ML030770369

Plant Modification Package 02-029, Aux Feed Mini Recirc Upgrade/Remove AF-117 Intervals.

ML030770369
Person / Time
Site:	Point Beach
Issue date:	09/06/2002
From:	Chapman R Nuclear Management Co
To:	Office of Nuclear Reactor Regulation
References
FOIA/PA-2003-0094
Download: ML030770369 (65)

Text

Point Beach Nuclear Plant PLANT MODIFICATION/MINOR PLANT CHANGE NO.:t 02-029 PLANT CHANGE INITIATION WO#

INITIATION

Title:

AUX FEED MINI RECIRC SAFETY UPGRADE I REMOVE AF-117 INTERNALS 0D QA [5 AQ [5 Non-QA 0 SR [I Non-SR Unit 1 5 Unit 2 5 Common 0 CHAMPS System Code: AF EWR: CR:

Project Objectives: UPGRADE THE SAFETY FUNCTION OF THE AUXILIARY FEEDWATER MINI.

RECIRC AOVS TO HAVE A SAFETY FUNCTION TO OPEN Proposed Scope: REMOVE INTERNALS OF AF-117 TO PREVENT COMMON MODE FAILURE OF AUX FEED MINI IRECIRC LINE Initiated By: Rob Chapman Date: 8/20/2002 CHANGE DETERMINATION YES NO Is the change Temporary? x If YES go to NP 7.3.1 Temp Mod Is this a Setpoint Only change? x If YES go to NP 7.3.8 Setpoints Is this an Equivalent change? ___ x If YES go to NP 9.3.3 SPEED Document change only? X If YES.determine if previously evaluated Does previous evaluation encompass change? X If YES proceed with document changes Commercial Facility Change? x If YES, determine if document updates are required.

For Commercial Facility Change Only:

If YES contact design supervisor. If Document Updates?

NO proceed outside of Engineering process controls.

Document below.

Is this small scope? X If YES perform Minor Plant Change If NO, it is a Plant Modification. Go to EAC for review and approval (NP 7.2.1)

If it is determined that this is not a Plant Change or Modification, document and/or attach justification. Also, attach document update checklist if necessary.

ENGINEERING CHANGE PROCESS TO USE:

Minor Plant PBF-1605a Revision 0 10/02101 Page I of 2

Temporary Modification - See NP 7.3.1. i A generally non-recurring physical change to operational plant systems, components, or equipment that exists for a short duration.

Equivalent Change - See NP 9.3.3.

A hardware change that results in the installation of an item, not identical to the original item, that does not result in a change to those bounded technical requirements that 1) ensure performance of design bases functions, or 2) ensure compliance with the plant licensing basis of either the item(s) or applicable interfaces.

Commercial Facility Change - See NP 7.2.6. Must answer NO to the following:

S1.Does the change impact licensing basis?

2. Will it impact the Electrical Distribution System?

3. Will the change affect HVAC or air systems within the Plant?

4. Does the change interface with existing fire suppression/detection systems or introduce any new combustibles?

5. Will the change impact Emergency Planning?

6. Have a seismic interaction with Plant Equipment?

7. Will the change benefit from the Design Change Process?

PBF-1605a Reviston 0 10/02/01 Page 2 of 2

(7. Point Beach Nuclear Plant PLANT DESIGN CHANGE CHECKLIST PLANT MODIFICATION/MINOR PLANT CHANGE NO.: 02-029

Title:

AUX FEED MINI RECIRC SAFETY UPGRADE / REMOVE AF-117 INTERNALS DESIGN SUPERVISOR Design Controls and Project Controls: (Ref. NP 7.2.1, Commentary, for completion of this section.)

Check Applicable Design Controls: Clarifications/Basis:

0] Design Input Checklist (PBF-1584) 0] DUC (PBF-1606) 0] Design Verification Notice (PBF-1583)

.] Calculations 0] Design Documentation (PBF-1585), or equivalent 0] Design Change In Progress DCNs El Engineering Change Requests El Specifications El El Check Applicable Project Controls: Clarifications/Basis:

13 Modification Team Required (indicate minimum groups to request)

El Conceptual Design Package Required El Budget Design Project (Impact) Number El Detailed Project Schedule IWP Required Assigned Modification Engineer: Roy Chapman Design Supervisor:

Date: 94/i07 PBF-1605 Revision 8 06/12/02 Pagce I of 4 Reference(s): NP 7.2.1. PBF-1583. PBF-15S4 NP7 2.2. PBF-1585. PBF-1606

Point Beach Nuclear Plant PLANT DESIGN CHANGE CHECKLIST PLANT MODIFICATION/MIOR PLANT CHANGE NO.: 02-029 CONCEPTUAL DESCRIPTION/REFERENCE INFORMATION (IF APPLICABLE)

GROUP HEAD CONCEPTUAL DESIGN REVIEW AND ACCEPTANCE [Check here if not required: Z]

Review conceptual design. Attach comments on NPBU Document Review Comment Sheet (PBF-1622 or equivalent)

Grou2 Acceptance Signature Date Comments Radiation Protection -_1=None D- Attached Fire Protection E-"] None El Attached Installing Organization [-_ None [-] Attached

_-"_ None E1 Attached "ElNone R- Attached

__-E None R- Attached

[-_ None E] Attached DesiEnSLSupervisor E-l None E-] Attached PBF-1605 Revision 8 06/12102 Page 2 of 4 Reference(s). NP 7.2.1. PBF-1583, PBF-1584 NP722.PBF-1535 PBF-1606

Point Beach Nuclear Plant PLANT DESIGN CHANGE CHECKLIST PLANT MODIFCATION/MINOR PL CHAGE NO.: 02-029 TRAINING AND PRA GROUP NOTIFICATION Training and PRA group notified of modification. TWR#: 012. 4o Modification Engineer: "-o OZc../

• 4 "A- Date: - $ .

FINAL DESIGN REVIEWS

-Review final design. Attach comments on Document Review Comment Sheet (PBF-1622 or equivalent)

Group Acceptance Signature Date Comments Oprtin None ['] Attached Fire Protection Engineer 9-6 7 iDýoe E] Attached Mechanical Maintenance VI'M,1 1112None ~IAttached Systems Engineering [DNone ElAttached El None El Attached ER None El Attached E- None Attached Tech. Review _ _ _ _re. _ _ _ _ _ _ _d--~

INDEPENDENT REVIEW OF INSTALLATION DOCUMENTS (IWP or Work Order Plan) List all IWP's and WO's used for installation IWP's/WO#(s) WO 0212107 All design and licensing requirements have been incorporated in the installation and testing document(s).

Reviewer: L-- k - _ _ _ _ _ _ _ Date: 9- -Oý RELEASE FOR INSTALLATION All design controls have been properly implemented and the project has been appropriately reviewed. All necessary documents are approved. This design is released for installation. Comments regarding release of this design are noted below:

Design Supervisor: /---4 I,24ý,, .4ýý,

Date: 44oz, COMMENTS PBF-1605 Revision 8 06/12/02 Page 3 of 4 Reference(s)' NP 7.2.1. PBF-1583, PBF-1584 NP 7.2 2. PBF-1585. PBF-1606

Point Beach Nuclear Plant PLANT DESIGN CHANGE CHECKLIST PLANT MODIFICATION1MINOR PLANT CHANGE NO.: 02-029 ACCEPTANCE Plant modification is installed, tested, and all documents required for acceptance are complete.

Modification Engineer: 9o C ,,,.,,. - Date: 01i \0-o CLOSEOUT Plant modification is complete, including submittal of all document updates in the Document Update Checklist (PBF-1606).

Reference change tracking numbers on PBF-1606 where appropriate (DCN numbers, FCR numbers, etc.).

Modification Engineer: Date:

Design Supervisor: Date:

NUCLEAR INFORMATION MANAGEMENT Microfilm the entire modification package.

PB F- 1605 Revision 8 06/12/02 Page 4 of 4 Reference(s): NP 7.2.1. PBF-1583. PBF-1584 NP 7.2 2. PBF-1585, PBF-1606

FINAL DESIGN DESCRIPTION MR 02-029 t September 12, 2002 AUX FEED MINI RECIRC SAFETY UPGRADE / Revision 1 REMOVE AF-1 17 INTERNALS UNIT 0 Purpose As committed to the NRC by letter NRC 2002-0068, the Auxiliary Feedwater minimum flow recirculation AOVs will be upgraded to have a safety-related function to open. This will provide an additional level of safety with regards to the potential common mode AFW pump failure originally identified by CR 01-3595 and LER 26612001 005-00. This modification will track all updates required to facilitate this change in safety classification.

Scope MR 02-029 will remove the internals for AF-117 to prevent a common mode active failure of all Auxiliary Feedwater pumps due to an isolation of the mini-recirculation line. This valve is non-QA, non-Seismic, non ASME class. Significant improvement in core damage probability (CDP) can be achieved by preventing this check valve from failing to open, and removing the internals is the best way to achieve this.

MR 02-029 will document the upgraded design basis of the Auxiliary Feedwater minimum flow recirculation AOVs (1/2AF-4002, AF-4007, AF-4014) and piping to support the safety-related function to provide a flow path for the AFW pumps to prevent overheating and hydraulic instabilities.

Desian Inputs

" PBNPLicensing Basis:

TS 3.7.5 - Auxiliary Feedwater System FSAR 9.7 - Instrument Air I Service Air FSAR 10.2 - Auxiliary Feedwater System FPER 5.2.2 - Safe Shutdown Systems and Equipment SSAR 2.3.1.4 - Reactor Heat Removal Function LER 266/2001-005-00 50.59 SCR 2002-0010-01

"* Correspondence:

Letter NPM 2002-0228, Dated 4/25/2002, Designation of Backup Pneumatics for AFW Mini-Recirculation Valves as Safety-Related Letter NRC 2002-0068, Dated 8/1212002, Reply to a Notice of Violation (EA-02-031)

"* CorrectiveAction:

CR 01-2278 CR 01-3595

"* Applicable Codes:

USAS B31.1 - 1967, Power Piping

"* PermanentDrawings:

BECH M-217 Sh. 1 - Auxiliary Feedwater BECH P-103 -Emergency Feedwater Pumps to Main Feedwater Lines 4" & 3" DB-3 Page 1 of 5

FINAL DESIGN DESCRIPTION MR 02-029 September 12, 2002 AUX FEED MINI RECIRC SAFETY UPGRADE / Revision I REMOVE AF-1 17 INTERNALS UNIT 0 BECH P-159 - Aux Feedwater from Condensate Return & Pump Recirc to CST 6" & 3" JG-4 ALOYCO A-46037 - #376-SP 150# Swing Check Valve PlantModifications:

MR 88-099 (*A/*B/*C/*D)

MR 01-144 MR 02-001 Design Description and Analysis

Background

CR 01-2278, CR 01-3595 and LER 266/2001-005-00 identified an issue that could cause a common mode failure of all Auxiliary Feedwater pumps. If an accident or event has occurred that has led to the loss of instrument air, then the AFW pump minimum recirculation control valves 1/2AF-4002 for 1/2P-29, AF-4007 for P-38A, and AF-4014 for P-38B will all fail closed. During this event, it will become necessary for operations to reduce auxiliary feedwater flow to control steam generator level and prevent overfilling, especially if all four auxiliary feedwater pumps are feeding the steam generators (which is likely in a Loss of Offsite Power event). Typically this is done by throttling down AFW pump flow as opposed to securing the pumps. If care is not taken to ensure that the minimum recirculation valves are open when the pump discharg; valves are shut, then the pumps will dead head and fail in a very short time due to overheating. After discovery of this issue, guidance was added to EOP 0, EOP 0.1, ECA 0.0, and AOP 5B to direct operations to verify adequate pump flow if instrument air has been lost before reducing flow to the steam generators, or to stop the pump.

The minimum recirculation flow AOVs (1/2AF-4002, AF-4007, AF-40r4) have a safety-related function to close to ensure adequate AFW flow to the steam generators during several events, with a seismic induced Loss of Normal Feedwater (LONF) and an Anticipated Transient Without Scram (ATWS) being the most limiting events. FSAR Section 10.2 also discusses the effects of a failure of a mini-recirc AOV to close and gives the flow that is diverted from the steam generators through the recirculation line as limited by the flow restricting orifices (1/2RO-4003, RO 4008, RO-4015).

These mini-flow recirc AOVs have never been classified as having a safety-related function to open to prevent pump damage. This has been described as a non-safety related function only, since the AFW pumps will always have forward flow to the steam generators on auto-start. These recirculation lines are most frequently utilized during pump testing, and on unit startup when the required flow is less than what is recommended for AFW pump cooling.

These recirculation line AOVs have an augmented quality function to be opened for Appendix R fires to support AFW pump operation, per SSAR 2.3.1.4.

Following the discovery of the potential common mode failure, backup pneumatic sources were installed for all minimum flow recirculation valves by MR 01-144 and MR 02-001 to provide additional assurance that the auxiliary feedwater pumps would not be damaged on a loss of instrument air. These modifications were an enhancement that reduced the core damage probability from a loss of instrument air and increased the time for an operator to take manual action to override the fail-closed action and open the valves. Instrument air accumulator tanks were installed by MR 02-001 for the 1/2P-29 mini-flow recirc valves (1/2AF-4002), and the existing nitrogen backup system for the MDAFP discharge valves (AF-4012/4019) was tied in by MR 01-144 for the P-38A/B recirc valves (AF-4007, AF-4014). These backup pneumatic sources will provide a safety-related motive force for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the TDAFP recirc valves (I/2AF-4002), and for 90 minutes for the MDAFP recirc valves (AF-4007/4014).

These modifications were installed primarily for the purpose of risk reduction, and the components installed were not originally classified as safety-related for MR 02-001. However, safety related controls were applied to the Parge 2 of 5

FINAL DESIGN DESCRIPTION MR 02-029 September 12,2002 AUX FEEI5 MINI RECIRC SAFETY UPGRADE / Revision 1 REMOVE AF-1 17 INTERNALS UNIT 0 installation and safety related parts were installed. Internal memo NPM 2002-0228 was subsequently issued during installation of MR 02-001, and documented the decision to conservatively upgrade the safety classification of the backup air accumulators and tubing for the 1/2AF-4002 valves to safety-related, while maintaining the open function for the AOVs as non safety-related. ECR 2002-0077 and a revision to SCR 2002-0010 were prepared to document this change.

Components for the nitrogen backup to the AF-4007/4014 valves installed per MR 01-144 were originally installed and scoped safety related, since the backup nitrogen for the MDAFP discharge valves (AF-4012/4019) was classified safety-related, and the connections to the mini-recirc AOVs became part of the safety-related pressure boundary.

Design Changes 0 Safety-Related Classification Uporade Letter NRC 2002-0068 dated 8/12/2002 states that PBNP will classify the open function for the AFW pump minimum flow recirculation valves as safety-related. The letter also states that because not all of the recirculation flow path is safety-related, operability of the AFW pumps will not be dependent on the availability of that flow path.

However, it has been conservatively decided to tie AFW pump operability to this recirculation line. Therefore, even though recirculation line downstream of the orifices is not safety-related, it is required to be inservice to consider the AFW pumps fully operable per TS 3.7.5.

The 1/2AF-4002, AF-4007, and AF-4014 will be classified as having a safety-related function to open to ensure adequate flow through the AFW pumps to prevent pump damage. The valves are already classified as safety related, due to their safety function to close to ensure adequate AFW flow to the steam generators. The backup pneumatic systems installed by MR 01-144 and MR 02-001 were already classified safety-related to support this function.

The current safety-related boundary for the recirculation lines is at the flow restricting orifices. This boundary will not change, and the recirculation piping downstream of these orifices will remain non safety-related. This piping feeds into the condensate storage tanks, which are non safety-related tanks at atmospheric pressure. This line can be credited to support a safety-function while not being classified safety-related because failure of the piping would be conservative in terms of AFW pump protection. The only credible failure of the piping that would cause AFW pump damage would be if check valve AF-1 17 failed to open (an active fiilure). Therefore the internals for this check valve will be removed. All other non-conservative failure modes for the recirc line are passive in nature.

Several manual valves exist in the recirculation lines, and all of these valves are currently red-locked open.

Mispositioning is not credible due to procedural controls in place (red locks), and a disk separation failure of a manual valve is considered passive.

a AF-1 17 Internals Removal To prevent a common mode active failure of the recirculation piping, the internals to check valve AF-1 17 will be removed. This valve has no specified function to isolate, since each recirculation line has a check valve upstream of the mini-recirc AOV, and it is unlikely that the system would be aligned in such a way that would generate a driving head to force water backwards through the line to the AFW pumps. It is likely that the check valve was installed to isolate relief valve AF-4035 from the CST, or to prevent heating steam condensate from entering the recirculation header. After the internals are removed, leakage of the AF-4035 relief valve could cause leakage of CST water into the turbine building, which would be a flooding concern. This, however, does not adversely affect the AFW system.

