1 Verify LHR and DNBR do not exceed limits by Continuously performing SR 3.2.1.1 and SR 3.2.4.1. PALO VERDE UNITS 1,2.3 3.1.11-1 AMENDMENT NO. 4-1 CEACs (Before CPC Upgrade) 3.3.3 ACTIONS CONDITION I REQUIRED ACTION ICOMPLETION TIME B. (continued) B.2 Verify all full tengthand par length or par strength control element assemD (CEA) groups are fully withdrawn and maintained fully withdrawn, except during Surveillance testing pursuant to SR 3.1.5.3 or for control, when CEA group #5 may be inserted to a maximum of 127.5 inches withdrawn. 4 hours I AND B.3 AND B.4 AND B.5 Verify the "RSPT/CEAC Inoperable" addressable constant in each core protection calculator (CPC) is set to indicate that both CEACs are inoperable. Verify the Control Element Drive Mechanism Control System is placed in "STANDBY MODE" and maintained in "STANDBY MODE," except during CEA motion permitted by Required Action B.2. Perform SR 3.1.5.1. 4 hours 4 hours Once per 4 hours (continued) AND PALO VERDE UNITS 1.2.3 3.3.3-2 AMENDMENT NO. 44,2

CEACs (After CPC Upgrade) 3.3.3 ACTIONS (continued) CONDITION [ REQUIRED ACTION ICOMPLETION TIME B. (continued) B.2.1 Verify the departure 4 hours from nucleate boiling ratio requirement of LCO 3.2.4, "Departure from Nucleate Boiling Ratio (DNBR)." is met. AND B.2.2 Verify all full (strengB 1 1-and prtm, -ength cor part strength)contrc l element assemDIy (CEA) groups are fully withdrawn and maintained fully withdrawn. except during Surveillance testing pursuant to SR 3.1.5.3 or for control, when CEA group #5 may be inserted to a maximum of 127.5 inches withdrawn. AND B.2.3 Verify the "RSPT/CEAC 4 hours Inoperable" addressable constant in each affected core protection calculator (CPC) is set to indicate that both CEACs are inoperable. AND (continued) PALO VERDE UNITS 1.2.3 3.3.3-6 AMENDMENT NO.

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palo Verde Nuclear Generating Station is located in Maricopa County, Arizona, approximately 50 miles west of the Phoenix metropolitan area. The site is comprised of approximately 4.050 acres. Site elevations range from 890 feet above mean sea level at the southern boundary to 1,030 feet above mean sea level at the northern boundary. The minimum distance from a containment building to the exclusion area boundary is 871 meters. 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 241 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO

2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Other cladding material may be used with an approved exemption. 4.2.2 Control Element Assemblies q n n . I s s roactor ocr; cfl8a contain 'l t tuH [,Wngl l anci iJ pa I cge n Gomt;ol W omont accmbilcz (CEAG).

Tho cantrol im-atrio aho-l -bo bcr3n rb lI

-T no-anel Alloy 625 uced ao G .:vr aw aborbcrS Qorr P Hmrio 0f Mth part length eontrol eleFment aoccmblico as appreveJ b, ticW-C .~~~ _Ag The reactor core shall contain 76 full strength and either 13 part length or 13 part strength control element assemblies (CEAs). The control section for the full strength CEAs shall be boron carbide with Inconel Alloy 625 cladding. For units that have part length CEAs, the control section shall be Inconel Alloy 625 in the lower half, followed by perforated stainless steel tubing over the next 40%, and boron carbide pellets with Inconel Alloy 625 clad over the last 10% of the control section. For units that have part strength CEAs, the control section shall be solid Inconel Alloy 625 slugs with Inconel Alloy 625 cladding.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Shutdown Margin - Reactor Trip Breakers Open for Specification 3.1.1.

2. Shutdown Margin - Reactor Trip Breakers Closed for Specification 3.1.2.

3. Moderator Temperature Coefficient BOL and EOL limits for Specification 3.1.4.

4.

Boron Dilution Alarm System for Specification 3.3.12.

5. CEA Alignment for Specification 3.1.5.

6.

Regulating CEA Insertion Limits for Specification or Part Stren th 3.1.7.

7.

Part en EA Insertion Limits for Specification 3.1.8.

8. Linear Heat Rate for Specification 3.2.1.

9. Azimuthal Power Tilt -

Tq for Specification 3.2.3.

10.

DNBR for Specification 3.2.4.

11.

Axial Shape Index for Specification 3.2.5.

12.

Boron Concentration (Mode 6) for Specification 3.9.1.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents: NOTE------------------------- The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). (continued) PALO VERDE UNITS 1,2.3 5.6-3 AMENDMENT NO. 4,-4. 18-

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 Core Operating Limits Report (COLR) (continued)

1.

"CE Method for Control Element Assembly Ejection Analysis," CENPD-0190-A. (Methodology for Specification 3.1.7. Regulating CEA Insertion Limits).

2.

"The ROCS and DIT Computer Codes for Nuclear Design." CENPD-266-P-A. [Methodology for Specifications 3.1.1. Shutdown Margin - Reactor Trip Breakers Open; 3.1.2. Shutdown Margin - Reactor Trip Breakers Closed; 3.1.4. Moderator Temperature Coefficient BOL and EOL limits; 3.1.7. Regulating CEA Insertion Limits and 3.9.1, Boron Concentration (Mode 6)].

3.

"Safety Evaluation Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR System 80. Docket No. STN 50-470, "NUREG-0852 (November 1981), Supplements No. 1 (March 1983). No. 2 (September 1983), No. 3 (December 1987) [Methodology for Specifications 3.1.2. Shutdown Margin - Reactor Trip Breakers Closed; 3.1.4, Moderator Temperature Coefficient BOL and EOL limits; 3.3.12, Boron Dilution Alarm System; 3.1.5. CEA Alignment; 3.1.7. Regulating CEA Insertion Limits; 3.1.8. Part Length A Insertion Limits and 3.2.3. Azimuthal Power Tilt - o at r t

4.

"Modified Statistical Combination of Uncertainties," CEN-356(V)-P-A and "System 80"" Inlet Flow Distribution," Supplement 1-P to Enclosure 1-P to LD-82-054, (Methodology for Specification 3.2.4. DNBR and 3.2.5 Axial Shape Index).

5.

"Calculative Methods for the CE Large Break LOCA Evaluation Model," CENPD-132. (Methodology for Specification 3.2.1, Linear Heat Rate).

6.

"Calculative Methods for the CE Small Break LOCA Evaluation Model," CENPD-137-P, (Methodology for Specification 3.2.1. Linear Heat Rate). (continued) PALO VERDE UNITS 1,2.3 5.6-4 AMENDMENT NO. F12, '37

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 Core Operating Limits Report (COLR) (continued)

7.

Letter: O.D. Parr (NRC) to F. M. Stern (CE), dated June 13. 1975 (NRC Staff Review of the Combustion Engineering ECCS Evaluation Model). NRC approval for:

5..5.b.6.

8.

Letter: K. Kniel (NRC) to A. E. Scherer (CE). dated September 27, 1977 (Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137. Supplement 1-P). NRC approval for 5.6.5.b.6.

9.

"Fuel Rod Maximum Allowable Pressure." CEN-372-P-A, (Methodology for Specification 3.2.1. Linear Heat Rate).

10.

Letter: A. C. Thadani (NRC) to A. E. Scherer (CE). dated April 10. 1990, ("Acceptance for Reference CE Topical Report CEN-372-P"). NRC approval for 5.6.5.b.9.

11.

'Arizona Public Service Company PWR Reactor Physics Methodology Using CASMO-4/SIMULATE-3," [Methodology for Specifications 3.1.1. Shutdown Margin - Reactor Trip Breakers Open: 3.1.2. Shutdown Margin - Reactor Trip Breakers C]osed: 3.1.4, Moderator Temperature Coefficient: 3.1.7. Regulating CEA Insertion Limits and 3.9.1. Boron Concentration (Mode 6)].

12.

"Technical Manual for the CENTS Code," CE-NPD 282-P-A, lor Part Strenoth r Volumes 1-3. [Methodology for Specifications 3.1.2. Shutdown Margin-Reactor Trip Breakers Closed; 3.1.4. erator Temperature Coefficient: 3.1.5. CEA Alignment: Len iL Regulating CEA Insertion Limits; 3.1.8. Part Length'CEA Insertion Limits and 3.2.3. Azimuthal Power Tilt-Tq].

13.

CENPD-404-P-A. "Implementation of ZIRLOC Cladding Material in CE Nuclear Power Fuel Assembly Designs.

c. The core operating limits shall be determined such that all applicable limits (e.g.. fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM. transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR. including any mid cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued) PALO VERDE UNITS 1.2.3 5.6-5 AMENDMENT NO. 4J-, 149-

ATTACHMENT 2 RETYPED TECHNICAL SPECIFICATION PAGES NOTE: The attached retyped Technical Specification pages for LCO 3.3.3 (only) are the associated pages for a pending change with the NRC for the approval of the replacement of Core Protection Calculator Systems (CPCS), submitted on 11/07102 (102-04864-CDMITNW/DWG - Request for Amendment to Technical Specifications: 3.2.4, Departure From Nucleate Boiling Ratio (DNBR), 3.3.1, Reactor Protective System (RPS) Instrumentation - Operating, 3.3.3, Control Element Assembly Calculators (CEACs))

PALO VERDE NUCLEAR GENERATING STATION IMPROVED TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.2 SL Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) -- Reactor Trip Breakers Open 3.1.2 SHUTDOWN MARGIN (SDM) -- Reactor Trip Breakers Closed 3.1.3 Reactivity Balance 3.1.4 Moderator Temperature Coefficient (MTC) 3.1.5 Control Element Assembly (CEA) Alignment 3.1.6 Shutdown CEA Insertion Limits 3.1.7 Regulating CEA Insertion Limits 3.1.8 Part Length or Part Strength CEA Insertion Limits 3.1.9 Special Test Exception (STE) -- SHUTDOWN MARGIN (SDM) 3.1.10 STE -- MODES 1 and 2 3.1.11 STE -- Reactivity Coefficient Testing 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Linear Heat Rate (LHR) 3.2.2 Planar Radial Peaking Factors (Fxy) 3.2.3 Azimuthal Power Tilt (Tq) 3.2.4 Departure From Nucleate Boiling Ratio (DNBR) 3.2.5 Axial Shape Index (ASI) PALO VERDE UNITS 1.2,3 i AMENDMENT NO. 4-1-7. PALO VERDE UNITS 1.2,3 i AMENDMENT NO. 14:,Z

Definitions 1.1 1.1 Definitions (continued) ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. K.-l Kn1is the K effective calculated by considering the actual CEA configuration and assuming that the fully or partially inserted full strength CEA of highest worth is fully withdrawn. I LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff),

that is captured and conducted to collection systems or a sump or collecting tank;

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE: or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System.

