ML030770750

10 CFR 50.59/72.48 Screening for Renewal of Internals from AF-117 and Upgrade Open Function of AFW Pumps mini-recirc Valves to Safety Related

ML030770750
Person / Time
Site:	Point Beach
Issue date:	10/10/2002
From:	Black D, Chapman R Nuclear Management Co
To:	Office of Nuclear Reactor Regulation
References
FOIA/PA-2003-0094 SCR 2002-0359-02
Download: ML030770750 (10)

Text

Point Beach Nuclear Plant 10 CFR 50.59/72.48 SCREENING (NEW RULE)

Title of Proposed Activity:

Associated Reference(s) #:

SCR 2002-0359-02 Venfy SCR number on all pages Page 1 Removal of Internals from AF-117 and Upgrade Open function of AFW pumps Mini-Recirc Valves to Safety-Related (MR 02-029)

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 Discharge 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), NRC 2002-0086, Clarification of Reply to a Notice of Violation (EA-02 031) NRC Special Inspection Reports No. 50-266/01-17 (DRS); 50-301/01-17 (DRS)

Prepared by:

Reviewed by:

David Black Name (Print)

Rob Chapman Name (Print)

I

&Si(-*

Date:

Date:

/cii262c 0 Z 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.

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 screening involves the following activities associated with MR 02-029:

- Remove internals for AF-1 17. the Auxiliary Feedwater Pump Common Mini-Recirc Header Check Valve to prevent a common mode active faflure in the recirc i etutn header.

- Upgrade the open function for the air-operated Auxiliaryfeedwater pump mini-recirc valves (J(2)AF-4002, AF-4007. AF 4014) to safety-related (also being done by ISTgroup for inservice test procedure changes under screenings SCR 2002-0321 and 2002-0339). The check valves upstream of the AO Vs (1/2AF-114, AF-J 15 and AF-116) are already classified as safety related in CHAMPS; they will now have a safery function to open. Inservice test procedure changes to verify the open function of the check valves are supported by screenings SCR 2002-0321 and 2002-0339.

- The backup nitrogen and air supply systems installed under modifications MR 01-144 and MR 02-001 as reviewed under screening SCR 2002-0010-01 now have a safety-related support function to provide a pneumatic source to open or maintain open these valves (l(2)AF-4002. AF-4007. AF-4014) in the event of loss of instrument air.

- State in the FSAR (FSAR 7.3, 9.7 and 10.2) and TechnicaiSpecqfcation Bases(S 3.7 5j that (1) the openfunctionforAuxiliary feedwater pump mini-recirc valves (1(2) AF-4002. AF-4007. AF-4014) is safety-related, and (2) the availability of a recirculation line downstream of the flow restricting orifices is required for AFW operability. The line downstream of the flow restricting orifices is nA safety-related since failure of the line is conservative. The Bases will indicate that mini-recirculation flowpaths have to be available for the AFWsystem to be operable. Specific discussion of the check valves function to open will not be included in the FSAR because it can be inferred from the discussion of the AOVs (avoids excessive detail in the FSAR).

PI3F-1515c Revision I 04/03/02 REC'D OCT 1 1 2002 Reercn,.c NP 5 I S

Point Beach Nuclear Plant SCR 2002-0359-02 10 CFR 50.59172.48 SCREENING (NEW RULE)

Verify SCR number on all pages Page 2 A simplified P&ID of the mini-recire header is shown below To Condensate Storage Tank T-248 From P-38A. P 388. & 2P-29 From 1 P-29 To Condensate Storage 2-~C-4Tank T-24A EVAL 2002-005 evaluated permanent procedure changes that were implemented in response to a condition (CR 01-3595) that was identified where, with a procedure-directed operator action to control steam generator level (which could be accomplished by reducing flov, through one or more AFWpumps), concurrent with a loss ofinstrument air (which would cause the AFW pumps' mini-recirculation valves to/ail close), the potential existedfor a simultaneousfailure of the multi stage high pressure AFW pumps due to very low or no flow through running AFW pumps. The procedure changes added instructions to the operators that if any AFW pump mini-recirc valve/ails shut OR annunciator COI A 1-9, INSTRUMENT AIR HEADER PRESSURE LOW is in alarm, then monitor and maintain minimum AFW flow or stop the affected AFWpump as necessary to control SIG levels. Minimum flow valuesfor each pump were also included in the procedures. The majority of the changes were associated with EOP and ECA foldout pages. The procedures were initially revised as a compensatory measure to support AFWpump operability. The 50.59 evaluation reviewed the procedure changes as a permanent change to the procedures as described in the FSAR to confirm consistency with the licensing basis. The permanent procedure changes restored the AFW pumps to full'operable status.