The heating steam condensate return to the CSTs is no longer used due to condensate water quality concerns.

This valve is an ALOYCO model #376-SP 150# class 3" swing check valve. The disk, clapper arm, and clapper arm shaft will be removed to prevent the valve from isolating flow. The existing cover and bolting will be reused.

Ptige 3 of 5

FINAL DESIGN DESCRIPTION MR 02-029

1. -'"*-1 September 12, 2002 AUX FEED MINI RECIRC SAFETY UPGRADE / Revision 1 "REMOVEAF-1 17 INTERNALS UNIT0 Licensing Basis Updates FSAR Section 10.2 will be revised to document the safety function of the AOVs to open. FSAR Section 9.7 will be revised to document the safety functions of the backup air sources. Tech Spec Bases B3.7.5 will be updated to clarify that the operability of the AFW pumps requires the minimum flow recirculation line to be inservice.

IST Updates Several changes are required within the IST program to reflect the new safety function of the AOVs to open. IT 08A, IT 09A, IT 10, IT 10A, and IT 10B will all be revised to provide stroke open time acceptance criteria for the air-operated valves. Additionally, a full stroke open exercise test of the recirculation check valves (1/2AF-1 14, AF 115, AF-116) was added to these procedures. A closure test of instrument air check valves I/2AF-173 was previously added to IT-0gA and IT-09A after the installation of MR 02-001, however, acceptance criteria *till now be added. The MDAFP Nitrogen backup check valves (AF-133/153) were already in the IST program for their safety function to support the AF-401214019 dicharge AOVs. The IST background document will be updated to document the safety-related open function for the AOVs and recirc check valves, as well as the isolation functions of the instrument air check valves.

Installation Both units may be in any condition, but the commoh recirculation line must be isolated to remove the AF-117 internals. While this line is out of service, the AFW pumps will be considered fully operable. Manual operator action will be credited to prevent pump damage by stationing a level 3 dedicated operator at the AF-4035 relief valve. If the minimum flow recirculation line is needed, and the relief valve does not open automatically while the line is isolated, then the dedicated operator will notify the control room that the recirculation flow path is not available. Although OM 3.26 allows a level 3 dedicated operator to perform other duties, this will not be allowed during this installation. A dedicated operator can be utilized to support AFW pump operability since there will not be a time requirement for the operator to take action, and the dedicated operator is only relaying information. This action for the dedicated operator will be required only when the control operator has taken action to reduce AFW flow.

AF-4035 was replaced by MR 88-099, which upgraded the size of the minimum recirculation line. The valve has a setpoint of 50 psig, and a capacity of 268 gpm (at 55 psig). This is enough flow capacity to provide cooling flow for

. all four AFW pumps. Calculation 2002-0026 and SCR 2002-0377 has been performed to document the acceptability of relying on this relief valve to support AFW pump operability while the common recirculation line is out of service for removal of the AF-1 17 internals.

Post Modification Testing Since no welding is being performed, and the internals removal requires only the disassembly of a mechanical joint (flanged connections), an initial service leakage test per ASME B31.1 is not required. A non-code leak check will be performed at normal operating conditions (with an AFW pump running) to check for gasket leakage.

Pdge 4 of 5

FINAL DESIGN DESCRIPTION MR 02-029 September 12,2002 AUX FEED MINI RECIRC SAFETY UPGRADE! Revision 1 REMOVE AF-1 17 INTERNALS UNIT 0 Desien Outputs

0 S

6 50.59 Safety Screening, PBF-1515c, SCR 2002-0359 (Modification MR 02-029) 50.59 Safety Screening, PBF-1515c, SCR 2002-0377 (Operability During Mod Installation)

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Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page I

(.,,:Title of Proposed Activity: Removal of Internals from AF-1 17 and Upgrade Open function of AFW pumps Mini-Recirc Valves to Safety-Related (MR 02-09)

Associated Reference(s) #: SCR 2002-0010-01, SCR 2002-0321 and 2002-0339; MR 01-144, AF-4007/4014 Backup Nitrogen Supply, MR 02-001, 1/2AF-4002 Backup Air Supply; EVAL 2002-005, Permanent Plant Changes to Address Simultaneous Failure of All AFW Pumps; FCR 02-019; SE 97-085, MR 97-038 AFW Motor Driven Pump Pressure Dischsrge Valve Modification: Fay Letter to NRC, NRC Bulletin 88-04, Potential Safety-Related Pump Loss, June 28, 1988; NPM 2002-0228, Designation Of Backup Pneumatics For AFW Mini- Recirculation Valves As Safety-Related, NRC 2002-0068, Point Beach Nuclear Plant, Units I and 2 Reply to a Notice Of Violation (EA-02-03 1) NRC Special Inspection Report No. 50-266/01-17 (DRS); 50-301/01-17 (DRS).

Prepared by: David Black (fav2 6&i Date: -D oL_

0 Name ( Print) Sigriature Reviewed by: Rob Chapman Date: .-- 02.

Name ( Print) S64ture PART I (50.59/72.48) - DESCRIBE THE PROPOSED ACTIVITY AND SEARCH THE PLANT AND ISFSI LICENSING 1

BASIS (Resource Manual 5.3.1)

NOTE: The "NMC 10 CFR 50.59 Resource Manual" (Resource Manual) and NEI 96-07. Appendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 10 CFR 50.59 and 10 CFR 72.48 screenings.

1.1 Describe the proposed activity and the scope of the activity being covered by this screening. (The 10 CFR 50.59 / 72.48 review of other portions of the proposed activity may be documented via the applicability and pre-screening process requirements in NP 5.1.8.) Appropriate descriptive material may be attached.

The scope of this screeninginvolves the following activities associatedwith MR 02-029:

- Remove internalsfor AF-117, the 4uxiliary FeedwaterPump Common Mini-Recirc Header Check Valve to prevent a common mode activefailure in the recircreturn header.

- Upgrade openfunctionfor all Auxiliaryfeedwaterpumpmini-recircvalves to safety-related(also being done by ISTgroupfor procedurechanges under screenings SCR 2002-0321 and 2002-0339.

-State in the FSAR (FSAR 10.2) and TechnicalSpecification Bases(B 3.7.4) that (I) the openfi.uctionfor all Auxiliary feedwaterpump mini-recirc valves is safety-related,and(2) the recirculationline downstream of the flow restrictingorifices has a safetyfunction and is requiredforAFW operability,but the line is not safety-relatedsincefailure of the line is conservative. The Bases will indicate that the mini-recirculationflowpathshave to be OPERABLEfor the AFWsystem to be operable.

A simplified P&ID of the mini-recircheader is shown below To Condensate Storage Tank T-24B From P.38A. P 385. & 2P-29 From 1 P-29 To Condensate Storage Tank T.2.1A 3*-jr-4 1'-dG-4 PBF-1515c R oI O-1/03,02 RcIicrnc: N1p5 1 .1

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR numberon all pages Page 2 EVAL 2002-005 evaluatedpermanentprocedurechanges that were implemented in response to a condition (CR 01-3595) that was identified where, with a procedure-directedoperatoraction to controlsteam generatorlevel (which could be accomplishedby reducingflow through one or more AFW pumps), concurrent with a loss of instrument air (which would cause the AFW pumps' mini-recirculationvalves to fail close), the potentialexistedfor a simultaneousfailure of the multi stage high pressureAFW pumps due to very low or no flow through runningAFW pumps. The procedurechanges added instructions to the operatorsthat ifanyAFWpump mini-recirc valve fails shut OR anminciator COI A 1-9, INSTRUMENT AIR HEADER PRESSURE LOW is in alarm, then monitor and maintain minimum AFW flow or stop the affected AFW pump as necessary to control SIG levels. Minimum flow values for each pump were also included in the procedures. The majority of the changes were associatedwith EOPand ECA foldout pages. The procedureswere initially revisedas a compensatory measure to support AFW pump operability. The 50.59 evaluationreviewed the procedure changes as a permanentchange to the procedures as describedin the FSAR to confirm consistency with the licensing basis. The permanentprocedure changes restoredthe AFW pumps tofully operablestatus.

This evaluation also was the basisfor changes to the FSAR to clarify that the mini-recirculationvalves require instrument air to function and that either a pump minimum flow is maintainedor pumps are secured if the valvefails or instrumentair is lost (FCR 02-019).

ScreeningSCR 2002-0010-01 reviewedmodifications to provide backup air sources to all AFW pump minimum flow recirculationvalves. These modifications were an enhancementthat reducedthe core damage probabilityfroma loss of instrumentair and increasedthe timefor an operatorto take manual action to override the valves open. Instrument air accumulatortanks were installedby MR 02-00lfor the 1/2P-29 valves (l/2AF-4002), and the existing nitrogen backup system for the MDAFP dischargevalves were tied in by MR 01-144for the P-3 8A/B valves (AF-4007.AF-4014).

PointBeach has made an NRC commitment to upgrade the openfitnctionfor all mini-recircvalves to safety-relatedas stated in NIMC letter NRC 2002-0068:

(7' "To fitrther improve the future effectiveness ofthe AFW system by providingadditionalpump protectionagainstlow flow, Point Beach is classifying the openfunction of the pump recirculation flow control valves, as safety-related This will provide a redundantmethod ofprovidingfor minimum AFW pump flow and consequently, pump cooling. As a result, testing andquality assurance requirementsrequiredforsafety-relatedfunctions will be appliedto the openfunction of these valves.

As discussedpreviously, internalpump cooling is designed to be provided by minimum forwardflow through thepumps. Classifyingthe recirculationflow control valves as safet-relatedwill provide greaterassurancethat minimum flow will be available to provide internalpump cooling The pneumatic backup supply to the recirculationflow control valves is limiting and therefore Point Beach will also continue to specify operatoraction to manually open these recirculationvalves. Similar to other plants, Point Beach has one common recirculationflowpathfrom all the AFW pumps to the condensatestorage tanks" The scope ofthis screening also includes FSAR changes and Technical Specification Bases changes to coincide with the mini-recirculationsafety upgrade describedabove.

12 Search the PBNP Current Licensing Basis (CLB) as follows: Final Safety Analysis Report (FSAR), FSAR Change Requests (FCRs) with assigned numbers, the Fire Protection Evaluation Report (FPER), the CLB (Regulatory) Commitment Database, the Technical Specifications, the Technical Specifications Bases, and the Technical Requirements Manual. Search the ISFSI licensing basis as follows: VSC-24 Safety Analysis Report, the VSC---4 Certificate of Compliance. the CLB (Regulatory)

Commitment Database, and the VSC-24 10 CFR 7P.212 Site Evaluation Report. Describe the pertinent design function(s),

performance requirements. and methods of evaluation for both the plant and for the cask, ISFSI as appropriate. Identify where the pertinent information is described in the above documents (by document section number and title). (Resource Manual 5.3.1 and NEI 96-07. App. B. B 2)

The .4uxiliar Feedwater t..IFO) system has theJbllowmngfuncttons describedin the licensing bas is.

a. To automaticallystart and ensure that adequateftedwateris sutpplied to the steain generators 1br heat removal during accidents ishcwh may result in a main steam .- e,.jr,valve opening (Lo.5s o] ,VormalFeedivater- includinz

.4 TI VS and Loss tf'.-IC to the Staton .4Azciliarwýl PI3F-151lc R¢.cretncL N' .I 'I Rc,,t,,i~n I O OI}-~},)2

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 3

b. To automaticallystart and provideflow to maintainsteam generatorlevels during accidents which require or result in rapidreactorcoolant system cooldown (Steam GeneratorTube Rupture and Rupture of a Steam Pipe).

c. To allow the isolation ofall lines to the rupturedsteam generator in the SGTR event.

d To provide sufficient feedwater to remove decay heatfrom both units for one hour during a station blackout (SBO) event (TDAFP only).

e. To provide sufficient flow to the steam generators to remove decay heat to achieve cold shutdown within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sfollowing a plantfire (Appendix R).

f To withstand a seismic event (i.e., the seismic Class 1 portions of the system) and to ensure that steam generatorlevels are maintainedduring a seismic event.

g. To provideflow to the steam generatorsduringplantstartup andshutdown, and duringhot shutdown or hot standby conditionsfor chemicaladditions andwhen operationof the mainfeedwater and condensatesystems is not warranted.

These modifications affect the operationof the minimum recirculationvalvesfor the auxiliaryfeedwaterpumps (1/2AF 4002, AF-4007, AF-4014). These valves have the following designfunctions:

1. To isolate the minimum recirculationline to ensure that the auTiliaryfeedwaterpumps deliver the requiredflow to the steam generatorsas needed to support thefollowing accidents or events: LONF. LOA C, MSLB, SGTR, A TWS, Appendix R, andSBO.

2. To open toprovideflow through the auxiliaryfeedwaterpumps to prevent hydraulic instabilitiesandto dissipatepump heat.

3. To maintainthe pressure boundaryintegrity of the auxiliaryfeedwater system.

FSAR 10.2.2, System Design and Operation,states: 'The auxiliaryfeedwater system consists of two electric motor-driven pumps, two steam turbine-drivenpumps, pump suction and dischargepiping,andthe controls andinstrumentationnecessary for operationof the system. Redundancy is provided by utili:ingtwo pumpingsystems, two different sources ofpowerfor the pumps, and two sources of water supply to the pumps. The system is categorikedas seismic Class I and is designed to ensure that a singlefault will not obstruct the system function."

CLB

References:

FSAR 7.2.3.2- Specific Controland ProtectionInteractions FSAR 7.3.3.4 - ManualAFW Flow Control DuringPlantShutdown FSAR Section 7.4.1 - AMSA C FSAR Section 10.1 - Steam and Power Conversion System FSAR Section 10 2 -Awriliary Feedwater FSAR Figure 10.2-1 Sheet I - Bech M-217 Sh. 1-A uxiliaryFeedwaterSystem FSAR Figure 10.2-1 Sheet 2- Bech M-217 Sit 2- Auxiliary FeedwaterSystem FSAR Section 14.1.10 - Loss ofNormal Feedwater FSAR Section 14. 1.11 - Loss ofAll AC Power to the StationAuxiliaries FSAR Section 14.2.4 - Steam GeneratorTube Rupture FSAR Section 14.2.5 - Rupture of a Steam Pipe FSAR Appendix A.I - Station Blackout FPER 5.2.2 - Safe Shutdown Systems and Equipment FPER5.2.5.2.3 - Auxiliary Feedwater Pump Room Tech Spec 3..".5- Azriliary Feedwater Tech Spec Bases B 3.7.5- A ucxiliary Feedwater NRC 2002-0068, Point Beach NuclearPlant, Units I and2 Reply to a Notice Of Violation (EA-02-031) NRC Special InspectionReport No. 50-266/01-17 (DRS); 50-301/01-17 (DRS) 1.3 Does the proposed activity involve a change to any Technical Specification? Changes to Technical Specifications require a License Amendment Request (Resource Manual Section 5.3.1.2).

Technical Specification Change: [] Yes 0 No Ifa Technical Specification change is required. explain what the change should be and %hy it is required.

PBr.1s51 ;

Revi61on I 0-4,03 02 R."reni.. NP 51l

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 4

" 1.4 Does the proposed activity involve a change to the terms, conditions or specifications incorporated in any VSC-24 cask Certificate of Compliance (CoC)? Changes to a VSC-24 cask Certificate of Compliance require a CoC amendment request.

El Yes 0 No If a storage cask Certificate of Compliance change is required, explain what the change should be and why it is required.

10 CFR 50.59 SCREENING PART 11 (50.59) - DETERMINE IF THE CHANGE INVOLVES A DESIGN FUNCTION (Resource Manual 5.3.2)

Compare the proposed activity to the relevant CLB descriptions, and answer the following questions:

YES NO QUESTION ED El Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?

0] El Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?

0] El Does the proposed activity involve SSCs whose failure could initiate a transient (e.g., reactor trip, loss of feedwater, etc.) or accident, OR whose failure could impact SSC(s) credited in the Safety Analyses?

0 El Does the proposed activity involve CLB-described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply.with, regulations, license conditions, orders or technical specifications?

"'. El 0] Does the activity involve a method of evaluation described in the FSAR?

El 0 Is the activity a test or experiment?. (i.e., a non-passive activity which gathers data)

El 0] Does the activity exceed or potentially affect a design basis limitfor afissionproduct barrier(DBLFPB)?

(NOTE: If THIS questions is answered YES, a 10 CFR 50.59 Evaluation is required.)