(continued) PALO VERDE UNITS 1.2,3 1.1-4 AMENDMENT NO.46

Definitions 1.1 1.1 Definitions (continued) RATED THERMAL POWER (RTP) REACTOR PROTECTIVE SYSTEM (RPS) RESPONSE TIME SHUTDOWN MARGIN (SDM) RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3876 MWt. The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC. SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All full strength CEAs (shutdown and regulating) are fully inserted except for the single CEA of highest reactivity worth, which is assumed to be fully withdrawn. With any full strength CEAs not capable of being fully inserted, the withdrawn reactivity worth of these CEAs must be accounted for in the determination of SDM and I

I

b. There is strength no change in part length or part CEA position.

I (continued) PALO VERDE UNITS 1,2,3 1.1-6 AMENDMENT NO. 449

CEA Alignment 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Control Element Assembly (CEA) Alignment LCO 3.1.5 APPLICABILITY: All full strength CEAs shall be OPERABLE, and all full strength and part length or part strength CEAs shall be aligned to within 6.6 inches (indicated position) of all other CEAs in their respective groups. MODES 1 and 2. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more CEAs A.1 Reduce THERMAL POWER 1 hour trippable and in accordance with misaligned from its the limits in the group by > 6.6 inches COLR. and < 9.9 inches. AND OR A.2 Restore CEA 2 hours One CEA trippable and alignment. misaligned from its group by > 9.9 inches. (continued) PALO VERDE UNITS 1.2.3 3.1.5-1 AMENDMENT NO. 19

CEA Alignment 3.1.5 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. Only one CEA position B.1 Restore at least two 6 hours indicator channel position indicator OPERABLE for one or channels to OPERABLE more CEA(s). status. OR B.2 Verify the CEA 6 hours Group(s) with the inoperable position AND indicators are fully withdrawn or fully Once per 12 inserted while hours maintaining the thereafter. insertion limits of LCO 3.1.6. LCO 3.1.7 and LCO 3.1.8. C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition A or B not met OR One or more full strength CEAs untrippable. D. Two or more CEAs D.1 Open the reactor trip Immediately trip pable and breakers. misaligned from their group by > 9.9 inches. I PALO VERDE UNITS 1,2,3 3.1.5-2 AMENDMENT NO. 44-;L

CEA Alignment 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify the indicated position of each full 12 hours strength and part length or part strength CEA is within 6.6 inches of all other CEAs in its group. SR 3.1.5.2 Verify that, for each CEA, its OPERABLE CEA 12 hours position indicator channels indicate within 5.2 inches of each other. SR 3.1.5.3 Verify full strength CEA freedom of 92 days movement (trippability) by moving each individual full strength CEA that is not fully inserted in the core at least 5 inches. SR 3.1.5.4 Perform a CHANNEL FUNCTIONAL TEST of each 18 months reed switch position transmitter channel. SR 3.1.5.5 Verify each full strength CEA drop time Prior to < 4.0 seconds. reactor criticality. after each removal of the reactor head I PALO VERDE UNITS 1,2.3 3.1.5-3 AMENDMENT NO. 44 Part Length or Part Strength CEA Insertion Limits 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 Part Length or Part Strength Control Element Assembly (CEA) Insertion Limits I LCO 3.1.8 APPLICABILITY: Thegprt length oripart strength CEA groups shall be limited to the insertion limits specified in the COLR. MODES 1 and 2. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Part length or part A.1 Restore part length or 2 hours strength CEA groups part strength CEA inserted beyond the groups to within the transient insertion limit. limit. OR A.2 Reduce THERMAL POWER 2 hours to less than or equal to that fraction of RTP specified in the COLR. B. Part length or part B.1 Restore part length or 2 hours strength CEA groups part strength CEA inserted between the groups to within the long term steady state long term steady state insertion limit and insertion limit. the transient insertion limit for intervals 2 7 effective full power days (EFPD) per 30 EFPD or 2 14 EFPD per 365 EFPD interval. (continued) PALO VERDE UNITS 1.2.3 3.1.8-1 AMENDMENT NO. 14

Part Length or Part Strength CEA Insertion Limits 3.1.8 I ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Verify part length or part strength CEA 12 hours group position. I PALO VERDE UNITS 1.2,3 3.1.8-2 AMENDMENT NO. 447-

STE-SDM 3.1.9 3.1 REACTIVITY CONTROL SYSTEMS 3.1.9 Special Test Exception (STE) - SHUTDOWN MARGIN (SDM) LCO 3.1.9 During performance of PHYSICS TESTS. the requirements of: LCO 3.1.2, LCO 3.1.6. LCO 3.1.7 "SHUTDOWN MARGIN (SDM)-Reactor Trip Breakers Closed"; "Shutdown Control Element Assembly (CEA) Insertion Limits", and "Regulating Control Element Assembly (CEA) Insertion Limits" may be suspended for measurement of CEA worth, provided shutdown reactivity equivalent to at least the highest estimated CEA worth (of those CEAs actually withdrawn) is available for trip insertion or the reactor is subcritical by at least the reactivity equivalent of the highest CEA worth. APPLICABILITY: MODES 2 and 3 during PHYSICS TESTS.

---

NOTE----------------------------

Operation in MODE 3 shall be limited to 6 consecutive hours. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any full strength CEA A.1 Initiate boration to 15 minutes not fully inserted and restore required less than the required shutdown reactivity. shutdown reactivity available for trip insertion. OR All full strength CEAs inserted and the reactor subcritical by less than the above required shutdown reactivity equivalent. I I PALO VERDE UNITS 1,2.3 3.1.9-1 AMENDMENT NO. 44,1

STE-SDM 3.1.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.9.1 Verify that the position of each CEA not 2 hours fully inserted is within the acceptance criteria for available negative reactivity addition. SR 3.1.9.2 Verify each full strength CEA not fully Within 7 days inserted is capable of full insertion when prior to tripped from at least the 50% withdrawn reducing SDM position. requirements to less than the limits of LCO 3.1.2 SR 3.1.9.3

---

NOTE-------------------

Only required to be performed in Mode 3. Verify that with all full strength CEAs 2 hours fully inserted. the reactor is subcritical within the acceptance criteria. I I PALO VERDE UNITS 1.2,3 3.1.9-2 AMENDMENT NO. 44 STE - MODES 1 and 2 3.1.10 3.1 REACTIVITY CONTROL SYSTEMS 3.1.10 Special Test Exceptions (STE) - MODES 1 and 2 LCO 3.1.10 During performance of PHYSICS TESTS, the requirements of: [CO 3.1.4, LCO 3.1.5, LCO 3.1.6. LCO 3.1.7. LCO 3.1.8, "Moderator Temperature Coefficient (MTC)"; "Control Element Assembly (CEA) Alignment"; "Shutdown Control Element Assembly (CEA) Insertion Limits"; "Regulating Control Element Assembly (CEA) Insertion Limits"; "Part Length or Part Strength CEA Insertion Limits": "Planar Radial Peaking Factors (Fxy)": "AZIMUTHAL POWER TILT (Tq)": "AXIAL SHAPE INDEX (ASI)"; and "Control Element Assembly Calculators (CEACs)" I LCO LCO LCO LCO 3.2.2, 3.2.3, 3.2.5, 3.3.3. may be suspended, provided THERMAL POWER is restricted to the test power plateau, which shall not exceed 85% RTP. APPLICABILITY: MODES 1 and 2 during PHYSICS TESTS. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Test power plateau A.1 Reduce THERMAL POWER 15 minutes exceeded. to less than or equal to the test power plateau. B. Required Action and B.1 Suspend PHYSICS 1 hour associated Completion TESTS. Time not met. PALO VERDE UNITS 1.2,3 3.1.10-1 AMENDMENT NO. 44,3-

STE - Reactivity Coefficient Testing 3.1.11 3.1 REACTIVITY CONTROL SYSTEMS 3.1.11 Special Test Exceptions (STE) - Reactivity Coefficient Testing LCO 3.1.11 During performance of PHYSICS TESTS, the requirements of: LCO 3.1.7, LCO 3.1.8, LCO 3.4.1. "Regulating Control Element Assembly (CEA) Insertion Limits"; "Part Length or Part Strength Control Element Assembly (CEA) Insertion Limits:" and "RCS Pressure, Temperature and Flow limits" (LCO 3.4.1.b, RCS Cold Leg Temperature only) I may be suspended, provided LHR and DNBR do limits in the COLR. not exceed the APPLICABILITY: MODE 1 with Thermal Power > 20% RTP during PHYSICS TESTS. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LHR or DNBR outside A.1 Reduce THERMAL POWER 15 minutes the limits specified to restore LHR and in the COLR. DNBR to within limits. B. Required Action and B.1 Suspend PHYSICS 1 hour associated Completion TESTS. Time not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.11.1 Verify LHR and DNBR do not exceed limits by Continuously performing SR 3.2.1.1 and SR 3.2.4.1. PALO VERDE UNITS 1,2.3 3.1.11-1 AMENDMENT NO. 114-

CEACs (Before CPC Upgrade) 3.3.3 ACTIONS CONDITION REQUIRED ACTION ICOMPLETION TIME B. (continued) B.2 AND B.3 AND B.4 AND B.5 Verify all full strength and part length or part strength control element assembly (CEA) groups are fully withdrawn and maintained fully withdrawn, except during Surveillance testing pursuant to SR 3.1.5.3 or for control, when CEA group #5 may be inserted to a maximum of 127.5 inches withdrawn. Verify the "RSPT/CEAC Inoperable" addressable constant in each core protection calculator (CPC) is set to indicate that both CEACs are inoperable. Verify the Control Element Drive Mechanism Control System is placed in "STANDBY MODE" and maintained in "STANDBY MODE," except during CEA motion permitted by Required Action B.2. Perform SR 3.1.5.1. 4 hours 4 hours 4 hours Once per 4 hours (continued) AND PALO VERDE UNITS 1.2.3 3.3.3-2 AMENDMENT NO. 44 CEACs (After CPC Upgrade) 3.3.3 ACTIONS (continued) CONDITION I REQUIRED ACTION ICOMPLETION TIME B. (continued) B.2.1 Verify the departure 4 hours from nucleate boiling ratio requirement of LCO 3.2.4, "Departure from Nucleate Boiling Ratio (DNBR)." is met. AND B.2.2 Verify all full strength and part length or part strength control element assembly (CEA) groups are fully withdrawn and maintained fully withdrawn, except during Surveillance testing pursuant to SR 3.1.5.3 or for control, when CEA group #5 may be inserted to a maximum of 127.5 inches withdrawn. AND B.2.3 Verify the "RSPT/CEAC 4 hours Inoperable" addressable constant in each affected core protection calculator CCPC) is set to indicate that both CEACs are inoperable. AND (continued) PALO VERDE UNITS 1,2.3 3.3.3-6 AMENDMENT NO.

