ML030870845
| ML030870845 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/18/2002 |
| From: | Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2003-0094 99-029*C | |
| Download: ML030870845 (101) | |
Text
Point-Beach Nuclear Plant PLANT MODIFICATION/MINOR PLANT CHANGE NO.: 99-029*C PLANT CHANGE INITIATION WO# 0202509 INITIATION Tide: AUX FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE ORIFICE 0 QA 0 AQ 0 Non-QA 0 SR 0 Non-SR Unitl 0 Unit 2 0 Common 0 CHAMPS System Code: AF EWR: 99-031 CR: 99-1391 Project Objectives: Eliminate excessive noise and vibration that occurs when operating 1P-29 on recirculation flow.
Proposed Scope: Replace IRO-4003 with a better type that will prevent cavitation.and reduce vibration in the minimum recirculation piping for 1P.29.
Initiated By: Alex Foltynowicz Date: 7/16/1999 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? If YES proceed with document changes Commercial Facility Change? X If YES, determine if document updates are required.
Phr i'nmmorw4al PeAloHt Mha",e Alnlw-Dcmn Updates?... -.. If NO YES contact proceed designofsupervisor. If outside Engineering 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:
L MinorPlantChanre 2.-IP-oz0 &ýý4 -1 C9 -0z PreparecL*. Date Engin eh"hg Group Lead: Date 1
OCT 2 4 2002 PBI-1605a oto/om, Rcvw, Page I of 2
Point Beach Nuclear Plant PLANT DESIGN CHANGE CHECKLIST PLAT MODIMATION&MOR PLA CHANGE NO.: 99-029"C
Title:
AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE DESIGN SUPERVISOR Design Controls and Project Controls. (Ref. NP 72.1, Commentary, for completion of this section.)
Check Applicable Design Controls: Clarifications/Basis:
0 Design Input Checklist (PBF-i584) 0] DUC (PBF-1606) 0] Design Verification Notice (PBF-1583) 0 Calculations 0] Design Documentation (PBF-1585), or equivalent 0 Design Change In Progress DCN's 0* Engineering Change Requests o] Specifications 01
[]
Check Applicable Project Controls: Clarifications/Basis:
. .- ... -ModWcation Team Reqvir-* (indicate minimun groups to request)
O] Conceptual Design Package Required O Budget Design Project (Impact) Number o] Detailed Project Schedule O] IWP Required Assigned Modification Egineer -itI CX4r.IP.% AL.-...
II1_
Design Supervisor Date: e -i8-o7t DMJý-
PBF-1605 Revision 6 10/02/0I Page I of 4 Refn*s): NP 7.1, PBF-1583. PBF-1584 NP 7.2.2) PBF-158S, PBF-1606
Point Beach Nuclear Plant PLN MOMAOMNRPATCAG O*9[2*
PLANT DESIGN CHANGE CHEMKIST .,TMDFIAONIOR tNGN." 9..2" CONCEPTUAL DESCRIPTION/REFERENCE INFORMATION (IF APPLICABLE)
GROUP HEAD CONCEPTUAL DESIGN REVIEW AND ACCEPTANCE (Check here if not required: 0]
Review conceptual design. Attach comrents on NPBU Document Review Comment Sheet (PBF-1622 or equivalent)
Grou* Acceotance Signatu'e Date Comments Radiation Protection [_ None 0 Attached Fire Protection _"] None El Attached Installing Organization [_ None El Attached
_ENone El Attached DE None 0 Attached DE None D] Attached
[E None E] Attached
- Desi Supervisor _ None [3 Attached PBF-1605 Revision 6 1(Y01 Page w
2 of 4 Itefrnce(s): NP 7.2.1. PBF-15B3, PBF-1584 NP 7.2.2 PBF- 1585. PBF-1606
Point Beach PLANT DESIGN Nuclear Plant CHANGE CHECKLIST [
- M~*XX O HNEN- 9-2" FINAL DESIGN REVIEWS Review final design. Attach comments on Document Review Comment Sheet (PBF-I622 or equivalent)
Acceptance Signature Dat Comments Radiiati~n Protection k~None ElAttached Fire Protection Engineer -70 -. Q 0el Attached Mechanical M~aintenanice -- None ElAttached oooNlone None Atce ElAttached Systemi Engineering /Q 4ZC2Z yuoeI Atce Site QA A/~~~9 ElAtce ElNone ElAttached SDM Technical ReviewElNn ElAtce INDEPENDENT REVIEW OF INSTALLATION DOCUMENTS (IWP or Work Order Plan) List all IWMs and WO's used for installation ri*O#(s) rfp 99.o29*C (WO 020209), WO 0202SO7 (Stub pieces), WO 02025S0 (Prefab)
All design and licensing requirements have been incorporated in the installation and testing document(s).
Reviewer. Date: V_________
RELEASE FOR INSTALLATION
'A e ,.z..trocr ve -bver. properly, IdeI dg- en'Inted end the project has been app t-elyrlve& AIL.t.
are approved. This design is released for installation. Comments regarding release of this design are noted below' Design Supervisor. / Z, y c Date:3& X COMMENTS PBF-1605 Revision6 10/0M2I0 Page 3 of 4 Refernce(s): NP 7.2.1. PBF-1583. PBF-1584 NP 7.2,. ,r,-158,.PBw-iou
Point Beach Nuclear Plant I PLANT MODMC.TIOWMOR PLM CHAGE NO.: 99-029*C PLANT DESIGN CHANGE CHECKLIST ACCEPTANCE Plant modification is installed, tested, and all documents required for acceptance are complete.
Modification Engineer~ '0%Z*6 S~ae Date: to-164 -o t.
6/
CLOSEOUT Plant modification is complete, including vibmittl 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: /a-- o ..
Design Supervisor. ... .Datm: 16 -2 1-0t/
NUCLEAR INFORMATION MANAGEMENT Microfilm the entire modification package.
PBF-1605 Revision 6 101I201 Page 4 of 4 Refence(s): NP 7.21, PBF-1583, PBF-1584 NP 7.2.2. PBF-1585. PBF- 1606
FINAL DESIGN DESCRMITION MR 99-029"C AUX FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE ORIFICE Revision 0 UNrr I March 13, 2002 PURPOSE The purpose of the proposed modification is to reduce piping line noise and vibration when operating the IP-29 Turbine-Driven Auxiliary Feedwater Pump (TDAFP) in the recirculation mode. The existing minimum flow recirculation line generates excessive vibration and noise, which has resulted in several socket weld failures. This has been attributed to turbulence and cavitation resulting from the flow condition through restrictive orifice IRO
- 4003. The reduction of piping line noise *and vibration will be accomplished by implementation of the recommendations of root cause evaluation RCE 99-081 and EWR 99-031, which is to replace the existing flow restricting orifice with a different type that will prevent cavitation.
In addition, as recommended by RCE 99-081 and CR 99-1391, a portion of the AF piping associated with RO will be replaced to facilitate oversized socket welds due to multiple occurrences of pinhole leaks. The purpose for the oversized socket welds is to offer a significant high-cycle fatigue improvement over the standard ASME Code socket welds in this vibration critical application.
SCOPE The scope of MR 99-029*C is to replace 1RO-4003 with a new pressure reducing orifice. The new orifice is a 600#
class globe valve with a cavitation reducing cage. In addition, pipe from the 900 elbow just downstream of IFE 4049 to the upstream socket-weld on 1AF-15 will be replaced. The replacement piping will be welded with socket welds that are oversized in a 2/1 configuration as described in EPRI technical reports TR-107455 and TR-1 11188.
This modification is classified as QA, Safety-Related (SR), seismic Class 1, although all piping downstream of 1R0-4003 is QA, non safety-related (AQ), seismic Class 1. The RO and modified piping are non-ASME Section XI class.
DESIGN INPUTS
- DG-M09, Revision 2, Design Requirements for Piping Stress Analysis, March 20,2000.
0 ASME.B31.1 -1992, Poweriin . -....
"* DG-M03, Revision 9, Bechtel Piping Class Summary, June 8, 2001.
"* Wisconsin Electric Power Company. Drawing GLD M-217, Sheet 1, QA Classification Diagram Auxiliary Feedwater System QA Classification Diagram, Point Beach Nuclear Plant - Unit I & 2, Revision 11.
"* Bechtel Drawing 6118 M-217 Sh. 1. Auxiliary Feedwater System, Revision 68.
- Bechtel Drawing P-159, Aux. F.W. From Heating Boiler Cads. Return & Pump Recirc. To Cads ST Tank 6" & 3"JO4 Unit 1.
- Bechtel Pipe Class DB-3, Auxiliary Feedwater Piping
" Flowserve Pressure Reducing Orifice Drawing 94-16249.2" 600# Globe Control Valve
"* FSAR Section 10.2, Auxiliary Feedwater System.
Page I of 5
FINAL DESIGN DESCRIPTION MR 99-029*C AUX FEED WATER PUMP 1P-29 FLOW RECIRC LINE RNevsUM ORIFICE Revision 0 UNIT I March 13,2002
- CR 99-1391, SCAQ on Potential Common Mode Failure Mechanism Affecting Welds In AFW Pump Recic. Line.
, Wisconsin Electric Power Company, Point Beach Nuclear Plant - RCE 99-081, "Socket - Weld Failures In Auxiliary Feedwater Pump Recirculation Piping".
S.Bechtel Specification No. 6118-M-6, Rev. 3, "Specification For Auxiliary Feedwater Pumps Point Beach Nuclear Plant Units 1 & 2 Wisconsin-Michigan Power Company, dated 10125168.
- EPRI TR-1 II188, "Vibration Fatigue Testing of Socket Welds". Interim Report, Decenhiber 1998.
- EPRI TR-107455, "Vibration Fatigue of Small Bore Socket-Welded Pipe Joints", Final Report, June 1997.
DESIGN DESCRIPTION AND ANALYSIS This modification will replace 1RO-4003 installed in the minimum recirculation line for AF pump 1P-29, with a new type of orifice. The presently installed RO was accredited with causing flow induced cavitation causing excessive noise and vibration in the Auxiliary Feedwater minimum recirculation piping system. The replacement [
RO will have the same function as the existing orifice, which is to provide pressure reduction and act as a pressure boundary for the AF system piping.
A comparison of the mechanical and flow performance characteristics of existing vs. new RO indicates that OT RO will provide an improved anti-cavitation characteristics and replacement RO is equal or better. The replacement thus will minimize hydrodynamic noise and vibration under liquid application. The original design requirements for I.
the RO are specified in the Bechtel Specification No. 61 18-M-6, Rev. 3, "Specification For Auxiliary Feedwater dated 102868. T.his Pumps Point Beach Nuclear Plant Units I & 2 Wisconsin-Michigan Power Company, sh.O, ..
-*ki*,t.ik,-,: idoes imt address- d. dconstiu-cto"-istec.fi:es.fothis.4i-.lce- It -pecifes .that-, ,.hp.u furnished with a pressure reducing orifice to be used in conjunction with the on-off control valve in the pump recirculation piping. The orifice shall be provided with ended weld connections for installation in AF piping. I the,
'flow through the orifice may cause erosion, special materials, such as 316 stainless steel, shall be used."
The design of the new RO is different than the presently installed orifice. The existing RO uses inner orifice plates to control the flow and pressure drop across the orifice. The new RO works in a similar manner except that control of flow and pressure drop is accomplished by directing the flow through the series of close-fitting cylindrical stages, each constructed with expansion holes and intersecting circumferential channels that restrict the flow. This flow the path of multiple restriction and enlargements reduces the pressure gradually across each trim cylinder, avoiding sharp pressure drop typical to conventional, single-throttling orifice.
(RO The orifices that were installed by MR 99.029*A/*B for the P-38A/B motor driven auxiliary feedwater pumps 4008/4015) were not adjustable. After installation of one orifice, the flow was lower than expected, and additional the holes needed to be drilled in the orifice inner cylindrical stage. It is for this reason that the orifices installed for turbine driven auxiliary feedwater pumps have the capability of being adjusted.
this In a letter dated 312/2001 from Flowserve, the stated minimum recirculation flow for IP-29 is 75 gpm, but then to be inspected after 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> of operation at this flow. If the pump is operated at 130 gpm, requires the pump Page 2 of 5
FINAL DESIGN DESCRIPTION MR 99-029"C AUX ORIFICE FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE iRevision 0 UNIT I March 13,2002 up to 1500 hours0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> of service can be accumulated before maintenance is required. Based on calculation N-91-032 and stated in FSAR Section 10.2.3, the current maximum flow through the recirculation line with the control valve failed open is 126 gpm. The new orifice will be set to approximately the same flow rate (between 120 and 130 gpm), even though that this is below the 1500 hour0.0174 days <br />0.417 hours <br />0.00248 weeks <br />5.7075e-4 months <br /> limit Increasing the flow would require additional analysis since it would reduce the available flow to the steam generators.
The replacement RO is contained in a 2"- 600# cast stainless steel globe valve body (ASME A 351 Type CFoM) and designed to the requirements of ASME B31.1 and ASME B16.34 - 1996 Edition. The working pressure is 1440 psig at 100 "F, meeting the Pipe Class 2"-DB-3 requirements. The flow rate of the RO can be adjusted during operation with the system pressurized. The RO will not be designed to shut-off flow. The adjusting device will be positively secured in its position using k lockwire attached to the stem and bonnet.
The currently installed RO, was designed and constructed by the Byron-Jackson Company (BICO), and installed under MR 88-099. The replacement RO was procured from the Flowsem-e Company under P.O.# 4500xxxxxx.
Included is a design and seismic report qualifying the RO for use in this application. A hydrostatic pressure test of the replacement RO shell was performed at the Flowserve facility in accordance with ASME B 16.34, except that the test pressure was maintained for at least 30 minutes.
In addition to RO replacement, some of the existing piping associated with RO will also be replaced. The piping to be replaced is shown on construction sketch SK-MR-99-029*C and includes a 900 piping elbow upstream of the IRO-4003 to the upstream socket weld on the isolation valve lAF-15. This piping replacement is being done to simplify the installation and to allow for the installation oversized socket welds. The replacement piping and RO will be joined by socket welds which are oversized in a "2/1" configuration, with an axial dimension approximately twice that of the radial dimension, as recommended by the EPRI technical reports. The oversized socket weld detail is shown on construction sketch SK-MR-99-029*C. All welds in the recirculation line up to valve 1AF-15 will be oversized, with the exception of the buttwelds at the 1FE-4049 flanges.
EPRI technical reports TR-107455 and TR-1 11188 address the issue ofhigh-cycle fatigue failure of socket welds at nuclear power plants. Significant research was performed on the subject, and a large number of failed welds from plants were examined. It was discovered both through analysis and examination that fatigue failures were less likely to occur when the axial size of the socket weld is nearly twice that of the radial size. As a result, these reports recommend installing these "2/1" socket welds in applications that experience high-cycle fatigue. Although the will be installed to prevent any future weld failures.
The piping to be replaced is classified as Pipe Class 2"-DB-3. This Pipe Class specifies carbon steel materials, however due to corrosion concerns the recirculation line piping was installed as stainless steel (per MR 88-099).
Thus, replacement piping and piping components will be also stainless steel.
The replacement piping material for the proposed modification is ASTM A-312 Grade TP 316. The replacement piping fittings material is ASTM A-182 Grade F 304. The replacement piping is 2" Schedule 80, and the fittings are 3000# class, which will meet the pressure and temperature ratings for Pipe Class 2"-DB-3 (1440 psig at 100 9F).
The replacement RO is heavier than existing one, and it will add approximately 40 lbs to the existing AF piping system. In addition, he replacement piping assembly will have a slightly different internal length of piping than the existing piping layout. However, face-to-face length of the replacement pipe spool piece will be exactly the same as the existing one. These differences between the existing and proposed piping configurations have been addressed by an addendum to Piping System Qualification Report WE-100070 which will demonstrate ASME B31.1 compliance of the modified piping.