If the answers to ALL of these questions are NO mark Part III as not applicable, document the 10 CFR 50.59 screening in the conclusion section (Part IV), then proceed directly to Part V - 10 CFR 72.48 Pre-screening Questions.

If any of the above questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.

1. The AF-117 check valve has an implicitfinction to open to allow minimum recirculationflowfromthe auxiliaryfeedwaterpumps to return to the condensatestorage tank(s). There is no direct discussion of thisfinction in the licensing basis, other than that the valve appears in FSAR Figure 10.2-1.

A search of plant documents did not yield any reference to a needfor valve AF-) 17 to close (or remainclosed). The following uses were consideredaspossiblefunctionsfor the closed position:

a. To prevent backflow of heating steam condensate from entering the mini-recirculation return header. Given that the min recirculationheader is not drainedand remainsfilled with water during and afterpump operation, migrationof heating steam condensate into the header in any stgniflcant amounts is unlikely Further,the plant no longer returns heatingsteam condensate to the condensate storagetank because of water quality concerns, and it is extremely unlikely that this plant designfeature will ever be used again. Regardless. this reasonfor preventing backtlow does not constitute a licensing basis clesignfinction Therefore this function will not be evaluatedfor adverse effects.

b To prevent backflow and sithonin! of vater from the condensate storage tanks through AF-4035 in the event that this valve onens due to hieh pressure in the line or due to a break in the header. The valve's nominalreliefsetting is 50 psi. For AF-4035 to open, either tn event occurs where multltple pumps start with high recircflowratesandpressurt:ethe line. or there is a blockage in the line. The safetv-relatedJlowrestrictingorifices in the recirc linefront each AFW puip hinnt the flow and pressurefrom each pump Further if the line is blocked either by mispostuomng nf/manual valves orfailure of the check valve to R: Nion I 0o-Im)3 ) R()crn1LC ,lp I

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 5 open, there is noflowpathfor backflow. The only credible wayfor the header to break is during a seismic event, in which case waterfrom the condensatestorage tank cannot be creditedfor accidentmitigation. The prevention of backflowfrom the condensatestorage tanks cannot be considereda licensing basisdesign function. Therefore thisfinction will not be evaluated for adverse effects.

2. The minimum recirculationvalves for the auxiliaryfeedwater pumps (1/2AF- 4002, AF-4007, AF-4014) have the designfunctions to isolate the minimum recirculationline to ensurethat the auxiliaryfeedwaterpumps deliver the requiredflowto the steam generatorsas needed to support the mitigation of accidents or events. The proposedactivity is to make the openfinctionfor these valves a safety-relatedfunction. The new componentsfor modifications MR 02-001 andMR 01-144 as discussed above were installedsafety-relateddue to their risk significance. These components were classifiedsafety-relatedbased on a PBNP managementdecision, as described in NPM 2002-0228. These modifications ensure a safety-relatedsupply of air or nitrogento ensure that the auxiliaryfeedwaterpumps have adequate cooling.

PART 111 (50.59) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (Resource Manual 5.3.3)

If ALL the questions in Part II are answered NO, then Part III is C] NOT APPLICABLE.

Answer the following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a 10 CFR 50.59 Evaluation is required; EXCEPT where noted-in Part 111.3.

II. 1 CHANGES TO THE FACILITY OR PROCEDURES YES NO QUESTION

[] 0] Does the activity adversely affect the designfunction of an SSC credited in safety analyses?

[

,"* 0] Does the activity adversely affect the method of performing or controlling the design function of an SSC credited in the safety analyses?

If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion as necessary):

1. The only function identifiedfor the AF-117.check valve was to open to allow mini-recirculationwaterfrom AFW pumps to return to the condensatestorage tank. It has no designfunction to close Therefore removing the check valve internals has no adverse effect on the openfunction andprecludes the valvefrom failing to open.

2. The change in designationof the open finction ofmini-recirculationvalves 1/2AF- 4002, AF-4007. AF-4014 as safety relatedhas no adverse affect on the valve'sfunction to open or close. The openfunction was a designfunction desribed in the FSAR. Designatingthe openfunction as safety-relatedrequiresadditionquality controland testing andshould.

therefore, make the openfunctions of the valve more reliable. The necessary backup air/nitrogen supplies were installed as safety-relatedso no additionalphysical changes to the plant are required SCR 2002-0010-01 assessedthe installation of these backup pneumatic supplies on the close finction of the valves andno adverse effects were identified. Therefore there is no adverse effect on the functions of these valves.

3. Changes will be made in the FSAR to state that the open functionfor all.4 uAxliaryfeedwater pump mini-recircvalves is safety-related The Technical Specification Bases will indicatethat the mini-recirculationflowpathshave to be OPERABLE for the AFW system to be operable. These changes only reflect the activities discussedabove, which have already been assessedand determinedthat there is no adverse effect on a designfunction FSAR changes will also be made to reflect the recirculationline downstream ofthe flow restrictingorifices has a safetyfiunction and is requiredfor

.4FWY operabiliot but the line is not safety-relatedsincefailure of the line is conservative. No physical changes were made to this line. If the linefails, a break or opening will be created which inay increaseoverallflow through the recirculationline. The service-water s.stem is the safet. -relateditatersupplyfor the.4FW system. The loss of CST inventory due to the linefailing andrecirculationwater spilling only affects the non-safety-relatedwater sourcefor the

,4FWV system. The requiredmintminn CST water is based on the station blackout(SBOJ event; SBO is not an event that would cause this line to fail (open). Therefore. based on the above, the licensing basis documents changes do not have an adverse effect on any designfunctions PBF-1lSc Rcviýaon I 0o-,03,02 R,!Cr01L. %,' 1 '1

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59172.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 6

!:)111.2 CHANGES TO A METHOD OF EVALUATION (If the activity does not involve a method of evaluation, these questions are Z NOT APPLICABLE.)

YES NO QUESTION El El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in the CLB?

[E [0 Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in the CLB?

If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers are NO. describe the basis for the conclusion (attach additional discussion, as necessary).

111.3 TESTS OR EXPERIMENTS If the activity is not a test or experiment, the questions in III.3.a and III.3.b are 0 NOT APPLICABLE.

a. Answer these two questions first:

YES NO QUESTION E- [I Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?

El El Are the SSCs affected by the proposed test or experiment isolated from the facility?

(7*,

If the answer to BOTH questions in V.3.a is NO, continue to 1II.3.b. If the answer to EITHER question is 'YES, then describe the basis.

b. Answer these additional questions ONLY for tests or experiments which do NOT meet the criteria given in III.3.a above.

If the answer to either question in III.3.a is YES. then these three questions are Z NOT APPLICABLE.

YES NO QUESTION El El Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the CLB?

l [El Does the activity utilize or control an SSC in a manner that is inconsistent with the analyses or descriptions in the CLB?

0l Dl Does the activity place the facility in a condition not previously evaluated or that could affect the capability of an SSC to perform its intended functions?

If any answer in III.3.b is YES, a 10 CFR 50.59 Evaluation is required. If the answers in I1.3.b are ALL NO, describe the basis for the conclusion (attach additional discussion as necessary):

, i ...... i i, V, I'

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on aH pages Page 7

". Part IV- 10 CFR 50.59 SCREENING CONCLUSION (Resource Manual 5.3.4).

Check all that apply:

A 10 CFR 50.59 Evaluation is [] required or Z NOT required.

A Point Beach FSAR change is 0 required or E] NOT required. If an FSAR change is required, then initiate an FSAR Change Request (FCR) per NP 5.2.6.

A Regulatory Commitment (CLB Commitment Database) change is El required or [0 NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.

A Technical Specification Bases change is [9 required or C] NOT required. If a change to the Technical Specification Bases is required, then initiate a Technical Specification Bases change per NP 5.2.15.

A Technical Requirements Manual change is C] required or Z NOT required. If a change to the Technical Requirements Manual is required, then initiate a Technical Requirements Manual change per NP 5.2.15.

10 CFR 72.48 SCREENING NOTE: NEI 96-07, Appendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 72.48 screenings.

PART V (72.48) - 10 CFR 72.48 INITIAL SCREENING QUESTIONS Part V determines if a full 10 CFR 72.48 screening is required tý be completed (Parts VI and VII) for the proposed activity.

i'.*'YES NO QUESTION C 0] Does the proposed activity involve IN ANY MANNER the dry fuel storage cask(s), the cask transfer/transport equipment, any ISFSI facility SSC(s), or any ISFSI facility monitoring as follows: Multi-Assembly Sealed Basket (MSB), MSB Transfer Cask (MTC), MTC Lifting Yoke, Ventilated Concrete Cask (VCC), Ventilated Storage Cask (VSC), VSC Transporter (VCST), ISFSI Storage Pad Facility, ISFSI Storage Pad Data/Communication Links, or PPCSIISFSI Continuous Temperature Monitoring System?

C 0] Does the proposed activity involve IN ANY MANNER SSC(s) installed in the plant specifically added to support cask loading/unloading activities, as follows: Cask Dewatering System (CDW), Cask Reflood System (CRF), or Hydrogen Monitoring System?

C 0] Does the proposed activity involve IN ANY MANNER SSC(s) needed for plant operation which are also used to support cask loading/unloading activities, as follows: Spent Fuel Pool (SFP), SFP Cooling and Filtration (SF),

Primary Auxiliary Building Ventilation System (VNPAB), Drumming Area Ventilation System (VNDRM),

RE-105 (SFP Low Range Monitor), RE-135 (SFP High Range Monitor), RE-221 (Drumming Area Vent Gas Monitor), RE-325 (Drumming Area Exhaust Low-Range Gas Monitor), PAB Crane, SFP Platform Bridge, Truck Access Area, or Decon Area?

C 0] Does the proposed activity involve a change to Point Beach CLB design criteria for external events such as earthquakes, tornadoes, high winds, flooding, etc.?

C 0] Does the activity involve plant heavy load requirements or procedures for areas of the plant used to support cask loading/unloading activities?

C] 0] Does the activity involve any potential for fire or explosion where casks are loaded, unloaded, transported or stored?

If ANY of the Part V questions are answered YES then a full 10 CFR 72.48 screening is required and answers to the questions in Part VI and Part VII are to be provided. If ALL the questions in Part V are answered NO. then check Parts VI and VII as not applicable. Complete Part VIII to document the conclusion that no 10 CFR 72.48 evaluation is required.

PG F. 15 ISc Rdicrenmc NI' . 1 ,

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59172.48 SCREENING (NEW RULE) Veif SCR nunmber on all pages Page 8

,-- PART VI (72.48) - DETERMINE IF THE CHANGE INVOLVES A ISFSI LICENSING BASIS DESIGN FUNCTION (If ALL the questions in Part V are NO, then Part VI is 0 NOT APPLICABLE.)

Compare the proposed activity to the relevant portions of the ISFSI licensing basis and answer the following questions:

YES NO QUESTION El El Does the proposed activity involve cask/ISFSI Safety Analyses or plant/cask/ISFSl structures, systems and components (SSCs) credited in the Safety Analyses?

El El Does the proposed activity involve plant, cask or ISFSI SSCs that support SSC(s) credited in the Safety Analyses?

E] nl Does the proposed activity involve plant, cask or ISFSI SSCs whose function is relied upon for prevention of a radioactive release, OR whose failure could impact SSC(s) credited in the Safety Analyses?

El El Does the proposed activity involve cask/ISFSl described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, CoC conditions, or orders?

El 0l Does the activity involve a method of evaluation described in the ISFSI licensing basis?

El El Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)

El El Does the activity exceed or potentially affect a cask design basis limitfor a fissionproduct barrier(DBLFPB)?

(NOTE: If THIS questions is answered YES, a 10 CFR 72.48 Evaluation is required.)

If the answers to ALL of these questions are NO, mark Parts VII as not applicable, and document the 10 CFR 72.48 screening in the conclusion section (Part VIII).

__. If any of the above questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s)

( involved.

PART VII (72.48) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (NEI 96-07, Appendix B, Section B.4.2. 1)

(If ALL the questions in Part V or Part VI are answered NO then Part VII is E] NOT APPLICABLE.)

Answer the following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a 10 CFR 72.43 Evaluation is required; EXCEPT where noted in Part VII.3.

VII.1 Changes to the Facility or Procedures YES NO QUESTION El El Does the activity adversely affect the designfinction of a plant, cask, or ISFSI SSC credited in safety analyses?

El El Does the activity adversely affect the method of performing or controlling the designfunction of a plant, cask, or ISFSI SSC credited in the safety analyses?

If any answer is YES, a 10 CFR 72.48 Evaluation is required. If both answers are 'NO. describe the basis for the conclusion (attach additional discussion, as necessary):

PBI-151.;c t) . - -I ININ iRC,.rCl.! NP ; I ý

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 9 VII2 Changes to a Method of Evaluation (If the activity does not involve a method of evaluation, these questions are Z NOT APPLICABLE.)

YES NO QUESTION El El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in a cask SAR?

El El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in a cask SAR?

If any answer is YES, a 10 CFR 72.48 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion, as necessary):

VII.3 Tests or Experiments (If the activity is not a test or experiment, the questions in VII.3.a and VII.3.b are 0 NOT APPLICABLE.)

a. Answer these two questions first:

YES NO QUESTION El El Is the proposed test or experiment bounded by other tests or experiments that are described in the cask ISFSI licensing basis?

El El Are the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISFSI facility?

If the answer to both questions is NO, continue to VII.3.b. If the answer to EITHER question is YES, then briefly describe the basis.

b. Answer these additional questions ONLY for tests or experiments which do not meet the criteria given in VII.3.a above.

If the answer to either question in VII.3.a is YES, then these three questions are 0 NOT APPLICABLE:

YES NO QUESTION El E] Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the ISFSI licensing basis?

El El Does the activity utilize or control a plant, cask or ISFSI facility SSC in a manner that is inconsistent with the analyses or descriptions in the ISFSI licensing basis?

El El Does the activity place the cask or ISFSI facility in a condition not previously evaluated or that could affect the capability of a plant, cask, or ISFSI SSC to perform its intended functions?

the basis for the If any answer in VII.3.b is YES, a 10 CFR 72.48 Evaluation is required. If the answers are all NO describe conclusion (attach additional discussion as necessary):

P13['-SISC qlcf'rtný NI1 '; I ý

Point Beach Nuclear Plant SCR 2002-0359 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pageps Page 10

.-.. PART VIII- DOCUMENT THE CONCLUSION OF THE 10 CFR 72.48 SCREENING "Checkall that apply:

A 10 CFR 72.48 Evaluation is E] required or Z NOT required. Obtain a screening number and provide the original to Records Management regardless of the conclusion of the 50.59 or 72.48 screening.

A VSC-24 cask Safety Analysis Report change is C] required or [] NOT required. If a VSC-24 cask SAR change is required, then contact the Point Beach Dry Fuel Storage group supervisor.

A Regulatory Commitment (CLB Commitment Database) change is E] required or Z NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.

A change to the VSC-24 10 CFR 72.212 Site Evaluation Report is El required or (Z NOT required. If a VSC-24 10 CFR 72.212 Site Evaluation Report change is required, then contact the Point Beach Dry Fuel Storage group supervisor.

P F-I SI-5c Rei:,on I ../0 3,O2 Rc ern, "41 .I N

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on alt pages Page 1

";'.Title of Proposed Activity: AFW System Operability during Removal of Internals from Check Valve AF-l 17 (MR 02-029)

Associated Reference(s) #: EVAL 2002-005, "Permanent Plant Changes to Address Simultaneous Failure of All AFW Pumps;" FCR 02-019; SCR 2002-0359, "Removal of Internals from AF-1 17 and Upgrade Open function of AFW pumps Mini-Recirc Valves to Safety-Related (MR 02-09)," Work Order Work Plan 0212107 - MR -02 029, Calculation 2002-0026, "Evaluation of AFW Recirculation Line Relief Valve AF-4035," OM 3.26, "Use of Dedicated Operators."

Prepared by: David Black _ _ _,, _ _ _ Date:

Name ( Print) Sigrhature Reviewed by: Rob Chapman Date: c7- O - G2 Name ( Print) 0nature PART I (50.59/72.48) - DESCRIBE THE PROPOSED ACTIVITY AND SEARCH THE PLANT AND ISFSI LICENSING BASIS (Resource Manual 5.3.1)

NOTE: The "NMC 10 CFR 50.59 Resource Manual" (Resource Manual) and NEI 96-07. Appendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 10 CFR 50.59 and 10 CFR 72.48 screenings.

"(*I.