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palo Verde Nuclear Generating Station is located in Maricopa County, Arizona, approximately 50 miles west of the Phoenix metropolitan area. The site is comprised of approximately 4,050 acres. Site elevations range from 890 feet above mean sea level at the southern boundary to 1,030 feet above mean sea level at the northern boundary. The minimum distance from a containment building to the exclusion area boundary is 871 meters. 4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 241 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO2) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Other cladding material may be used with an approved exemption. 4.2.2 Control Element Assemblies The reactor core shall contain 76 full strength and either 13 part length or 13 part strength control element assemblies (CEAs). The control section for the full strength CEAs shall be boron carbide with Inconel Alloy 625 cladding. For units that have part length CEAs. the control section shall be Inconel Alloy 625 in the lower half, followed by perforated stainless steel tubing over the next 40%. and boron carbide pellets with Inconel Alloy 625 clad over the last 10% of the control section. For units that have part strength CEAs, the control section shall be solid Inconel Alloy 625 slugs with Inconel Alloy 625 cladding. (continued) PALO VERDE UNITS 1,2,3 4.0-1 AMENDMENT NO. 44 Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Shutdown Margin - Reactor Trip Breakers Open for Specification 3.1.1.

2. Shutdown Margin - Reactor Trip Breakers Closed for Specification 3.1.2.

3. Moderator Temperature Coefficient BOL and EOL limits for Specification 3.1.4.

4. Boron Dilution Alarm System for Specification 3.3.12.

5. CEA Alignment for Specification 3.1.5.

6. Regulating CEA Insertion Limits for Specification 3.1.7.

7. Part Length or Part Strength CEA Insertion Limits for Specification 3.1.8.

8. Linear Heat Rate for Specification 3.2.1.

9. Azimuthal Power Tilt - Tq for Specification 3.2.3.

10.

DNBR for Specification 3.2.4.

11.

Axial Shape Index for Specification 3.2.5.

12.

Boron Concentration (Mode 6) for Specification 3.9.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. specifically those described in the following documents:

---

NOTE-------------------------

The COLR will contain the complete identification for each of the Technical Specification referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements). (continued) PALO VERDE UNITS 1.2.3 5.6-3 AMENDMENT NO. 44

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 Core Operating Limits Report (COLR) (continued)

1. "CE Method for Control Element Assembly Ejection Analysis," CENPD-0190-A, (Methodology for Specification 3.1.7. Regulating CEA Insertion Limits).

2. "The ROCS and DIT Computer Codes for Nuclear Design,"

CENPD-266-P-A, [Methodology for Specifications 3.1.1, Shutdown Margin - Reactor Trip Breakers Open; 3.1.2. Shutdown Margin - Reactor Trip Breakers Closed; 3.1.4. Moderator Temperature Coefficient BOL and EOL limits; 3.1.7, Regulating CEA Insertion Limits and 3.9.1. Boron Concentration (Mode 6)].

3. "Safety Evaluation Report related to the Final Design of the Standard Nuclear Steam Supply Reference Systems CESSAR System 80, Docket No. STN 50-470, "NUREG-0852 (November 1981). Supplements No. 1 (March 1983), No. 2 (September 1983). No. 3 (December 1987) [Methodology for Specifications 3.1.2, Shutdown Margin - Reactor Trip Breakers Closed; 3.1.4. Moderator Temperature Coefficient BOL and EOL limits; 3.3.12, Boron Dilution Alarm System; 3.1.5. CEA Alignment; 3.1.7. Regulating CEA Insertion Limits; 3.1.8. Part Length or Part Strength CEA Insertion Limits and 3.2.3, Azimuthal Power Tilt - Tq].

4. "Modified Statistical Combination of Uncertainties,"

CEN-356(V)-P-A and "System 80.. Inlet Flow Distribution." Supplement 1-P to Enclosure 1-P to LD-82-054, (Methodology for Specification 3.2.4, DNBR and 3.2.5 Axial Shape Index).

5. "Calculative Methods for the CE Large Break LOCA Evaluation Model," CENPD-132, (Methodology for Specification 3.2.1, Linear Heat Rate).

6. "Calculative Methods for the CE Small Break LOCA Evaluation Model," CENPD-137-P. (Methodology for Specification 3.2.1, Linear Heat Rate).

(continued) PALO VERDE UNITS 1.2.3 5.6-4 AMENDMENT NO. 43
-

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.5 Core Operating Limits Report (COLR) (continued)

7. Letter:

O.D. Parr (NRC) to F. M. Stern (CE), dated June 13, 1975 (NRC Staff Review of the Combustion Engineering ECCS Evaluation Model). NRC approval for:

5..5.b.6.

8. Letter: K. Kniel (NRC) to A. E. Scherer (CE), dated September 27, 1977 (Evaluation of Topical Reports CENPD-133, Supplement 3-P and CENPD-137, Supplement 1-P).

NRC approval for 5.6.5.b.6.

9. "Fuel Rod Maximum Allowable Pressure," CEN-372-P-A, (Methodology for Specification 3.2.1, Linear Heat Rate).

10.

Letter: A. C. Thadani (NRC) to A. E. Scherer (CE). dated April 10, 1990, ("Acceptance for Reference CE Topical Report CEN-372-P"). NRC approval for 5.6.5.b.9.

11.

"Arizona Public Service Com pany PWR Reactor Physics Methodology Using CASMO-4/SIMULATE-3," [Methodology for Specifications 3.1.1. Shutdown Margin - Reactor Trip Breakers Open: 3.1.2, Shutdown Margin - Reactor Trip Breakers C]osed; 3.1.4. Moderator Temperature Coefficient; 3.1.7. Regulating CEA Insertion Limits and 3.9.1, Boron Concentration (Mode 6)].

12.

"Technical Manual for the CENTS Code," CE-NPD 282-P-A. Volumes 1-3. [Methodology for Specifications 3.1.2. Shutdown Margin-Reactor Trip Breakers Closed; 3.1.4. Moderator Temperature Coefficient; 3.1.5. CEA Alignment; 3.1.7. Regulating CEA Insertion Limits; 3.1.8, Part Length or Part Strength CEA Insertion Limits and 3.2.3. Azimuthal Power Tilt-Tq].

13.

CENPD-404-P-A. "Implementation of ZIRLOh Cladding Material in CE Nuclear Power Fuel Assembly Designs.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM. transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR. including any mid cycle revisions or supplements.

shall be provided upon issuance for each reload cycle to the NRC. (continued) PALO VERDE UNITS 1.2.3 5.6-5 AMENDMENT NO. 444

ATTACHMENT 3 ASSOCIATED CHANGES TO TECHNICAL SPECIFICATION BASES (for information only)

SDM - Reactor Trip Breakers Open B 3.1.1 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM) - Reactor Trip Breakers Open BASES BACKGROUND The reactivity control systems must be redundant and capable of holding the reactor core subcritical when shutdown under cold conditions. in accordance with GDC 26 (Ref. 1). Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, the SDM defines the degree of subcriticality that would be obtaindimmedi iately following the insertion of all full leint1,"{strength) control element assemblies (CEAs). assuming the single tUCEOfthghest reactivity worth is fully withdrawn with Reactor Trip Breakers open. This reactivity worth is credited in establishing the required SDM. The system design requires that two independent reactivity control systems be provided, and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable CEAs and soluble boric acid in the Reactor Coolant System (RCS). The CEA System provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding acceptable fuel design limits, assuming that the CEA of highest reactivity worth remains fully withdrawn. The soluble boron system during operation and all and maintain the reactor can compensate for fuel depletion xenon burnout reactivity changes, subcritical under cold conditions. During power operation, SDM control is ensured by operating with the shutdown CEAs fully withdrawn and the regul ating CEAs within the limits of LCO 3.1.7. "Regulating Control Element Assembly (CEA) Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration. (continued) PALO VERDE UNITS 1.2.3 B 3.1.1-1 REVISION 0

SDM - Reactor Trip Breakers Open B 3.1.1 BASES (continued) APPLICABLE occurs as a result of the post trip return to power, and SAFETY ANALYSES THERMAL POWER does not violate the Safety Limit (SL) (continued) requirement of SL 2.1.1. In addition to the limiting MSLB transient, the SDM requirement for MODES 3. 4. and 5 must also protect against:

a.

Inadvertent boron dilution;

b. Startup of an inactive reactor coolant pump (RCP); and

c.

CEA ejection. Each of these is discussed below. In the inadvertent boron dilution analysis, the amount of reactivity by which the reactor is subcritical is determined by the reactivity difference between an initial subcritical boron concentration and the corresponding critical boron concentration. The initial subcritical boron concentration assumed in the analysis corresponds to the minimum SDM requirements. These two values (initial and critical boron concentrations), in conjunction with the configuration of the Reactor Coolant System (RCS) and the assumed dilution flow rate, directly affect the results of the analysis. For this reason the event is most limiting at the beginning of core life when critical boron concentrations are highest. The startup of an inactive RCP will not result in a "cold water" criticality, even if the maximum difference in temperature exists between the SG and the core. Although this event was considered in establishing the requirements for SDM, it is not the limiting event with respect to the specification limits. In the analysis of the CEA ejection event, maintaining SDM ensures the reactor remains subcritical following a CEA ejection and, therefore, satisfies the radially averaged enthalpy acceptance criterion considering power redistribution effects. SHUTDOWN MARGIN is the amount by which the core is subcritical, or would be subcritical immediately following a reactor trip, considering a single malfunction resulting in the highestworth CEA failing to insert. With any full +engt strength CEAs not capable of being fully inserted, the withdrar rctivity worth of these CEAs must be accounted for in the determination of SDM. (continued) PALO VERDE UNITS 1.2.3 B 3.1.1-3 REVISION 12

SDM - Reactor Trip Breakers Closed B 3.1.2 B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.2 SHUTDOWN MARGIN (SDM) - Reactor Trip Breakers Closed BASES BACKGROUND The reactivity control systems must be redundant and capable of holding the reactor core subcritical when shut down under cold conditions, in accordance with GDC 26 (Ref. 1). Maintenance of the SDM ensures that postulated reactivity events will not damage the fuel. SDM requirements provide sufficient reactivity margin to ensure that acceptable fuel design limits will not be exceeded for normal shutdown and anticipated operational occurrences (AOOs). As such, SDM defines the degree of subcriticality that would be obtained immediately following the insertion of all full strength control element assemblies (CEAs), assuming the singleCEA of highest reactivity worth is fully withdrawn. The system design requires that two independent reactivity control systems be provided. and that one of these systems be capable of maintaining the core subcritical under cold conditions. These requirements are provided by the use of movable CEAs and soluble boric acid in the Reactor Coolant System (RCS). The CEA System provides the SDM during power operation and is capable of making the core subcritical rapidly enough to prevent exceeding the acceptable fuel design limits, assuming that the CEA of highest reactivity worth remains fully withdrawn. The soluble boron system during operation and all and maintain the reactor can compensate for fuel depletion xenon burnout reactivity changes. subcritical under cold conditions. During power operation. SDM control is ensured by operating with the shutdown CEAs fully withdrawn and the regulating CEAs within the limits of LCO 3.1.7. "Regulating Control Element Assembly (CEA) Insertion Limits." When the unit is in the shutdown and refueling modes, the SDM requirements are met by means of adjustments to the RCS boron concentration. (continued) PALO VERDE UNITS 1,2.3 B 3.1.2-1 REVISION 0