Page 3 of 5
FINAL DESIGN DESCRIPTION MR 99-029"C AUX ORIFICE FEED WATER PUMP iP-29 MINIMUM FLOW RECIRC LINE Revision 0 UNIT 1 March 13,2002 In addition, the flow characteristic of the replacement RO and its affect on the associated plant calculations was evaluated. This evaluation was documented in Addendum N-91-031-00-A to Calculation No. N-91-031, "1 & 2 P 29 Mini - Recirc Line System Characteristics", Rev.0 and Addendum N-91-032-00-A to Calculation No. N-91-032, "Comparison of Nominal Flow Rates from 2P-29 to 2HX-1A and 2HX-lB with the Recirc Line Open", Rev. 0. The results of this evaluation found that the slight differences in the flow characteristic between existing and replacement pressure reducing orifices is acceptable and does not significantly alter the above calculations results.
None of the above changes is introducing a new, unknown equipment to PBNP. Furthermore, replacement components awe passive in nature when the system is operational and will be designed, installed and tested in accordance with existing procedures and controls.
To implement this modification, the portion of the AF piping will be cut at the socket weld at valve IAF-15 and disconnected at the IFE-4049 flange. This disassembly is shown on Sketch SK-MR-99-029*C. Piping, and pipe components removed will not be reused for this modification. The only exception is the IFE-4049 flange and its associated pipe stub up to the first 900 elbow. This assembly will be inspected and then reused. To assure high quality of socket welds, a replacement piping spool piece (containing the new RO) will be fabricated in the shop in accordance with details provided by Construction Sketch SK-MR-99-029C.
Implementation of this modification will reduce the possibility for line noise and vibration when operating this line in the recirculation mode.
Design pressure, operating pressure, design temperature and other pertinent design parameters for RO are specified in the Data Sheet attached to purchase order P003467.
No procedure changes result from this modification. This is a physical replacement of a RO and associated portion of the auxiliary feedwater system. There will be no additional components added or operating modes changes that will require operating procedure changes.
Welding for this modification will be performed in accordance with welding procedure WPM 2.P-1--GT and WPM 2.PS-GT.
The RO will be tested at a calibrated flow test facility. The RO will be adjusted accordingly during this test to pass
--a~opj t9 those the . A-9y~Isni fter installation, flow will be verified by other multiple flow instruments, and the orifice can"e adjusted accordingly, f necessary.
NDE requirements for the Pipe Class affected by the proposed modificon are specified in DG-M02 and the original code of construction, USAS B31.1 - 1967. They require the finished socket welds to receive a Visual Examination (VI). The affected existing welds have a history of failure, therefore, in addition to VT of the final socket welds, root welds will receive VT and Liquid Penetrant Examination (PT). Piping socket welds shall be examined utilizing the acceptance criteria of ASME B31.1 - 1992.
As required by ASME B31.1. an initial service leakage test will be performed at normal operating pressure and temperature (with the 1P-29 auxiliary feedwater pump running).
I.
Page 4 of 5
FINAL DESIGN DESCRIPTION MR 99-029*C AUX FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE ORIFICE Revision 0 UNIT I March 13, 2002 D)FSGN OUTPUT The Installation Work Plan IWP 99-029"C will be prepared to identify installation requirements including pre operational conditions, installation testing and post installation testing requirements. In addition, a 10 CFR 50.59 Safety Review (SCR 2001-0981) has been prepared to evaluate the proposed change to PBNP.
The following caledations were prepared to address the proposed modification:
- Addendum to WE Piping System Qualification Report WE-100070
- Addendum N-91-031-00-A to WE Calculation No. N-91-03 1,Rev. 0
- Addendum N-91-032-00-A to WE Calculation No. N-91-032, Rev. 0 The following Installation Work Plan Is associated with this modification:
"WP 1 99;029*C (WO 0202509), Aux Feed Water Pump IP-29 Minimum Flow Recirc Line Orifice - Unit 1
- WO 0202507, Welding of stub pieces onto orifice for offsite flow testing
- WO 0202508, Prefab work for MR 99-029*C The following construction sketch Is associated with this modification:
0 SK-MR-99-029*C, Auxiliary Feedwater System Orifice 1RO-4003 Replacement, Unit 1 Other documents:
4 10 CFR 50.59172.48 Safety Review, SCR 2001-0981
- Flowserve Drawing 94-16249 6 Fire Protection Conformance Checklist, PBF-2060, PBF-2060e
. eI.
Pap 5 of 5
DOCUMENT UPDATE CHECKLIST Plant Modification/Minor Plant Change No. MR 99-029*C Work Order No.:
- I iVIR2 DOCUMENTATION UPDATE SHEET AND CLOSEOUT CHECKLIST Required For Release Acceptance Closeout N/A (Completion) (Submittal)
A. TRAINING
- 1. Copy Submitted to Training (Design Description)
- 2. TWRGenerated(TWR# O C'-. )tef.SM1MLC1.I
- 3. Simulator Changes Initiated (SDR # )
- 4. Plant Status Updateflust In Time Training B. FINAL DESIGN ORGANIZATION
- 1. Drawings
- a. Design Change In Progress DCN's Initiated
- b. Construction sketches Issued
- c. Revised Drawings Issued for Priority I and 2 Control Room Drawings - Logics, P&IDs, 499 series elementaries.
- d. Revised Drawings Issued for Work Control Center Drawings P&IDs
- e. Revised Drawings Issued for I&C Drawings - Reactor Protection and Safeguards Elementaries.
- f. Master Ifata Book - Control Room Work Control Center, and Local Panel - PBF-2093
- g. DCN's released for incorporation
- h. Sketches Voided - PBF-1592
- 2. Specifications (Conformed at Closeout, ref NP 9.21)
- 3. Component Instruction Manuals (for issue, revision, deletion) x PBF-1586
- 4. Cable and Raceway Data Schedule Revisions - PBF-0091 T. 5. Environmental Qualification Documentation Updates - Ref. NP I
ti-A 7.7.1
- 6. Seismic Qualification Updates NP 7.7.2
- 7. Calculations or engineering evaluations added/deleted I revised I PBF-1608 X 8. DBD Revisions - PBF-1653
- 9. PSA Models and Documentation - PBF-1626
-. - - -. 1 PBF.1606 Revision 6 1o0/2/0i Page I of 4 Reference(s): NP 7.2.1
DOCUMENT UPDATE CHECKLIST Plant Modification/Minor Plant Change No. MR 99-029*C Work Order No.:
Hor7 ENEi "IDOCUMENTATIONUPDATE SHEEr AND CLOSEOUT CHECKIuST Required For Release Acceptance Closeout (Completion) (Submittal)
S*.. x.; C. LICENSING (Conformed at Acceptance)
X x-.. A* 1. Technical Specification - change; specify section(s) affected and
_ _ _._ _change request number.
X ___ I_2. Tech Spec BasslaTechnical Requirements Manual X ** :" .*! -*aluminurn inventory list with FSAR update.
x . FSAR -change; NP 5.76. Report major changes to the containment x 4. EPER - FHAR - SSAR Revisions - NP 5.11 X a. Safe0 Shutdown Analysis Management 7?
System Revisions - NP X _ 5. Offsite Dose Calculation Manual (ODC)
X _,___ 6. Radiological Effluent Control Manual (RAM X ____ M____ ____ 7. Emergency Plan and EPIPs X ________ -8. Notification to Security for Security plan update
- 9. Report major changes to radwaste treatment system withannual
- ______ ___ __ ~FSAR update pecr REQM 1.6.3 V'@r mz D. CHAMPS DATABASE x 1. Equipment Identification - additions assigned from CHAMPS
- 2. Permanent Labeling - labels on new equipment; PBF-9900
- 3. Temporary Labeling - labels on new equipment; PBF-2074
- 4. Equipment Record - update to CHAMPS coordinator specify chanze(s): PBF-9922 PBF-9925. PBF-1023
- 6. Unused material removed from modification bin.
E. OPERATIONS
- 1. Abnormal Operating, Normal Operating, System Operating, and Refueling Procedures - PBF-0026a
- 2. Operating Instructions and Checklists - PBF-0026a
- 3. Alarm Response and RMS Alarm Setpoint and Response Books
.1 PBF-0026a A. T.@tine - T. 1? APT nthir - P1F-flfl2
-"___ 5. EOPs, ECAs, CSPs, SAMG's - PBF-0026a
- 6. Periodic Surveillances - PBF-9920 j7. Fire Protection Procedures - PBF-0026a
...... * - .. *1 9 .
Tank LeelBo T e v el" -.
PBF-O026a PBF- 1606 Revisbn 6 10/02/01 Page 2 of 4 Refcracc(s): NP 7.2.1
DOCUMENT UPDATE CHECKLIST Plant Modification/Minor Plant Change No. MR 99-029*C Work Order No.:
Ti DOCUMEqTAION UPDATE SHEET AND CLOSEOUT CHECKLIST Required For NIA Release Acceptance Closeout (Completion) (Submittal)
F. MAINTENANCEWI&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-992M9920 x 5. Ensure station batteries' load profile changes are incorporated into the appropriate discharge test RMPs.
x 6. Lubrication Manual (NP 7.3.11)
- 0. SECURITY x 1. Security Procedures H. ENGINEERING/MISC.
x 1. ISI Procram x 2. IST Program x 3. M*scellaneous FX ECTlCleaning program
- 4. Reactor Engineering Instructions - change; specify section(s) x affected.
x ______ ______affected.
- 5. Reactor Engineering Procedures - change; specify section(s) x _____"__request
_ 6. Software Control - specify system affected and sokwe change number.
x 7. Component maintenance program.
x 8. Governing calculations and models (e.g., SW model, DC loading, EDO loading, piping analysis, structural loading, etc.).
x 9. Design Guidelines (ref. NP 7.12)
L OTHER (CHEKI HP, ETC.)
X 1. Other (lisc. Procedures, etc.)
011 I. ECRs I. ECR Final Resolution completed and approved by Design
_.___,, ___;___ __ _ -,Supervisor.
- 2. ECR Imvlementation completed.
P.---
+/-
. -; w ______________ I
- 2. ECR limlexnentation comvlcted.
PBF-1606 Revision 6 t0O02/01 Page 3 of4 PReftiwgs): NP 7.2.1
Prior to Prior to Prior to Change No.
Section Specific Uvdates Reqnired Aceuetance Closeout (if Apnlicable) /IB/ Date A.1 Design description submitted to Relas Dl
_ training. El ZS ,4ý A. e-n,(%
A.2 TWR generated.
El A.4 Plant Status Update / J1T training .n El notified of modification.
B.l.b SK-MR-99-029*C issued.
Dl B.lI.h SK-MR-99-029*C voided.
0 0 B.3 Update to FLOWCO CIM (01708).
B.7 Addendum to N-91-031.
0 0 B.7 Addendum to N-91-032. El Dl VlO-.z M& ta)A j 4cc-.
B.7 TR 01.103 approved.
B.8 Revision to DBD-01, Auxiliary Feedwater. 0 OE1*5..%-W O~a-Ac B.10 Notify EPIX coordinator.
D.2 Permanent label for 1RO-4003.
El 0 Nux4 p,, :)iaai dO4'-o1 D.3 Temporary label for 1RO-4003.
Dl 0D: El 1 ?4-Ntn-'Ck V'dM I-W
-J-,r *,. I-,'1VAI'~& %W04-6 2£tt A..C._C-
- D.4 CHAMPS updated.
H.8 Addendum to WE-100070. D 0 -,4*SUI U-*"-o 1C. im.c LI Vew drawing FLOWCO 94-16249. El1 0 9c t-oh1-W jd7-r- O~
L. DCN to BECH P-103 (if necessary). El 0 o 00 ZOOL.-i+g. O 1 .x -212. z Oz -.
El El El E ___
El n El El __
El E ___
I.
PBF-1606 Revision 6 10/02A)1 Page 4 of 4 PP44frzence(s): NP 7.2.1
I Point Beach Nuclear Plant SCR 2001-0981 10 CFR 50-59f72.48 SCREENING (NEW-RULE) Ver*fy SCR umuzi nan pVq=
Page 1 de of Proposed Activity: MR 99-029"C/*D - AUX. FEED WATER PUMP 1/2P-29 MINIMUM FLOW RECIRC. LINE Associated Reference(s) #: MR 99-029"C/*D, EWR 99-031, CR 99-1391, ASME B31.1, RCE 99-081, MR 99-029*A/*B Prepared by: Rob Chapman , Date: l 2.- -1o Name (Print) C ature Reviewed by: John P. Schroeder /.IDatcL:.a Name ( Print) Signature PART I (50.-917248) - 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. Avpendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 10 CFR 50-9 and 10 CFR 7248 screenings.
LI Describe the proposed activity and the scope of the activity being covered by this screening. (The 10 CFR 50.59172.48 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.
Engineering Work Request (EWR)99-031 was Initiated requesting the evaluation of high noise level and vibration present In the Auxiliary Feedwater (AF) pump recirculation lines during their operation In minimum recirculation mode. This evaluation had determined that the Installed flow restricting orifices (ROs), are cavitating and causing excessive noise and vibration In the associated piping. In addition, Condition Report CR §% 91`Was ffitiated to address the Issue of pinhole leaks In the socket welds which have developed at the existing ROs. To improve the socket weld's cycle fatigue response over standard ASME Code socket weld profile in vibration critical application the root cause evaluation RCE 99-081 recommended replacing the orifices to prevent cavitation, and Increasing the size of the socket welds. This modification was already performed for the motor driven auxiliary feedwater pumps (P-38AJB) by MR 99-029*A/*B.
The purpose of the proposed modifications is to minimize piping line noise and vibration and preclude socket weld failure when operating the pump on minimum recirculation mode. MR 99-029"C and MR 99-029'D will replace the existing orifices IRO-4003 and 2R0-4003 In the AF system with improved design orifices. In addition, portion of the AF piping associated with RO will be replaced to simplify the installation and to facilitate Increasing the socket weld size. Piping will be replaced upstream of the orifice Include some elbows up to and Including the upstream weld on the AF pump recirculation line Isolation valve lAP-IS for pump 1P-29 and valve 2AF-53 for pump 2P-29.
The replacement ROs differ from presently Installed ROs. The existing ROs have orifice plates to reduce the flow and pressure through the unit. The replacement ROs work In a similar manner except that control of flow and pressure drop Is accomplished by directing the flow through the series of dose-fitted cylindrical stages, each constrlcted with expansion holes and intersecting rcumferential channels that restrict the flow.
These cylinders are placed in a 600# class globe valve body with a valve stem and disk that allows adjustment of the flow setting after Installation. This trim will not allow complete shutoff. The flow will be set to a nominal value of between 120 and 130 gpm, which is essentially the same as the existing orifice.
(.T'h" A seismic analysis and report, to determine that the orifice wim operate during and after seismic event was determined by WE Seismic Qualification Group not to be required for these ROs. This determination was based on rugged design of the ROs body and pressure reducing componentL PBF-1LR Reiinc 02d1n llO 0 4, 2001 R~
Point Beach Nuclear Plant SCR 2001-0981 10 CFR 5OS9f/2.48 SCREENING (NEW RULE) vWXY SCR Rab o-al pq=
Page 2 The proposed modifications will meet design, material and construction standards of the existing installation.
The Implementation of the proposed modifications, will not affect the overall performance of the AF system, operation or function of the AF pumps 1P-29 and 2P-29 and the ability of AF system to perform Its Intended safety functions.
Post modification testing will Include a visual exam (VT) of all replaced piping socket welds. Piping welds will be examined In accordance with ASME B31.1 - 1992. Performance of this exam is required by both the original piping specification, Bechtel M-78, and the original code of construction, USAS B3L1- 1967.
Additional NDE will be performed on the root welds for additional assurance of weld quality. B3L1 also requires that post modification testing Include an Initial service leak test at normal system operating pressure and temperature, wirch will be performed with the pump running. In addition, a functional test and verification of the flow through the replacement ROs will also be performed.
The proposed modification MR 99-029*D Is scheduled to be Installed during U2R25, and MR 99.029"C is scheduled to be Installed during U1R27. These modifications will be Installed while the unit Is In Mode 4, S or 6, when the turbine driven AFW pumps IP.29 and 2P-29 are not be required to be in service per LCO 3.7S.