I Describe the proposed activity and the scope of the activity being covered by this screening. (The 10 CFR 50.59 /72.48 "reviewof other portions of the proposed activity may be documented via the applicability and pre-screening process requirements in NP 5.1.8.) Appropriate descriptive material may be attached.

The scope ofthis screening involves the isolation (dangertagging) of the AFW pump recirculationline in order to remove the internals ofcheck valve AF- 117.

The NMC guidancefor the new 10 CFR 50.59 rule (NMC Resource Manual) states that modification activities (i.e., interim conditions) are to be assessed as maintenance activities under 10 CFR 50.65 (a)(4) of the MaintenanceRule. However, the guidance also states that the status of systems, structures,andcomponents during these type of activitiesshould be addressed to ensure operability in accordancewith TechnicalSpecifications. This scope of this screening is to assess operability ofthe AFW pumps during the periodthe line is isolated.

In preparationfor removing the internals ofAF-117, this check valve will be isolatedby closing manual valves on both sides ofAF-117. Since the line where AF-) 17 is located is common to allfour AFW pumps, an alternatemini-recircflowpath is desirablefor AFW system operability. The header upstream ofAF-Il 7 and the manual isolationvalve AF-I includes relief valve AF-4035 which is installedin the line to prevent over-pressuri:ationof the line. The nominalreliefsetting is 50psig with an accumulation of 5 psig. The valve is not classified as a safety-relatedvalve..

The 50.59 evaluation EVAL 2002-005 establishedthat manual control ofAFWflow to maintain steam generatorlevel following a transient or accident was part of the plant'slicensing basis. This was done in response to a condition that was identified where,following a loss of instrument air,all the AFW mini-recirccontrolvalves willfail (closed) and may not be detected by the operator. Normal practicefor controlling AFWflow is to throttle or close the discharge valve even up to the point of no forwardflow through the pump. In this situation, the operator would have to establish sufficientforwardflow to maintainpump cooling or secure the pump(s) as needed There was no proceduraldirection in the EOPs to do eitherof these actions. EVAL 2000-005 evaluated the procedurechanges which added directions to operatorsto respondto a recirc valve failure or a loss of instrument air,and the impacts on the AFW pump and motors if the AFW pumps had to be secured and restartedto maintainSG level. It was determined that the changes did not require NRC approval.

PBF-1515c

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 2 In order to maintain a recirculationflowpathfor the AFW pumps while removing the internals ofAF-117, arecirculation flow path will be providedduringthe time the recirculationheader back to the condensate storage tank (CST) isolated. The flow path that will be credited is through the reliefvalve AF-4035 which is located upstream of the AF-! valve. AF-) will be closed to isolate the AF-) 17 valve. If the AFW pumps start during the time the headeris isolated,then the line will become pressurizedbecause the line is isolated,andreliefvalve AF-4035 should lift. Calculation2002-0026, "EvaluationofAFW RecirculationLine Relief Valve AF-4035, " evaluates the reliefcapacity of the pipingnetwork of the individualpump recirc lines to the headerand through the reliefvalve. This calculationdetermined that adequate the reliefvalve can pass the requiredflowfor cooling allpumps AF-4035 is not a safety-relatedvalve and there is no redundantvalve. Ifa transientor accident occurs and this valve fails to function, the control room operators will have to maintainadequateforwardflow in the runningpumps and secure other pumps as required.. In order to alert the control room ifthe reliefvalvefails to open, a Level 3 DedicatedOperatorper OM 3.26, "Use ofDedicatedOperators"will be stationednear the AF-4035 valve during the time the recirculationline is isolated.The Level 3 Dedicated Operatorwill remain in constant radio communicationswith the control room, andthe operator'sonly function is to monitor actuationofAF-4035, and to notify the control room if the valve fails to open after AFW pumps start. 1IfAF-4035fails to relieve (open), the dedicated operatorwill notify the controlroom immediately while the pumps are still being cooled byforwardflow, and the control room operatorswill know that when they reduce AFWflow to controlsteam generatorlevel they will have to maintain the requiredminimum forwardflow in AFW pumps or secure pumps as necessary as directed by the EOPs as discussed above.

A discharge hose will be attachedto AF-4035, andthe hose will be routed to an appropriatelocation. The hose will be securedto prevent movement in the event of aAFWpump auto start. These measures will prevent damage to otherplant equipment should the reliefvalve open after an AFW pump auto start.

-(1.2 Search the PBNP Current Licensing Basis (CLB) as follows: Final Safety Analysis Report (FSAR), FSAR Change Requests (FCRs) with assigned numbers, the Fire Protection Evaluation Report (FPER), the CLB (Regulatory) Commitment Database, the Technical Specifications, the Technical Specifications Bases, and the Technical Requirements Manual. Search the ISFSI licensing basis as follows: VSC-24 Safety Analysis Report, the VSC-24 Certificate of Compliance, the CLB (Regulatory)

Commitment Database, and the VSC-24 10 CFR 72.212 Site Evaluation Report. Describe the pertinent design function(s),

performance requirements, and methods of evaluation for both the plant and for the cask/ISFSI as appropriate. Identify where the pertinent information is described in the above documents (by document section number and title). (Resource Manual 5.3.1 and NEI 96-07, App. B, B.2)

The Auxiliary Feedwater(AFW) system has thefollowingfunctions describedin the licensing basis:

a. To automaticallystart and ensure that adequatefeedwater is supplied to the steam generatorsfor heat removal during accidents which may result in a main steam safety valve opening (Loss ofNormal Feedwater- including A TWS, andLoss ofAC to the Station Auxiliaries).

b. To automaticallystart andprovide flow to maintain steam generatorlevels during accidents which require or result in rapidreactorcoolantsystem cooldown (Steam GeneratorTube Rupture and Rupture of a Steam Pipe).

c. To allow the isolationof all lines to the rupturedsteam generatorin the SGTR event.

d To provide sufficientfeedwater to remove decay heat from both units for one hour during a station blackout (SBO) event (TDAFP only).

e. To provide sufficientflow to the steam generators to remove decay heat to achieve cold shutdown within 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sfollowing aplantfire (Appendix R).

f To withstand a seismic event (i.e., the seismic Class I portions of the system) and to ensure that steam generatorlevels are maintainedduring a seismic event.

g. To provideflow to the steam generatorsduringplant startup and shutdown, and duringhot shutdown or hot standby conditionsfor chemical additionsand when operation of the mainfeedwater and condensatesystems is not warranted The minimum recirculationlinesfor the auxiliaryfeedwater pumps and the recirculationheaderfor the auxiliaryfeedwater pumps have the designfunction to provide recirculationflowpathsfrom the auxiliaryfeedwater pumps to prevent hydraulic instabilitiesand to dissipatepump heat.

CLB References FSAR 7.2.3.2- Specific Control and Protectionfnteradtions

ý01 --

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages I Page 3 FSAR 7.3.3.4 - Manual AFW Flow Control DuringPlantShutdown FSAR Section 7.4.1 - AMSA C FSAR Section 10. 1 - Steam and Power Conversion System FSAR Section 10.2 -Auxiliary Feedwater FSAR Figure 10.2-1 Sheet I - Bech M-217 Sh. I - Auxiliary FeedwaterSystem FSA R Figure 10.2-1 Sheet 2 - Bech Mf-217 Si 2 - A uxiliary FeedwaterSystem FSAR Section 14. 1. 10 - Loss ofNormal Feedwater FSAR Section 14.1.11 -Loss ofAllAC Power to the Station Auxiliaries FSAR Section 14.2.4 - Steam GeneratorTube Rupture FSAR Section 14.2.5 - Rupture of a Steam Pipe FSAR AppendL A. 1 - Station Blackout FPER 5.2.2 - Safe Shutdown Systems and Equipment FPER 5.2.5.2.3 -Auxiliary FeedwaterPump Room Tech Spec 3.7.5 - A muxiliary Feedwater Tech Spec Bases B 3.7.5 - Auxiliary Feedwater 13 Does the proposed activity involve a change to any Technical Specification? Changes to Technical Specifications require a License Amendment Request (Resource Manual Section 5.3.1.2).

Technical Specification Change: El Yes 0 No If a Technical Specification change is required, explain what the change should be and why it is required.

1.4 Does the proposed activity involve a change to the temis, conditions or specifications incorporated in any VSC-24 cask Certificate of Compliance (CoC)? Changes to a VSC-24 cask Certificate of Compliance require a CoC amendment request.

- ]Yes ED No If a storage cask Certificate of Compliance change is required, explain what the change should be and why it is required.

-_ 10 CFR 50.59 SCREENING PART 11 (50.59) - DETERMINE IF THE CHANGE INVOLVES A DESIGN FUNCTION (Resource Manual 5.3.2)

Compare the proposed activity to the relevant CLB descriptions, and answer the following questions:

YES NO QUESTION 0 C3 Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?

0] El Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?

0] El Does the proposed activity involve SSCs whose failure could initiate a transient (e.g., reactor trip, loss of feedwater, etc.) or accident, OR whose failure could impact SSC(s) credited in the Safety Analyses?

0] El Does the proposed activity involve CLB-described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications?

El 0 Does the activity involve a method ofevaltation described in the FSAR?

El 0] Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)

El 0] Does the activity exceed or potentially affect a design basis limitfor afissionproduct barrier(DBLFPB)?

(NOTE: If THIS questions is answered YES, a 10 CFR 50.59 Evaluation is required.)

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 4 I. t.

Sffthe answers to AL.L ofthese questions are NO, mark Part III as not applicable, document the 10 CFR 50.59 screening in the conclusion section (Part IV), then proceed directly to Part V - 10 CFR 72.48 Pre-screening Questions.

If any of the above questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.

The minimum recirculationlinesfor the auxiliaryfeedwaterpumps and the recirculationheaderfor the auxiliaryfeedivaterpumps have thefinction ofprovidingrecirculationflowpathsfromthe auxiliaryfeedwaterpumps to prevent hydraulicinstabilitiesandto dissipate pump heat. Hydraulic instabilitiesare prevented by the presence offlow restrictingorifices in the individualAFW pump recirculationlines, so thefunction ofmaintainingindividualpump coolingand the impact ofdiverted CST water are the concerns.

PART III (50.59) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (Resource Manual 5.3.3)

If ALL the questions in Part II are answered NO, then Part III is E] NOT APPLICABLE.

Answer the following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a 10 CFR 50.59 Evaluation is required; EXCEPT where noted in Part II1.3.

111.1 CHANGES TO THE FACILITY OR PROCEDURES YES NO QUESTION

[] 0] Does the activity adversely affect the designfunction of an SSC credited in safety analyses?

1] 0] Does the activity adversely affect the method of performing or controlling the designfunction of an SSC credited in the safety analyses?

If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion as necessary):

The current licensingbasis methodfor maintainingand controllingadequate AFW pumpflowforpump cooling was establishedin EVAL 2002-005 andassociatedFSAR change request FCR 02-019. 50.59 evaluationEVAL 2002-005 evaluated the procedure changes which added directionsto operatorsto respondto a recirculationvalvefailure or a loss of instrument air,and the impacts on the AFW pump and motors ifthe AFWpumps had to be securedand restartedto maintain SG level. The proposedmethod of maintainingAFW system operabilityutilizing reliefvalve AF-4035 and having a dedicatedoperatorto notify the control room if the valvefails to openprovides prompt notification to the control room operatorsthat the recirculationheader is notfunctional,and therefore there is no adverse effect on manual control ofAFW pump forwardflow to ensure AFW pump cooling is maintained If the recirculationheaderreliefvalve opens, recirculationwaterflow will be divertedfrom the CST. The loss of CST inventory due to the valve opening and recirculationwater spilling only affects the non-safety-related water sourcefor the AFW system. The requiredminimum CST water is based on the station blackout (SBO) event. In a SBO event, only the steam -driven AFW pump is availableto provide AFWflow to maintainsteam generatorlevel. TS Bases B 3.7.6, Condensate Storage Tank (CS7) states, "...the minimum amount of water in the CST assures the capabilityto maintain the unit in MODE 3for at least one hour concurrentwith a Loss of all AC power, while then allowingsufficient operatoraction time to transfer AFWsuction to the service water system." A concurrentsinglefailure or design basis accident need not be assumedduring a station blackout event (FSAR Appendix A. 1, Station Blackout). The closing of the recirculationvalve can be creditedto preventflow diversion. RequiredAFWflowfor thefirst hour is approximately 200 gpm (see FSAR 14. 1.1, for example) which is well above the value requiredfor pump cooling. After thefirst hour, service water will be the source of AFW water which is essentiallya limitless source. Therefore, based on the above, maintainingAFWV operabilit in the manner described above does not have an adverse effect on the designfitnctions ofprovidingsufficient water to the steam generators orfor ensuringpump cooling Pn1u*.1;1'

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages i Page 5 1.. 11I.2 CHANGES TO A METHOD OF EVALUATION (If the activity does not involve a method of evaluation, these questions are Z NOT APPLICABLE.)

YES NO QUESTION El [I Does the activity use a revised or different method of evaluation for performing safety analyses than that described in the CLB?

0l El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in the CLB?

If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion, as necessary).

111.3 TESTS OR EXPERIMENTS Ifthe activity is not a test or experiment, the questions in III.3.a and III.3.b are 0 NOT APPLICABLE.

a. Answer these two questions first:

YES NO QUESTION E] El Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?

Cl El Are the SSCs affected by the proposed test or experiment isolated from the facility?

If the answerto BOTH questions inV.3.a is NO continue to III.3.b. If the answer to EITHER question is'YES, then describe the basis.

b. Answer these additional questions ONLY for tests or experiments which do NOT meet the criteria given in III.3.a above.

If the answer to either question in III.3.a is YES, then these three questions are Z NOT APPLICABLE.

YES NO QUESTION El El Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the CLB?

[E El Does the activity utilize or control an SSC in a manner that is inconsistent with the analyses or descriptions in the CLB?

El El Does the activity place the facility in a condition not previously evaluated or that could affect the capability of an SSC to perform its intended functions?

If any answer in III.3 .b is YES, a 10 CFR 50.59 Evaluation is required. If the answers in III.3.b are ALL NO describe the basis for the conclusion (attach additional discussion as necessary):

Pnc.s t 4:-

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages I Page 6 a.;*.:art IV - 10 CFR 50.59 SCREENING CONCLUSION (Resource Manual 5.3.4).

Check all that apply:

A 10 CFR 50.59 Evaluation is [] required or 0 NOT required.

A Point Beach FSAR change is 0] required or [] NOT required. If an FSAR change is required, then initiate an FSAR Change Request (FCR) per NP 5.2.6.

A Regulatory Commitment (CLB Commitment Database) change is E] required or [D NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.

A Technical Specification Bases change is [I required or 0 NOT required. If a change to the Technical Specification Bases is required, then initiate a Technical Specification Bases change per NP 5.2.15.

A Technical Requirements Manual change is I[ required or 0D NOT required. If a change to the Technical Requirements Manual is required, then initiate a Technical Requirements Manual change per NP 5.2.15.

10 CFR 72.48 SCREENING NOTE: NET 96-07. Appendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 72.48 screenings.

PART V (72.48) - 10 CFR 72.48 INITIAL SCREENING QUESTIONS Part V determines if a full 10 CFR 72.48 screening is required to be completed (Parts VI and VII) for the proposed activity.

NO QUESTION SYES El 0 Does the proposed activity involve IN ANY MANNER the dry-fuel storage cask(s), the cask transfer/transport equipment, any ISFSI facility SSC(s), or any ISFSI facility monitoring as follows: Multi-Assembly Sealed Basket (MSB), MSB Transfer Cask (MTC), MTC Lifting Yoke, Ventilated Concrete Cask (VCC), Ventilated Storage Cask (VSC), VSC Transporter (VCST), ISFSI Storage Pad Facility, ISFSI Storage Pad Data/Communication Links, or PPCSIISFSI Continuous Temperature Monitoring System?

El 0] Does the proposed activity involve IN ANY MANNER SSC(s) installed in the plant specifically added to support cask loading/unloading activities, as follows: Cask Dewatering System (CDW), Cask Reflood System (CRF), or Hydrogen Monitoring System?

El 0] Does the proposed activity involve IN ANY MANNER SSC(s) needed for plant operation which are also used to support cask loadinglunloading activities, as follows: Spent Fuel Pool (SFP), SFP Cooling and Filtration (SF),

Primary Auxiliary Building Ventilation System (VNPAB), Drumming Area Ventilation System (VNDRM),

R.E-105 (SFP Low Range Monitor), RE-135 (SP High Range Monitor), RE-221 (Drumming Area Vent Gas Monitor), RE-325 (Drumming Area Exhaust Low-Range Gas Monitor), PAB Crane, SFP Platform Bridge, Truck Access Area, or Decon Area?