This evaluation also was the basisfor changes to the FSAR to clarify that the mini-recirculation valves require instrument air to/unction and that either a pump minimum flow is maintained or pumps are secured if the valve/ails or instrument air is lost (FCR 02-019).

Screening SCR 2002-0010-01 reviewed modifications to provide backup air sources to allFtVpump minimumflow recirculation valves These modifications were an enhancement that reduced the core damage probability from a loss of instrument air and increased the timefor an operator to take manual action to override the valves open Instrument air accumulator tanks were installed by MR 02-001for the 1/2P-29 valves (1/2AF-4002), and the existing nitrogen backup system for the MDAFP discharge valves were tied in by MR 01-144/for the P-38AfB valves.AF-4007. AF Point Beach has made an NRC commitment to upgrade the open/unction/or all mini-recirc valves to safety-related as stated in NMC letter NRC 2002-0068:

"To further improve the future effectiveness ofthe AFW system by providing additional pump protection against low flow, Point Beach is classifying the open/unction ofthe pump recirculation flow control valves, as safety-related This will provide a redundant method ofprovidingfor minimum AFW pump flow and consequently, pump cooling. As a result, testing and quality assurance requirementr required/for safety-relatedfunctionr will be applied to the open/function ofthese valves As discussed previously, internal pump cooling is designed to be provided by minimum forwardflow through the pumps. Classifying the recirculation flow control valves as safety-related will provide greater assurance that minimum flow will be available to provide internal pump cooling. The pneumatic backup supply to the recirculation flow control valves is limiting and therefore Point Beach will also continue to specify operator action to manually open these recirculation valves Similar to other plants, Point Beach has one common recirculationflowpathfrom all the AFW pumps to the condensate storage tanks" The scope of this screening also includes FSAR changes and Technical Specification Bases changes to coincide with the mini-recirculation safety upgrade described above rev isio I 5c Revision 1 04/03/02 Reterncric NP 5 1 8

Point Beach Nuclear Plant 10 CFR 50.59/72.48 SCREENING (NEW RULE)

SCR 2002-0359-02 Verify SCR nuxnbcr on all pages Page 3 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 the following functions described in the licensing basts:

a.

To automatically start and ensure that adequate feedwater is supplied to the steam generators for heat removal during accidents which may result in a main steam safety valve opening (Loss of Normal Feedwater - including ATWS, and Loss ofAC to the Station Auxiliaries).

b.

To automatically start and provide flow to maintain steam generator levels during accidents which require or result in rapid reactor coolant system cooldown (Steam Generator Tube Rupture and Rupture.of a Steam Pipe).

c.

To allow the isolation of all lines to the ruptured steam generator in the SGTR event.

d To provide sufficient feedwater to remove decay heat from both units for one hour during a station blackout

-(SBO) event (FDAFP only)....

e.

To provide sufficient flow to the steam generators to remove decay heat to achieve cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a plant fire (Appendix R).

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

g To provide flow to the steam generators during plant startup and shutdown, and during hot shutdown or hot standby conditions for chemical additions and when operation of the main feedwater and condensate systems is not warranted These modifications affect the operation of the minimum recirculation valves for the auxiliaryfeedwater pumps (1/2AF 4002 AF-4007, AF-4014). These valves have the following design functions:

1.

To isolate the minimum recirculation line to ensure that the auxiliaryfeedwater pumps deliver the requiredflow to the steam generators as needed to support the following accidents or events: LONF, LOA C, MSLB. SGTR. A TWS, Appendix R, and SBO.