Upon completion of each modification, the new Installed RO will perform the same function as the existing orifices IRO-4003 and 2RO-4003.
12 Search the PBNP Current Licensing Basis (CLB) as follows: Final Safety Analysis Report (OSAR), FSAR Change Requests (FCRs) with assigned mnnbers, the Fire Protection Evaluation Report (FPER), the CLB (Regulatory) Commitment Database, the Technical Specifications (both Custom and Improved), tde 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 caskAISFI as appropriate. Identify where the pertinent information is described in the above documents (by document 0 section number and title). (Resource Manual 5.3.1 and NEI 96-07. App. B, B.2)
"* FPER,Auxifiry FeedwaterSystem, F'gre6.6- 4a.
"* FSAR Section 1., GeneralDesign Crteria
"* FSAR Section 10.1, Steam andPower Conversion System
"* FSAR Section 10.2, AMiary FeedwaterSystem
"* FSAR Section 14.1.9, Loss of FxternalElectricLoad
"* FSAR Section 14.1d10, Loss ofNormalFeedwater
"* FSAR Section 14.111, Loss ofAIZAC Powerto the Auxlaries S............... .....F,.AR~Secdqor'1424,* Ste,-'n (rrno*7tebo feT.tur*e ... .........
..... ...... ~..,-*.... -**-.-..*. ..... ..
- TS 3.75,AFW System The flow restricting orifices for the turbine driven auxiliary feedwater pumps (11RO-4003) have the, C following design functions, as described In FSAR Section 10.2. Irv !*
" They serve to restrict the recirculation Dow for the pumps to ensure adequate auxiliary feedwater flow to ,
the steam generators In the event that the minimum flow recirculation control valve (IJZAF-4002) falls to dose.
" They ensure adequate flow and pressure drop through the auxiliary feedwater pumps when they are operated in r tion mode, thus preventing low flow instabilities and excessive fluid temperatures.
" They passively maintain the auxiliary feedwater system pressure boundary Integrity.
..L3 Does the proposed activity involve a change to any Custom or Improved Technical Specification (ITS)? Changes to C Technical Specifications require a License Amendment Request (Resource Manual Section 53.12.).
Technical Specification Change: 0 Yes 0 No PBF-i l.*
Revision 0 10/24101 Refernce: NP5.t.
Point Beach Nuclear Plant SCR 2001-0981 10 CFR 50.9172.48 SCREENING (NEW RULE) vc* sCR n r pags Page 3 If a Technical Specification change is required, explain what the change should be and why it is required.
L4 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.
OYes ONo 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 II (50.59) -DETERMINE IF THE CHANGE INVOLVES A DSIGN FUNCTION (Resource Manual 5.3.2)
Compare the proposed activity to the relevant CLB descriptions, and answer the following questions:
YES NO QUESTION
-0D 0 Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?
0 0 Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?
0 0 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 Does the proposed activity involve CLE-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?
EO 0R Does the activity involve a methodofevaluationdescribed in the'FSAR?
O3 0 Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)
[o 0 Does the activity exceed or potentially affect a design basislimitfor afissionproductbarrier(DBLFPB)?
(NOTE IfTHIS questions is answered M a 10 CFR 50.59 Evaluation is required.)
d answers to ALL of these questions are KO_, mark Partma Einot applicable, document the 10 CFR 50.59 `smreeni~ngvin- thie conclusion section (Part IV), theen proceed direty to Part V- 10 CFR 72.48 Pre-screening Questions.
If any of the above questions am marked MES. identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.
The flow restricting orifices for the turbine driven auxiliary feedwater pumps (lfRO-4003) have the following design functions that are affected by MR 99-029"C/*D:
"* They serve to restrict the recirculation flow for the pumps to ensure adequate auxiliary feedwater flow to the steam generators In the event that the minimum flow recirculation control valve (LfAF-4002) falls to dose.
t
"* They ensure adequate flow and pressure drop through the auxiliary feedwater pumps when they are operated In rechrcnlation mode, thus preventing low flow Instabilities and excessive fluid temperatures.
- o They passively a-int-in the auxiliary feedwater system pressure boundary Integrity.
PBF--I$1c Revision 0 0/24/0I We==m*NP5.l.9
Point Beach Nuclear Plant . SCR 2001-0981 10 CFR so.59//2.48 SCREENING (NEW RULE) Vzfc SCR mumber an lpags Page 4 PART 111 (50.9) - DETERMINE WHETHER THE ACMIVITY INVOLVES ADVERSE EFFECTS (Resource Manual 5.3.3)
If ALL the questions in Part II am answered Q, then Part III is 0 NOT APPLICABLE.
Answer the following questions to determine if the activity has an adverse effect on a design function. Any WS answer means that a 10 CFR 50.59 Evaluation is required; EX where noted in Part 111.3.
III CHANGES TO THE FACILITY OR PROCEDURES YES NO QUESTION
[3 0 Does the activity adversely affect ft designfiction of an SSC credited In safety analyses?
o 0 Does the activity adversely affect the method of performing or controlling the designfuncdon of an SSC credited in the safety analyses?
If any answe is = a 10 CFR 50-59 Evaluation is required. If both answers are &Q, describe the basis for the conclusion (attach additional discussion as necessary):
The replacement of flow restricting orifices IIZRO-4003 by MR 99-029*CI"D will not adversely affect their design functions. Although the new orifices are of a different type, they will perform the same functions to allow flow to maintain TDAFP operability when in recrculation mode and to restrict flow If the recirculatlon control valve fails open. The method of performing these functions Is slightly different, and the capability will be added to adjust the flow, but this will not adversely affect these desfgn.(uEons. The new orifices will provide essentially Identical flow through the recirculation line, but with Improved flow characteristics that will prevent cavitation. The orifice bodies are designed to ASME standards and have ratings that exceed that of the auxiliary feedwater piping. Non-destructive examination of the new welds and functional testing of the orifice will ensure that all design basis requirements are met.
These orifices are not explicitly required in an accident analysis to be able to pass service water, since the 4%.
recirculation control valve would be closed when the pump is aligned to the steam generator. However, it is possible that when the pump Is aligned to the service water systen-supply after the condensate storage tanks have been drained, service water could be pumped through the recirculation lines. To preclude the chance of dogging the orifice trim, the flow is directed from the outside of the stages Inward. The holes In the outer stage are the smallest, and they get progressively larger In the Inner stages. This causes the largest differential pressure to exist at the outer stages at locations with the smallest holes, which will reduce the potential for debris accumulation inside the orifice.
E1.2 CHANGES TO A METHOD OF EVALUATION (If the activity does not involve a method of evaluation, these questions are 0 NOT APPLICABLE.)
YES NO QUESTION 0 03 Does the activity use a revised or different method of evaluation for performing safety analyses than that desc=ibed in the CLB?
0 El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described Inthe CLB?
If any answer is YiS a 10 CFR 50.59 Evaluation is required. If both answers are K0 describe the basis for the conclusion (attach additional discussion, as necessary)
PBF-1515c
Reference:
NP 5.18 Revision 0 10/24/01
Point Beach Nuclear Plant . SCR 2001.0981 10 CFR 50.9/72.A8 SCREENING (NEW RULE) SR a =
Page 6
- NOT 10 CFR 72.48 SCREENING NOTE: NET 96-07, Appendix B. Guidelines for 10 CFR 72A8 Implementation should be used for guidance to determine the proper responses for 72A8 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
[3 Does the proposed activity involve IN ANY MANNER the dry fuel storage cask(s), the cask transfr/transport equipment, any ISMI facility SSC(s), or any ISESI 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/&SFSI Continuous Temperature Monitoring System?
[3 10 Does the proposed activity involve IN-ANY MANNER SSC(s) installed in the plant specifically added to support cask loading/unloading activities, as foblows: Cask Dewatering System (CDW), Cask Reflood System (CRF), or Hydrogen Monitoring System?
[3 [] Does the proposed activity involve IN ANY MANNER SSC(s) needed for plant operation which ame 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?
-$ 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.?
o[ Does the activity involve plant heavy load requirements or procedures for areas of the plant used to support cask loading/unloading activities?
O I] 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 MQ, then check Parts VI and VII as not applicable. C lete Part VIII to do ument 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 AM the questions in Part V are &Q then Part VI is ID NOT APPLICABLE.)
Compare the proposed activity to the relevant portions of the ISFS licensing basis and answer the following questions:
YES NO QUESTION O1 -1 Does the proposed activity involve ckJISFS1 Safety Analyses or plantcaskfiSFSl structures, systems and components (SSCs) credited in the Safety Analyses?
o] Does the plbposed activity involve plant, cask or ISFSI SSCs that support SSC(s) credited in the Safety Analyses?
o 0 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?
o [3 Does the proposed activity involve caskIISFSI described SSCs or procedural controls that perform functions that are
- required by, or otherwise necessary to comply with, regulations, Hcense conditions, CoC conditions, or orders?
0 0 Does the activity involve a method ofteyluation described in the ISFSI licensing basis?
O 0 Is the activity a test orexperiment? (Le., a non-passive activity which gathers data)
PBF-1515c Revision 0 102.40i Refcrence: NP 5.1.8
Point Beach Nuclear Plant - SCR 2001-0981 10 CFR 5o..9r72A8 SCREENING (NEW RULE) VaifY SC number on all pages Page 7 3 l Does the activity exceed or potentially affect a cask design basis lfinitforafifsionproductbarier(DBLFPB)?
S
)0 (NOTE: If THIS questions is answered YES, a 10 CFR 72.48 Evaluation is required.)
If the answers to ALL of these questions are fQ, mark Parts VII as not applicable., and document the 10 CFR72.48 screening in the conclusion section (Part VIII).
If any of the above questions arc marked M_., identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.
PART VII (72A8) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS M 96-07, Appendix B, Section BA.2.I)
(If ALL the questions in Part V or Part VI are answered Q., then Part VII is N 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; EX where noted in Part VIL3.
VILl Changes to the Facility or Procedures YES NO QUESTION
[3 Does the activity adversely affect the design fincton of a plant, cask. or ISFSI SSC credited in safety analyses?
[o - Does the activity adversely affect the method of performing or controlling the designfinctionof a plant, cask, or ISFSI SSC credited in the safety analyses?
If any answer Is Y a 10 CFR 72.48 Evaluation is required. If both answers 'areKt, describe tih basis for the conclusion (attach additional discussion, as necessary):
VIL2 Changes to a Method of Evaluation (lIfthe activity does not involve a method of evaluation, these questions are 0 NOT APPLCABLE.)
YES NO QUESTION 0 [1 Does the activity use a revised or different method of evaluation for performing safety analyses than that described in a cask SAR?
0 0 Does the activity s 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 N.Q, describe the basis for the conclusion (attach additional discussion, as necessary):
PBFisio1c Rcs'iio1 0 10240
Reference:
N4P5.1.
Point Beach Nuclear Plant . SCR 200140981 10 CFR 50..5912.48 SCREENING (NEW RULE) Vedrif SCR umbr an anl M=
Page 8 VI 3 Tests or Experiments (If the activity is not a test or experiment, the questions in VIL3.a and VIL3.b are 0l NOT APPLICABLE.)
- a. Answer these two questions first YES NO QUESTION O El Is the proposed test or experiment bounded by other tests or experiments that are described In the cask ISFSI ficensing basis?
o 0 Arm the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISM facility?
If the answer to both questions is 0-, continue to VI.3.b. If the answer to E WER question is ] then briefly describe the basis.
- b. Answer these additional questions O for tests or experiments which do not meet the criteria given in VIL3.a above.
If the answer to either question in VIL3,a is M then these three questions are " NOT APPLICABLE:
YES NO QUESTION o] 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?
o o Does the activity utilize or control a plant, cask or ISSI facility SSC in a manner that is inconsistent with the analyses or descriptions in the ISFSI licensing basis?
[1 0" 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 VIL3.b is _YES aI0 CFR 72.48 Evaluation is required. If the answers are all K describe the basis for the conclusion (attach additional discussion as necessary):
PART VIII- DOCUMENT THE CONCLUSION OF THE 10 CFR 72.48 SCREENING Check all that apply:
A 10 CFR 72.48 Evaluation is [] required or r"1 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 [I NOT required. If a VSC-24 cask SAR change is required, then contact the Point Beach Dry Fuel Storage group supervisor.
A Regulatory Commitment (CLE Commitment Database) change. is El requiredor 0l 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 [] required or [E 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.1515c Revision 0 10/24/01 Referencc NP 5.1.9
NUCLEAR POWER BUSINESS UNIT DESIGN VERIFICATION NOTICE Tide of Document AUX FETE WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE
... Document No. MR 99-029*C Rev. 0 Date 2-18-02 Design Verification Method: Design Review [Alternate Caics; [lQualification Testing UPDATES TO THIS FORM COVERED BY EXISTING SCR 97-410 REVIEWER CHECKLIST CONSIDERATIONS:
NoQ llA
- 1. Were the inputs correctly selected and incorporated into design?
- 2. Ar assumptions uecessary to perform the desin activity adequately described and reasonable? -Where necessary, are the assumptions identified for subsequent revcrifications when the detailed design activities are completed?
/
- 3. Are the appropriate quality and quality assurance requirements specified?
J/
- 4. Are the applicable codcs, standards, and regulatory requirements including issue and V
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?
/
., 1It. Have adequate maintenance features and requirements been specified?
- 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?
- 15. Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisfactorily accomplished?
16." Have adequate pre-operational (IST, PMT, ISI snubber, etc.), subsequent periodic test, and inspection requirements been appropriately specifi6d, including acceptance criteria? ,
- 17. Are adequate handling, storage, cleaning, and shipping requirements specified? -
- 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?
I-.
- 21. Has a field inspection been done?
- 22. Have i .on ot4er systems been identified?
COM LW:LeJ None [1 Attached (Use Form PBF-1633)
Design Prepared By-. Rob Chapman Date
",,',-.Reviewed By: Jeff Novak D
'Approval By:
Dt .3 /)VO7 1 1 PBF-1583 Pap I oflI Re&cenc: NP 7.2.2 Revision I 08/08/97
Point Beach Nuclear Plant DESIGN INPUT CHECKLIST Adifcation or'Temporary Modification Number. MR 99-O29.C Tide: AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE 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. ADl 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 97411.)
APPLIES TO DESIGN YES no A. General codes, standards, regulatory requirements, and design criteria.
Identify and address design criteria as appropriate.)
QDC I - Oualitv Standards: This modifation will be instailedQA,1R, andall new components will be verified to be of suffiem quality to ensure the performanceoftheirsrfetyfunctions.
GDC 2 - PerfomanceStandards: All component are installed seismic Class 1, andthus will perform theirsafetyfinctions during a design basisearthquake.
GDC3- FireProtection: The RO will perform itsjnctionto support the safe shutdown of Unit I in the Appendix R scenario.
GDC$5-RecordsReauirement This modicationpackage will satisfy the recordsrequiremen=
GDC 37- EngineeredSafety FeaturesBasis for Design: This RO was chosen with an appropriateC*
to ensure that the IP-29 auxiliaryfeedwaterpump will provide sufficientflow to the Unit I steam generatorsas neededL GDC41 - EngineeredSety FeaturesPerformanceCapability: Thi RO was chosen with an appropriateC, to ensure that the IP-29auxiliaryfeedwaterpumpwillprovide sufficientflow to the Urdrt I v'tea-- generatr;;a neeeed rThe RO vidl be tasted aifter installatidnto ver65, appropriateflow.. -
GDC42- EnineeredSafety FeaturesComaonents Camablitv: Thi RO hasdesign ratingsthatmeet thatof thisportionof the auxiliaryfeedwatersystem, andhas been designed to pass a certainflow.
Non-destructiveexaminationandpressuretestingfollowing modification installadonwill verify that all new welds were installedcorrectly. Flow testing will verif theflow setting.
- 2. Are any design requirements contained in commitments affected? (Reference CL.B database and the Safety Evaluation/Screening associated with this change.)
Reference 50.59 screening 2001-0981.