El 0] Does the proposed activity involve a change to Point Beach CLB design criteria for external events such as earthquakes, tornadoes, high winds, flooding, etc.?

El 0] Does the activity involve plant heavy load requirements or procedures for areas of the plant used to support cask loadinslunloading activities?

El z Does the activity involve any potential for fire or explosion where casks are loaded, unloaded, transported or stored?

If ANY of the Part V questions are answered YES, then a full 10 CFR 72.43 screening is required and answers to the questions in Part VI and Part VII are to be provided. If ALL the questions in Part V are answered NO, then check Parts VI and VII as not applicable. Complete Part VIII to document the conclusion that no 10 CFR 72.48 evaluation is required.

flf. I ; 4

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 7

-'.?ART VI (72.48) - DETERMINE IF THE CHANGE INVOLVES A ISFSI LICENSING BASIS DESIGN FUNCTION (If ALL the questions in Part V are NO then Part VI is [D NOT APPLICABLE.)

Compare the proposed activity to the relevant portions of the ISFSI licensing basis and answer the following questions:

YES NO QUESTION El El Does the proposed activity involve cask/ISFSI Safety Analyses or plant/caskIISFSI structures, systems and components (SSCs) credited in the Safety Analyses?

0l E3 Does the proposed activity involve plant, cask or ISFSI SSCs that support SSC(s) credited in the Safety Analyses?

El El Does the proposed activity involve plant, cask or ISFSI SSCs whose function is relied upon for prevention of a radioactive release, OR whose failure could impact SSC(s) credited in the Safety Analyses?

[E E] Does the proposed activity involve cask/ISFSI described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, CoC conditions, or orders?

El El Does the activity involve a method of evaluation described in the ISFSI licensing basis?

El El Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)

El El Does the activity exceed or potentially affect a cask design basis limitfor afissionproduct barrier(DBLFPB)?

(NOTE: If THIS questions is answered YES, a 10 CFR 72.48 Evaluation is required.)

If the answers to ALL of these questions are NO, mark Parts VII as not applicable, and document the 10 CFR 72.48 screening in the conclusion section (Part VIII).

Ifany of the above questions are marked YES. identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.

PART VII (72.48) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (NEI 96-07, Appendix B, Section B.4.2.1)

(If ALL the questions in Part V or Part VI are answered NO then Part VII is [] NOT APPLICABLE.)

Answer the following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a 10 CFR 72.48 Evaluation is required; EXCEPT where noted in Part VII3.

VII.1 Changes to the Facility or Procedures YES NO QUESTION El El Does the activity adversely affect the designfunction of a plant, cask, or ISFSI SSC credited in safety analyses?

El El Does the activity adversely affect the method of performing or controlling the designfunction of a plant, cask, or ISFSI SSC credited in the safety analyses?

If any answer is YES, a 10 CFR 72.48 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion, as necessary):

PlBF-- 51 5

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 8

!tlI.2 Changes to a Method of Evaluation (If the activity does not involve a method of evaluation, these questions are Z NOT APPLICABLE.)

YES NO QUESTION El El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in a cask SAR?

El [3 Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in a cask SAR?

If any answer is YES. a 10 CFR 72.48 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion, as necessary):

VII.3 Tests or Experiments (If the activity is not a test or experiment, the questions in VII.3.a and VII.3.b are ED NOT APPLICABLE.)

a. Answer these two questions first:

YES NO QUESTION D El Is the proposed test or experiment bounded by other tests or experiments that are described in the cask ISFSI licensing basis?

."l El Are the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISFSI facility?

If the answer to both questions is

• continue to VII.3.b. If the answer-to EITHER question is YES, then briefly describe the basis.

b. Answer these additional questions ONLY for tests or experiments which do not meet the criteria given in VII.3.a above.

If the answer to either question in VII.3.a is YES. then these three questions are 0 NOT APPLICABLE:

YES NO QUESTION El El Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the ISFSI licensing basis?

El El Does the activity utilize or control a plant, cask or ISFSI facility SSC in a manner that is inconsistent with the analyses or descriptions in the ISFSI licensing basis?

El El Does the activity place the cask or ISFSI facility in a condition not previously evaluated or that could affect the capability of a plant, cask, or ISFSI SSC to perform its intended functions?

If any answer in VII.3.b is YES, a 10 CFR 72.48 Evaluation is required. If the answers are all NO, describe the basis for the conclusion (attach additional discussion as necessary):

PBF-1515c Rer,-nee NqP-, I It

Point Beach Nuclear Plant SCR 2002-0377 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 9 I..?ART VIII - DOCUMENT THE CONCLUSION OF THE 10 CFR 72.48 SCREENING Check all that apply:

A 10 CFR 72.48 Evaluation is E] required or 0 NOT required. Obtain a screening number and provide the original to Records Management regardless of the conclusion of the 50.59 or 72.48 screening.

A VSC-24 cask Safety Analysis Report change is [] required or 0Z NOT required. Ifa VSC-24 cask SAR change is required, then contact the Point Beach Dry Fuel Storage group supervisor.

A Regulatory Commitment (CLB Commitment Database) change is E] required or (0 NOT required. Ifa Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.

A change to the VSC-24 10 CFR 72.212 Site Evaluation Report is [] required or 0 NOT required. If a VSC-24 10 CFR 72.212 Site Evaluation Report change is required, then contact the Point Beach Dry Fuel Storage group supervisor.

PBF-15 l .c Rdlcren*.e "NP'4 1 S,...... I AlfV );'111

Point Beach Nuclear Plant DESIGN INPUT CHECKLIST Sdification or Temporary Modification Number: MR 02-029

Title:

AUX FEED MINI RECIRC SAFETY UPGRADE / REMOVE AF-117 INTERNALS INSTRUCTIONS: Consider the basic functions of each structure, system, and component, (SSC), when answering the questions. The designer shall check the appropriate box for each design input or section. All inputs that apply to the design shall be explained. The explanation may be documented on this checklist or in the design summary. The reviewer shall review the checklist, and any differences between the designer and the reviewer should be addressed. This checklist addresses most design concerns, but is not all encompassing. Any additional concerns should be addressed in the design summary.

(Updates to this form covered by SCR 97-411.)

APPLIES TO DESIGN YES NO A. General codes, standards, regulatory requirements, and design criteria.

1. Are any of the PBNP FSAR general design criteria applicable? (Reference FSAR, Section 1.3.

Identify and address design criteria as appropriate.)

GDC1, Quality Standards- The AOVs are alreadysafety related. No physical changes are requiredto these valves. The AF-117 valve is not safety related.

GDC2, PerformanceStandards- The recirculationpiping is already analyzed as seismic Class 1 up to the manual valves (AF-1527/40/53),and is non-seismic downstream to the CSTs. This classificationupgrade will not affect the seismic analysis.

GDC 4, Sharing of Systems - The recirculationpiping is common to both units. The check valve internalsare being removed to prevent a single activefailurefrom affecting allfour of the AFW pumps.

GDC 5, Records Requirement - This modiftcation package meets this requirementsfor records.

GDC37, Engineered Safety Features Basisfor Design -The safety upgrade will improve the ability of the AFW pumps to perform their emergency coolingfunction.

GDC38, Reliability and Testability of EngineeredSafety Features- The recirculationline allows for testing of the AFW pumps without sendingflow to the steam generators.

GDC41, EngineeredSafety FeaturesPerformance Capability - The recirculationline will now supportthe operation of the AFW pumps to prevent a common mode failureof all AFWpumps.

The AF-117 check valve internalswill be removed to support this.

2. Are any design requirements contained in commitments affected? (Reference CLB database and the 0 n Safety Evaluation/Screening associated with this change.)

i*., Upgrade is being performed in accordancewith letter NRC 2002-0068, which committed PBNPto i - upgradethe open function of the AFWpump mini-recircAOVs to safety related. See 50.59 SCR 2002-0359 and SCR 2002-0377.

3. Meet State of Wisconsin Administrative Code requirements? (Refer to ILHR 41.42, PSC 114, and 0 other sections as appropriate for requirements.)

4. Meet existing DNR permits or require DNR approval? (Contact WE Environmental Department.) El 0

5. Consider the effect of design and accident conditions, such as pressure, temperature, fluid chemistry, El and radiation on components, including internal elastomers and material coating compatibility PBF- 1584 P-1cf. I ný I I R etrent e NP722

DESIGN INPUT CHECKLIST APPLIES TO DESIGN YES NO (Changes in design parameters may impact Environmental Qualification.)

6. Incorporate new types/models of equipment not presently used at PBNP? (Contact EPIX coordinator.) El 0

7. Affect accessibility of any equipment? Consider interim conditions, future maintenance, and in-service inspection. (Reference CIMs and drawings for manufacturer's clearance requirements.)

8. Require breaching a High Energy Line Break (HELB) barrier? (Reference NP 8.4.16) If yes, EQ engineer review required.

9. Consider operating experience from PBNP and industry events. (Reference DG-G04 for operatingCE experience reviews and NPRDS, NODIL, CHAMPS, INPO Keywords, or other databases.)

Discussionwith similarvintage plants has confirmed that a single non-safety relatedrecirculation line has been used to supportAFW pump operation.

10. Consider failure effects on structures, systems, and components: (Contact the NSA-PSA group for guidance and scope).

a. The design discusses those events/accidents which the system/components are to withstand? 0 El The AFW recirculationline is requiredto supportAFW pump operation during events where AFW is requiredto provide reactorheat removal, Most of the recirculationpiping is non Seismic, as are the CSTs, but afailure would be conservative in terms of recircflow.

'"

• b. The failure effect of the system/components: (Reference the NSA-PSA Group, Operating Experience, & IEEE-352-1975.) El 9 How components may fail, and the effect of the failure on the system and related systems?

• What mechanisms might produce failures? (Consider both equipment and human induced failures.)

• How a failure would be detected?

• What provisions are included to compensate for the failure?

AF-117 internals are being removed toprevent a single activefailurefrom making alIAFW pumps inoperable.

11. Does the design add or remove components in containment?

a. Change the amount of exposed aluminum or zinc in containment? (Reference DG-G07 and El El FSAR Section 5.6.)

b. Introduce materials into containment that could affect sump performance or lead to equipment degradation? (Reference DG-G07.)

c. Decrease free volume of containment? El El

d. Require addition or modification of a containment penetration boundary? (Consult the E containment system engineer.)

e. Require painting in containment? (Reference MI 36.3.) El El PBF-1584 Revtsion t0 081I9/02 Paee 2 of 13 Reference NP 7 2 2

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO (7-" 12. Consider potential for fuel failure?

ED

a. Affect fuel handling equipment?

El 1:1

b. Present the potential for introducing foreign material/debris into the RCS or connected systems?

El []

c. Affect core barrel flow patterns? ("Baffle jetting" concerns) o3

13. Meet requirements to abandon equipment if applicable. (Reference NP 7.1.5)

B. Mechanical requirements. (Contact Mechanical Design Engineering for guidance.)

1. Have applicable ASME Boiler & Pressure Vessel codes or other standards been identified?

(Reference the applicable specification. In addition, safety-related components should be reconciled with DG-M16, and QA components should be reconciled with ANSI N45.2.)

Installation will beperformed in accordance with USAS B31.1 -1967.

2. Affect or add components/systems to ASME Section XI class 1, 2, or 3 equipment? (Reference PBNP CHAMPS, CBD drawings, and IST Coordinator. If YES, follow NP 7.2.5, Repair/Replacement El 0 Program.)

3. Require State of Wisconsin Administrative Code permits/approvals? (Reference NP 7.4.9, Wisconsin Administrative Code for Boilers and Pressure Vessels or the Authorized Inspector.)

El

4. Consider component performance requirements such as capacity, rating, output? El 0]

5. Consider hydraulic requirements such as pump net positive suction heads, allowable pressure drops, 0]

allowable fluid velocities and pressures, valve trim requirements, packing/seal requirements?

El 1:1 0]

6. Provide vents, drains, and sample points to accommodate operational, maintenance and testing needs?

7. Require service water? (Both essential and nonessential service water loads are modeled, and load changes must be evaluated. Contact the SWAP Coordinator.)

El

8. Require the addition of check valves? (Reference DG-M13 for selection guidance.)

9. Require and evaluate any additional loading on instrument or service air, circ, fire protection, or demineralized water, or other system?

PBF-1584 Reviston 10 08119/02 Page 3 of 13 Reference NP 7 2.2

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO

\k- J

10. Evaluate any additional loading on HVAC systems or affect ventilation flow during or after installation? (This will require an EQ review for potential updates to EQSS, EQML & EQMR.) [] [

I1. Affect ventilation barriers, including containment, primary auxiliary building, or control room? [] [

12. Require insulation? (Reference WE specification PB-485 for insulation, and NP 1.9.10 for asbestos [] [

control.)

13. Require lubrication? (Reference Lubrication Manual.) [] [

14. Require an independent means of pressure relief? (Reference B31.1.) El [

El

15. Affect the assigned system design pressure or temperature?

[] [

16. Involve cobalt-laden materials into the RCS or into systems that supply the RCS? (Reference NP 4.2.29, "Source Term Reduction Program.")

[] [

17. Are new materials and their coatings/plating compatible with system chemistry and disposal systems

(. (NP 8.4.15)?

[] [

18. Affect embedded or buried piping?

El [

C. Electrical requirements. (Contact Electrical Design Engineering for guidance.)

1. Consider design conditions such as ampacity, voltage drop?

[] [

2. Consider component and system performance requirements, such as current, voltage, or power?

El [

3. Consider redundancy, diversity and separation requirements of structures, systems and components?

(Reference DG-E07 for separation of electrical circuits.)

El [

4. Comply with protective relaying requirements of equipment and systems?

El [

5. Selection of overcurrent devices for proper protection and coordination? (Reference DG-E04 for selection of molded case circuit breakers.)

El [

6. Affect available fault current at any bus?

PBF-1584 n .. : ._- ,, ,oi n,.,l-, PR _gc 4 of I I Reterence NP 7 2 2

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO

7. Assure that all added cables meet fire retardancy requirements? (Reference FPER Section 4.1.8, El0 IEEE 383.)

El 0]

8. Be compatible with existing electrical insulation and wiring?

El 0]

9. Affect ampacity of existing cables?

El 0]

10. Maintain UL (or equivalent) listings?

11. Alter the voltage harmonic distortion content or change the non-linear loading (i.e., the addition of switching power supplies, the alteration of the circuit's power factor, etc.) on a vital or sensitive instrument bus?

12. Add new raceways? (Reference DG-E03 for electrical raceway sizing and DG-E02.) El 0]

El 0]

13. Add cables to existing electrical raceways?

El 0]

14. Be routed through fire wrapped raceways?

0]

15. Affect the station grounding or lightning protection system?

El El 0]

16. Make any vital circuit susceptible to ground?

El 0]

17. Affect emergency diesel loading? (Contact Electrical Design Analysis group for guidance.)

El 0]

18. Add more station battery loading?

0]

19. Add load to a vital bus?

El El

20. Add load to a non-vital bus?

0]

El

21. Be compatible with service transformer capacity?

coupling

22. Consider electromagnetic interference between new/existing equipment and electromagnetic interactions between circuits?

PBF-1584 Pace 5 of 13 Reterence NP 7 2 2 v-.; in nvoivn'

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO

(-,- 23. Affect embedded conduits or buried cables, including the station grounding system? El D. Instrumentation and control requirements. (Contact I&C Design Engineering for guidance.)

1. Consider design conditions such as pressure, temperature, fluid chemistry, amperage, voltage? E 0

2. Have the instruments been properly selected for the application? El 0

3. Have sufficient instruments for operators to monitor the process? El 0

4. Have appropriate instrument scales? El 0

5. Have the instruments, control switches, and indicating devices been appropriately located for human E0 0 factors (both for operations and maintenance)? (Reference DG-GO 1.)

6. Have alarms for off-normal conditions? El 0

7. Be capable of or require remote and/or local operation? El 0

8. Be capable of or require manual and/or automatic operation? El 0

9. Require calibration and maintenance requirements for the instruments to be specified? El 0

10. Have specified the instruments with proper range and accuracy? El 0

11. Address solid state vulnerability to RFI? El 0

12. Consider software and programming/programmable settings of digital or electronic equipment? El 0

13. Affect logic circuits or associated GL 96-01 review/required testing? Contact I&C System Engineering group.

PBF-1584 Pnae 6 of 13 Reteren.c NP7 2 2 Revisio)n' l10 "8,19,02.;

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO

(>.. Structural requirements. (Contact Civil Design Engineering for guidance.)