SDM - Reactor Trip Breakers Closed B 3.1.2 BASES (continued) APPLICABLE The startup of an inactive RCP will not result in a SAFETY ANALYSES "cold water" criticality. even if the maximum difference in (continued) temperature exists between the SG and the core. Although this event was considered in establishing the requirements for SDM, it is not the limiting event with respect to the specification limits. In the analysis of the CEA ejection event. SDM alone cannot prevent reactor criticality following a CEA ejection. At temperatures less than 500 F. the Kiirequirement ensures the reactor remains subcritical and, therefore, satisfies the radially averaged enthalpy acceptance criterion considering power redistribution effects. Above 500 F. Doppler reactivity feedback is sufficient to preclude the need for a specific KN-1 requirement. The function of SHUTDOWN MARGIN is to ensure that the reactor remains subcritical following a design basis accident or anticipated operational occurrence. During operation in MODES 1 and 2. with keff greater than or equal to 1.0. the transient insertion limits of Specification 3.1.3.6 ensure that sufficient SHUTDOWN MARGIN is available. SHUTDOWN MARGIN is the amount by which the core is subcritical, or would be subcritical immediately following a reactor trip, considering a single malfunction resulting in the hi ghestworth CEA failing to insert. With any full 4e^iqh strength CEAs not capable of being fully inserted, the withdraw retivity worth of the CEAs must be accounted for in the determination of SDM. SHUTDOWN MARGIN requirements vary throughout the core life as a function of fuel depletion and reactor coolant system (RCS) cold leg temperature (TCold). The most restrictive condition occurs at EOL, with Tcold at no-load operating temperature, and is associated with a postulated steam line break accident and the resulting uncontrolled RCS cooldown. In the analysis of this accident, the specified SHUTDOWN MARGIN is required to control the reactivity transient and ensure that the fuel performance and offsite dose criteria are satisfied. (continued) PALO VERDE UNITS 1.2.3 B 3.1.2-4 REVISION 12

CEA Alignment B 3.1.5 BASES BACKGROUND (continued) The CEAs are arranged into groups that are radially symmetric. Therefore, movement of the CEAs does not introduce radial asymmetries in the core power distribution. The shutdown and regulating CEAs provide the required reactivity worth for immediate reactor shutdown upon a reactor trip. The regulating CEAs also provide reactivity (power level) control during normal operation and transients. Their movement may be automatically controlled by the Reactor Regulating System. Part length or part strength CEAs are not credited in the safety analyses for s u ing own the reactor, as -are the regulating and shutdown groups. The part length(or part stre gh CEAs are used solely for ASI control. The axial position of shutdown and regulating CEAs is indicated by two separate and independent systems, which are the Pulse Counting CEA Position Indication System (described in Ref. 4) and the Reed Switch CEA Position Indication System (described in Ref. 5). The Pulse Counting CEA Position Indicating System indicates CEA position to the actual step. if each CEA moves one step for each command signal. However, if each CEA does not follow the commands, the system will incorrectly reflect the position of the affected CEA(s). This condition may affect the operability of COLSS (refer to Section 3.2. Power Distribution Limits for the applicable actions) and should be detected by the Reed Switch Position Indication System through surveillance or alarm. Although the Reed Switch Position Indication System is less precise that the Pulse Counting CEA Position Indicating System, it is not subject to the same error mechanisms. (continued) PALO VERDE UNITS 1.2.3 B 3.1.5-2 REVISION 12

CEA Alignment B 3.1.5 BASES (continued) APPLICABLE CEA misalignment accidents are analyzed in the safety SAFETY ANALYSES analysis (Ref. 3). The accident analysis defines CEA misoperation as any event. with the exception of sequential group withdrawals, which could result from a single malfunction in the reactivity control systems. For example. CEA misalignment may be caused by a malfunction of the CEDM, CEDMCS, or by operator error. A stuck CEA may be caused by mechanical jamming of the CEA fingers or of the gripper. Inadvertent withdrawal of a single CEA may be caused by an electrical failure in the CEA coil power programmers. A dropped CEA could be caused by an opening of the electrical circuit of the CE ding coil for a full lenghstrengthpart length (or part strength CEA. The acceptance criteria for addressing CEA inoperability or misalignment are that:

a.

There shall be no violations of:

1.

specified acceptable fuel design limits, or

2.

Reactor Coolant System (RCS) pressure boundary integrity: and

b.

The core must remain subcritical after accident transients. Three types of misalignment are distinguished. During movement of a group, one CEA may stop moving while the other CEAs in the group continue. This condition may cause excessive power peaking. The second type of misalignment occurs if one CEA fails to insert upon a reactor trip and remains stuck fully withdrawn. This condition requires an evaluation to determine that sufficient reactivity worth is held in the remaining CEAs to meet the SDM requirement with the maximum worth CEA stuck fully withdrawn. If a CEA is stuck in the fully withdrawn position, its worth is added to the SDM requirement, since the safety analysis does not take two stuck CEAs into account. The third type of misalignment occurs when one CEA drops partially or fully into the reactor core. This event causes an initial power reduction followed by a return towards the original power due to positive reactivity feedback from the negative moderator temperature coefficient. Increased peaking during the power increase may result in erosion of DNB margin. (continued) PALO VERDE UNITS 1.2.3 B 3.1.5-3 REVISION 0

CEA Alignment B 3.1.5 BASES APPLICABLE SAFETY ANALYSES (continued) Analysis considers the case of a single CEA withdrawn approximately 10 inches from a bank inserted to its insertion limit. Satisfying limits on departure from nucleate boiling ratio (DNBR) bounds the situation when a CEA is misaligned from its group by 6.6 inches. The effect of any misoperated CEA on the core power distribution will be assessed by the CEA calculators, and an appropriately augmented power distribution penalty factor will be supplied as input to the core protection calculators (CPCs). As the reactor core responds to the reactivity changes caused by the misoperated CEA and the ensuing reactor coolant and Doppler feedback effects, the CPCs will

• nitiate a low DNBR or high local power density trip signal i pecified acceptable fuel design limits (SAFDLs) are app ched.

&inec T CEA drop incidents result in the most rapid approach to SAFDLs caused by a CEA misoperation. The accident analysis analyzed a single full 4N 4strength) CEA drop, a single part length CEA drop, and a part length LEA subgroup Adrop drop. The most rapid approach to the DNBR SAFDL or the fuel ciar centerline melt SAFDL is caused by a single full enthstrength CEA drop. A part strength CE would cause a sim reactivity response although with less magnitude due to i strength CEAs han more significant rev worth. of a the full ring a activity In the case of the full lethIstrength CEA drop. a prompt decrease in core average power and a distortion in radial power are " itially produced, which when conservatively coupled result in cal power and heat flux increases, and a decrease in DNBR. s the dropped CEA is detected, core power and an appropriately augmented power distribution penalty factor are supplied to the CPCs. For plant operation within the DNBR and local power density (LPD) LCQOs DNBR and LPD trips can normally be avoided on a dropped 4-finger CEA. For a part length or art strength)CEA subgroup drop. a distortion in poweristri J.ttom. an a decrease in core power are roduced. As the position the dropped part length or Dart strength CEA subgroup is an appropriate power distribution penalty actor is supplied to the CPCs, and a reactor trip signal on low DNBR is generated ;For the part length CEA drop, both core average power three dimensional peak to average power density i ase promptly. As the dropped part length CEA is etected, core power and an appropriately augmented power distribution penalty factor are supplied to the CPCs. (continued) PALO VERDE UNITS 1,2,3 B 3.1.5-4 REVISION 0

CEA Alignment B 3.1.5 BASES (continued) APPLICABLE SAFETY ANALYSES CEA alignment satisfies Criteria 2 and 3 of 10 CFR 50.36 (c)(2)(ii) LCO The limits on part length or part strength, shutdown, and regulating CEA alignments ensure that tne assumptions in the safety analysis will remain valid. The requirements on OPERABILITY ensure that upon reactor trip, the CEAs will be available and will be inserted to provide enough negative reactivity to shut down the reactor. The OPERABILITY requirements also ensure that the CEA banks maintain the correct power distribution and CEA alignment. The requirement is to maintain the CEA alignment to within 6.6 inches between any CEA and all other CEAs in its group. Failure to meet the requirements of this LCO may produce unacceptable power peaking factors, DNBR, and LHRs. or unacceptable SDMs, all of which may constitute initial conditions inconsistent with the safety analysis. I APPLICABILITY The requirements on CEA OPERABILITY and alignment are applicable in MODES 1 and 2 because these are the only MODES in which neutron (or fission) power is generated, and the OPERABILITY (e.g., trippability) and alignment of CEAs have the potential to affect the safety of the plant. In MODES 3, 4, 5. and 6, the alignment limits do not apply because the reactor is shut down and not producing fission power. In the shutdown modes, the OPERABILITY of the shutdown and regulating CEAs has the potential to affect the required SDM, but this effect can be compensated for by an increase in the boron concentration of the RCS. See LCO 3.1.2. "SHUTDOWN MARGIN (SDM) - Reactor Trip Breakers Closed," for SDM in MODES 3, 4. and 5, and LCO 3.9.1. "Boron Concentration," for boron concentration requirements during refueling. (continued) PALO VERDE UNITS 1.2,3 B 3.1.5-5 REVISION 7

CEA Alignment B 3.1.5 BASES (continued) ACTIONS A.1 and A.2 A CEA may become misaligned, yet remain trippable. In this condition, the CEA can still perform its required function of adding negative reactivity should a reactor trip be necessary. f one..or more CEAs (regulating, shutdown. &F part length, art strength) are misaligned by 6.6 inches and s 9.9 inches but trippable, or one A misaligned by > 9.9 inches but trippable, continued operation in MODES 1 and 2 may continue, provided, within 1 hour, the power is reduced in accordance with the limits in the COLR, and within 2 hours salianmnt is restored. Regulating and part length or part strength)CEA alignment can be restored by either aligning the mis aigeA ULA to within 6.6 inches of its group or aligning the misaligned CEA's group to within 6.6 inches of the misaligned CEA(s). Shutdown CEA alignment can be restored by aligning the misaligned CEA(s) to within 6.6 inches of its group. Xenon redistribution in the core starts to occur as soon as a CEA becomes misaligned. Reducing THERMAL POWER in accordance with the limits in the COLR ensures acceptable power distributions are maintained (Ref. 3). For small misalignments (< 9.9 inches) of the CEAs, there is:

a. A small effect on the time dependent long term power distributions relative to those used in generating LCOs and limiting safety system settings (LSSS) setpoints;

b. A negligible effect on the available SDM; and

c. A small effect on the ejected CEA worth used in the accident analysis.