- 3. Meet State of W'isonsin Administrative Code requirements? (Refer to LHR 41.42, PSC 114, and 0 other sections as appropriate for requirements.)
- 5. Consider the effect of design and accident conditions, such as pressure, temperature, fluid chemistry, o and radiation on components, including internal elastomers and material coating compatibility.
PBF-1584 Revision9 11/SA)0 Page 1 of 13 Refcmm: NP 7.2.2
"APPLIES TO DESIGN DESIGN INPUT CHECKLIST (Changes in design parameters may impact Environmental Qualification.)
.. Replacement orifice, piping.andfatings all have design ratingsthat are adequate to ensure that the auxiliaryfeedwatersystem will perform it requiredsafety finction during a design basisaccident.
- 6. Incorporate new types/models of equipment not presently used at PBNP? El 0
- 7. Affect accessibility of any equipment? Consider interim conditions, future maintenance, and in-service inspection. (Reference CIMs and drawings for manufacturces clearance requirements.) 0 r' 1RO-4003 Is very close to the auxitiaryfedwvaterpump local instrumentationrack JRl-3& The replacementRO, although larger,will not interfere with the observation of the indicators, orfor opening the rackfor calibnralo.
- 8. Require breaching aHigh Energy Line Break (EIE) barrier? (Ref=enceNP 8A.16) Ifyes, EQ engineer review required.
- 9. Consider operating experience from PBNP and industry events'. (eference DO-G04 for operating experience reviews and NPRDS, NODIL, CHAMPS, INPO Keywords, or other databases.)
Industry experience andEPRI researchwas consideredwhen determiningthe appropriateactionsto preventfuture weld failuresin the auxiliaryfeedwater minimumflow recirculationpiping. EPRI has recommendedwelds with an axialsize twice that of the radialsize (2/1 conflguration)
- 10. Consider failure effects on structures, systems, and components: (Failure analysis is only required for maintenance rule systems. Contact the NSA-PSA group for guidance and scope.)
- a. The design discusses those events/accidents which the system/components are to withstand? r] El
- b. The failure effect of the system/components: (Reference the NSA-PSA Group, Operating Experience, & IEEE-352-1975.)
"* How components may fail, and the effect of the failure on the system and related systems?
"*What mechanisms might produce failures?
"* How a failure would be detected?
"*What provisions are included to compensate for the failure?
- 11. Does the design add or remove components in containment? 0
- a. Change the amount of exposed aluminum in containment? (Reference DG-G and FSAR Section 5.6.)
- b. Change the amount of exposed zinc in containment? (Reference DO-GO?.) E' El
- c. Introduce matrs into containment that could affect sump performance or lead to equipment El degradation? (Reference DG-G07.)
- d. Decrease free volume of containment? El El
- e. Require addition or modification of a containment penetration boundary? (Consult the 13 o containment system engineer.)
PBF-1584 Rcvhdcn9 u1s.ro Page 2 of 13 Rcren NP7.2.2
Point Beach Nuclear Plant . SCR 2001-0981 10 CFR 50.59/72.48 SCREENING (NEW RULE) vify SCR mb'en aU pg Page 5 1..3 TESTS OR ExPER]MENTS If the activity is not a test or experiment, the questions in ML3.a and 11L3.b are 0 NOT APPLICABLE.
- a. Answer these two questions first:
YES NO QUESTION O Dl Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?
o] ] Are the SSCs affected by the proposed test or experiment isolated from the facility?
If the answer to BOTFI questions in V.3.a is &0, continue to [1L3.b. If the answer to EITHER question is X then describe the basis.
- b. Answer these additional questions ONLY for tests or experiments which do VOT meet the criteria given in I.3.a above.
If the answer to either question in I1.3.a is Y. then these three questions are El NOT APPLICABLE.
YES NO QUESTION
[3 [3 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?
[o E3 Does the activity utilize or control an SSC in a manner that is inconsistent with the analyses or descriptions in the CLE?
El [3 Does the activity placeitsthe facility in a condition not previously evaluated or that could affect the capability of an SSC to perform intended functions?
If any answer in 11.3.b is ES. a 10 CFR 50.59 Evaluation is required. If the answers in 113.b are ALL W, describe the basis for the conclusion (attach additional discussion as necessary):
Part IV - 10 CFR 50.59 SCREENING CONCLUSION (Resource Manual 5.3.4).
ceklthat applry A 1 CFR S0.S9 Evaluationis [r] qredorNOTrequired.
A Point Beach FSAR change is 13 required or ED NOT required. If an FSAR change is required, then initiate an FSAR Change Request (FCR) per NP 5.2.7 A Regulatory Commitment (CLB Commitment Database) change is 0 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 [3 required or C9 NOT required. If a change to the Technical Specification Bases is required, then initiate a Technical Specification Bases change per NP 52.15.
A Technical Requirements Manual change is [: required or 0 NOT required. If a change to the Technical Requirements Manual is required, then initiate a Technical Requirements Manual change per NP 5.7-15.
RIs.151ns0
Reference:
NP 5.1.
1Revidan 0 10a=4/1
. APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO t
- f. Require painting in containment? (Refernmce M136.3.) 01 0]
- 12. Consider potential for fuel failure? 0
- a. Affect fuel handling equipment? 0 0
- b. Present the potential for introducing foreign material/debris into the RCS or connected systems?
0]
0
- c. Affect core barrel flow patterns? ('Baffle jetting" concerns) 0 0
- 13. Meet requirements to abandon equipment if applicable. (Reference NP 7.1.5) 0]
B. Mechanical requirements. (Contact Mechanical Design Engineerng 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 0 with DG-M16, and QA components should be reconciled with ANSI N45,2.)
This modicationis being designedand installedin accordancewith ASME B3.1 - 1992. ThU S replacementorifice body has been constructedto 8w standardsofASME B16.34 - 1996 (98 addenda).
- 2. Affect or add components/systems to ASME Section XM class 1, 2, or 3 equipment? (Reference PBNP CHAMPS, CBD drawings, and IST Coordinator. If YES, follow*NP 7.2.5, RepairlReplacement 0 0 Program.)
- 3. Require State of Wisconsin Administrative Code permits/approvals? (Reference NP 7.4.9, Wisconsin 03 0 Administrative Code for Boilers and Pressure Vessels or the Authorized Inspector.)
- S S' r-..
- 4. Consider component performance requirements such as capacity, rating, output? 0 0 The replacementorfice, piping,andfitings all have design ratingsof 1440 psig at 100 T orgreater (ratingsfor pipe class DB-3).
- 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?
The replacementRO has been designed with a C, rangeof 1.4 to 4.1 with a pressuredrop of approximately 1400 psi to provide the desiredflow throughthe recirculationline.
0 0]
- 6. Provide vents, drains,Ynd sample points to accommodate operational, maintenance and testing needs?
0
- 7. Require service water? (Both essential and nonessential service water loads are modeled, and load
- -... changes must be evaluated. Contact the SWAP Coordinator.)
- 8. Require the addition of check valves? (Reference DG-M13 for selection guidance.) 0]
PBF-1584 Rcvision 9 1It/o35o Page 3 of 13
Reference:
NP 7.2.2
PPLIES TO DESIGN DESIGN INPUT CHECKLIST YNO
- 9. Require and evaluate any additional loading on instrument or service air, circ, fire protection, or demineralized water, or other system?
- 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, EQUL & EQMR.)
- 11. Affect ventilation barriers, including containment, primary auliary building, or control room?
o []
- 12. Require insulation? (Reference WE specification PB-485 for insulation, and NP 1.9.10 for asbestos controL) 13
- 13. Require lubrication? (Reference Lubrication Manual.) 0
- 14. Require an independent means of pressure relief? (RCference B31.1.) 13 0] 0
- 15. Affect the assigned system design pressure or temperaurc?
The replacementorifice,piping, andfittings all have design ratingsthat are equalto orgreaterthan thatfor the pipe clas (DB-3).
- 16. Involve cobalt4aden materials into the RCS or into systems that supply the RCS? (Reference NP 42.29, -Source Term Reduction Program.")
0] 0
- 17. Are new materials and their coatings/plating compatible with system chemistry and disposal systems (NP 8.4.15)?
[] 0]
All new component are stainlessstree which isappropriatefor use in the auxiliaryfeedwatersystem.
- 18. Affect embedded or buried piping?
0] []
C. Electrical requirementL (Contact Electrical Design Engineering for guidance.)
I. Consider design conditions such as ampacity, voltage drop? 0] 0]
- 2. Consider component and system performance requirements, such as current, voltage, or power?
0] 0]
t
- 3. Consider redundancy, diversity and separation requirements of structres, systems and components? 0] 0]
(Reference DG-E07 for separation of electrical circuits.)
- 4. Comply with protective relaying requirements of equipment and systems?
0] 0]
PBF-1584 Rmisio 9 I1UoI Page4 of 13 Ref*e=: NP 7.22
- APPLIES TO DESIGN DESIGN INPUT CHECKLIST
- 5. Selection of overcurrent devices for proper protection and coordination? (Reference DG-E04 for selection of molded case circuit breakers.) 0l
- 6. Affect available fault current at any bus? El
- 7. Assure that all added cables meet fire retardancy requirements? (Reference FPER Section 4.1.8, IEEE 383.) El 0
- 8. Be compatible with existing electrical insulation and wiring? E3
- 9. Affect ampacity of existing cables? El
- 10. Maintain UL (or equivalentY 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 circuits power factor, etc.) on a vital or sensitive 0]
instrument bus?
El
- 12. Add new raceways? (Reference DG-E03 for electrical raceway sizing and DG-E02.) 0]
13 IJ
- 13. Add cables to existing electrical raceways? 0]
- 14. Be routed through fire wrapped raceways? E3 E] -0 0]
- 15. Affect the station grounding or lightning protection system?
El 0]
- 16. Make any vital circuit susceptible to ground? El El 0]
- 17. Affect emergency diesel loading? (Reference DG-E06 for diesel load change evaluation.)
0]
- 18. Add more station batterloading? []
0]
- 19. Add load to a vital bus?
0]
- 20. Add load to a non-vital bus?
PBF-1584 Revison 9 11/09501 Page 5 of 13 Refw=c NP 7.2.2
APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES NO Eo 0
- 21. Be compatible with service transformer capacity?
- 22. Consider electromagnetic interference between new/existing equipment and electromagnetic coupling interactions between circuits?
ol 0
- 23. Affect embedded conduits or buried cables, including the station grounding system? 0]
D. Instrumentation and control requirements. (Contact I&C Design Engineering for guidance.)
0]
- 1. Consider design conditions such as pressure, temperature, fluid chenstry, amperage, voltage? El EJ 0]
- 2. Have the instruments been properly selected for the application?
El 0]
- 3. Have sufficient instruments for operators to monitor the process?
0]
- 4. Have appropriate Jnstrument scales?
0
- 5. Have the instrumnents, control switches, and indicating devices been appropriately located for human 0]
factors (both for operations and maintenance)? (Reference DG-GO1.)
0]
- 6. Have alarms for off-normal conditions?
0]
- 7. Be capable of or require remote and/or local operation?
0 0]
El
- 8. Be capable of o require manual and/or automatic operation? -- .
0]
- 9. Require calibration and maintenance requirements for the instruments to be specified?
0]
- 10. Have specified the instruments with proper range and accuracy?
0 0]
- 11. Address solid state vuln1ility to RFI?
0l 0]
- 12. Consider software and programminglprogrammable settings of digital or electronic equipment?
0]
"13. Affect logic circuits or associated GL 96-01 review/required testing? Contact I&C System "Engineeringgroup.
PBF-1584 ip-A.,-a mmuQII'*ni Paze 6 of 13 Refcrencc NP 7.2.2
APPLIES TO DESIGN DESIGN INPUT CHECKLIST 0E NO Structural requirements. (Contact Civil Design Engineering for guidance.)
- 1. Affect or scope seismically qualified equipment (Class 1 or 2) and therefore require a seismic ED []
qualification evaluation? (Reference NP 7.7.2, *Seismic Qualification of Equipment.9)
All components are being installedseismi Class 1. An addendum to calculation WE-100007 will addressthe additionalweight added andensure the seismic adequacy ofthe auxiaryfeedwater piping.
- 2. Affect seismic boundaries? 11 0]
0]
- 3. Affect stress calculations of pipe? (Reference DG-M09.)
An addendumto WE-100007 will address the additionalweight ofthe replacementoraice to ensure 0]
thatall pipingand supportstressesare below the code allowable values. El 0]
- 4. Affect the loading or require chahges to existing equipment foundations? 0]
0]
- 5. Affect wall stress calculations for pressurized concrete cubicles or structures?
0]
0]
- 6. Require analysis of non-seismic components placed over or adjacent to seismic components?
0]
El
- 7. Add items which span between two separate seismic areas/buildings? (The effect of the relative 0]
movement must be addressed.) 0]
0]
- 8. Require clearance review for seismic movement or thermal expansion considerations?
0]
- 9. Require a floor or wall loading analysis? (Reference Bechtel C-dwgs.)
0]
- 10. Rl -iireIhe-addifbn of new supports, hangers, or f6ondadions or add weight Wbt betwtcn existing . 0]
supports, hangers, embeds, or foundations during installation or post-installation? (Reference 0]
DG-M09 and DG-M1O for pipe support.)
The replacement orifice is heavier than the original.An addendum to WE-100007 will addressthe additionalweight on pipe supporu.
0]
- 11. Add new or add load to seismically qualified raceways? (Reference NP 7.7.2, "Seismic Qualification of EquipmenL')
0]
C, 0]
- 12. Modify, attach to, or locte within the proximity of masonry block walls? (Reference IEB 80-11 Block Wall Program.)
0]
- 13. Acquire core drills, expansion anchors, or re-bar cuts? (Reference DG-CO1 for expansion anchor design and installation.) 0]
- 14. Create an external or internal missile hazard?
PBF-1584 Page 7 of 13 Refercnce: NP 7.22 Rcvision9 I1/05101
"APPLIES TO DESIGN DESIGN INPUT CHECKLIST YE~S O "R15. Consider wind and storm loading on external structures? 11 0 0]
- 16. Require protection from high energy line break jet? (Refer to FSAR Appendix A.2.)
EJ
- 17. Consider dynamic requirements such as live loading, vibration, and shockmpact? 0 F. Programs
- a. Affect 1ST acceptancecriteria or calculations? (Contact Component Engicering.)
- b. Require classification of new components? (Reference DG-G06 for system, component. and par classification.)
0]
- c. Affect QA-scope systems or boundaries? (Contact Site Programs Engineering Support for Q-USL)
The orifice iLthe boundry between a QAA&R portionofthe system, anda QA/AQ portionof the system. However, all components are being Installedas QAISP The boundary is not being moved.
- d. Require special personnel/equipment qualifications not proceduralized at PBNP (i.e.,
underwatcr welding)? 0]
- e. Require material certification or other certification to ensure quality equal to or better than the 0]
affected SSC? (These requirements need to be specified in the specification or purchase requiisition.) -... . . . . . . . .*
All new components(orifice,piping,ftings) are QAI, and will be certf'iedto be of quality construction.
- f. Have all design requirements, such as pressure or current rating, been reviewed against lot descriptions or been specified on purchase requisitions/specifications? 11 See Bill ofMaterialfor lot numbers IPOnumbers. AU components have been verfied to have adequatedesign ratings.
- 2. Fioe protection considerations:
- a. Affect access to aIlre zone' fire protection equipment or Appendix R safe shutdown equipment, El including manual fir fighting activities? (Reference Section 5+/--1 of Design Guide DO-FOIl) 0]
- b. Affect a fire barrier? (Reference NP 8.4.11 and Fire Barrier Drawings WE PBC-218 El 0]
Sheets 1-20, Section 5=.2 of Design Guide DG-Fo1)
El 0]
C. Affect a fire protection system or its performance? (Reference Section 5.23 of Design Guide PBF-1584
- 3cvhioa9 ILWOI Page 8 of 13 Re&== NP 7.2.2
- APPLIES TO DESIGN DESIGN INPUT CHECKLIST DG-FO1)
- d. Increase or decrease permanent combustible loading in a room? (Reference Section 5.2.4 of 0 Design Guide DG-FO1)
- 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 0 El Guide DO-FO0)
ThM mod icatoniwill affect a portionof the azxiliaryfeedwatersystem, which &va sofe shutdown system describedIn Section 2 of the SSAR. The RO is in theflow pathfor safe shutdown shown on APPR M-217 h, 1.