1. Affect or scope seismically qualified equipment (Class I or 2) and therefore require a seismic qualification evaluation? (Reference NP 7.7.2, "Seismic Qualification of Equipment.")

El

2. Affect seismic boundaries? 0

3. Affect stress calculations of pipe? (Reference DG-M09.) El The removal of the internalswill have a negligible effect on the piping analysis caused by the slight weight reduction.

4. Affect the loading or require changes to existing equipment foundations? El 0]

5. Affect wall stress calculations for pressurized concrete cubicles or structures?

El 0]

6. Require analysis of non-seismic components placed over or adjacent to seismic components?

El 0]

7. Add items which span between two separate seismic areas/buildings? (The effect of the relative El 0]

movement must be addressed.)

8. -Require clearance review for seismic movement or thermal expansion considerations?

[] 0]

9. Require a floor or wall loading analysis? (Reference Bechtel C-dwgs.)

El 0]

10. Require the addition of new supports, hangers, or foundations or add weight to or between existing supports, hangers, embeds, or foundations during installation or post-installation? (Reference El DG-M09 and DG-M10 for pipe support.)

11. Add new or add load to seismically qualified raceways? (Reference NP 7.7.2, "Seismic Qualification of Equipment.")

12. Modify, attach to, or locate within the proximity of masonry block walls? (Reference IEB 80-11 Block Wall Program.)

13. Require core drills, expansion anchors, or re-bar cuts? (Reference DG-C0l for expansion anchor El design and installation.)

14. Create an external or internal missile hazard?

El

15. Consider wind and storm loading on external structures?

El PBF-1584 p ......... M MOM* Paue 7 of 13 Reference NP 7 2 2

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO

' 16. Require protection from high energy line breakjet? (Refer to FSAR Appendix A.2.) El 0

17. Consider dynamic requirements such as live loading, vibration, and shock/impact? El 0 F. Programs

1. ASME Section XI and QA considerations:

a. Affect IST acceptance criteria or calculations? (Contact Component Engineering.) 0 El The 1I2AF-4002, AF-4007, and AF-4014 valves will now have acceptancecriteriafor stroking open. Check valves 1/2AF-114,AF-115, and AF-116 have acceptance criteriato passflow. Check valves 1/2AF-173 will be given leakage acceptancecriteria.

b. Require classification of new components? (Reference DG-G06 for system, component, and El 0 j(Y part classification.)

c. Affect QA-scope systems or boundaries? (Contact Site Programs Engineering Support for 0 El Q-List.)

The safety-relatedboundary in the recirculationpiping is not being moved. The boundary Awill remain at theflow restrictingorifices (112R0-4003, RO-4008, RO-4015). All components upstream are currently safety related (QA Code 04). QA Code 06 will be added to the Q.listfor the mini-recircAOVs and check valves, since they now have afunction to provide cooling waterflow to the AFW pumps.

d. Require special personnel/equipment qualifications not proceduralizecdat PBNP (i.e., El 0 underwater welding)?

e. Require material certification or other certification to ensure quality equal to or better than the affected SSC? (These requirements need to be specified in the specification or purchase El 0 requisition.)

f. Have all design requirements, such as pressure or current rating, been reviewed against lot El 0 descriptions or been specified on purchase requisitions/specifications?

2. Fire protection considerations:

a. Affect access to a fire zone, fire protection equipment or Appendix R safe shutdown equipment, El 0 including manual fire fighting activities? (Reference Section 5.2.1 of Design Guide DG-F01)

b. Affect a fire barrier? (Reference NP 8.4.11 and Fire Barrier Drawings WE PBC-218 El 0 Guide DG-F01)

Sheets 1-20, Section 5.2.2 of Design C. Affect a fire protection system or its performance? (Reference Section 5.2.3 of Design Guide El 0 DG-F01)

PBF-1584

..1. ... "RIMM. of I RclerenLC NP 7 2 2

--- 8 *4 ...... I I **

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YE...S N._O

d. Increase or decrease permanent combustible loading in a room? (Reference Section 5.2.4 of Design Guide DG-F01)

e. Based on Section 2 and Appendix A of the SSAR, will the change add to, delete from, or affect the performance of safe shutdown systems or equipment? (Reference Section 5.2.5.1 of Design fl Guide DG-F01)

The AFW system is a safe shutdown system creditedfor Appendix R. The mini-recircAOVs are creditedto be opened in this scenarioto provide AFW pump cooling. Upgradingthe safety function of the AOVs and removing the AF-117 internalsimproves this capability.

f. Based on Sections 3, 4, and Appendix C of the SSAR, will the change affect a cable associated with safe shutdown equipment, a safe shutdown power supply, or the physical location of a safe El 0 shutdown cable? (Reference Section 5.2.5.2 of Design Guide DG-FO1)

g. Based on Table 1-1, Section 5 and Appendix D of the SSAR, will the change affect fire area analysis and compliance with Appendix R separation criteria or the conditions of an approved Appendix R exemption for any PBNP Fire Area? (Reference Section 5.2.5.3 of Design Guide DG-FO1, Table 3.2-2 of DBD T-40)

b. Will the change add, remove, or affect the performance of any emergency lighting required for compliance with Section III.J of Appendix R? (Reference Section 5.2.6 of Design Guide DG- El F01)

Will the change add, remove, or affect the performance of any plant communications system relied upon for fire fighting or safe plant shutdown? (Reference Section 5.2.7 of Design Guide El 0 DG-F01)

j. Will the change affect the Reactor Coolant Pump Oil Collection System? (Reference El 09 Section 5.2.8 of Design Guide DG-FO1)

k. Will the change affect the Fire Protection Manual?

1. Will the change affect any of the Supporting Documents listed in the SSAR (Section 6.0) or the El 0 FHAR (Section 4.0)?

If any of the questions a through j are answered "yes", an evaluation must be performed using the applicable sections of the FPCC checklist, PBF-2060 per Section 5 of Design Guide DG-FO0.

3. Flooding protection considerations:

A flooding analysis should be performed if any of the following questions are applicable and answered yes. (Reference Section 4.3 of DG-C02.)

a. Modify potential flooding sources or add new potential flooding sources to a flood zone and thereby increase the direct and/or indirect flooding vulnerability of essential equipment'?

By removing the AF-117 internals,it is now possiblefor a failure of the AF-4035 relief valve to drain the CSTs andpotentially create a floodingproblem in the Unit I turbine building.

PBF-1584 Revimon 10 0,JI19/02 Page 9 of 13 Relerence NP7 2 2

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES N__O However, theflow rate would be limited through the reliefvalve, depending on the nature of thefailure. Even if the relief valve was passing its design capacity (268 gpm), this water "wouldadequately drainfrom the 8' Unit I turbine building throughfloor drains,subsoil drains,and under doors to the exterior. This flooding would not be as severe as afailure of a service water or circulatingwaterpipe, and can therefore be consideredto be bounded by these events.

b. Degrade existing flood barriers or flood mitigation features providing unanalyzed pathway for flooding to propagate? (Reference Section 3.2 of DG-C02.)

c. Involve the opening of potential flood sources anywhere at the station? (Installation procedures need to address inadvertent flooding. Reference DG-C02, Section 4.4.)

The recirculationline connects to the CSTs. If leakage exists past the AF-2 or AF-9 valves, then the potentialexists for drainingthe CST into the Unit I turbine building. Precautions will be added to the work plan.

d. Reduce the capacity to isolate or cope with flooding? (Reference Sect. 4.2 of DG-C02.) El 0

e. Change plant drainage/backfill requirements? Q

f. Locate essential equipment or supporting systems where it would be susceptible to flooding?

(Flooding conditions may also impact Environmental Qualification.)

r". 4. Environmental considerations:

a. Be subject to adverse environmental conditions during storage or construction? (Reference [i. 0 NP 9.5.2.)

b. Require freeze protection or affect existing freeze protection? El 0

c. Locate safety-related or post accident monitoring equipment in a HARSH environment? El 0 (Reference NP 7.7.1.)

d. Require Environmental Qualification (EQ)? (Reference NP 7.7.1 for EQ qualification.) El 0

e. Be attached to an EQ system/component? (This will require an EQ review for potential updates El 0 to EQSS, EQML & EQMR. Reference EQ master list.)

f. Change environmental parameters (e.g., pressure, temperature, radiation, humidity)? (Reference El 0 Electrical Equipment."

NP 7.7.1, "Environmental Qualification of

5. Radiation Protection (RP) and ALARA considerations: (Reference DG-G03, "ALARA Consideration Guideline for Design & Installation.)

The areas mentioned below are normally within the RCA, but radiological concerns should be considered for SSC outside the RCA also.

PBF-1584 V-.- in nQ,1iwrn Pa.e 10 of 13 Reference NP 722

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YE...S N...O

a. Affect any SSC in an RWP required area, a contaminated area, or a radiation area, including tI4 opening of a system that may be a radiological concern?

b. Will the change generate excessive radwaste or highly radioactive/contaminated waste? El 0

c. Remove any plant equipment from a potentially contaminated system (including BOP systems)? El 0

d. Result in an anticipated increase in operational or maintenance exposures?

(Consider equipment rearrangement to reduce plant life dose?)

e. Result in an expected exposure of greater than I rem for any individual during installation of the change?

L. Result in an anticipated collective exposure of greater than 2 rem for the installation of the change?

If questions d, e, or f apply and are answered yes, then an ALARA review shall be performed.

(Reference NP 4.2.1, Plant ALARA Program.)

,'.: 6. Chemistry considerations:

a. Require or affect established chemistry limits? (Contact system engineer and review chemistry procedures.)

b. Require any routine chemical analyses? (Contact system engineer and review chemistry El 0 procedures.)

c. Require chemical additives? (Contact PBNP Chemistry.) El 0

d. Do new fluids/chemicals need to be evaluated for TRI (Toxic Release Inventory), Control Room habitability, CHES, critical applications, or special disposal requirements? (Contact El 0 Chemistry/Chemical Engineering.) Reference OE 11400, RG 1.78 and NP 3.1.6.

G. Installations

1. Installation requirements/plant conditions have been determined? 0 El The common recirculationline will be taken out of servicefor the check valve removal The AFW pumps will be consideredoperable during this installationby taking creditfor relief valve AF-4035 to open andprovide adequateflow for pump cooling (perCalculation2002.0026). Additionally, a level 3 dedicated operatorwill be present to monitorthe relief valve, and notify the control room if it fails to open.

2. Consider test and inspection requirements, including the conditions under which they will be performed? (Reference NP 7.4.1 for pressure test requirements, NP 7.4.3 for post-maintenance and 0 El modification NDE rcquirements, NP 1.2 5 for special test procedures, and OM 4.2.2 for in-service PBF-1584 v.. nicfrll Paue I I o" 13 Reference NP722

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO tests.)

A non-ASME code leak check will be performedon the check valve gasket.

3. Have post-installation acceptance criteria been properly specified to test the intended function of the E]

component(s)/system?

4. Comply with all WE lifting and rigging requirements? (Reference WE Safety Manual, PBNP Safe Load Path procedures, and NP 8.4.7.)

5. Consider ALARA for installation activities? (i.e., shielding, monitoring water level, etc.) El 0

6. Require special handling, shipping, or environmental conditions for storage or construction? 0 (Reference NP 9.5.2 for material storage.)

7. Consider transportability requirements such as size and shipping weight limitations. El 0

8. Require spare parts or special non-standard items or tools? El 0

9. Will any added components introduce chemical contaminants to the system? (i.e., preservative coating El 0

4. on valves, coatings on weld rod can also introduce contaminants)

10. Consider personnel requirements and limitations, including the qualification.and number of personnel available for plant operation, maintenance, testing and inspection, and permissible personnel radiation El 0 exposures?

11. Operational requirements under various conditions, such as plant startup, normal plant shutdown, plant emergency operation, special or infrequent operation, and system abnormal or emergency operation.

a. Require new procedures or procedure changes? (Reference NP 1.2.5.) 0 El Revisions to several IT proceduresarerequiredto add acceptancecriteriafor stroking the AOVs open, add an exercise testfor the recirc check valves, and add acceptance criteriafor the instrument airaccumulatorAF-173 check valves.

b. Potentially impact other systems, components, or structures during installation? 0 El The AFW system will be affected when the recirculationline is taken out of service.

c. Present installation impacts on plant operations (i.e., fire watches, etc.)? 0 El Per CM 3.26, a level 3 dedicated operatorwill be used to notify the control room if relief valve AF-4035fails to open automaticallyas needed.

12. Access and administrative requirements for plant security: If any security requirements are applicable, notify Security.

a. Create an opening >96 in. 2 in any wall, ceiling, or other barrier? El 0

b. Require work within 20' of fence? El 0 PBF-1584 Page 12 of 13 Reference NP 7 2 2 Revision 10 08/19/02

APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO YES NOf

c. Affect security equipment and documents, including those containing safeguards information?

(Contact Security for design development requirements and design concurrence.)

d. Affect access controls? El 0

13. Safety requirements:

a. Affect safety equipment and thereby create personnel hazards (i.e., removal of handrails)? El 0

b. Introduce hazardous material into the plant? (Reference NP 1.9.1.) Ei 0

c. Affect evacuation routes or escape provisions from enclosures? Ei 0

d. Meet OSHA regulations? (Reference Wisc. Electric Safety Manual and OSHA 29 CFR 1910.) 0 0 Installationwill be performed in accordancewith the safety manual

e. Move any energy sources? If yes, verify installation document covers move, including E transferring danger tags.

f. Require that equipment be grounded? El 0 Designed by: Rob Chapman Date: _ -,___-_

Reviewed by: Kevin Krause Date: 9 " o PBF-1584 Revision 10 08119/02 Page 13 of 13 Reference NP 7.2 2

NUCLEAR POWER BUSINESS UNIT DESIGN VERIFICATION NOTICE Title of Document AUX FEED MINI RECIRC SAFETY UPGRADE / REMOVE AF-117 INTERNALS

-Document No. MR 02-029 Rev. 0 Date 9/5/2002 Design Verification Method: Design Review [ [Alternate Caics L.. Qualification Testing UPDATES TO THIS FORM COVERED BY EXISTING SCR 97-410 REVIEWER CHECKLIST CONSIDERATIONS:

Yes No N/A

1. Were the inputs correctly selected and incorporated into design?

2. Are assumptions necessary to perform the design activity adequately described and reasonable? Where necessary, are the assumptions identified for subsequent reverifications when the detailed design activities are completed?

3. Are the appropriate quality and quality assurance requirements specified?

4. Are the applicable codes, standards, and regulatory requirements including issue and addends properly identified and are their requirements for design met?

5. Have applicable construction and operating experience been considered?

6. Have the design interface requirements been satisfied?

7. Was an appropriate design method used?

8. Is the output reasonable compared to inputs?

9. Are the specified parts, equipment and processes suitable for the required application?

10. Are the specified materials compatible with each other and the design environmental conditions to which the material will be exposed?

Have adequate maintenance features and requirements been specified?

11.

12. Are accessibility and other design provisions adequate for performance of needed maintenance and repair?

13. Has adequate accessibility been provided to perform the in-service inspection expected to be required during the plant life?

14. Has the design properly considered radiation exposure to the public and plant personnel? 7 _

15. Are the acceptance criteria incorporated in the design documents sufficient to allow 16.

verification that design requirements have been satisfactorily accomplished?

Have adequate pre-operational (IST, PMT, ISI, snubber, etc.), subsequent periodic test, 7

and inspection requirements been appr ately pecifed, including acceptance criteria?

17. Are adequate handling, storage, cleaning, an shipping requirements specified? U,- _ _

18. Are adequate identification requirements specified?

19. Are requirements for records adequately specified?

20. Will the change remain within the analyzed or specified capabilities of any affected equipment?

21. Has a field inspection been done?

22. Have impacts on other systems been identified?

COMMENTS: 2 None D Attached (Use lornmPBF-1633)

"DesignPrepared By: Rob Chapman Date C/--

Reviewed By: Kevin Krause -D ate

• Approval By: 49,13'A4 Date PBF-1583 Revision 1 08/08/97 Page I of I Reference- NP722

DOCUMENT UPDATE CHECKLIST Plant Modification/rvlinor Plant Change No.02-029 i Work Order No.:

DOCUMENTATION UPDATE SHEET AND CLOSEOUT CHECKLUST Required For N/A Release Acceptance Closeout (Completion) (Submittal)

A. TRAINING X 1. Copy Submitted to Training (Design Description)

X 2. TWR Generated (TWR # o2.- -2.L" ) Ref. SIMGL CI.1 X 3. Simulator Changes Initiated (SDR # 02- o0-'4 )

X 4. Plant Status Update/Just In Time Training B. FINAL DESIGN ORGANIZATION

1. Drawings X - a. Design Change In Progress DCI's Initiated X .' J b. Construction sketches Issued

.,, X c. Revised Drawings Issued for Priority I and 2 Control Room

_____ __ -Drawings - Logics, P&IDs, 499 series elementaries.

x ' d. Revised Drawings Issued for Work Control Center Drawings P&IDs X e. Revised Drawings Issued for I&C Drawings - Reactor Protection and Safeguards Elementaries.