2.

To open to provide flow through the auxiliary feedwater pumps to prevent hydraulic instabilities and to dissipate pump heat.

3.

To maintain the pressure boundary integrity of the auxiliary feedwater system.

FSAR 10.2.2, System Design and Operation. states: "The auxiliaryfeedwater system consists of two electric motor-driven pum,,ps;,'o-stec.m turbi'.e-driven pumps, pump suction and discharge piping, atnd-te controls andjzntnrmentoti.on necg.sary_

for operation of the system. Redundancy is provided by utilising two pumping systems, two different sources of power for the pumps, and two sources of water supply to the pumps. The system is categorized as seismic Class I and is designed to ensure that a single fault will not obstruct the system function."

CLB

References:

FSAR 7.2.3.2-Specific Control and Protection Interactions FSAR 7.3.3.4 - Manual AFW Flow Control During Plant Shutdown FSAR Section 7. 4. 1 - A'ASA C FSAR Section 10 1 - Steam and Power Conversion System FSAR Section 10.2 -Auxiliary Feedwater FSAR Figure 10 2-1 fheet I - Bech M-217 Sh. I - Auxiliary Feedwater System FSAR Figure 10.2-1 Sheet 2-Bech M-217 Sh 2-Auxiliary Feedwater System FSAR Section 14. 1. 10 - Loss of Normal Feedwater FSAR Section 14 1. 11 - Loss ofAlI AC Power to the Station Auxiliaries FSAR Section 14.2.4 - Steam Generator Tube Rupture FSAR Section 14 2.5 - Rupture ofa Steam Pipe FSAR Appendix A I - Station Blackout FCR 02-019 FPER 5 2 2 - Safe Shutdown Systems and Equipment PBF-IS15c Reviston I 04/03/02 Reference" NP 5 I S C

Point Beach Nuclear Plant SCR 2002-0359-02 10 CFR 50.59/72.48 SCREENING (NEW RULE)

Verfy SCR number on all page Page 4 FPER 5.2.5.2.3 -Auxiliary Feedwater Pump Room Tech Spec 3.7.5 - Auxiliary Feedwater Tech Spec Bases B 3.7.5 - Auxiliary Feedwater NRC 2002-0068, Point Beach Nuclear Plant. Units I and 2 Reply to a Notice Of Violation (EA-02-03.l) NRC Special Inspection Report 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 :

E] 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 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.

E]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

E]

Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?

Z E]

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

[

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?

"DdE-he proposed activity involve CLB-described SSCs or procedural cuntrou6"that perform furrctions that are required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications?

E]

0 Does the activity involve a method of evaluation described in the FSAR?

El

[]

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

El

[]

Does the activity exceed or potentially affect a design basis limit for afission product 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), thin 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 implicit function to open to allow minimum recirculationflowfrom the auxiliaryfeedwater pumps to return to the condensate storage tank(s). There is no direct discussion of this function in the licensing basis, other than that the valve appears in FSAR Figure 10 2-'i.

PD F-1515c Rcvision I 04/03/02 Reterence NP 5 1 S

Point Beach Nuclear Plant SCR 2002-0359-02 10 CFR 50.59/72.48 SCREENING (NEW RULE)

Vrfy SCR number on al pag Page 5 A search of plant documents did not yield any reference to a needfor valve AF-117 to close (or remain closed). The following uses

-were considered as possible functions for the closed position:

a. To prevent backflow of heatinz steam condensate from entering the eini-recirculation return header. Given that the min recirculation header is not drained and remains filled with water during and after pump operation, migration of heating steam condensate into the header in any significant amounts is unlikely. Further, the plant no longer returns heating steam condensate to the condensate storage tank because of water quality concerns, and it is extremely unlikely that this plant design feature will ever be used again. Regardless. this reason for preventing back flow does not constitute a licensing basis design function.