E 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 E" 0 shutdown cable? (Reference Section 52-5.2 of Design Guide DG-FOI)
- 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 El 0o Appendix R exemption for any PBNP Fire Area? (Reference Section 52.5.3 of Design Guide DO-FOl, Table 3.2-2 of DBD T-40)
IL. Will the change add, remove, or affect the performance of any emergency lighting required for compliance with Section I1J of Appendix R? (Reference Section 5.2.6 of Design Guide DG- ." 0 F01)
- i. 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-FO1)
- r. Will the c*angliffect gieReactorC0olantPa Oil olle*t*oSystem? (Refce-e .. . .
Section 5.2.8 of Design Guide DG-FOl)
- k. Widl the change affect the Fire Protection Manual?
L Will the change affect any of the Supporting Documents listed in the SSAR (Section 6.0) or the El 0 PHAR (Section 4.0)?
I If any of the questions a tihoughj are answered "yes", an evaluation must be performed using the applicable sections of the FPCC checklist, PBF-2060 per Section 5 of Design Guide DO-FO1.
PBF-1584 Revision 9 11A/0i Page 9 of 13 Ref=== NP 72.2
APPLIES TO DESIGN DESIGN INPUT CHECKLIST YES no_
- 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? 0
- b. Degrade existing flood barriers or flood mitigation features providing unanalyzed pathway for flooding to propagate? (Reference Section 3.2 of DG-CO2.)
0]
0]
C. Involve the opening of potential flood sources anywhere at the station? (Installation procedures 0]
need to address inadvertent flooding. Reference DG-CO2, Section 4.4.)
Valve IAF-I1 will be isolatingthe condensatestorage tanksfrom the auzxiaryfeediaterpump room during*u*allion. The WPfor tds modificatonwill have appropriatecaution steps to 0]
ensure th IAF-I5 is Isolatingproperly, and thatappropriatestepsare taken fleakagepast the valve isseen.
- d. Reduce the capacity to isolate or cope with flooding? (Reference Sect. 4.2 of DG-C02.)
0]
0]
C. Change plant drainagedbackfill requirements?
0]
0]
0 Locate essential equipment or supporting systems where it would be susceptible to flooding? 0]
(Flooding conditions may also impact Environmental Qualification.)
0]
- 4. Environmental considerations:
- a. Be subject to adverse environmental conditions during storage or construction? (Reference 0]
NP 9.5.2.)
...... Re.u*
b*..- e~ze pro6 nonaffect ei freek6 p.&ecti&i. .
0D 0]
- c. Locate safety-related or post accident monitoring equipment in a HARSH environment? 0]
(Reference NP 7.7.1.)
0]
0]
0]
- c. Be attached to an UQ system/component? CThis will require an EQ review for potential updates 0]
to EQSS. EQML & EQMR. Reference EQ master list.)
[]
- f. Change environmental parameters (e.g., pressure, temperature, radiation, humidity)? (Reference 0]
NP 7.7.1. "Environmental Qualification of Electrical Equipment."
- 5. Radiation Protection (RP) and ALARA considerations: (Reference DG-G03, "ALARA Consideration PBF-1584 Rcviswa9 11RoSioI Page 0 of 13 Referefe=: NP 7.2.2
APPLIES TC) DESIGN DESIGN INPUT CHECKLIST NO-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.
- a. Affect any SSC in an RWP required area, a contaminated area, or a radiation area, including El 0l opening of a system that may be a radiological concern?
- b. Will the change generate excessive radwaste or highly radioactive/contaminated waste? 0 0]
- c. Remove any plant equipment from a potentially contaminated system (including BOP systems)? 0 0]
- d. Result in an anticipated increase in operational or maintenance exposures?
(Consider equipment rearrangement to reduce plant life dose?)
0]
C. Result in an expected exposure of greater than 1 Rem for any individual during Installation of the change?
f Result in an anticipated collective exposure of greater than 2 Rem for the installation of the change?
Q') I questions d, e, or f apply and are answered yes, then an ALARA review shall be performed.
(Reference NP 4.23, "ALARA Review Procedure.u)
- 6. Chemistry consideations:
- a. Require or affect established chemistry limits? (Contact system engineer and review chemistry 0]
procedures.)
b... .. u..ir.lyioutih.*
- y T.e i fidijiases (C~otact Rys.mýeagineer and review ch try ... "
procedures.)
0]
- c. Require chemical additives? (Contact PBNP mistry) 0l
- d. Do new fluids/chemicals need to be evaluated for TRI (Toxic Release Inventory), CHES, critical applications, or special disposal requirements? (Contact Chemistry/Chemical 0 Engineering.)
G. Installations
- 1. Installation requirments/plant conditions have been determined? 01 Plant wiU be in Mode 4. 5. or 6 during uallation, with the IP-29 auxiliaryfeedwaterpumpout of
-.j service- See 1WP 99-029"C
- 2. Consider test and inspection requirements, including the conditions under which they will be 0 R7 performed? (Reference NP 7.4.1 for pressure test requirements, NP 7.4.3 for post-maintenance and PBF-1584 Ref=: NP 72.2 2 ~-*.nn a imI 'Pare I1 of 13
APPLIES TO DESIGN DESIGN INPUT CHECKLIST modification NDE requirements, NP 12.5 for special test procedures, and OM 4.2.2 for in-service tests.)
VWand PT1" eminationswill be performed on the weldi. An initd service leak teas at normal operatingpressureand temperaturewill be performedafter the installation. Flow measurements will be taen to verify the correctsetting of the RO.
- 3. Have post-installation acceptance criteria been properly specified to test the intended function of the component(s)/system?
NDE acceptancecriteriawill be perASME B31.1 - 1992. No viible leakage can be seen during the pressuretest. Flow will be ver*-d to be between 120 and 130 gpm. See IWP 99.029**
- 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? - 0
- 9. Will any added components introduce chemical contaminants to the system? (Le., preservative coatin 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 rl 0
- 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 0
- b. Potentially impact other systems, components, or structures during installation? El 0
- c. Present installation impacts on plant operatons (i.e., fire watches, et-)? 0 0 Firewatiches arerequiredduring welding in the auxi*aryfeedwaterpump room.
- 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? 0 0D PBF-1584 Pagte 12 of 13 12f3efrnace NP 7.2.2
- APPLIES TO DESIGN DESIGN INPUT CHECKLIST
- b. Require work within 20' of fence? 0 o] 0
- c. Affect security equipment and documents, including those containing safeguards information?
(Contact Security for design developm ent requirements and design concurrence.)
o] 0
- d. Affect access controls?
- 13. Safety requirements:
- a. Affect safety equipment and thereby create personnel hazards (i.e., removal of handrails)? 11 0
- b. Introduce hazardous material into the plant? (Reference NP 1.9.1.)
0] 0]
13 C. Affect evacuation routes or escape provisions from enclosures? 0] 0]
0
0] 0]
The inkualladonwill be performedin accordancewith th WE andAWC safey manuals.
- e. Move any energy sources? If yes, verify installation document covers move, including 0] 0]
transferring danger tags.
Designed by:. Rob Chapman Deat: 3 1-'1-0-..
Reviewed by: Jeff Novak ___ ~~Date: / &'~
PBF-1584 Revision 9 1105/01 Page 13 of 13
Reference:
NP 7.2.2
Point Beach Nuclear Plant FIRE PROTECTION CONFORMANCE CHECKLIST MR Number MR 99-029*C Unit I X Unit 2 Common Facilities System Auxiliary Feedwater - AF Location Aux Feed Pump Room, IP-29 Cubicle AFFECTED FIRE ZONE(S)-FIRE AREAS (see FPER Sect. 9) Fire Zone 304S 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 103.1, 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 cpmpleting 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 A,.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.
Completeeappropriate sub-forms, based upon ibe nature of 6i hd adfidea*n' the FPCC iplicaibility iaý t't 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-F01 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 (Le., 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 clocklist 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 Revision 3 O6A08A)I Page 1 of 2
FIRE PROTECTION CONFORMANCE CHECKLIST 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 103.1, 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.
Fare Protection Conformance Checklist Applicability Matrix Applicable? Section Pvsmn Input Action Checklist Section Yes No 0 [] 1.0 Plant Access F.2.a Complete & attach PBF-2060a 00,* 2.0 Fir. Barriers F.2.b Complete & attach PBF-2060b 0 0] 3.0 Fire Protection Systems F.2.c Complete & attach PBF-2060c F.2.d Complete & attach PBF-2060d 0l 0 4.0 Combustible Loading/Fire Hazards D 00 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
.0 Including Associated Circuits 0 D 7.0 Fire Area Analysis, F.2.g Complete & attach PBF-2060e Including Exemptions/Evaluations Emergency Lighting F.2.h Complete & attach PBF-2060f 0 0 8.0 Plant Communications F.2.i Complete & attach PBF-2060g 0 0 9.0 Reactor Coolant Pump F.2.j Complete & attach PBF-2060h 0 0 10.0 Oil Collection System Conformance checklist (including all applicable attachments) completed.
By: a! eg Date: "3--o- .
Szoo PBF-2060 Revision 3 06MI8/O1 Page 2 of 2
Point Beach Nuclear Plant FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5, 6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION
-44Complete 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 52.5.1 of Design Guide DO-FOD) 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.
SYes, go to 5.11, complete actions and resume at 52.
0No, go to5.3 Comments:
5.2 Widl 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)
" Yes, go to 5.11, complete actions and resume at 5.3
[] No, go to 53 Comments:
53 Does the modification require deletion of a safe shutdown component? (SSAMS Database, SSEL Module, SSAR Section 2)
E] Yes, go to 5.11, complete actions and resume at 5.4.
SNo, 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)
[ Yes, go to 5.11, complete actions and resume at 5.5
[ No, go to 5-5 Comments:
PBr-2060c Revisio 0 6080 Page I of 6
FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5, 6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 5.5 Does the modification add/delctc/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)
E] 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: Operation of the auxiliary feedwater system will not be affected. The replacement RO will be set to the same flow rate as the current RO.
5.7 Does the modification violate the safe shutdown systems performance goals as presented in FPER Section 7.2 and SSAR Scion 2?
El] Yes, go to 5.11, complete actions and resume at 5.8 0 No, go to5.8 Comments:
on 5.8 Does the modification affect any mechanical sub- or support components of safe shutdown components not listed the safe shutdown equipment list? (e.g., SOVs, check valves, etc.) (See CHAMPS Appendix R listingX. 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.
El Yes, go to 5.11. complete actions, resume at 5.9
[ No, go to 5.10 Comments:
5.9 Does the modification to the sub- or support component affect the operability of its associated safe shutdown equipment? (iLe., Failure of a support component that results in failure of a safe shutdown component)
[] Yes, gh to 5.11. complete actions, resume at 5.10 D- No, go to 5.10 Comments:
PBF-2060e Revision 0 061V8Pg1 Pagp 2 of 6
FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5, 6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 5.10 Does the modification add/delete/revise 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)
E Yesgoto5.11
[ 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-FO0.
RESUME checklist completion.
Safe Shutdown System(s), Components, Sub- or Support Component(s): IRO-4003 6.0 -SAFE SHFUTDOWN 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)
I] Yes, go to 6.10, complete actions and resume at 6.2
[I No, go to 6.2 Comments:
6.2 Does the modification require deletion of a safe shutdown cable? (Ref. SSAMS Circuit Analysis Module, SSAR Sections3,'4, and 5) .
El Yes, go to 6.10, complete actions and resume at 6.3 SNo, go to 63 Comments:
t PBF-2060c Revision 0 06MI)ae Pagp 3 of 6
FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5, 6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 6.3 Does the modification revise an existing sife 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.
I] Yes, go to 6.10, complete actions and resume at 6A
[ 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)
[ Yes, go to 6.10, complete actions and resume at 6.5
[ 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 DO-F01, Appendix R Highlighted Single Line Drawings, SSAR Section 3)
[ Yes, go to 6.6 J0 No, go to 6.7 Comments:
6.6 Will adequite eeýi coordination betweenthe safe shutdown power suppiyjfeeder treaer mid ihe added or revised component breaker or fuse exist? (Ref. Section 525.2 of Design Guide DG-FOland SSAR Section 3)
-] Yes, go to 6.7 L' No, go to 6.10, complete actions and resume at 6.7 Comments:
PBF-2060c Revision 0 06108101 Page 4 of 6
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?
SYes, go to 6.8 ED No, Safe Shutdown Cables Section Complete, go to 7.1 Comments:
6.8 Will fte new or revised cables be equipped with properly designed circuit breakers, fuses or some kind of current limiting device? (Ref. SSAR Section 3)
E] Yes, Safe Shutdown Cables Section Complete, go to 7.1 El No, go to 6.9 Comments:
6.9 Widl the new or revised cables share a common enclosure (raceway, panel etc.) with safe shutdown cables? (Ref.
Section 5.25.2 of Design Guide DG-FOland SSAR Section 3)
-' Yes, go to 6.10, complete actions
[] 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 5.2.52 of Design Guide DG-F01. RESUME checklist completion.
Comments:
7.0 FIRE AREA ANALYSIS. INCLUDING EXEMPTIONS/EVALUATIONS (Ref. Section 52.5.3 of Design Guide DO-FOl) 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 M.G.2 of Appendix R of Appendix R, local manual actions, repairs, etc.). (Pef. SSAP, Section 5) t El Yes, go to 7.5. complete actions and resume at 7.
2
[I No, goto 7.3 Comments:
PBF-2060c R-visi=O 06oV0801 Page 5 of 6
FIRE PROTECTION CONFORMANCE CHECKLIST SECTIONS 5, 6, & 7 - APPENDIX R SAFE SHUTDOWN EVALUATION 410 7.2 Is compliance with the separation criteria for redundant safe shutdown capability in Section HLG of Appendix R affected by the change? (Ref. SSAR Table DBD T-40, Table 3-2.2)
[0 Yes, go to 7.5, complete actions and resume at 7.3
'INo, go to 7.3 Comments:
7.3 Is the modification proposed to be implemented in a fire zonelarea for which an Appendix R Exemption or FPEE is in place (Ref. DBD T-40, Table 3-2., FHA, Table 1-1 of the SSAR)
[ Yes, go to 7A
[ No, Fire Area Analysis Section is complete, go to Section 8 or next applicable Section.
Co*mýents: Exemptions 6 and 18 for fire zone 304 (aux feed pump room).
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 reqvirements of Appedix R, the bmsis for en Approved AppendiX R rxempdipP, .
such as a Fire Protection Engineering Evaluation. List the basis affected and refer to Section 5.2.53 of Design Guide DG-FOL. RESUME checklist completion.
Bases:
PBF-2060c Revision 0 0610&10 1 "IPage 6of 6
SIWPNUMBLYO: 99-029*C I SPage 1__ of vj 99-029"C INSTAI!LATION WORK PLAN INSTALLA17ION WORK PLAN PBNP MINOR PROCEDURE El CheckAs Applicable WORK ORDER WORK PLAN 0 FOR MODIFICATION # MR 99-029*C WO# 0202509 INSTALLATION WORK PLAN TITLE AUX FEED WATER PUMP 0P-29 A[NUM FLOW RECIRC LINE ORMCE IRO-4003 UNrrI [D QA-SCOPE E] NON QA-SCOPE Originator'l x a idý2 ~ *- . Date 7-Il-o2Z Reviewer 2f44L ----'
Finl Design Group Head Qmuaty Date Installation Grup Hled h I 1::'
Date .. "
Mana Operaitions o NOTE Canges to this work plan must be done with the concurrence of the responsible or team engineer and the installation surpevisor, or as dlineated within the IWP.