With a large CEA misalignment (2 9.9 inches), however, this misalignment would cause distortion of the core power distribution. This distortion may, in turn, have a significant effect on the time dependent, long term power distributions relative to those used in generating LCOs and LSSS setpoints. The effect on the available SDM and the ejected CEA worth used in the accident analysis remain small. Therefore, this condition is limited to the single CEA misalignment, while still allowing 2 hours for recovery. (continued) PALO VERDE UNITS 1.2,3 B 3.1.5-6 REVISION 0

CEA Alignment B 3.1.5 BASES ACTIONS C.1 (continued) known to be untrippable), the unit is required to be brought to MODE 3. By being brought to MODE 3. the unit is brought outside its MODE of applicability. When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should be commenced. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. If a full !em (Ttirength) CEA is untrippable, it is not available for reactivity insertion during a reactor trip. With an untrippable CEA, meeting the insertion limits of LCO 3.1.6. "Shutdown Control Element Assembly (CEA) Insertion Limits," and LCO 3.1.7. "Regulating Control Element Assembly (CEA) Insertion Limits," does not ensure that adequate SDM exists. Therefore, the plant must be shut down in order to evaluate the SDM required boron concentration and power level for critical operation. Continued op is allowed with untrippable part lengthor part strength CEAs if the alignment and insertion limits are met. Continued operation is not allowed with one or more full length CEAs untrippable. This is because these cases are indicative of a loss of SDM and power distribution, and a loss of safety function, respectively. D.1 Continued operation is not allowed in the case of more than one CEA misaligned from any other CEA in its group by > 9.9 inches. For example, two CEAs in a group misaligned from any other CEA in that group by > 9.9. inches, or more than one CEA group that has a least one CEA misaligned from any other CEA in that group by > 9.9 inches. This is indicative of a loss of power distribution and a loss of safety function, respectively. Multiple CEA misalignments should result in automatic protective action. Therefore, with two or more CEAs misaligned more than 9.9 inches, this could result in a situation outside the design basis and immediate action would be required to prevent any potential fuel damage. Immediately opening the reactor trip breakers minimizes these effects. (continued) PALO VERDE UNITS 1,2,3 B 3.1.5-8 REVISION 1

CEA Alignment B 3.1.5 BASES SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that individual CEA positions are within 6.6 inches (indicated reed switch positions) of all other CEAs in the group at a 12 hour Frequency allows the operator to detect a CEA that is beginning to deviate from its expected position. The specified Frequency takes into account other CEA position information that is continuously available to the operator in the control room, so that during actual CEA motion, deviations can immediately be detected. SR 3.1.5.2 OPERABILITY of at least two CEA position indicator channels is required to determine CEA positions. and thereby ensure compliance with the CEA alignment and insertion limits. The CEA full in and full out limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions. SR 3.1.5.3 Verifying each full +e~§*' gth CEA is trippable would require that each CE ped. tinMUe' 1 and 2 tripping each full e strength CEA would result in radial or wer tilts, or oscillati'e I n hrefore individual full +e¶9h ~strength CEAs are exercised every to provide increased confi t all full +en;;thstrength) CEAs continue to be trippable. even if they are not regularly ripped. A movement of 5 inches is adequate to demonstrate motion without c he alignment limit when only one full 4eg st rength CEA is being moved. The 92 day Frequency takes into c eraion other information available to the operator in the control room and other surveillances being performed more frequently. which add to the determination of OPERABILITY of the CEAs (Ref. 3). Between required performances of SR 3.1.5.3, if a CEA(s) is discovered to be immovable but remains trippable and aligned, the CEA is considered to be OPERABLE. At anytime, if a CEA(s) is immovable, a determination of the trippability (OPERABILITY) of that CEA(s) must be made, and appropriate action taken. (continued) PALO VERDE UNITS 1.2.3 B 3.1.5-9 REVISION 0

CEA Alignment B 3.1.5 BASES SURVEILLANCE REQUIREMENTS (continued) SR 3.1.5.4 Performance of a CHANNEL FUNCTIONAL TEST of each reed switch position transmitter channel ensures the channel is OPERABLE and capable of indicating CEA position. Since this test must be performed when the reactor is shut down, an 18 month Frequency to be coincident with refueling outage was selected. Operating experience has shown that these components usually pass this Surveillance when performed at a Frequency of once every 18 months. Furthermore, the Frequency takes into account other factors, which determine the OPERABILITY of the CEA Reed Switch Indication System. These factors include:

a.

Other, more fre uentl performed surveillances that help to verify 8PERABILITY;

b. On-line diagnostics performed automatically by the CPCs. CEACs, and the Plant Computer which include CEA position comparisons and sensor validation; and

c. The CHANNEL CALIBRATIONs for the CPCs (SR 3.3.1.9) and CEACs (SR 3.3.3.4) input channels that are performed at 18 month intervals and is an overlapping test.

SR 3.1.5.5 Verification of full +emgthstrength CEA drop times determines that the maximum CEA drop time permitted is consistent with the assumed drop time used in the safety analysis (Ref. 3). Measuring drop times prior to reactor criticality, after reactor vessel head removal, ensures the reactor internals and CEDM will not interfere with CEA motion or drop time, and that no degradation in these systems has occurred that would adversely affect CEA motion or drop time. Individual CEAs whose drop times are greater than safety analysis assumptions are not OPERABLE. This SR is performed prior to criticality due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power. The 4 second CEA drop time is the maximum time it takes for a fully withdrawn individual full 4enh.strength CEA to reach its 90% insertion position when electrical power is interrupted to the CEA drive mechanism with RCS T. d greater than or equal to 5500F and all reactor coolanf pumps operating. (continued) I PALO VERDE UNITS 1,2,3 B 3.1.5-10 REVISION 5 PALO VERDE UNITS 1,2.3 B 3.1.5-10 REVISION 5

CEA Alignment B 3.1.5 BASES The CEA drop time of full ieM strength)CEAs shall also be demonstrated through measurement prior to reactor criticality for specifically affected individual CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs. REFERENCES

1.

10 CFR 50, Appendix A. GDC 10 and GDC 26.

2.

10 CFR 50.46.

3.

UFSAR, Section 15.4.

4.

UFSAR, Section 7.7.1.3.2.3.

5.

UFSAR, Section 7.5.1.1.4. PALO VERDE UNITS 1,2, 3 B 3.1.5-11 REVISION 12

Regulating CEA Insertion Limits B 3.1.7 BASES BACKGROUND (continued) event of a CEA ejection accident, and the shutdown and regulating bank insertion limits ensure the required SDM is maintained. Operation within the subject LCO limits will prevent fuel c adding failures that would breach the primary fission product barrier and release fission products to the reactor coolant in the event of a LOCA. loss of flow, ejected CEA, or other accident requiring termination by a Reactor Protection System trip function. APPLICABLE SAFETY ANALYSES The fuel cladding must not sustain damage as a result of normal operation (Condition I) and anticipated operational occurrences (Condition II). The accep or the regulating CEA insertion, part length or art strength CEA insertion, ASI. and I LCOs preclude core power dions rom occurring that would violate the following fuel design criteria:

a. During a large break LOCA. the peak cladding temperature must not exceed a limit of 2200 0F, 10 CFR 50.46 (Ref. 2);

b. During CEA misoperation events, there must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition:

c. During an ejected CEA accident, the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 3):

and

d. The CEAs must be capable of shutting down the reactor with a minimum required SDM, with the highest worth CEA stuck fully withdrawn, GDC 26 (Ref. 1).

Regulating CEA position, ASI, and Tq are process variables that together characterize and control the three dimensional power distribution of the reactor core. Fuel cladding damage does not occur when the core is operated outside these LCOs during normal operation. However, fuel cladding damage could result, should an (continued) PALO VERDE UNITS 1,2.3 B 3.1.7-3 REVISION 0

Part Length CEA Insertion Limits B 3.1.8 B 3.1 REACTIVITY CONTROL SYS B 3.1.8 Part Length ntrol Element Assembly (CEA) Insertion Limits BASES BACKGROUND The insertion limits of the part length EAs are initial assumptions in the safety analyses for A misoperation events. The insertion limits directly a fect core power distributions. The applicable criteria r these power distribution design requirements are 10 C 50. Appendix A, GDC 10, "Reactor Design" (Ref. 1), and 10 FR 50.46. "Acceptance Criteria for Emergency Core Co ing Systems for Light Water Nuclear Plants" (Ref. 2). Limi s on part length, CEA insertion have been established, and al CEA positions are monitored and controlled during power o ation to ensure that the power distribution defined the design power peaking limits is preserved. l rt The part length are used for axial power sha control of the reactor. The positions of the part lengt CEAs are manually controlled. They are capable of changing reactivity very quickly (compared to borating or diluting). The power density at any point in the core must be limited to maintain specified acceptable fuel design limits. including limits that preserve the criteria specified in 10 CFR 50.46 (Ref. 2). Together, LCO 3.1.7, "Regulating Control Element Assembly (CEA) Insertion Limits"; LCO 3.1.8: LCO 3.2.4, "Departure From Nucleate Boiling Ratio (DNBR)": and LCO 3.2.5. "AXIAL SHAPE INDEX (ASI)," provide limits on control component operation and on monitored process variables to ensure the core operates within the linear heat rate (LHR) (LCO 3.2.1. "Linear Heat Rate (LHR)"): planar eaking factor (F.) (LCO 3.2.2. "Planar Radial Peaking Pactors (Fj')"); and LCO 3.2.4 limits in the COLR. Operation within the limits given in the COLR prevents power eaks that would exceed the loss of coolant accident (LOCA) limits derived by the Emergency Core Cooling Systems analysis. Operation within the Fy and departure from nucleate boiling (DNB) limits given in the COLR prevents DNB during a loss of forced reactor coolant flow accident. (continued) PALO VERDE UNITS 1.2.3 B 3.1.8-1 REVISION 0

Part LengtI CEA Insertion Limits Ir B 3.1.8 for Part StrengthI BASES BACKGROUND (continued) The establishment of limiting safety system settings and LCOs requires that the expected long and short term behavior of the radial peaking factors be determined. The long term behavior relates to the variation of the steady state radial peaking factors with core burnup; it is affected by the amount of CEA insertion assumed, the portion of a burnup cycle over which such insertion is assumed, and the expected power level variation throughout the cycle. The short term behavior relates to transient perturbations to the steady state radial peaks due to radial xenon redistribution. The magnitudes of such perturbations depend upon the expected use of the CEAs during anticipated power reductions and load maneuvering. Analyses are performed, based on the expected mode of operation of the Nuclear Steam Supply System (base loaded, maneuvering, etc.). From these analyses, CEA insertions are determined, and a consistent set of radial peaking factors are defined. The long term (steady state) and short term insertion limits are determined, based upon the assumed mode of operation used in the analyses: they provide a means of Preserving the assumptions on CEA insertions used. Te long and short term insertion limits of LCO 3.1.8 are specified for the plant, which has been designed primarily for base loaded operation, but has the ability to accommodate a limited amount of load maneuvering. APPLICABLE The fuel cladding must not sustain damage as a result of SAFETY ANALYSES normal operation (Condition I) and anticipated operational occurrences (Condition II). The regulating CEA insertion, part lengtNCEA insertion, ASI, and Tq LCOs preclude core poweg.*titributions from occurring that would violate the lor part stren owing fuel design criteria:

a. During a large break LOCA, the peak cladding temperature must not exceed 2200OF (Ref. 2);

b.