Therefore this function will not be evaluated for adverse effects.

b. To prevent backflow and siphoning of water from the condensate storace tanks throuO AF-4035 in the event that this valve opens due to hitch pressure in the line, or due to a break in the header. The valve's nominal reliefsetting is 50 psi. ForAF-4035 to open. either an event occurs where multiple pumps start with high recircflowrates and pressurize the line, or there is a blockage in the line. The safety-related flow restricting orifices in the recirc line from each AFW pump limit the flow and pressure from each pump. Further, if the line is blocked either by mispostioning of manual valves orfailure of the check valve to open, there is noflowpathfor backflow. The only credible way for the header to break is during a seismic event, in which case water from the condensate storage tank cannot be credited for accident mitigation. The prevention ofbackflow from the condensate storage tanks cannot be considered a licensing basis design function. Therefore this function will not be evaluated for adverse effects

2. The minimum recircidlation valves for the auxiliaryfeedwater pumps (l/2AF-4002. AF-4007. AF-4014) have the design functions to isolate the minimum recirculation line to ensure that the auxiliaryfeedwater pumps deliver the requiredflow to the steam generators as needed to support the mitigation of accidents or events. The proposed activity is to make the open function for these valves and the associated pump-side check valves a safety-relatedfunction. The new components for modifications MR 02-001 and MR 01-144 as discussed above were installed safety-related due to their risk significance. These components were classified safety-related based on a PBNP management decision, as described in NPM 2002-0228. These modifications ensure a safety related supply of air or nitrogen to ensure that the auxiliaryfeedwater pumps have adequate cooling.

PART Il1 (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 [] 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.

III.!

CHANGES TO THE FACILITY OR PROCEDURES YES NO QUESTION E1 0

Does the activity adversely affect the design function of an SSC credited in safety analyses?

I

[]

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):

1. The only function identified for the AF-117 check valve was to open to allow mini-recirculation water from AFW pumps to return to the condensate storage tank. It has no design function to close. Therefore removing the check valve internals has no adverse effict on the open function and precludes the valve from failing to open

2. The change in designation of the open function of mini-recirculation valves 1/2,4F-4002. AF-4007, AF-4014 as safety related has no adverse affect on the valve's function to open or close. The open function was a design function desribed in the FSAR. Designating the open function as safety-related requires addition quality control and testing and should, therefore, make the open functions of the valve more reliable The necessary backup air/nitrogen supplies were installed as safety-related so no additional physical changes to the plant are required SCR 2002-0010-01 assessed the installation of these backup pneumatic supplies on the close function of the valves and no adverse effects were identified Therefore there is no adver3e effect on the functions of these valves.

PBF-1515c Rcvision I 04/03102 Rcicrcnce NP 5 1 S

Point Beach Nuclear Plant SCR 2002-0359-02 10 CFR 50.59/72.48 SCREENING (NEW RULE) vcn-y SCR number on all pages Page 6 3 Changes will be made in the FSAR to state that the open function for all Auxiliaryfeedwater pump mini-recirc valves is safety-related The Technical Specification Bases will indicate that the mini-recirculationflowpaihs have to be OPERABLE for the AFW system to be operable. These changes only reflect the activities discussed above, which have already been assessed and determined that there is no adverse effect on a design function FSAR changes will also be made to reflect the recirculation line downstream of the flow restricting orifices has a safety function and is required for AFW operability, but the line is not safety-related since failure of the line is conservative. No physical changes were made to this line. If the line fails. a break or opening will be creat'ed which may increase overallflow through the recirculation line. The service-water system is the safety-related water supplyfor the AFW system. The loss of CST inventory due to the line failing and recirculation water spilling only affects the non-safety-related water source for the AFW system. The required minimum CST water is based on the station blackout(SBO) 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 design functions.

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

YES NO' QUESTION

[:

El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in the CLB?

[I 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 If the activity is not a test or experiment, the questions in I1.3.a and I1I.3.b are 0 NOT APPLICABLE.

a. Answer these two questions first:

YES NO

. QUESTION

[E

[E Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?

[E E]

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

If the answer to BOTH questions in V.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 111.3.a above.

If the answer to ei her 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?