RDG-viin5 Revision I BESTCOPYRAVALABLE
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNITI February 20, 2002 1.0 SCOPE 1.1 The scope of this installation work plan is to replace a portion of the existing AF line 2"- DB-3, including pressure reducing otiifce (1RO-4003). The replaced piping and new RO will be welded back utilizing oversized socket welds.
1.2 The purpose of this modification is to reduce piping line noise and vibration when operating Auxiliary Feedwater (AF) pump 1P-29 in the recirculation mode. The presently installed RO is cavitating, causing excessive noise and pipe vibration. The purpose for oversized socket welds is to offer a significant high cycle fatigue improvement over standard ASME Code socket welds.
1.3 The approach of this installation is as follows:
1.3.1 Pre-fabricate new section of pipe as shown on sketch SK-MR-99-029*C (pre fab work done under WO 0202508).
1.3.2 Isolate and drain the affected piping.
1.3.3 Remove portion of the existing AF line 2"-DB-3.
1.3.4 Install new sections of AF line 2"_-DB-3.
1.3.5 Perform VT and PT exams on all new welds.
1.3.6 Functionally test the mini-recirc. line to verify operability.
I -- ý_-- I -
1.3.7 Perform inservice leak check of new/modified piping and welds.
1.4 This installation is scoped as QA, safety-related work.
QA Scope Clarification:
The piping downstream of 1RO-4003 is QA, non safety-related (AQ) scope.
1.5 Ingtallation of this ]WP will be performed while Unit 2 is in Mode 4, 5, or 6. The 1P-29 auxiliary feedwater pump will be out of service.
1.6 This modification will not affect any ASME Section XI pressure boundaries. An R/R/M is not required.
Page 2 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC IcNE ORIFICE UNIT I February 20,2002 1.7 Support Requirements
1.7.1 Operations
Support to install and remove danger tags, system draining, and post maintenance and operability testing.,
1.72 NDE Group: Perform visuallpenetrant (VT/PT) examinations and pipe thickness measurements where specified in this IWP.
1.7.3 Security
Performs fire watch duties as directed by Operations.
1.7.4 QC: Perform inspections as required.
1.7.5 Engineering
Support NDE evaluations and post maintenance testing.
RE: Rob Chapman x7636 pager 0114 Home Tel. 920-429-9146 1.7.6 Mechanical Maintenance: Perform removal and installation of orifice, piping, and supports.
2.0 PRE-INSTALLATION REQUIREMENTS 2.1
References:
2.1.1 Construction sketch:
- a. SK-MR-99-029*C, "Auxiliary Feedwater System Orifice 1RO-4003
- - ... ..'s. . . =-placanent,1Units 1 &-2. . . .. . ., -, -,.
2.1.2 Vendor/Contractor drawing:
- a. Flowserve drawing 94-16249 2..1.3 Applicable Codes and Standards:
- a. USAS B31.1 - 1967
- b. ASMEB31.1-1992 2..1.4 Supplemental Procedures:
- a. MI 32.1 Flange and Closure Bolting
- b. MI 32.8 Guidelines for Opening Piping Systems Page 3 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW I RECIRC LINE ORIFICE
-UNIT I February 20,2002
- c. MI 32.11 Installation and Reuse of Swagelok Fittings
- d. NP 1.9.6 Plant Cleanliness, Storage, and Inspection Program
- e. NP 1.9.9 Transient Combustible Control
- f. NP 1.9.13 Ignition Control Procedure
- g. NP 1.9.15 Danger Tag Procedure
- h. NP 8.4.10 Exclusion of Foreign Material from Plant Components and Systems
- i. NP 8.5.2 CHAMPS Equipment Database Usage and Control
- j. 01 62B Turbine-Driven Auxiliary Feedwater System (2P-29)
- k. PBF-9142 Bolting-Torque And Loading
- 1. WPM 2.P8-GT Welding Procedure for Stainless Steels Group P-8 GTAW-Pipe Diameters Over I" OD
- m. WPM 2.Pl-8-GT Carbon Steels ASME Group P-I to Austenitic Stainless Steel ASME Group P-8 GTAW Pipe Diameters Over I" OD Responsible Engineer has assured that all references listed above are approved and the appcikble trduine*---have been Incorporatedinto the IWP.: The tfocrcncea 6wetithcr with the Installation Group, attached, or are readily available to the Installation Group.
RE~z~ZZ...........Date___________
Page 4 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT 1 Februa 20,2002 2.2 Background References (those references not needed to perform work):
2.2.1 Drawings
- a. Bechtel Drawing M-217, Sh. 1, P&ID of Auxiliary Feedwater System
- b. Bechtel Drawing P-103, Stress Isometric of Emergency Feedwater Pumps to Main Feedwater Lines 4" & 3"-DB-3 2.3 Installation Preparation Activities
, 2.3.1 A Bill of Material (BOM) is attached to this IWP 6r is included on the Construction sketch.
2.3.2 The Responsible Engineer has assured that all materials on the BOM are on site, available for the modification, and QA released.
2.3.3 The Responsible Engineer has verified that all calculation Addenda specified on PBF-1 606 have been approved prior to the start of work.
2.3.4 New CHAMPS label for IRO-4003 is required and has been requested.
The RE has assured that all of the above Installation Preparation Activities are complete.
REDate ~4"1.
2.4 Pre-Installation Discussions 2.4.1 A pre-installation discussion with the Installation Group representative, the Testing Group re iseulve- and the Acceptance Group representative has 2..2 A field walkdo*ii been performed, if necessary, to verify that all aspects of the procd&= ybe oedas intended. / j 2.4.3 A commitment has been obtained from Security to support fire watch requirements. Record the responsible group below.
Fire Watch SupportGroup1() Gra -- Date Page 5 of 2/P
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM ROW RECIRC LINE ORIFICE UNIT _ February 20, 2002 2.4.4 Foreign material exclusion (FME) shall be controlled per NP 8.4.10, Exclusion of Foreigz1 `erial n Plant Components and Systems, and PBF 9158, MEC* qtfdi. /" . . /ý RFAS' 2.5 Personnel Safety Concerns m s The following precautionary personnel safety requirements are recommended for this IWP:
2.5.1 Caution should be exercised when lifting or rigging components.
2.5.2 This installation will take place in an area of increased fire awareness.
Installation personnel shall take precautions against fire hazards. Care should be taken no to allow combustibles to extend from the IP-29 cubicle to the adjacent AFP cubicles.
2.5.3 Care should be taken during welding to prevent the halon system in the auxiliary feedwater pump room from actuating. This may be accomplished by removing from service any fire sensors in the 1P-29 cubicle or in the area of the welding.
2.5.4 Aux. Feedwater pump IP-29 minimum recirculation line does not have drain connections. To allow this line to drain, the downstream flange for metering orifice 1FE-4049 will be broken open. The safety cautions of M 32.8,
"-elineaotiogpe g gysthmshallt utilized.g..
The Install *on Su or is aw of the above listed safety con s.
Page 6 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT I February 20, 2002 2.6 Identification of Permits Required 2.6.1 Work Order 0202509 for this IWP has been written and submitted to CHAMPS. The Work Order number has been recorded on the IWP coversheet.
2.6.2 Ignition control permit is required for welding and shall be obtained by the Installation Supervisor when needed.
The Install Supervisor has assured that all necessary permits for this is Date f7 2.7 Pre-Installation Work NOTE: The following work will be performed under other work orders.
2.7.1 Pre-fabricate piping assembly, including replacement orifice 1RO-4003, in accordance with the Construction sketch SK MR-99-029*C under WO 0202508.
2.8 Operational Installation Prerequisites 2.8.1 This installation will be performed during the Unit 1 Reactor
........... being-in Hot-Shutdwv.rn (Moe4.4), C o^ Slihtdown (Mode.5)-, .
and/or Refueling (Mode 6) operating condition. If the installation is performed with Unit I in any other mode of operation, appropriate action statements per LCO 3.7.5 shall be performed.
NOTE: The following step Indicates a RECOMMENDEDDanger Tag Series. This may be altered depending on the plant conditions or other work being performed on the auxiliary feedwater system as determined by OPS.
2.8.2 Prepare a Danger Tag Series to isolate the 1P-29 minimum recirculation line from flow element IFE-4049 to valve lAF 15.
Recommended tag series:
Page 7 of 24
NUCLEAR POWER BUSINESS UNIT ]WP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNITI February 20, 2002
- a. Valve IAF-00015, Pump IP-29 Mini Recirc Outlet, CLOSED
- b. Valve lAF-04002, Pump 1P-29 Mini Recirc Control, CLOSED
- c. Valve 1MS-00126, P-29 AFP Steam Supply Inlet CLOSED DANGER TAG SERIES: A ?-71 1.4 W.. fw 0 t 2.8.3 Hang the Danger Tag Series prepared in Step 2.8.2.
OPS/Date 2.8.4 Release For Installation All of the above operational installation prerequisites have been met a~dit is acceptable to proceed with the installation.
DSS -qv
- W Date 't/t 4 I&Time 111 0 t
Page 8 of 24
NUCLEAR POWER BUSINESS UNiT ]WP 99-029*C.
WO 0202509 INSTALLATION WORK PLAN WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT I February 20,2002 3.0 INSTALATION 3.1 Installation Description NOTE: The following Is a detailed step-by-step listing of the actions necessary to perform this IWP. The steps are to be performed In a logical work order. Work can be performed In an order other than as written at the discretion of the Responsible Engineer or the Installation Supervisor.
CAUTION Aux. Feed. pump IP-29 minimum recirculation line does not have a drain connections. To allow system to drain, downstream flange of metering orifice IFE- 4049 shall be broken open. Safety caution of M1 32.8, Guidelines for Opening Piping Systems shall be observed.
3.1.1 Install temporary supports on recirc piping iffnecmss .
-AMTIDa?
3.1.2 Drain and vent Line 2"-DB-3 by breaking the flange at orifice 1FE-4049, IF necessary. Control drainage as well as possible by using hoses and catch basins.
-- A*
Noe ystem ma aeenieie&ii i IFE-4049 instrumentconnections. If this Is the case, then breakingthe flange Is not necessary. J Note: Coordinatewith Operationsas required MT/D&*
CAUTION If leakage past isolation valve 1AF-15 Is seen, STOP work and contact the job supervisor or RE Immediately. T%,e&- %S ,no sa&i' btwo IAP-ir #4 4,. crrs.
'I flY 4LW & fen, 4{o C4-,4r,A pqL Page 9 of 24
NUCLEAR POWER BUSINESS UNIT 1WP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT1I February 20, 2002 CAUTION Take extreme care to apply all possible fire protection precautions in the AF pump 1P-29 cubicle.
3.1.3 Disconnect instrument tubing from 1FE-4049 downstream flange.
MT/Date NOTE: Note exact orientationof IFE-4049orficeplate before removing. Recordany orientationinformation below, and retainthe orificefor reinstallationlater.
3.1.4 Remove the piping assembly, including orifice IRO-4003, as required per Construction sketch SK-MR-99-029*C. Record orifice orientation information below.
Note: The U-bolt for the support just downstream of IFE 4049.y have to be removed.
MTIDate 3.1.5 Perform FRE cleanliness inspection and install temporary FRE covers on all unattended open pipe ends. The guidelines of the FME "Checklist", PBF-9158, shall apply to this IWP.
Mr/Date Page 10 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMPL IP-29 MINIMUM FLOW RECIRC LINE ORIFICE February 20, 2002 UNIT 1 IIII I I I .... M 3.1.6 Cut the 1FE-4049 flange and pipe stub from piping assembly removed in Step 3.1.4 at the elbow closest to the flange.
Flange and stub piece can be re-used. Save all other pipe for examination by Engineering.
MTDt FME HOLD POINT 3.1.7 Prior to installation, perform FME cleanliness inspection of the IFE-4049 flange with pipe stub, and the upstream piping.
MT/Date 3.1.8 Install the 1FE-4049 orifice and flange including pipe stub cut out in Step 3.2.4. Use new gaskets (BOM Item 4) Check orientation of orifice before installing.
SMT/Date OC HOLD POINT 3.1.9 Torque bolts using a staggered pattern. Torque the bolts to 255 Ft-Lbs (+/-) 12 Ft-Lbs in accordance with PBF-9142.
QC SHALL witness final torquing of bolt;, ., .g M&TE: (2 -Z6 !o I Cal. Due Date: *-0-.
Final "As Left torque value:.
QC ~ki~i2ILA~t. Dat 4/ZIdz.
CAUTION 1AF-15 Is an isolation tagout boundary valve. It will not be possible to open this valve while welding. Take steps as necessary to keep the Xalve from getting too hot.
FME HOLD POINT 3.1.10 Prior to installation, perform FME cleanliness inspection of M.U( -'1-ao the piping assembly pre-fabricated in Step 2.7.1.
%.MT/Date Page II of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW I SRECIRC LINE ORIFCE UNIT 1 February 20,2002 I I o - M I..
3.1.11 Install the piping assembly pre-fabricated in Step 2.7.1 in accordance with Construction sketch SK-MR-99-029*C and the weld map (Attachment C). Perform fitup only with tack welds, and check that fitup meets ASME B3 1.1 requirements.
(B31.1 allows an internalmisalignmentof/16")
Note.: Someof the welds below wil befitup under WO 0202508.
By: _Tr P." Q*A"
- Date:_______
Weld#1 By: Aý-.
Weld #2 By: P Dater 9.ZT-o2-Weld #3 By. 'S "L Date: 2 Weld #4 Date: gq -
By: ME" "& a#*r Weld #5
(',a Ild S .2e,(
3.1.12 Perform root welds in accordance with Construction sketch dd,e - C o SK-MR-99-029*C and the weld map (Attachment C). r L-1
÷#,.
C'nmn ,. ,j, 4
,,nu, kdI0*otl wtill 1ip welJ~d uiiilr Wti 0202508.
"Weld#1 By_. P* _"_
By: _X__-
Date:
Date:
"Z- Q7-o Weld #2
-Weld#3. By: " 'Date:
Weld #4 By. Z ,A Date: A--OX Weld #5 BT."'Xc Date: ct 0-3J L
Page 12 of 24
NUCLEAR POWER BUSINESS UNIT 1WP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP I-P49 MJNIMUM FLOW RECIRC LINE ORIFICE UNITiI February 20, 2002 NDE HOLD Point 3.1.13 Perform visual examination (VT) of all field root welds. The acceptance criteria for piping welds are ASME B31.1 - 1992 and Constrction sketch SK-MR-99-029*C.
Note: Some of the welds below will be examined under WO 020250&
Weld#1: Unsat Weld #2: L2Sat Unsat2 uref 1. , f t0 "
Weld#3: :]Sat :Unsat ,3 UnsaY)'
Weld #4:
Weld #5: Unsat Note: If an unsat inspection condition Is Identified, NDE will not sign off this step until the appropriate evaluation and /or rework along with re-inspection has been accomplished.
/NDEfDate
ý"
- ., ., - - ..-, -1. - I Page 13 of 24
NUCLEAR POWER BUSINESS UNIT ]WP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE ORIHCE UNIT 1 February 20, 2002 KDE HOLD Point 3.1.14 Perform penetrant examination (PT) of all field root welds.
The acceptance criteria for piping welds are ASME B31.1 1992 and construction sketch SK-MR-99-029*C.
Note: Some of the welds below wig be examined under WO 0202508.
Weld #1: at Unsat * *-J ?-,r:?. fq
- irJL~
Weld #2: at Unsat, Weld #3: Li at Unsat 9 " -.. /f..
Weld #4: L4Sat Weld #5: _at Unsat Note: If an unsat inspection condition is Identified, NDE will not sign off this step until the appropriate evaluation and /or rework along with re-inspection has been accomplished.
- -NDE/Date 3.1.15 Perform rmal welds in accordance with Construction sketch SK-MR-99-029*C and the weld map (Attachment C).
Note: Some of the welds below will be welded under WO 0202508.