During a loss of forced reactor coolant flow accident, there must be at least a 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition;

c.

During an ejected CEA accident, the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 3); and (continued) PALO VERDE UNITS 1.2.3 B 3.1.8-2 REVISION 0

Part Length CEA Insertion Limits 'I B 3.1.8 lor Part Stren;th BASES APPLICABLE

d.

The CEAs must be capable of shutting down the reactor SAFETY ANALYSES with a minimum required SDM, with the highest worth (continued) CEA stuck fully withdrawn, GDC 26 (Ref. 1). Regulating CEA position, part length CEA position. ASI. and Tq are process variables that togeth r characterize and control the three dimensional power istribution of the reactor core. Fuel cladding damage does not occur when the core is operated outside these LCOs during ormal operation. However, fuel cladding damage coul result, should an accident occur with simultaneous v lation of one or more of these LCOs. Changes in the power istribution can cause increased power peaking and corres nding increased local LHRs. nith - The part length serti Vm s sati fy Criterion 2 of 10 CFR 50.36 (c)(2)(ii).J~e~~ lengtACEAs are required due to the potential pkigfQtr violations that could occur if part lengthbCEs dinsertion limits. LCO The limits on part len h4EA insertion, as defined in the COLR, must be maintai ed because they serve the function of preserving power di ribution. APPLICABILITY The part length nsertion limits shall be maintained with the reactor in MODES 1 and 2. These limits must be maintained, since they preserve the assumed power distribution. Applicability in lower MODES is not required. since the power distribution assumptions would not be exceeded in these MODES. (continued) PALO VERDE UNITS 1,2.3 B 3.1.8-3 REVISION 0

Part Lengt T EA Bor Part StrengthI BASES (continued); Insertion Limits B 3.1.8 ACTIONS 7x/

1) Transient insertion limits;

2) Between the long term (steady-state) insertion limit and the transient Insertion limit for:

a) 7 or more effective full power days (EFPD) out of any 30 EFPD period; b) 14 EFPD or more out of any 365 EFPD period. A.1, A.2 and B.1 If the part length groups are inserted beyond theffollowing limits state) incartian limit and the tram3iemt limt f-7 rmoe Q atterns b g n o eue ra e assume Tor long utnu owedeto co nt n e e o this limit, the peaking factors assumed as initial conditions in the accident analysis may be invalidated (Ref. 4). Restoring the CEAs to within limits or reducing THERMAL POWER to that fraction of RTP that is allowed by CEA group position, using the limits specified in the COLR, ensures that acceptable peaking factors are maintained. Since these effects are cumulative. actions are provided to limit the total time the part lengthLZEAs can be out of limits in any 30 EFPD or 365 EFPD per

d.

Since the cumulative out of limit times are in d an dditional Completion Time of 2 hours is reasonable or estoring the part length to wit the allowed by ts. C.1 When a Required Acti on cannot be completed wi thin the required Completion Time, a controlled shutdown should commence. A Completion Time of 6 hours is reasonable, based on operating experience, for reaching Mode 3 from full power conditions in an orderly manner and without challenging plant systems. (continued) PALO VERDE UNITS 1.2.3 B 3.1.8-4 REVISION 0

Part Length CEA Insertion Limits V B 3.1.8 lo attrenh BASES (continued) SURVEILLANCE REQUIREMENTS SR 3.1.8.1 Verification of each part length group position every 12 hours is sufficient to detect CEA positions that may approach the limits, and provide the operator with time to undertake the Required Action(s). should insertion limits be found to be exceeded. The 12 hour frequency also takes into account the indication provided by the power dependent insertion limit alarm circuit and other information about CEA group positions available to the operator in the control room. REFERENCES

1.

10 CFR 50, Appendix A. GDC 10 and GDC 26.

2.

10 CFR 50.46.

3.

Regulatory Guide 1.77, Rev. 0, May 1974.

4.

UFSAR. Section 15.4. PALO VERDE UNITS 1.2.3 B 3.1.8-5 REVISION 0

STE-SDM B 3.1.9 BASES (continued) ACTIONS A.1 With any CEA not fully inserted and less than the minimum required reactivity equivalent available for insertion, or with all CEAs inserted and the reactor subcritical by less than the reactivity equivalent of the highest worth withdrawn CEA. restoration of the minimum shutdown reactivity requirements must be accomplished by increasing the RCS boron concentration. The required Completion Time of 15 minutes for initiating boration allows the operator sufficient time to align the valves and start the boric acid pumps and is consistent with the Completion Time of LCO 3.1.2. In the determination of the required combination of boration flow rate and boron concentration, there is no unique requirement that must be satisfied. Since it is imperative to raise the boron concentration of the RCS as soon as possible, the boron concentration should be a highly concentrated solution, such as that normally found in the refueling water tank. The operator should borate with the best source available for the plant conditions. In determining the boration flow rate the time in core life must be considered. For instance, the most difficult time in core life to increase the RCS boron concentration is at the beginning of cycle, when boron concentration may approach or exceed 2000 ppm. Assuming that a value of 1% Ak/k must be recovered and a boration flow rate of 26 gpm, it is possible to increase the boron concentration of the RCS by 100 ppm in approximately 35 minutes with a 4000 ppm source. If a boron worth of 10 pcm/ppm is assumed, this combination of parameters will increase the SDM by 1% Ak/k. These boration parameters of 26 gpm and 4000 ppm represent typical values and are provided for the purpose of offering a specific example. SURVEILLANCE SR 3.1.9.1 lor art streath REQUIREMENTS Verification of the positiom of each partiaNly or fully withdrawn full ie1gti¶ strength ep part lengtkiCEA is necessary to ensure that the minimum negative reactivity requirements for insertion on a trip are preserved. A 2 hour Frequency is sufficient for the operator to verify that each CEA position is within the acceptance criteria. (continued) PALO VERDE UNITS 1.2.3 B 3.1.9-5 REVISION 7

STE-MODES 1 and 2 B 3.1.10 BASES BACKGROUND (continued) PHYSICS TESTS procedures are written and approved in accordance with established formats. The procedures include all information necessary to permit a detailed execution of testing required to ensure that design intent is met. PHYSICS TESTS are performed in accordance with these procedures and test results are approved prior to continued power escalation and long term power operation. Examples of PHYSICS TESTS include determination of critical boron concentration, CEA group worth, reactivity coefficients, flux symmetry, and core power distribution. APPLICABLE SAFETY ANALYSES It is acceptable to suspend certain LCOs for PHYSICS TESTS because fuel damage criteria are not exceeded. Even if an accident occurs during PHYSICS TESTS with one or more LCOs suspended, fuel damage criteria are preserved because the limits on power distribution and shutdown capability are maintained during PHYSICS TESTS. Reference 5 defines requirements for initial testing of the facility, including PHYSICS TESTS. Requirements for reload fuel cycle PHYSICS TESTS are defined in ANSI/ANS-19.6.1-1985 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits of all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. As long as the linear heat rate (LHR) remains within its limit, fuel design criteria are preserved. In this test, the following LCOs are suspended: LCO 3.1.4, LCO 3.1.5. LCO 3.1.6, LCO 3.1.7, LCO 3.1.8, "Moderator Temperature Coefficient (MTC)"; "Control Element Assembly (CEA) Alignment"; "Shutdown Control Element Assembly (CEA) Insertion Limits": "Regulating Control lement Assembly (CEA) Insertion Li mits (FL". "Part Length~for Part Strength) Control Eleme Insertion Li-mi-ts ; "Planar Radial Peaking Factors"; "AZIMUTHAL POWER TILT (T )"; "AXIAL SHAPE INDEX (ASI) ; and "Control Element Assembly Calculators (CEAC nt Assembly (CEA) LCO LCO LCO LCO 3.2.2, 3.2.3, 3.2.5, 3.3.3, s)". (continued) PALO VERDE UNITS 1,2.3 B 3.1.10-2 REVISION 0

STE-Reactivity Coefficient Testing B 3.1.11 BASES BACKGROUND (continued) The PHYSICS TESTS requirements for reload fuel cycles ensure that the operating characteristics of the core are consistent with the design predictions and that the core can be operated as designed (Ref. 4). PHYSICS TESTS procedures are written and approved in accordance with established formats. The procedures include all information necessary to permit a detailed execution of testing required to ensure that design intent is met. PHYSICS TESTS are performed in accordance with these procedures and test results are approved prior to continued power escalation and long term power operation. Examples of PHYSICS TESTS include determination of critical boron concentration, CEA group worth, reactivity coefficients, flux symmetry, and core power distribution. APPLICABLE SAFETY ANALYSES It is acceptable to suspend certain LCOs for PHYSICS TESTS because fuel damage criteria are not exceeded. Even if an accident occurs during PHYSICS TESTS with one or more LCOs suspended, fuel damage criteria are preserved because the limits on power distribution and shutdown capability are maintained during PHYSICS TESTS. Reference 5 defines requirements for initial testing of the facility, including PHYSICS TESTS. Requirements for reload fuel cycle PHYSICS TESTS are defined in ANSI/ANS-19.6.1-1985 (Ref. 4). Although these PHYSICS TESTS are generally accomplished within the limits of all LCOs, conditions may occur when one or more LCOs must be suspended to make completion of PHYSICS TESTS possible or practical. This is acceptable as long as the fuel design criteria are not violated. As long as the linear heat rate (LHR) and DNBR remain within its limits, fuel design criteria are preserved. In this test, the following LCOs are suspended: LCO 3.1.7. LCO 3.1.8, LCO 3.4.1. "Regulating Control Element Assembly (CEA) Insertion Limits": "Part Length or Part Strength)Control Element Assembly (CEA) Insertion Limitst; and "RCS Pressure, Temperature, and Flow Limits" (LCO 3.4.1.b, RCS Cold Leg Temperature only). (continued) PALO VERDE UNITS 1,2.3 B 3.1.11-2 REVISION 0