X f. Master Dita Book - Control Room, Work Control Center, and Local Panel - PBF-2093 SXg. DCN's released for incorporation X "- h. Sketches Voided - PBF-1592 X 2. Specifications (Conformed at Closeout, ref NP 9.2.1)

X 3. Component Instruction Manuals (for issue, revision, deletion)

PBF-1586 X 4. Cable and Raceway Data Schedule Revisions - PBF-0091 X 5. Environmental Qualification Documentation Updates - Ref. NP

-. - 7.7.1 X - 6. Seismic Qualification Updates NP 7.7.2

- 7. Calculations or engineering evaluations added/deleted I revised PBF-1608 X 8. DBD Revisions - PBF-1653 X 9. PSA Models and Documentation - PBF-1626

10. EPIX Update - report Equipment changes/additions to the EPIX X Coordinator.

PBF-1606 Revision 6 10/02/01 Pag-e I of 4 Reference(s) NP 7 2 1

DOCUMENT UPDATE CHECKLIST Plant ModificationlMinor Plant Change No.02-029

[ Work Order No.:

SOrI IMFNTATION I IPnATF.MAFT*T ANnr ! £0FrT l rrHrKi jT Required For Release Acceptance Closeout (Completion) (Submittal)

C. LICENSING (Conformed at Acceptance)

X 1. Technical Specification - change; specify section(s) affected and change request number.

X 2. Tech Spec Basis/Technical Requirements Manual X 3. FSAR - change; NP 5.2.6. Report major changes to the containment aluminum inventory list with FSAR update.

X 4. FPER - FHAR - SSAR Revisions - NP 5.2.11 "x a. Safe Shutdown Analysis Management System Revisions - NP X _______:_____5. .... Offsite Dose Calculation Manual (ODCM)

X ___ -- 6. Radiological Effluent Control Manual (RECM) x,"-: - 7. Emergency Plan and EPIPs X 8. Notification to Security for Security plan update X 9. Report major changes to radwaste treatment systems with annual FSAR update per RECM 1.6.3 D. CHAMPS DATABASE X 1. Equipment Identification - additions assigned from CHAMPS X 2. Permanent Labeling - labels on new equipment; PBF-9900 X 3. Temporary Labeling - labels on new equipment; PBF-2074 x 4. Equipment Record - update to CHAMPS coordinator specify change(s); PBF-9922 X 5. Spare parts stocking and scrapping inputs into CHAMPS; PBF-9925, PBF-1023 X _'"-'"_" 6. Unused material removed from modification bin.

t , E. OPERATIONS

1. euligPoedrs-PB-0 Abnormal Operating, Normal Operating, System Operating, and

--. Refueling Procedures - PBF-0026a X 2. Operating Instructions and Checklists - PBF-0026a

- 3. Alarm Response and RMS Alarm Setpoint and Response Books X_-_____-__ PBF-0026a X 4. Testing - TS, IT, ORT, other - PBF-0026a X 5. EOPs, ECAs, CSPs, SAMGs - PBF-0026a X 6. Periodic Surveillances - PBF-9920 X 7. Fire Protection Procedures - PBF-0026a

8. EOP Setpoints, EOP Instrument Uncertainty Calculations PBF-8001

9. Tank Level Book - PBF-0026a PBF-1606 Pace 2 nf 4 Reference(s) NP 7 2 I

.1 DOCUMENT UPDATE CHECKLIST Plant Modification/Minor Plant Change No.02-029 t Work Order No.:

DOCUMENTATION UPDATE SHEET AND CLOSEOUT CHECKLIST Required For NIA Release Acceptance Closeout (Completion) (Submittal)

F. MAINTENANCE/I&C X 1. Maintenance Procedures/Instructions - PBF-0026a X 2. ICPs - PBF-0026a X 3. Setpoint Document - PBF-8001 x 4. Preventative Maintenance - initiate/revise CHAMPS callups; PBF-9921/9920 x - - 5. Ensure station batteries' load profile changes are incorporated into the appropriate discharge test RMPs.

X -, 6. Lubrication Manual (NP 7.3.11)

..- G. SECURITY x ""-' 1. Security Procedures H. ENGINEERING/MISC.

X 1. ISI Program X 2. IST Program X 3. Miscellaneous-HX ECT/Cleaning program "x 4. Reactor Engineering Instructions - change; specify section(s)

-. affected.

x " 5. Reactor Engineering Procedures - change; specify section(s) affected.

x 6. Software Control - specify system affected and software change

_______ _ request number.

- 7. Component maintenance programs.

x 8. Governing calculations and models (e.g., SW model, DC loading, EDG loading, piping analysis, structural loading, etc.).

X 9. Design Guidelines (ref. NP 7.1.2)

I. OTHER (CHEM, HP, ETC.)

X 1. Other (Misc. Procedures, etc.)

J. ECRs

1. ECR Final Resolution completed and approved by Design Supervisor.

2. ECR Implementation completed.

PBF-1606 Rere,'nr' e¢*" NP7 2 I

Prior to Section Specific Updates Required Prior to Prior to Change No.

Acceptance Closeout (if Applicable) /BY/ Date Release A.1 Design description submitted to LI training.

A.2 TWR 02-240 F] LI LI [2 t",i.*.4Z . *,. °-;-

c.,1 oz.- P*c.*.c A.3 Simulator Change Initiated LI LI A.4 JIT notified of modification D1 "CAt "S.- t* S 2- r B.1.a DCN toBECH M-217 Sh. I LI B.1.a DCN to ALOYCO A-46037 LI LI B.l.c Control room drawings issued FI LI Dq B.L.d WCC drawings issued ELI LI B.l.g DCN to ALOYCO A-46037 issued.

z LI B.7 Calculation 2002-0026 approved.

Lq ~ACV_

B.8 DBD-01, AFW, revised LI B.8 DBD-06, IA/SA, revised LI

. --9. FZ3 L-,o-..

B.9 PSA updated El . w- *C._

C.2 FSAR 9.7 revised LI LI "C.2 FSAR 10.2 revised C.3 TS B 3.7.5 revised L2 D.4 CHAMPS records updated LI D:

EA4 IT 08A revised LI [I E.4 IT 09A revised nI LI LI E.4 IT 10 revised

[L] LI LI]

E.4 IT 10A revised LI LI E.4 IT OB revised LI LI F.4 Delete callup M-A6 (for AF- 17)

LI Q~eJ rz "" ,, , O-3-02. *c.r, H.2 DST Background Doc Updated LI 1.1 New procedure IT 0oC issued LI 1.1 New procedure IT 09C issued T.1 CwvAPS cd-Lis-r (T209issue PBF-1606 Revtmon6 10/01/01 Pag~e 4 of 4 Referenccs) NP 7 2 1

Section Specific Updates Required Prior to Prior to Prior to Change No.

Release Acceptance Closeout (if Applicable) /By/ Date

• >')A. 1 Design description submitted to E[

training.

A.2 TWR 02-240 F71 _-'

A.3 Simulator Change Initiated "' ']

A.4 JIT notified of modification [] I __[--]

B.l.a DCN to BECH M-217 Sh. I ka- D' B.I.a DCN to ALOYCO A-46037 M- [-_

B.l.c Control room drawings issued M 0 B. .d WCC drawings issued [I L_ _ _ _

B.l.g DCN to ALOYCO A-46037 issued. [-0_ ___['_

B.7 Calculation 2002-0026 approved. R" 0-_

B.8 DBD-01, AFW, revised L' L- __

B.8 DBD-06, IA/SA, revised E- [_ _ _ _

B.9 PSA updated -] [7 [_

C.2 FSAR 7.3 revised -' -I C.2 FSAR 9.7 revised C.2 FSAR 10.2 revised - [I [_

C.3 TS B 3.7.5 revised LI L- __

D.4 CHAMPS records updated [] [] [_

E.4 IT 08A revised [- __ __--_

E.4 IT 09A revised L L___

E.4 IT 10 revised [] __ __-'_

E.4 IT I0A revised LI L__ _ _

E.4 IT 10B revised LI [__F_

F.4 Delete callup M-A6 (for AF-1 17) "- L__["_

H.2 IST Background Doc Updated El [-1 [__

1.1 New procedure IT 08C issued EI [L] __

1.1 New procedure IT 09C issued LI LI __

PBF-1606 I,- 'II

Point Beach Nuclear Plant FIRE PROTECTION CONFORMANCE CHECKLIST

"" MR Number MR 02-029 Unit 1 Unit 2 Common Facilities X System AUXILIARY FEEDWATER Location 26' El, Control Building AFFECTED FIRE ZONE(S)-FIRE AREAS FZ 320 / A01-E (see FPER Sect. 9)

PURPOSE The Fire Protection Conformance Checklist (FPCC) was developed to help evaluate the impact of plant modifications, procedural changes, and tests on the plant fire protection program and safe shutdown capability for compliance with 10 CFR 50 Appendix R and other plant fire protection license commitments.

The FPCC also provides the screening criteria to ensure that a 10 CFR 50.59 safety evaluation is performed on activities that affect the design basis of fire protection equipment or plant's capability to achieve and maintain safe shutdown for any design basis fire. If the FPCC screening indicates the plant fire protection or safe shutdown design basis will be affected, a 10 CFR 50.59 screening shall be performed per NP 10.3.1, Authorization of Changes, Tests, and Experiments (10 CFR 50.59), with consideration of the FPCC information, to determine if an unreviewed safety question is involved. The design basis fire is the accident to be considered in the 10 CFR 50.59 evaluation. The FPCC becomes part of the documentation supporting the 10 CFR 50.59 screening and safety evaluation.

The FPCC is comprised of this main form, PBF-2060 and sub-forms, PBF-2060a through h that address different topical areas of the PBNP fire protection program. The intent of multiple sub-forms is to eliminate unnecessary burden in completing forms for areas of fire protection clearly not affected by a particular change. Based upon the nature of the change (as identified by answers to the 6,...questions on the Design Input Checklist PBF-1584), the applicable sections on the FPCC will be filled out. The appropriate sections on the FPCC to be filled out shall be indicated below on the FPCC Applicability Matrix. The applicable sections that are completed will be attached to the main form PBF-2060 and included with the plant change package.

INSTRUCTIONS

1. Complete the FPCC Applicability Matrix below, based on the nature of the change and answers to questions on the Design Input Checklist PBF-1584 for the applicable change.

2. Complete the appropriate sub-forms, based upon the nature of the change as defined on the FPCC applicability matrix. It is not necessary to complete sub-forms for areas of fire protection that are clearly not affected by the subject change.

3. Use the paragraphs in Section 5 of Design Guide DG-FO1 that correspond to the FPCC sections for additional information and guidance when answering the questions in the checklist.

4. Consider requirements for a 10 CFR 50.59 screening by reviewing the RESULTS section below.

5. Ensure that the appropriate documents required for update (i.e., FPER, FHA, SSAR, SSAMS, FPDS, Calculations, FPEEs, etc.) are properly identified for future revision in the governing document update procedures. This includes documents that must be updated for changes that could potentially adversely affect fire protection conformance, as well as changes that are determined by the checklist not to adversely affect fire protection conformance (but still require document updates).

6. Sign and date the FPCC. If the NPBU Fire Protection Engineer is not the preparer of the FPCC, then the Fire Protection Engineer shall review, sign and date, the FPCC.

PBF-2060 Rev-  %"(yifl'n 1 Paie I of 2

FIRE PROTECTION CONFORMANCE CHECKLIST RESULTS If the completion of any FPCC screening from Sections 1.0 - 10.0 on forms PBF-2060a through 2060h indicates the modification has potential adverse impact, then the plant fire protection or safe shutdown design basis may be affected. A 10 CFR 50.59 screening must be performed per NP 10.3.1, Authorization of Changes, Tests, and Experiments (10 CFR 50.59),

with consideration of the FPCC information to determine if an unreviewed safety question is involved, the design basis fire is the accident to be considered in the 10 CFR 50.59 evaluation. The FPCC becomes part of the documentation supporting the 10 CFR 50.59 screening and safety evaluation.

Inform the NPBU Fire Protection Engineer if fire protection program commitments or compliance with 10 CFR 50, Appendix R will be affected.

Fire Protection Conformance Checklist Applicability Matrix Applicable? Section To2ic Design Input Action Yes No Checklist Section 1.0 Plant Access F.2.a Complete & attach PBF-2060a 2.0 Fire Barriers F.2.b Complete & attach PBF-2060b E3 N 3.0 Fire Protection Systems F.2.c Complete & attach PBF-2060c Combustible

[]

0 0[

4.0 Loading/Fire Hazards F.2.d Complete & attach PBF-2060d 0] 0 5.0 Safe Shutdown Systems F.2.e Complete & attach PBF-2060e and Equipment 6.0 Safe Shutdown Cables, F.2.f Complete & attach PBF-2060e Including Associated Circuits El ED 7.0 Fire Area Analysis, F.2.g Complete & attach PBF-2060e Including Exemptions/Evaluations El0 8.0 Emergency Lighting F.2.h Complete & attach PBF-2060f E50 9.0 Plant Communications F.2.i Complete & attach PBF-2060g El ED 10.0 Reactor Coolant Pump F.2.j Complete & attach PBF-2060h Oil Collection System Conformance checklist (including all applicable attachments) completed.

Comments: Removal ofAF-117 internalsand upgrading the AFW mini-recircAOVs to have an open safety-relatedfunction will improve AFW pump reliability.

By: Date: 0i-Z-Date PBF-2060 1.2 0tfftQff'l!

Point Beach Nuclear Plant FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5, 6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION Complete the evaluation (Sections 5.0, 6.0, and 7.0) and attach to form PBF-2060.

APPENDIX R SAFE SHUTDOWN EVALUATION 5.0 SAFE SHUTDOWN SYSTEMS AND EQUIPMENT (Ref. Section 5.2.5.1 of Design Guide DG-FOI) 5.1 Does the modification require addition of a safe shutdown component? Is the new component located within the Appendix R flowpath boundaries shown in the Appendix R Highlighted P&IDs, SSAMS Database, SSEL Module, or the SSAR Section 2, Safe Shutdown Logic Diagrams in Appendix B of SSAR.

LII Yes, go to 5.11, complete actions and resume at 5.2 Z No, go to 5.3 Comments:

5.2 Will the new component support other safe shutdown systems or component(s)? (Refer to SSAMS Database, SSEL Module, SSAR Section 2, Safe Shutdown Logic Diagrams in Appendix B of SSAR)

M' Yes, go to 5.11, complete actions and resume at 5.3 E' No, go to 5.3 Comments:

5.3 Does the modification require deletion of a safe shutdown component? (SSAMS Database, SSEL Module, SSAR Section 2)

M Yes, go to 5.11, complete actions and resume at 5.4.

0 No, go to 5.4 Comments:

5.4 Does the modification require a design change to a safe shutdown component? (SSAMS Database, SSEL Module, SSAR Section 2)

E] Yes, go to 5.11, complete actions and resume at 5.5 Z No, go to 5.5 Comments:

PBF-2060e

FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5, 6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 5.5 Does the modification add/delete/revise safe shutdown equipment to the system flow path or boundary isolation from interconnecting systems? (the Appendix R Highlighted P&IDs, SSAMS Database, SSEL Module, and the SSAR Section 2)

Z Yes, go to 5.11, complete actions and resume at 5.6

"] No, go to 5.6 Comments:

5.6 Does the modification affect the operation of a system relied upon for post-fire safe shutdown (e.g., changes in system flow rate, change in normal positions, etc. See SSAMS, SSEL Module, SSAR Section 2)?

[ Yes, go to 5.11, complete actions and resume at 5.7 I"' No, go to 5.7 Comments:

5.7 Does the modification violate the safe shutdown systems performance goals as presented in FPER Section 7.2 and SSAR Section 2?

E] Yes, go to 5.11, complete actions and resume at 5.8 Z No, go to 5.8 Comments:

5.8 Does the modification affect any mechanical sub- or support components of safe shutdown components not listed on the safe shutdown equipment list? (e.g., SOVs, check valves, etc.) (See CHAMPS Appendix R listing)(. If it is a support component for safe shutdown equipment, then it should be considered a safe shutdown component for the purposes of review for impact.