A.I /
Weld #1 By:-uLT/'L Date. . 4v By: Date: .t_
Weld #2 Weld #3 By: Date: _____
Weld #4 By: Date:
Weld #5 By: Date: a IT'oe~
01 Page 14 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT 1 February 20,2002 NDE HOLD Point 3.1.16 Perform visual examination (VT) of all E final welds. The acceptance criteria for piping welds are ASME B31.1 - 1992 and Construction sketch SK-MR-99-029*C.
Note: Some of the welds below will be examined under WO 0202508.
Weld #1: *at Unsat Weld #2: IISat UnsaD 1 ,jIJL z,,2-A 4o Weld #3: ISat Unsat ,f O/oLSO Weld #4: .*Sat Unsat Weld #5: L[ISat EUnsat .9(TI-ADz.
Note: If an unsat inspection condition Is identified, NDE will not sign off this step until the appropriate evaluation and /or rework along with re-Inspection has been accomplished.
3.1.17 Torque bolts on pipe support next to IFE-4049 to a nominal 6 ft-lbs.
M&TE: &ff(:7-?l - e.5 Cal. Due Date: (r-IV-615 Final "As Left" torque value: / /
.Z,9 9CY&
MTIDate 3.1.18 Restore fire protection for AF system pump room as necessary and notify Construction to release fire watch.
OP fDate 3.1.19 Mark-up Construction sketch SK-MR-99-029*C to indicate new as-installed configuration of this modification. Rcr- I9--V-0-L REDate Page 15 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNITI February 20,2002 3.1.20 Hang new label on 1RO-4003. 30-6%
P-4. qMate__+/-
3.2 Clean up Remove all construction debris, tools, and material from the work area.
Ensure all work areas meet PBNP housekeeping expectation.
Date 3.3, Installation Complete 3.3.1 As-Built Description This IWP was installed bv 07n. M ,&PR- , le-.r Date 7. -
The installation was performed in accordance with this ]WP and drawings (list revisions):
ECR(s) No. Aft-41, CR(s) No.
Other comments 1ja"_
Attach any additional documentation of the as-built description to this IWP.
3.3.2 List all calibrated equipment used during installation of this modification on the work order.
3.3.3 The installation of this IWP is complete. It has been installed in accordance with this IWP and all associated ECRs.
t RE -Date 73- ...
Date_______
Page 16 of 24
NUCLEAR POWER BUSINESS UNIT 1WP 99-029"C.
WO 0202509 INSTALLATION WORK PLAN WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE February 20,2002 UNIT I 4.0 TETMG 4.1 Testing Information 4.1.1 The acceptance tests for this modification are:
- a. Initial Service Leak Test performed at normal operating pressure and temperature (with pump running)
- b. Functional Test of new orifice 1RO-4003 to verify acceptable recirculation flow and reduced vibration.
4.1.2 The intent of the testing is to:
- a. Verify the functional performance of new orifice.
- b. Verify the integrity of the modified piping.
- c. Verify that all new welds associated with modified piping are leak tight.
- d. Satisfy the pressure testing requirements of NP 7.4.1.
4.1.3 Acceptance criteria for the testing is as follows:
- a. Piping and fittings within the modified piping boundary shall not show any evidence of structural distress (bulging or deformation) at normal AF system operating temperature and pressure.
iping and fittings within the modified piping boundary shall. not show any" ib.
"evidenceof through-wall ka& any iev elds at n0i AFýstem.
operating temperature and pressure.
- c. New orifice IRO-4003 flow reading must be between 120 and 130 gpm.
- d. With full flow through the orifice, there should not be excessive noise or vibration.
4.2 Pre-Test Reouirements 4.2.1 Remove the Danger Tags that were hung in Step 2.8.3. OPSIDate 4.2.2 Fill and vent auxiliary feedwater system and prepare IP-29 for operation as described in OI-62B. OPS/1ate 4.2.2 Install ultrasonic flowmeters on the mini-recirc piping to Page 17 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNITI February 20,2002 facilitate flow testing.
UT#1: MTE CIeFfA- o0 0 Cal Due: 7-\xv-o3 e-.
I0 11/h : M _____
Cal Due:
RElDate 4.3 Release for Testing 4.3.1 The auxiliary feedwater pump 1P-29 is available as required to be started for testing. Testing can start.
nser ý JZt-.Date 404 /d/, Timne 4.4 Testing Note: Sound and vibrationdata may be collectedas directed by the System Engineerwhile thepump Is runningduring 01-62B.
4.4.1 Start 1P-29 per OI-62B (or other procedure).
OrPS at 4.42 While pump is running, take flow readings below:
Rlow Reading IFf.--4 9.9 . *3. (,__,,
ENGIDate Flow Reading UT # I 11.ft gp1. P-H-0.
ENG/Date FlowReadingUT#2 ffIA gpm. fc.x- 1o-14-o02 ENG/Date Acceptance acrteria:The flow reading must be between 120 and 130 gpm. If 1RO-4003 Performance Test is unacceptable, manually adjust the 1RO-4003 per Attachment B.
SFJr-of gu "
- \ *,*d 1*- L~eL z Page 18 of 24
NUCLEAR POWER BUSINESS UNiT IWP 99-029*C.
INSTALLATION WORK PLAN4 WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT1 Februar 20,2002 4.4.3 When flow is acceptable, lock orifice with safety wire (provided by vendor).
RE/Date 4.4.4 Mark orifice position using indicating plate. ( EI~eo-rL RFJDate CAUTION Use caution when examining piping and welds. The system operating pressure may be in excess of 1300 Psig.
4.4.5 Perform the Initial Service Leak Test of the new piping and welds while the pump IP-29 is running in recirculation mode.
Examine for leakage all new joints. A 10 minute hold time is required. Record results of the Initial Service Leak Test below:
Leak Test: k'sat IlUnsat NO- y -v- * -- mI-As 4 Acceptance criteria: No visual evidence of weeping or leaking Acceptance criteria. No visual evidence of weeping Or leaking RFJOPS Date 2 -
Page 19 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW I RECIRC LINE ORIFICE UNIT I February 20,2002 4.4.6 Evaluate the vibration and noise created by the orifice with full flow. Document results below.
I Acceptance criteria: No excessive noise or vibration.
Date __ - l_ ....
4.5. Testing Results 4.5.1 Attach any additional testing documentation to this lWP.
4.5.2 List all calibrated equipment used during testing of this modification.
4.5.3 The testing is completed and all Acceptance Criteria have been met.
- - -_Date .. ...-
Pag.e 20 of 24
NUCLEAR POWER BUSINESS UNIT 1WP 99-029*C.
INSTAILATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP IP-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT 1 February 20,2002 5.0 RESTORATION 5.1 pre-A2Mtance 5.1.1 The following items must be completed prior to acceptance:
- a. All ECRs have final approvals.
- b. All update items required prior to acceptance on PBF-1606 have been completed.
- c. All testing described above has been satisfactorily completed.
All of the above items have been completed.
RE ~Date P 5.2 System Restoration 5.2.1 Close out any remaining tagouts and permits for this.1WP.
5.2.2 Aux Feedwater pump IP-29 is reidy for release for operation.
DSS . , _ " Date Time 6D . AC"CEPT.-A-*.. ._ ..- . . . .. ,--. . .,.. ..... -. .. ..... . .. ........ .. ...... ....
6.1 Verify systems and components affected by this modification are placed in an appropriate condition for present plant configuration. A 6.2 If fire rounds in progress, then discontinue fire rounds. &4 ,
6.3 FInal Acceptance TPis installation and the associated modification have been installed and tested and are acceptable.
DSS 4 7~ Date/~
Return completed IWP and modification to Responsible Engineer Page 21 of 24
4UCLEAR I IER BUSINESS UNIT IWP 99-C C NSTALLATION WORK PLAN WO .0202509 WORK PLAN WUX FEED WATER PUMP I P-29 MINIMUM FLOW
'%ECIRC LINE ORIFICE JNIT I t. Febrntry 20, 2002 .
op.-
ATTACHMENT A iB ILL OF MATERIAL Item SR Size Description Model Number Oty Stock ID No. Comments 1 S 2" Pipe, Sch. 80 5ft 901-7551- Seamless Stainless Steel, ASTM A-312, Type TP 316 2 S 2" Elbow, 90 deg., 30001b, 1 901-5019 ASTM A-182, Grade F-304 socket weld 3 S 2" Pressure Reducing Flowserve 94-16249 1 100-1198 ASTM A-351 Type CF8M body, A-479 Device Type 316 bonnet, 600# class, Socket Weld ends 4 N 2" Gasket, Flexitallic 2 915-4223 Flexitallic, 1500 Lb., Blue Asbestos and 304 S.S.
BESTCOPYAVA Ijj.5
,! Pagee 22 or 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029"C.
INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT I February 20, 2002 ATTACHMENT B 1RO-4003 Manual Adjustment Steps (Reference drawing 94-16249)
B.I. Remove safety wire (Item 260).
B.2. Loosen jam nut (Item 244).
B.3. Rotate stem as necessary with the pin nut (Item 235P) to adjust flow - clockwise will reduce flow.
B.4. Tighten jam nut (Item 244).
B.5. Retest per Section 4.4.
B.6. If test is satisfactory, then install safety wire (Item 260).
Page 23 of 24
NUCLEAR POWER BUSINESS UNIT IWP 99-029*C.
V INSTALLATION WORK PLAN WO 0202509 WORK PLAN AUX FEED WATER PUMP 1P-29 MINIMUM FLOW RECIRC LINE ORIFICE UNIT I February 20,2002 ATTACHMENT C WELD MAP IAF-15 IRO-4003 1FE-4049 114
- I.- . . - ý . - ý- ". ýz.,:- ý.
BEST COPYAVAILABLE Notes:
I - No oversizing needed on buttweld at flange. Pipe section welded to flange may be re-used.
2- Weld #1 is P1 to P8. All other welds are P-8 to P-8.
3 -All welds shall be oversized (2xl) per SK-MR-99-029*C.
4 - Add field weld flags to this weld map, based on how the piping was Installed.
Page 24 of 24
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"CONDITIONS .
DANGER TAG NA
".....SCOPE......... ...............
- -'a-. ANGEIRTAG-- N/A . "
"- REFERENCES "
MMITATIONS Qualified welder to WPM 2.P8-GT AND PRECAUTIONS Store completed pre-fab under W.O.# 0202509 TOOLS IJN 9017551 2" Pipe. Sch. 80 S.S. ASTM A-312 Type 316 M.ATERAND UN 9015019 2- Elbow, 90 deg.. 3000#. Socket Weld MATERIALS orifice. SIR# 2856 L- 1L 7. P003467 RO-04003 new valve type QUALITY CONTROL QC REVIEW OF WORK PLAN (independent QC review required on OA classified work order only) NA if non-QA work order Z C. l .o2.
Any change in scope requires WO WP review by QC inspector. aQC Date INSP.
. . .. . . . S U PP O R T
-SUPPORT- 0-" Ch~emis*ry... **--. .*-=* .
O Engineering o HP o _ i&C o Maintenance 0 NDE VT Root pass PT Root pass VT all final welds o Operations O* QC BS~PAL!L Qz Security GESTCOP;AVAIL .
O Crane -TB 0 PAB 0 Polar [I Other 1-- o Other
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SEqulprnent 1D: RO-04003 PREFAB SHOP.WORK:..
Equipment
Description:
. IP-29 AFP MINI RECIRC ORIFICE.
Work Pla'.OrIghnator: --.. Mike Desroches X6919 -:.
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Supervisor I Job Leade" 'tcondudt pre-job brief usi'rý"PBF-92117 (Mtnca-I&C)*rO*M3. (OFs)'-"
NOTE:' Pre-job brief may'require attendance of other workgmroups Involved Inthe wo:rk. activity:.
... *_: ~... . . . ..
PRE-JOB BRIEF COMPLETED Supervisor or Job Leader
-- I .
NOTES NOTE: The steps In this work plan may be performed in any logical order.
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 I 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: When replacing parts, compare the old part to the new part to verify it is an acceptable replacement
"*OTE: "If work scope changes, an R/JRIM form may be required for parts replacemnent* .repl'.. i
- NOTE: Any pen and ink change to work plan requires initial and date by the change.
NOTE: Write WO number on top/header of any supplemental pages added to work package. i.e., forms, procedures.
checklists...
t.
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Equiprment Description':. IP-29 AFPIMINI RECIRC ORIFICE. -.:.' " .. .
.Work Plan OrigInator.'. 0. Mike Desroches X6919.. ... Date: June 12,2002.' .
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- l No. . .- ... . CAUTION... .--.. .. . 2; A,.d. :7 Hold Step*Work.Plan Desc pti.- Wor Date
"-Weld sizes are not typical, reference SK dwg."7 Welder 1. General Information Step Caution: Reference weld map below and SK-MR-9-0*29*C to determine which Weld sizes are portion of the piping can be prefabricated and inspected.'Welds # 2. 3. &
not typical, 4, can be made as prefabs. (Note: Weld 4 is being done to increase weld reference SK size). " - ..
dwg.
AF-1 5 2
3
"- FE-4049 45 NOTE: All welds to be oversized. The length of the weld shall be at least W, which is 2X the.thickness of %"at the socket joints. See DETAIL 1 and Note of SK dwg. for clarification. /*- */;,r' MT DATE
- 2. Obtain field measurements as required to allow prefab work on piping. .//(L. L'"*J MT DATE CAUTION Root Pass VT & PT Inspection is required after each root pass weld.
BEST COPY AVAILABLE
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- Onl a qi.a'fied.weldeir tb WPM 2.pB Hon lds.s per welders 2&'.*'
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checklist and wg. SK-MR-99-029.*C De ilandotes".
- r- Weld #2* Welder l.D.4,/ 1L.Z .. : Root Pass 0f Weld # 3.. Welder I.DI. P'--" ". Root Pass I3*
,. Weld # 4 Welder I.D..,'; R Root Pass 0 ,
. " ".: . A. ,
- 4. Perform'a VT exam of root pass for any completed vwelds below.
Reference weld map SK-MR-99-029*C for weld sizes.
Applicable code per the following:
VT - ASME B31.1 - 1992 Pipe Class:* DB-3 Welds (Ref. Weld Map) VT EXAM NDE ExaTniner V Weld #2 -r" Sat 0 Unsat
- Weld #3 (Z Sat 0 Unsat C; I.b
- Weld #4 2-Sat 0] Unsat I?
NDE DATE
- 5. Perform a PT exam of root pass for any completed welds below.
Applicable code per the following:
PT - ASME B31.1 -1992 Pipe Class: DB-3 "Welds(Ref. Weld Map) PT EXAM NDE Examiner
.. W*dd'2 n .. 0. at QJ n]it .. I-U
. Weld #3 Sat 0 Unsat A_44 Weld #4 U-Sat 0 Unsat "liid 9- f,'
NDE DATE
- 6. A qualified welder to WPM 2.P8-GT to perform final welds as applicable, checking off welds as completed. Oversized socket welds are required.
All welds to be performed as per welders checklist and dwg. SK-MR-99
- 029"C, DETAIL 1 and Notes.
- Weld # 2 Welder I.D. L) AI-/Final Pass
.Weld #3 Welder I.D. Final Pass MT DATE
, Weld # 4 ' Welder .D.-__=_ Final Pass [*
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Equipme'ntlD: . RO-04003 PREFB SHOPWORK- P .
EquipmentDesc.iption: -1P-29 AFP MINI RECIRC ORIFICE Work Plan.Originator;: Mike Desroches X6919 .:9 -. " -
Date: June 12, 2002
.. Point Hold Se , Work Plan . . ....-.. , , W ..De comleed Pon.nrd
, 7...-. Perform'a VT exam of final passlfo .ar4 letedwldS b elow. . . f*nrp
. Applicable code per the following:* .. .
VT.- ASME 631.1 - 1992 Pipe Class: DB-3 Welds (Ref. Weld Map) VT EXAM NDE Examiner
- Weld#2 Qf'at 0'1 Unsat I*iZ e; "I.
- Weld #3 O'Sat [I Unsat "'. . -4 ,.".