STE-Reactivity Coefficient Testing B 3.1.11 BASES APPLICABLE SAFETY ANALYSIS (continued) PHYSICS TESTS meet the criteria for inclusion in the Technical Specifications, since the component and process variable LCOs suspended during PHYSICS TESTS meet Criteria 1, 2. and 3 of 10 CFR 50.36 (c)(2)(ii). lor Part Strenaih LCO This LCO permits Part Length 4EAs and Regulating CEAs to be positioned outside of their normal group heights and insertion limits, and RCS cold leg temperature to be outside its limits during the performance of PHYSICS TESTS. These PHYSICS TESTS are required to determine the isothermal temperature coefficient (ITC), MTC, and power coefficient. The requirements of LCO 3.1.7. LCO 3.1.8. and LCO 3.4.1, (for RCS cold leg temperature only) may be suspended during the performance of PHYSICS TESTS provided COLSS is in service. APPLICABILITY This LCO is applicable in MODE 1 with THERMAL POWER > 20% RTP because the reactor must be critical at THERMAL POWER levels > 20% RTP to perform the PHYSICS TESTS described in the LCO section. ACTIONS A. 1 With the LHR or DNBR outside the limits specified in the COLR. adequate safety margin is not assured and power must be reduced to restore LHR and DNBR to within limits. The required Completion Time of 15 minutes ensure prompt action is taken to restore LHR or DNBR to within limits. (continued) PALO VERDE UNITS 1.2.3 B 3.1.11-4 REVISION 0

LHR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 Linear Heat Rate (LHR) BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to the initial values assumed in the accident analyses. Operation within the limits imposed by this LCO limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and release fission products to the reactor coolant in the event of a Loss Of Coolant Accident (LOCA), ejected Control Element Assembly (CEA) accident, or other postulated accident requiring termination by a Reactor Protection System (RPS) trip function. This LCO limits the damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable bounding conditions at the onset of a transient. [or Dart sth Methods of controlling the power distribu ion include:

a. Using full strengthf lei part length EAs to alter the axial power iistrl~u ion;

b. Decreasing CEA insertion by boration, thereby improving the radial power distribution; and

c. Correcting off optimum conditions (e.g.. a CEA drop or misoperation of the unit) that cause margin degradati ons.

The core power distribution is controlled so that, in conjunction with other core operating parameters (e.g., CEA insertion and alignment limits), the power distribution does not result in violation of this LCO. The limiting safety system settings and this LCO are based on the accident analyses (Refs. 1 and 2), so that specified acceptable fuel design limits are not exceeded as a result of Anticipated Operational Occurrences (AOOs), and the limits of acceptable consequences are not exceeded for other postulated accidents. Limiting power distribution skewing over time also minimizes xenon distribution skewing, which is a significant factor in controlling the axial power distribution. (continued) PALO VERDE UNITS 123 .2. B 3.2.1-1 REVISION 0

LHR B 3.2.1 BASES BACKGROUND (continued) In addition to the monitoring performed by the COLSS, the RPS (via the CPCs) continually infers the core power distribution and thermal margins by processing reactor coolant data, signals from excore neutron flux detectors. and input from redundant reed switch assemblies that indicate CEA positions. In this case, the CPCs assume a minimum core power of 20% RTP because the power range excore neutron flux detecting system is inaccurate below this power level. If power distribution or other parameters are perturbed as a result of an AOD, the high LPD or low DNBR trips in the RPS initiate a reactor trip prior to exceeding fuel design limits. The LHR ASI F initiaY and DNBR algorithms are valid within the limits on and Tq. These limits are obtained directly from core or reload analysis. APPLICABLE SAFETY ANALYSES The fuel cladding must not sustain damage as a result of normal operation or AOOs (Ref. 4). The power distribution and CEA insertion and alignment LCOs prevent core power distributions from reaching levels that violate the following fuel design criteria:

a. During a LOCA. peak cladding temperature must not exceed 22001F (Ref. 5):

b. During a loss of flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 4);

c. During an ejected CEA accident, the fission energy input to the fuel must not exceed 280 al/gm (Ref. 6);

and Aor Part strength I

d. The control rods (excluding part lengthl4ods) must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 7).

(continued) PALO VERDE UNITS 1,2.3 B 3.2.1-3 REVISION 0

F B32 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.2 Planar Radial Peaking Factors (F.) BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to the initial values assumed in the accident analyses. Operation within the limits imposed by this LCO either limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and release fission products to the reactor coolant in the event of a Loss Of Coolant Accident (LOCA), loss of flow accident. ejected Control Element Assembly (CEA) accident, or other postulated accident requiring termination by a Reactor Protection System (RPS) trip function. This LCO limits damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at the onset of a transient. For part strength Methods of controlling the power distrib/tion include:

a. Using full{(

ngthW part lengt CEAs to alter the axial power distribution;

b. Decreasing CEA insertion by boration. thereby improving the radial power distribution; and

c. Correcting off optimum conditions (e.g., a CEA drop or misoperation of the unit) that cause margin degradations.

The core power distribution is controlled so that, in conjunction with other core operating parameters (CEA insertion and alignment limits), the power distribution does not result in violation of this LCO. Limiting safety system settings and this LCO are based on the accident analyses (Refs. 1 and 2). so that specified acceptable fuel design limits are not exceeded as a result of Anticipated Operational Occurrences (AOOs), and the limits of acceptable consequences are not exceeded for other postulated accidents. Limiting power distribution skewing over time also minimizes xenon distribution skewing, which is a significant factor in controlling axial power distribution. Power distribution is a product of multiple parameters, various combinations of (continued) PALO VERDE UNITS 12,3 .2. B 3.2.2-1 REVISION 0

F BASES APPLICABLE

b.

During a loss of flow accident, there must be at least SAFETY ANALYSES 95% probability at the 95% confidence level (the (continued) 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 4):

c. During an ejected CEA accident. the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 6);

and lor art strenathI

d. The control rods (excluding part length4 ods) must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 7).

The power density at any point in the core must be limited to maintain the fuel design criteria (Refs. 4 and 5). This result is accomplished by maintaining the power distribution and reactor cool ant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses (Ref. 1) with due regard for the correlations between measured quantities, the power distribution, and the uncertainties in the determination of power distribution. Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate (LHGR) so that the peak cladding temperature does not exceed 22000F (Ref. 5). Peak cladding temperatures exceeding 22000F cause severe cladding failure by oxidation due to a Zircaloy water reaction. The LCOs governing LHR. ASI. CEAs. and RCS ensure that these criteria are met as long as the core is operated within the ASI and F limits specified in the COLR. and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core. Operation within the limits for these variables ensures that their actual values are within the ranges used in the accident analyses (Ref. 1). Fuel cladding damage does not occur because of conditions outside the limits of these LCOs for ASI. Fe, and Tq during normal operation. However, fuel cladding damage results if an accident occurs from initial conditions outside the limits of these LCOs. This potential for fuel cladding damage exists because changes in the power distribution can (continued) PALO VERDE UNITS 1,2.3 B 3.2.2-4 REVISION 0

B 3.2.3 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.3 AZIMUTHAL POWER TILT (Tq) BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to the initial values assumed in the accident analyses. Operation within the limits imposed by this LCO either limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and release fission products to the reactor coolant in the event of a Loss Of Cool ant Accident (LOCA). loss of flow accident, ejected Control Element Assembly (CEA) accident, or other postulated accident requiring termination by a Reactor Protection System (RPS) trip function. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at the onset of a transient. rartstrennthI Methods of controlling the power distribulion include:

a. Using full',Fregtfh ).e. part length EAs to alter the axial power distriutlon;

b.

Decreasing CEA insertion by boration, thereby improving the radial power distribution; and

c.

Correcting off optimum conditions, (e.g.. a CEA-drop or misoperation of the unit) that cause margin degradations. The core power distribution is controlled so that. in conjunction with other core operating parameters (e.g., CEA insertion and alignment limits), the power distribution does not result in violation of this LCO. The limiting safety system settings and this LCO are based on the accident analyses (Refs. 1 and 2). so that specified acceptable fuel design limits are not exceeded as a result of Anticipated Operational Occurrences (AOOs) and the limits of acceptable consequences are not exceeded for other postulated accidents. Limiting power distribution skewing over time also minimizes xenon distribution skewing, which is a significant factor in controlling axial power distribution. (continued) PALO VERDE UNITS 1.2.3 B 3.2.3-1 REVISION 0

Tq B 3.2.3 BASES APPLICABLE

b.

During a loss of flow accident, there must be at least SAFETY ANALYSES 95% probability at the 95% confidence level (the (continued) 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 4);

c.

During a CEA ejection accident, the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 6); and g Dart strength

d.

The control rods (excluding part lengtH4 ods) must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 7). The power density at any point in the core must be limited to maintain the fuel design criteria (Ref. 1). This result is accomplished by maintaining the power distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analysis (Ref. 2) with due regard for the correlations between measured quantities. the power distribution. and uncertainties in the determination of power distribution. Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate (LHGR) so that the peak cladding temperature does not exceed 22000F (Ref. 1). Peak cladding temperatures exceeding 22000F cause severe cladding failure by oxidation due to a Zircaloy water reaction. The LCOs governing LHR. ASI, CEAs. and RCS ensure that these criteria are met as long as the core is operated within the ASI and F limits specified in the COLR. and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core. Operation within the limits of these variables ensures that their actual values are within the range used in the accident analyses (Ref. 1). (continued) PALO VERDE UNITS 1.2.3 B 3.2.3-4 REVISION 0

DNBR B 3.2.4 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.4 Departure from Nucleate Boiling Ratio (DNBR) BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to the initial value assumed in the accident analyses. Specifically, operation within the limits imposed by this LCO either limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and release fission products to the reactor coolant in the event of a Loss Of Coolant Accident (LOCA). loss of flow accident, ejected Control Element Assembly (CEA) accident, or other postulated accident requiring termination by a Reactor Protection System (RPS) trip function. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at the onset of a transient. r ot h I Methods of controlling the power distribu ion include:

a. Using full, Er t)h, part length CEAs to alter the axial power dist on;

b. Decreasing CEA insertion by boration, thereby improving the radial power distribution; and

c. Correcting off optimum conditions (e.g., a CEA drop or misoperation of the unit) that cause margin degradations.

The core power distribution is controlled so that. in conjunction with other core operating parameters (e.g., CEA insertion and alignment limits), the power distribution does not result in violation of this LCO. The limiting safety system settings and this LCO are based on the accident analysis (Refs. 1 and 2). so that specified acceptable fuel design limits are not exceeded as a result of Anticipated Operational Occurrences (AOOs) and the limits of acceptable consequences are not exceeded for other postulated accidents. Limiting power distribution skewing over time also minimizes the xenon distribution skewing, which is a significant factor in controlling axial power distribution. (continued) PALO VERDE UNITS 1.2,3 B 3.2.4-1 REVISION 0

DNBR B 3.2.4 BASES APPLICABLE

b.