1 Yes, go to 5.11, complete actions, resume at 5.9

[- No, go to 5.9 Comments:

5.9 Does the modification to the sub- or support component affect the operability of its associated safe shutdown equipment? (i.e., Failure of a support component that results in failure of a safe shutdown component)

[ Yes, go to 5.11, complete actions, resume at 5.10

-- No, go to 5.10 Comments:

PBF-2060e

FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5, 6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 5.10 Does the modification add/deletelrevise any electrical sub or support components which support the identified safe shutdown component(s) (e.g., power supplies, relays, switches, motor operators)? (Ref. Associated Circuit Analysis

- SSAR Section 3)

I- Yes, go to 5.11 Z No, go to 6.1 Comments:

5.11 The addition/deletion/revision of safe shutdown components, sub- or support components affects the safe shutdown analysis and must be evaluated for impact on Appendix R compliance and documentation impacts in Sections 6.0 and 7.0. List the equipment and the affected systems and refer to Section 5.2.5.1 of Design Guide DG-F01.

RESUME checklist completion.

Safe Shutdown System(s), Components, Sub- or Support Component(s):

AF- 117 internals are being removed per MR 02-029. This check valve will no longer be able to isolate backflow. This is more conservative in terms of forward flow through the AFW common recirculation line, and has no affect on the safe shutdown analysis. This check valve will be less likely to fail closed, and reduces the likelihood of failure of the AFW pumps.

6.0 SAFE SHUTDOWN CABLES, INCLUDING ASSOCIATED CIRCUITS (Ref. Section 5.2.5.2 of Design Guide DG-F01) 6.1 Does the modification require addition of a safe shutdown cable, including cables which could spurious operation of safe shutdown equipment (i.e., through interlocks and interfacing relays and contacts)? (Ref. Section 3 of the SSAR, Section 5.2.5.2 of Design Guide DG-F01)

"' Yes, go to 6.10, complete actions and resume at 6.2

[ No, go to 6.2 Comments:

6.2 Does the modification require deletion of a safe shutdown cable? (Ref. SSAMS Circuit Analysis Module, SSAR Sections 3, 4, and 5)

EI Yes, go to 6.10, complete actions and resume at 6.3

] No, go to 6.3 Comments:

PBF-2060e

FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5,6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 6.3 Does the modification revise an existing safe shutdown cable, such that safe shutdown equipment functionality, either during normal/emergency equipment operation, or when subjected to a fire-induced circuit failure, could be impacted? This includes changes that could impact the ability to transfer equipment control from one operating location to another and changes which affect circuit protective device performance.

-' Yes, go to 6.10, complete actions and resume at 6.4 I No, go to 6.4 Comments:

6.4 Does the modification require a change to the routing of an existing safe shutdown cable? This includes actual physical routing changes and changes in CARDS/SSAMS to correct routing discrepancies. This may involve changes to the cable endpoint, changes to the cable endpoint location, changes to the raceways in which a cable is routed, or changes to the fire zones through which a raceway is routed. (Ref. SSAMS Cable and Raceway Module, CARDS)

E'I Yes, go to 6.10, complete actions and resume at 6.5 I No, go to 6.5 Comments:

6.5 Does the modification require addition or revision of a circuit connected or to be connected to safe shutdown power supply? (Ref. Section 5.2.5.2 of Design Guide DG-F01, Appendix R Highlighted Single Line Drawings, SSAR Section 3)

El Yes, go to 6.6 i No, go to 6.7 Comments:

6.6 Will adequate electrical coordination between the safe shutdown power supply feeder breaker and the added or revised component breaker or fuse exist? (Ref. Section 5.2.5.2 of Design Guide DG-FOland SSAR Section 3)

E] Yes, go to 6.7

[' No, go to 6.10, complete actions and resume at 6.7 Comments:

PBF-2060e

FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5,6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 6.7 Does the modification require addition or revision of any non-safe shutdown circuits?

L" Yes, go to 6.8

[ No, Safe Shutdown Cables Section Complete, go to 7.1 Comments:

6.8 Will the new or revised cables be equipped with properly designed circuit breakers, fuses or some kind of current limiting device? (Ref. SSAR Section 3)

M Yes, Safe Shutdown Cables Section Complete, go to 7.1

-" No, go to 6.9 Comments:

6.9 Will the new or revised cables share a common enclosure (raceway, panel etc.) with safe shutdown cables? (Ref.

Section 5.2.5.2 of Design Guide DG-FOland SSAR Section 3)

-" Yes, go to 6.10, complete actions i-] No, Safe Shutdown Cables Section Complete, go to 7.1 Comments:

6.10 The modification impacts the safe shutdown circuit analysis and must be evaluated further in Section 7.0 for impact on Appendix R compliance and documentation updates. List the safe shutdown circuits and associated components and refer to Section 5.2.5.2 of Design Guide DG-F01. RESUME checklist completion.

Comments:

7.0 FIRE AREA ANALYSIS, INCLUDING EXEMPTIONS/EVALUATIONS (Ref. Section 5.2.5.3 of Design Guide DG-F01) 7.1 Do the changes to the safe shutdown systems/equipment (from Section 5.0 of the FPCC), safe shutdown circuits or the physical routing of the cables (from Section 6.0 of the FPCC) result in a change to the potential consequences of a fire in any plant fire area? This includes changes that could result in the addition/deletion/modification of a compliance strategy for a piece of safe shutdown equipment for any fire area (such as availability of redundant equipment outside of the fire area, separation in accordance with Section III.G.2 of Appendix R of Appendix R, local manual actions, repairs, etc.). (Ref. SSAR, Section 5)

L" Yes, go to 7.5, complete actions and resume at 7.2

[ No, go to 7.3 Comments:

PBF-2060e Revision 0 06/08/01 Page 5 o f 6

FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5,6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 7.2 Is compliance with the separation criteria for redundant safe shutdown capability in Section III.G of Appendix R affected by the change? (Ref. SSAR Table DBD T-40, Table 3-2.2)

El] Yes, go to 7.5, complete actions and resume at 7.3 E] No, go to 7.3 Comments:

7.3 Is the modification proposed to be implemented in a fire zone/area for which an Appendix R Exemption or FPEE is in place (Ref. DBD T-40, Table 3-2.2, FHA, Table 1-1 of the SSAR)

[ Yes, go to 7.4 I"- No, Fire Area Analysis Section is complete, go to Section 8 or next applicable Section.

Comments:

7.4 Does the modification violate or potentially change the basis for the Appendix R exemption or FPEE? (Ref.: DBD T-40, Table 3.2-2, FHA, Table 1-1 of the SSAR)?

E] Yes, go to 7.5

] No, Section 5, 6, and 7 checklists complete Comments:

7.5 The modification impacts the Fire Area Analysis and potentially violates the basis for compliance with the separation requirements of Appendix R, the basis for an Approved Appendix R exemption, or technical evaluation such as a Fire Protection Engineering Evaluation. List the basis affected and refer to Section 5.2.5.3 of Design Guide DG-FO1. RESUME checklist completion.

Bases:

PBF-2060e Revision 0 06/08,01 Pave 6 of 6

NUCLEAR POWER BUSINESS UNIT WO WORK PLAN Work Control Document: 0212107- MR 02-029 UNIT: PB0 Equipment ID: AF-1 17 Equipment

Description:

AFP COMMON MINI RECIRC HEADER CHECK Work Plan Originator: Rob Chapman x7636 pager 0114 Date: September 5, 2002

_______,___,WORK SCOPE WORK SCOPE The purpose of MR 02-029 is to correct a potential common mode active failure of the AFW pump and common recirculation line, as part of the upgrade of the mini-recirc AOVs to have a safety-related PURPOSE function to open.

The scope of this installation work plan is to remove the internals from AF-1 17.

The AF-1 17 check valve is located on the 26' elevation of the Control Building, west of the CSTs. The AF-4035 relief valve is located on the 8' elevation of the Unit 1 turbine building, west of the south Aux Feed tunnel doors.

AF-1 17 is non-QA, non Seismic, non ASME Class. An RRM is NOT required for an increase in work scope.

INITIAL None of the AFW pumps shall be running.

CONDITIONS DANGERTAG Prepare a tag series to isolate the AFW common recirc line at valve AF-1 17. The tag series must not SCOPE isolate AF-4035. The following valves shall be tagged SHUT:

"* AF-1, AFP RECIRC HEADER ISOLATION

"* AF-2, T-24A CST MINI RECIRC ISOLATION

"* AF-9, T-24B CST RECIRC ISOLATION

• AF-10, T-24A/B CST HEATING RETURN HEADER ISOLATION The following valves shall be tagged OPEN:

"* BE-95, T-24A/B CST FEED

"* BE-96, T-24A/B CST FEED DRAIN THE AF-4035 RELIEF VALVE WILL BE CREDITED AS THE MINI-RECIRC FLOW PATH. A LEVEL 3 DEDICATED OPERATOR WILL BE STATIONED TO MONITOR OPERATION OF THIS RELIEF VALVE, AND NOTIFY THE CONTROL ROOM IF IT FAILS TO OPEN AUTOMATICALLY.

THE AFW PUMPS WILL BE FULLY OPERABLE DURING THE INSTALLATION OF THIS MODIFICATION.

DANGER TAG BECH M-217 Sh. 1 REFERENCES WEST 684J971 Sh. 2 PBF-9169 Page 1 of 7 H 0fATATFILESWODSWR 02-029 - AF Recwrc Satety Uograde -AF1 17%0212107 doc Rplvm*inn q4 (31111P1n

NUCLEAR POWER BUSINESS UNIT WO WORK PLAN Work Control Document: 0212107 - MR 02-029 UNIT: PB0 I Equipment ID: AF-117 Equipment

Description:

AFP COMMON MINI RECIRC HEADER CHECK Work Plan Originator: Rob Chapman x7636 pager 0114 Date: September 5, 2002

,t-W0K SCOPEj--

,

LIMITATIONS

• All pre-work staging activities shall be completed prior to tagging the recirc line out of service.

AND This will include attaching a hose to the relief valve discharge. Once the system is tagged out, PRECAUTIONS operation of the AFW pumps will cause the AF-4035 valve to open. As long as AF-1, AF-2, and AF-9 are open, there is reasonable assurance that the relief valve will not open with the AFW pumps running on recirc. Therefore, attaching and removing the hose shall NOT be performed while tags are hung. OPS shall verify that AF-1, AF-2, and AF-9 are locked open prior to attaching the hose to the AF-4035 relief valve.

• Caution should be taken when taking the bonnet off the check valve due to water present in the piping. This line may be filled with water initially, and should drain when the bonnet is cracked loose. Leakage from the CSTs past AF-2 or AF-9 could cause water to build up inside the check valve body.

• Scaffolding will be required to access the AF-4035 relief valve (8' TB1) and the AF-117 check valve (26' CB).

a STEPS IN THIS WORK PLAN SHALL BE PERFORMED INTHE ORDER SPECIFIED.

TOOLS

• 3" 150# Class flange with hose connection AND

• 3" hose (minimum), approximately 40 ft MATERIALS

• Tools as needed to disassemble the check valve and remove the internals a Gasket for ALOYCO 367 3" 150# check valve, SIR P102-024756, LIN 917-5440 (QTY 1)

• Screw, Hex Head Cap, Y2*-13(UNC) x 2 Y2", Grade B8, L/N 903-0397 (As needed)

• Nut, Hex, Y2"-13(UNC), Grade 8, L/N 903-0569 (As needed)

XA QC REVIEW OF WORK PLAN (independent QC review required on QA classified work order only) NA if non-QA work order Any change in scope requires WO WP review by QC inspector.

PBFo9169 Page 2 of 7 H DATATFILES'OOSWIR 02-029 - AF Reciec Safety Upgrade - AFI t7I0212107 doc Revison 3 01/12/01

NUCLEAR POWER BUSINESS UNIT WO WORK PLAN Work Control Document: 0212107- MR 02-029 UNIT: RBO Equipment ID: AF-117 Equipment

Description:

AFP COMMON MINI RECIRC HEADER CHECK Work Plan Originator:. Rob Chapman x7636 pager 0114 Date: September 5, 2002

~

- ~ ~-- .. SUPPORT-SUPPORT El Chemistry 0 Engineering Rob Chapman x7636 pager 0114 El HP El I&c 0 Maintenance Performing internals removal.

O] NDE 0 Operations Hang tags, level 3 dedicated operator aQC Witness torquing El Security El Crane [ITB El PAB El Polar El Other 0 Other Component Engineer / Aaron Guenther x6422 PBF-9169 Page 3 of 7 H \DATAWILES MODSV.¶R 02-029 - AF Recirc Safety Upgrade - AF1 7"0212107 doc

NUCLEAR POWER BUSINESS UNIT WO WORK PLAN Work Control Document: 0212107 - MR 02-029 UNIT: PB0 Equipment ID: AF-117 Equipment

Description:

AFP COMMON MINI RECIRC HEADER CHECK Work Plan Originator: Rob Chapman x7636 pager 0114 Date: September 5, 2002 NOTE: The steps in this work plan must be performed in the order specified.

FME: Tools and equipment shall be checked for loose parts and debris and temporary covers should be installed for foreign material exclusion (FME) of system/components per Exclusion of Foreign Material from Plant components and Systems, NP 8.4.10.

NOTE: IF inspections or discrepancies require modifications to Work Scope:

THEN STOP work, place equipment in SAFE condition, and NOTIFY Supervision.

NOTE: The Control Room / the Work Control Center / and the watchstander (as appropriate) shall be informed of the status of jobs which:

bring in alarms, affect indications, and other work being performed on operating equipment.

NOTE: All workers shall perform all Danger Tagging requirements as defined in NP 1.9.15 NOTE: Any pen and ink change to work plan requires initial and date by the change and engineering concurrence.

NOTE: Write WO number on top/header of any supplemental pages added to work package, i.e., forms, procedures, checklists...

PBF-9169 Page 4 of 7 H %DATATILES\MOOSWR02-029 - AF Recrc Safely Uplrade - AFt 17%0212107 doc Revision 3 01/12/01

NUCLEAR POWER BUSINESS UNIT WO WORK PLAN Work Control Document: 0212107- MR 02-029 UNIT: PB0 Equipment ID: AF-117 Equipment

Description:

AFP COMMON MINI RECIRC HEADER CHECK Work Plan Originator: Rob Chapman x7636 pager 0114 Date: September 5, 2002 I

PBF-9169 Page 5 of 7 H ATATFILESWODSWMR 02-029 - AF Recirc Safety Upgrade. AFi1T7\212107 doc Revision 3 01112/01

NUCLEAR POWER BUSINESS UNIT WO WORK PLAN Work Control Document: 0212107 - MR 02-029 UNIT: PBO Equipment ID: AF-117 Equipment

Description:

AFP COMMON MINI RECIRC HEADER CHECK Work Plan Originator: Rob Chapman x7636 pager 0114 Date: September 5, 2002 Hold Point -.- Workln Description ,

7. Perform a visual inspection of the valve body for evidence of corrosion and/or erosion, and notify component engineering if any discrepancies are noted (Aaron Guenther, x6422).

MT DATE

8. Remove clapper arm shaft / pin (Item 5) from clapper arm and cover hanger.

Remove clapper arm (Item 3) and entire disk assembly from the cover hanger.

DO NOT DISCARD INTERNALS. CONTACT Rob Chapman at x7636.

MT DATE FME 9. Prior to installing the valve in the system perform a final FME closure exam. Inspect accessible piping for dirt and debris, and remove from piping. Ensure that all removed parts are accounted for. Document the closure exam below:

FME Closure Exam: [3 SAT E] UNSAT MT DATE QC Hold Point 10. Re-install the check valve cover using a new gasket.

Torque bolting to 45 ft-lbs ___ __ DATE Actual Value: MT DATE MTE Serial Number. Cal Due Date:

QC DATE

11. Clear tag series.

OPS DATE

12. Secure the Level 3 Dedicated Operator.

OPS DATE

13. The AFW pump common minimum recirculation line is available.

OPS DATE PBF-9169 Page 6 of 7 HADATAFILESMOOSWMR 02-029 - AF Recirc Safely Upgrade oAF 117\021 2107 doc Revision 3 01/12/01

NUCLEAR POWER BUSINESS UNIT WO WORK PLAN Work Control Document: 0212107 - MR 02-029 UNIT: PBO Equipment ID: AF-117 I Equipment

Description:

AFP COMMON MINI RECIRC HEADER CHECK Work Plan Originator: Rob Chapman x7636 pager 0114 Date: September 5, 2002 I

PBF-9169 Page 7 of 7 H DATATIESWMODSWR 02-029

• AF Recirc SaFely Upgrade - AF 117\212107doc Revision 3 01/12/01