El El Does the act.iyity utilize or control an SSC in a manner that is inconsistent with the analyses or descriptions in the CLB?

PBF-1515c Revision I 04/03/02 Rcfcrcnce. NP 5.1 S

Point Beach Nuclear Plant SCR 2002-0359-02 10 CFR 50.59172.48 SCREENING (NEW RULE)

Veify SCR number on all pages Page 7 E]

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 1Il.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):

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

Check all that apply:

"A 10 CFR 50.59 Evaluation is El required or ED NOT required.

"A Point Beach FSAR change is Z required or El NOT required. If an FSAR change is required, then initiate an FSAR Change Request (FCR) per NP 5.2.6.

A P.egf,'tory Commitment (CLB Commitment Database) change is El required or Z NOT required. If a Regulatory..

Commitment Change is required, initiate a commitment change per NP 5.1.7.

A Technical Specification Bases change is 0 required or [] 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 [E required or [D 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.

YES NO QUESTION

-Does the proposed activity involve IN ANY MANNERth~e.dry fuel swt-rge-.c-sk.(s),_tm.*askj sr/tanspqr 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 PPCS/ISFSI Continuous Temperature Monitoring System?

El 1

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 ED 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 Atdiliary 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?

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 E0 Does the activity involve plant heavy load requirements or procedures for areas of the plant used to support cask loading/unloading activities?

PBF-1 5 1 Sc Revision I 04/03/02 Reference NP 5 1.3

Point Beach Nuclear Plant SCR 2002-0359-02 10 CFR 50.59/72.48 SCREENING (NEW RULE) verify SCR number on all pages Page 8 El

[]

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.

PART VI (72.48) - DETERMINE IF THE CHANGE INVOLVES A ISFSI LICENSING BASIS DESIGN FUNCTION (If AL.L the questions in Part V are NO, then Part VI is Z 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/ISFSI 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?

Ml F

Dcs thz Frcpnscd a*tivity involve plant, cask or ISFSI SSCs whose filnction is relied uipon for prevention ofa radioictive release, OR whose failure could impact SSC(s) credited in the Safety Analyses?

E]

F]

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?

F]

F]

Does the activity involve a method ofevaluation described in the ISFSI licensing basis?

[]

[]

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

E]

E]

Does the activity exceed or potentially affect a cask design basis limit for afission product 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.

FPART- -VIII (7-.I8) --DETET.... LNE WHETHE.R THE ACTIVITY INVOLVES ADVERSE EFFECTS (N El 9 6-7, Appendix B, Section B.4.2. 1)

(if ALL the questions in Part V or Part VI are answered NO then Part VII is Z 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 VII.3.

VII. I Changes to the Facility or Procedures YES NO QUESTION

[]

F]

Does the activity adversely affect the design function of a plant, cask, or ISFSI SSC credited in safety analyses?

El

[]

Does the activity adversely affect the method of performing or controlling the design/unction 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):

PBIs-15i5c Revision I 04103/02 Reference NP 5 1 3

Point Beach Nuclear Plant SCR 2002-0359-02 10 CFR 50.59/72.48 SCREENING (NEW RULE) vrify SCR number on all pages Page 9 VII.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 E]

0l 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):

~ ~)

I-VII.3 Tests or Experiments (If the activity is not a test or experiment, the questions in VII.3.a and VlI.3.b are [

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 cask ISFSI licensing basis?

E]

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 quebtions 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 [D 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?

E]

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?

E]

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 ISFS[ SSC to perform its intended functions?

I If any answer in VII.3.b is YES, a 10 CFR 72.48 Evaluati6n is required. If the answers are all NO, describe the basis for the conclusion (attach additional discussion as necessary):

PBF-I515c Revision I 04/03/02 Refereni.e NP 5 1 -4

Point Beach Nuclear Plant 10 CFR 50.59/72.48 SCREENING (NEW RULE)

SCR 2002-0359-02 Verify SCR number on all pages Page 10 PART 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 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 [] required or Z 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 [] required or ED 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 El required or 0D 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-IS I 5c Revision I 04/03/02 Refercnce NP 5 1 3