During a loss of flow accident, there must be at least SAFETY ANALYSES 95% probability at the 95% confidence level (the (continued) 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 4):

c.

During an ejected CEA accident, the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 6); and or part strengthI

d.

The control rods (excluding part length' ods) must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 7). The power density at any point in the core must be limited to maintain the fuel design criteria (Ref. 4). This is accomplished by maintaining the power distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses (Ref. 1) with due regard for the correlations between measured quantities. the power distribution, and uncertainties in the determination of power distribution. Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate (LHGR) so that the peak cladding temperature does not exceed 2200'F (Ref. 4). Peak cladding temperatures exceeding 22000F may cause severe cladding failure by oxidation due to a Zircaloy water reaction. The LCOs governing LHR, ASI, CEAs, and RCS ensure that these criteria are met as long as the core is operated within the ASI and F limits specified in the COLR. and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core. Operation within the limits for these variables ensures that their actual values are within the range used in the accident analyses (Ref. 1). (continued) PALO VERDE UNITS 1.2.3 B 3.2.4-4 REVISION 0

ASI B 3.2.5 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.5 AXIAL SHAPE INDEX (ASI) BASES BACKGROUND The purpose of this LCO is to limit the core power distribution to the initial values assumed in the accident analysis. Operation within the limits imposed by this LCO either limits or prevents potential fuel cladding failures that could breach the primary fission product barrier and release fission products to the reactor coolant in the event of a Loss Of Cool ant Accident (LOCA). loss of flow accident, ejected Control Element Assembly (CEA) accident, or other ostulated accident requiring termination by a Reactor Protection System (RPS) trip function. This LCO limits the amount of damage to the fuel cladding during an accident by ensuring that the plant is operating within acceptable conditions at the onset of a transient. Methods of controlling the axial power di Iribution include:

a. Using full rength.).eia part lengthCEAs to alter the axial power distrT Utl Tn;

b. Decreasing CEA insertion by boration, thereby improving the axial power distribution; and

c. Correcting off optimum conditions (e.g., a CEA drop or misoperation of the unit) that cause margin degradations.

The core power distribution is controlled so that, in conjunction with other core operating parameters (CEA insertion and alignment limits), the power distribution does not result in violation of this LCO. The limiting safety system settings are based on the accident analyses (Refs. 1 and 2). so that specified acceptable fuel design limits are not exceeded as a result of Anticipated Operational Occurrences (AOOs) and the limits of acceptable consequences are not exceeded for other postulated accidents. Limiting power distribution skewing over time also minimizes xenon distribution skewing, which is a significant factor in controlling axial power distribution. (continued) PALO VERDE UNITS 1,2,3 B 3.2.5-1 REVISION 0

ASI B 3.2.5 BASES APPLICABLE SAFETY ANALYSES (continued)

b. During a loss of flow accident, there must be at least 95% probability at the 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 4);

c. During an ejected CEA accident, the on energy input to the fuel must not exceed 280 caf/gm (Ref. 6);

d. The control rods (excluding part lengtr rods) must be capable of shutting down the reactor with a minimum required SDM with the highest worth control rod stuck fully withdrawn (Ref. 7).

The power density at any point in the core must be limited to maintain the fuel design criteria (Refs. 4 and 5). This is accomplished by maintaining the power distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses (Ref. 1) with due regard for the correlations among measured quantities, the power distribution, and uncertainties in the determination of power distribution. Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate (LHGR) so that the peak cladding temperature does not exceed 22000F (Ref. 5). Peak cladding temperatures exceeding 22000F may cause severe cladding failure by oxidation due to a Zircaloy water reaction. The LCOs governing LHR, ASI, and RCS ensure that these criteria are met as long as the core is operated within the ASI and F limits specified in the COLR, and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core. Operation within the limits for these variables ensures that their actual values are within the range used in the accident analysis (Ref. 1). Fuel cladding damage does not occur from conditions outside these LCOs during normal operation. However, fuel cladding damage results when an accident occurs due to initial conditions outside the limits of these LCOs. This potential for fuel cladding damage exists because changes in the power distribution can cause increased power peaking and correspondingly increased local LARs. (continued) PALO VERDE UNITS 1.3 .2. B 3.2.5-4 REVISION 0

ASI B 3.2.5 BASES ACTIONS A.1 The ASI limits specified in the COLR ensure that the LOCA and loss of flow accident criteria assumed in the accident analyses remain valid. If the ASI exceeds its limit, a Completion Time of 2 hours is allowed to restore the ASI to within its specified limit. This duration gives the operator sufficient time to reposition the regulating or part lepgtbeCEAs to reduce the axial power imbalance. The 1or Dart stren thMLapSHtDe of any potential xenon oscillation is significantly reduced if the condition is not allowed to persist for more than 2 hours. B.1 If the ASI is not restored to within its specified limits within the required Completion Time, the reactor continues to operate with an axial power distribution mismatch. Continued operation in this configuration induces an axial xenon oscillation, and results in increased LHGRs when the xenon redistributes. Reducing thermal power to

• 20% RTP reduces the maximum LHR to a value that does not exceed the fuel design limits if a design basis event occurs. The allowed Completion Time of 4 hours is reasonable, based on operating experience, to reduce power in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.5.1 REQUIREMENTS The ASI can be monitored by both the incore (COLSS) and excore (CPC) neutron detector systems. The COLSS provides the operator with an alarm if an ASI limit is approached. Verification of the ASI every 12 hours ensures that the operator is aware of changes in the ASI as they develop. A 12 hour Frequency for this Surveillance is acceptable because the mechanisms that affect the ASI, such as xenon redistribution or CEA drive mechanism malfunctions, cause slow changes in the ASI, which can be discovered before the limits are exceeded. (continued) PALO VERDE UNITS 1,2,3 B 3.2.5-6 REVISION 0

RPS Instrumentation - Operating B 3.3.1 BASES APPLICABLE Design Basis Definition (continued) SAFETY ANALYSES 12, 13. Reactor Coolant Flow - Low The Reactor Coolant Flow Steam Generator #1-Low and Reactor Coolant Flow Steam Generator #2-Low trips provide protection against an RCP Sheared Shaft Event. A trip is initiated when the pressure differential across the primary side of either steam generator decreases below a variable setpoint. This variable setpoint stays below the pressure differential by a preset value called the step function, unless limited by a preset maximum decreasing rate determined by the Ramp Function, or a set minimum value determined by the Floor Function. The setpoints ensure that a reactor trip occurs to limit fuel failure and ensure offsite doses are within 10 CFR 100 guidelines.

14.

Local Power Density - High The CPCs perform the calculations required to derive the DNBR and LPD parameters and their associated RPS trips. The DNBR - Low and LPD - High trips provide plant protection during the following AOOs and assist the PSF systems in the mitigation of the following accidents. The LPD - High trip provides protection against fuel centerline melting due to the occurrence of excessive local power density peaks during the following AO0s: Decrease in Feedwater Temperature; Increase in Feedwater Flow: Increased Main Steam Flow (not due to the steam line rupture) Without Turbine Trip; r Units that have Uncontrolled CEA Withdrawal From Low Power; partlengthCEAs) Uncontrolled CEA Withdrawal at Power: and CEA Misoperation; Single Part Length CEA Drops> For the events listed above (except CEA Misoperation; Single Part Length CEA Drop). DNBR - Low will trip the reactor first, since DNB would occur before fuel centerline melting would occur. (continued) PALO VERDE UNITS 1.2.3 B 3.3.1-15 REVISION 6

RPS Instrumentation - Operating B 3.3.1 BASES APPLICABLE Design Basis Definition (continued) SAFETY ANALYSES

15.

Departure from Nucleate Boiling Ratio (DNBR) - Low The CPCs perform the calculations required to derive the DNBR and LPD parameters and their associated RPS trips. The DNBR - Low and LPD - High trips provide plant protection during the following AOOs and assist the PSF systems in the mitigation of the following accidents. The DNBR - Low trip provides protection against core damage due to the occurrence of locally saturated conditions in the limiting (hot) channel during the following events and is the primary reactor trip (trips the reactor first) for these events: Decrease in Feedwater Temperature; Increase in Feedwater Flow; Increased Main Steam Flow (not due to steam line rupture) Without Turbine Trip: Increased Main Steam Flow (not due to steam line rupture) With a Concurrent Single Failure of an Active Component; Steam Line Break With Concurrent Loss of Offsite AC Power; Loss of Normal AC Power; Partial Loss of Forced Reactor Coolant Flow; Total Loss of Forced Reactor Coolant Flow; Single Reactor Coolant Pump (RCP) Shaft Seizure; Uncontrolled CEA Withdrawal From Low Power; Uncontrolled CEA Withdrawal at Power: orPartStren h CEA Misoperation; Fulil Ltlfr CEA CEA Misoperation; Part LengthEA Subgroup Drop; Primary Sample or Instrument Line Break; and Steam Generator Tube Rupture. In the above list, only the steam generator tube rupture, the RCP shaft seizure, and the sample or instrument line break are accidents. The rest are AO0s. (continued) PALO VERDE UNITS 1.2,3 B 3.3.1-16 REVISION 0

RCS Pressure. Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued) The transients analyzed for include loss of coolant flow events and dropped or stuck Control Element Assembly (CEA) events. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.7, "Regulating CEA Insertion Limits": LCO 3.1.8. "Part Length WEA Insertion Limits"; LCO 3.2.3. "AZIMUTHAL POWER TILT (Tq \\ and LCO 3.2.5. "AXIAL SHAPE INDEX (ASI). lor Part Stre~nZh The RCS DNB limits satisfy Criterion 2 of 10 CFR 50.56(c)(2)(ii) I LCO This LCO specifies limits on the monitored process variables - RCS pressurizer pressure. RCS cold leg temperature, and RCS total flow rate - to ensure that the core operates within the limits assumed for the plant safety analyses. Operating within these limits will result in meeting the DNBR criterion in the event of a DNB limited transient. The LCO numerical value for minimum flow rate is given for the measurement location but has not been adjusted for instrument error. Plant specific limits of instrument error are established by the plant staff to meet the operational requirements of minimum flow rate. APPLICABILITY In MODE 1 for RCS flow rate. MODES 1 and 2 for RCS pressurizer pressure, Mode 1 for RCS cold leg temperature, and MODE 2 with Keff Ž 1 for RCS cold leg temperature. the limits must be maintained during steady state operation in order to ensure that DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough so that DNBR is not a concern. (continued) PALO VERDE UNITS 1.2.3 B 3.4.1-2 REVISION 7}}