- Weld #4 [2 Sat 0 Unsat M%.G.. '-1C"z. /NDE DATE FME- 8. Provide FME covers over completed section of prefabricated piping. ? (" , q ' '
MT DATE
- 9. Adequately label prefabbed section of piping and store in the QA cage. - ,.
under W*O.# 0202509 component RO-04003. (UI R27 work)
MT DATE Caution: Do not use tape on piping. "_ _
PMT will be performed on field installation work order, 0202509. No PMT required for this prefab work.
OPERATIONS RETURN TO 1 N/A Prefab work only. 4, ts- -7?7 '//q2q/
SERVICE DATE OPS TESTING POST-JOB BRIEF Conduct post-job debrief using PBF-9218 (Mtn and I&C) or OM 3.29 (OPs). Document lessons learned, good ppd6ces.
problems encountered. etc. on feedback form. Debrief should include all applicable work groups.
POST-JOB DEBRIEF COMPLETED ,.peorvisor J Leader L , "* ate FEEDBACK Fill out feedback form attached to work package (maintenance group use PBF-9929) M
_______ q MT DATE t
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WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT AUX. FEEDWATER SYSTEM PRESSURE REDUCING DEVICE DATA SHEET I OF 2 IDENT.
1 DESCRIPTION: ADJUSTABLE PRESSURE REDUCING DEVICE (APRD) 2 QUANTITY REQUIRED: TAG NO.: P & ID NO: BECH. 6118 M-217 SH. 1 1RO-04003 a3 CooS cASs: ASMEB16.34-1996 Edition (1998 Addenda) 4 SAFETY RELATED: YES (DEVICE BODY & WELDS- SEISMIC CLASS: YES - 1 100% 1 SERVICE 5 FLUID: 1. NORMAL OPERATION: FILTERED &DEMINERAUZED WATER
- 2. EMERGENCY OPERATION: SERVICE WATER DESIGN MINIFLOW 6 PUMP DISCHARGE PRESSURE (PSIG):
1440 1370 7 APRD INLET PRESSURE (PSIG): 1440 1350 8 APRD OUTLET PRESSURE (PSIG): 50 15 9 TEMPERATURE (F)32-120 32-120 10 PREDICTED FLOW RATE: BY VENDOR 11 SERVICE CONDITION: FLOW CONTROL AND PRESSURE REDUCTION 12 PUMP RATED FLOW - APRD ASSOCIATED: 400 GPM 13 REQUIRED APRD FLOW (MIN/MAX): 120/130 GPM 14 PIPING & FITTINGS AP UP TO INLET OF APRD: 20 PSIG 16 MAX. ALLOWABLE SOUND LEVEL, 75 DBA (THREE FT AWAY FROM THE APRD UNNSULATED)
BODY 15 AP--RO bobY SizE- - 2 ....
16 APRO DESIGN RATING (ANSI PRESSURE CLASS): 600#
17 APRD BODY MATERIAL A351 CFBM 18 END CONNECTIONS: SOCKET WELD 19 CONNECTING PIPE SIZE/SCHEDULE: 2 INJ SCH. 80 20 BODY LENGTH END-TO-END: BY VENDOR 7 WEIGHT: BY VENDOR REVISION - FEAE Y DATE.
APPOVE
__j_4_
REVIEWED BY:
.BY:
DATE t° Pot?
ATlil ey
WISCONSIN ELECTRIC POWER COMPANY POINT BEACH NUCLEAR PLANT AUX. FEEDWATER SYSTEM PRESSURE REDUCING DEVICE DATA SHEET 2 OF 2 ADDITIONAL REQUIREMENTS
- 1. Adjustable pressure reducing device (APRD) assembly shall be designed to withstand seismic loading equivalent to 3.0 g in the horizontal direction and 2.0 g in the vertical direction. When exposed to the above loading the APRD shall be capable of performing all its functions.
Vendor shall furnish seismic analysis and design report for Purchaser's review.
- 2. Hydrostatic testing of the APRD body shall be conducted in accordance with ASME/ANSI B 16.34 except that the test pressure shall be maintained for at least 30 minutes.
- 3. All ,materialsin contact with the working fluid shall be austenitic stainless steel.
- 4. Castings and wrought materials procured for the manufacture of valve body and trim shall be in accordance with applicable ASTM and ASME specifications and Certified Material Test Reports shall be furnished.
- 5. All welding shall be in accordance with ASME Section IX requirements.
- 6. Examination requirements for pressure retaining parts of the APRD shall be in accordance with ASME/ANSI B 16.34, Section 8.0. In addition, radiographic examination shall be performed in accordance with ASME Section V, Article 2.
- 7. The APRD flow rate adjusting device shall be equipped with means to secure APRD position using a lockwire attached to the stem and bonnet.
- 8. Fabrication drawing, indicating APRD parts list and their associated ASME/ANSI standards, shall be submitted for NMC approval prior to APRD fabrication.
- 9. Provide Certificate of Compliance attesting that the APRD is designed in accordance with P.O.
specified requirements.
- 10. Provide eight (8) copies of Instruction Manuals for APRD including Parts List and Part Numbers.
REVISION PREPARED BY: DATE:
REVIEWED BY: 7- DATE:.
APPROVED BY: '0'e'. Ia ,,zK DATE:., / / .
.1.
It
Chapman, Rob From: VanderVelde, Brian Sent: Monday, March 04,2002 11:11 AM To: Chapman, Rob
Subject:
MR 99-029"C/*D MM Comments
- 1. Do we have all the parts on site for this mod? Are they OA receipted already and released with OAR numbers?
- 2. Who Inthe planning department will get the work package statused and released? CHAMPS indicates 0202509 is still status 10. There Isalso the pre-fab package 0202508 that Is still status 10. The pre-fab package should be scheduled In the 12-week process several months Inadvance of the Unit I outage. .
- 3. Will the piping be completely drained? Will there be an Issue of water leaking through a valve and Interfering with welding?
- 4. Are you submitting CHAMPS updates for any valves, equipment, etc. that will be different that what Is stated for the specific equipment Id's?
- 5. For your Information: WP-7 and WP-8 now have a complimentary procedure that Is Inthe new Welding Program Manual format WP-7 and WP-8 will be maintained for an indefinite period of time. We can still reference them. The new procedures are WPM 2.P1-8-GT and WPM 2.Pr-GT.
- 6. Will the orifice come ready to Install, or Isthere any pre-Installation testing Involved for the orifice?
- 7. The note after Step 3.2.5 noting the exact orientation of I FE-4049, should be moved to just prior to Step 3.2.5. The second sentence of the note Is an action statement and should have a signoff step Itself or Incorporated Into another sign off step.
- 8. Work with the planner to develop a Welder's Checklist Otherwise, It looks workable.
Should Ihold on to the copy you gave me or toss it?
Thanks, Brian I .. . , - .. - .. .. , .. . ..-. -
- _. .. * . .. - . . , -** . - *" '*.- 4 ' * - .
I
Page 1 of 1 Point Beach Nuclear Plant RECORD REVIEW - REQUEST FOR COMMENTS Record Type: MR 99-029*C Record ID:
Unit: CI 0 1 [1 2 Record Date: Draft COMMENTS RESOLUTION (INCLUDE SIGNATURE & DATE) (INCLUDE SIGNATURE & DATE)
Step 3.2.9 Consider having QC also verify orientation of the orifice Step 3.2.10 1AF-15 QC taking Temp Stick readings on the valve?
Qft-4b. 7,*.w
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How are we going to satisfy the caution? What temp do we have to keep the valve under ? Is 44-4,m l/,I 3.2.12 f14.4~Vj~
B31.1 internal misalignment is 1/16", easier for the welder and planner if they do not have to look this up.
32.13 Root portion of fillets? Are we using code case N-416? , % - A...-
If not, I am unsure of the need to call-out roots for all cudor ,-,0 60.
welds. (Butt welds only? - specify which welds receive a root inspection)
SK-MR-99-029*C It would help to predetemine the weld size and place it on the sketch for the welder and NDE inspector to follow, Jeff P ce-Ice-c V
ROUTE COMPLETED FORM TO NP-FILE I RECORDS MANAGEMENT PBF-1622 Revision 2 09t19/01 1;L. Aq7
Chapman, Rob From: Chapman, Rob Sent: Monday, October 14,2002 12:25 AM To: Glaser, Jill Cc: Chapman, Rob SubjectL EPIX update for MR 99-029"C MR 99-029*C replaced the Unit I turbine driven AFP (1 P-29) mini flow recirculation orifce (1RO-4003).
The new orifice Is a Valtek Mark I pressure reducing device, and is essentially a globe valve without a handwheel and some cavitation control trim Inside.
If you need any more Info for EPIX, please let me know.
Thanks.
Rob Rob Chapman MechanicalDesign Engineering PointBeach NuclearPlant Ph 920-755-7636 Fax 920-755-7410 1
POINT BEACH NUCLEAR PLANT MSSM 2000-023 MANAGER'S SUPERVISORY May 12,2000 STAFF MEETING Page 1 April 18,2000 1300-1515 hrs.
Members: W. J. Herrman (Chairman)
R. F. Homak (ENG) B. J. O'Grady (OPS) J. T. Flannagan (MTN)
G. A. Corell (CHEM)
VisitorstGuests:
A. B. Beach M. Flynn* M. B. Arnold T. Cottengim*
D. desRochers* C. M. Jilek* 3. Stanford* A. Foltynowicz*
T. G. Malanowski* J. P. Schroeder* S. G. Gucwa* *Part-time
- 1. In accordance with TS Amendments 190/195 and FCR #99-061 the MSS reviewed and AprP-rv th fUit ,o r oeod'upressi:
Ping Mproc AOP-lB, Unit 1, Reactor Coolant Pump Malfunction, Revision 11. (Permanent)
A0P-lB, Unit 2, Reactor Coolant Pump Malfunction, Revision 11. (Permanent)
IT 80, Unit 1, Main and Radwaste Steam Valves (Quarterly), Revision 20. (Permanent)
IT 85 Unit 2, Main Steam Valves (Quarterly), Revision 19. (Permanent)
NP 1.9.15, Tagging Procedure, Revision 12. (Permanent)
Mr. Stanford presented the significant changes proposed to NP 1.9.15. A major change from past practice relates to the installation of padlocks when operating permits are used.
Mr. Flannagan noted the daily sign in and off requirement for the protected worker log (PWL). He questioned the feedback received from the Maintenance organization on this specific change. Mr. Stanford said the Maintenance department received training on this process. Everything is done through SOMS-NT. This is accomplished via various PCs throughout the plant. Only the supervisor will have to go to the Work Control Center; all others can sign in/out at any PC with SOMs capability. Mr. Flannagan voiced his concern with contractors working here during the outage and their access to a PC. He is concerned with lack of productivity because of waiting lines at PCs for people signing on and off jobs.
Mr. Stanford said other options were evaluated including a link to the Security computers and also a link with personnel badges; however, these options were considered to be not as feasible as the selected method. Mr. Flannagan asked whether SOMS was in place at this time. Mr. Stanford said SOMS is scheduled for use on May 12, 2000, which is also when NP 1.9.15 will be issued.
The MSS concurred with the changes and approved the procedure. The MSS recommended that future revisions to this NP be reviewed by the tagging committee and not the MSS.
Messrs. O'Grady, Hornak, Corell and Flannagan felt the formation of the tagging committee was positive and it should be utilized as the future review body of changes to the NP.
REOD MAY 2 5 2000
POINT BEACH NUCLEAR PLANT MSSM 2000-023 MANAGER'S SUPERVISORY May 12,2000 STAFF MEETING Page 2 Step 2.3 had been added in the past as the method to ensure MSS review occurred based on a request by the Operations manager who is now the plant manager. This recommendation will be made to the plant manager for his consideration. The step will remain as is until this is resolved by the plant manager.
0137, Shifting of Instrument Supply Bus Feeders, Revision 32. (Cancel)
Mr. Ettien said 0137 is canceled and replaced by I&2-SOP-Y-001. The MSS approved the cancellation.
IPT-MS-1, Unit 1, Main Steam System Pressure Test - Outside Containment, Revision 0.
(New Procedure) 2PT-MS-l, Unit 2, Main Steam Systm"mPi6s Test-- Outside Contalinmeint, Revision 0.
(New Prbcedure) 1PT-MS-2, Unit 1, Main Steam System Pressure Test - Inside Containment, Revision 0.
(New Proedure) 2PT-MS-2, Unit 2, Main Steam System Pressure Test - Inside Containment, Revision 0.
(New Procedure)
Mr. O'Grady questioned the acceptance criteria for the new pressure tests (PTs).
Mr. Cottengim said Section 6.0 contains this criteria and Steps 5.7 and 5.8 discuss the methodology for data reviews. Mr. O'Grady asked how packing leaks impact these tests.
Mr. Cottengim said the tests still can be acceptable and repairs made if the leak is identified.
The MSS had no further questions. The MSS approved the new PTs. Also, in accordance with NP 1.6.5, the MSS determined that future revisions to these four new PTs can be accomplished via a QR and not the MSS.
0-SOP-Y-001, 120 V Vital Instrument Inverters, Revision 0. (New Procedure)
The MSS approved the new procedure. Also, in accordance with NP 1.6.5, the MSS determined that future revisions to this procedure can be reviewed and approved by a QR, not the MSS.
1-SOP-Y-001, Shifting 120 V RPS/Safeguards Instrument Buses, Revision 0.
(New Procedure)
The MSS approved the new procedure. Also, in accordance with NP 1.6.5, the MSS determined that future revisions to this procedure can be reviewed and approved by a QR, not the MSS.
POINT BEACH NUCLEAR PLANT MSSM 2000-023 MANAGER'S SUPERVISORY May 12,2000 STAFF MEETING Page v
3
,-- N mm T ,
- n =nT 2-SOP-Y-001, Shifting 120 V RPS/Safeguards Instrument Buses, Revision 0.
(New Procedure)
The MSS approved the new procedure. Also, in accordance with NP 1.6.5, the WSS determined that future revisions to this procedure can be reviewed and approveo Iby a QR, not the MSS.
- 2. In accordance with NP 1.2.6, the following screenings were accepted as non-IF E by the MSS:
IPT-MS-1, Unit 1, Main Steam System Pressure Test - Outside Containment, ! evision 0.
2PT-MS-I, Unit 2, Main Steam System Pressure Test - Outside Containment, l evision 0.
1PT-MS-2, Unit 1, Main Steam System Pressure Test - Inside Containment, R vision 0.
2PT-MS-2, Unit 2, Main Steam System Pressure Test - Inside Containment, R"vision 0.
0-SOP-Y-001, 120 V Vital Instrument Inverters, Revision 0.
I-SOP-Y-001, Shifting 120 V RPS/Safeguards Instrument Buses, Revision 0.
2-SOP-Y-001, Shifting 120 V RPS/Safeguards Instrument Buses, Revision 0.
- 3. In accordance with TS Amendments 190(195 and FCR #99-061 the MSS reviei ,ed the following meeting minutes. Minor changes were recommended and incorporat d and the minutes subsequently issued:
MSSM 2000-019, meeting held on March 21,2000.
MSSM 2000-020, meeting held on March 28, 2000.
MSSM 2000-021, meeting held on April 4,2000.
- 4. In accordance with TS Amendments 190/195 and FCR #99-061 the MSS revieN'ed and recommended approval of the following safety evaluations, unless otherwise nc -ed. No unreviewed safety questions were identified during MSS review:
SE 0QQQ-0-02, MR 98-002*A, Safety Assessment System (SAS)/Plant Proc Iss Computer System (PPCS) Replacement Modification - Pre-Parallel Run Test on New PP( S.
Mr. desRochers presented the SE revision that addresses deenergizing RMS-C'I I and/or RMS-CT2. This is a change in design scope that is necessary to allow interface testing.
Mr. Corell asked if compensatory measures exist for this circumstance. Mr. de, ?.ochers said RMSASRB 2.0 contains this information.
POINT BEACH NUCLEAR PLANT MSSM 2000-023 MANAGER'S SUPERVISORY May 12,2000 STAFF LMETING Pae4 Mr. O' Grady noted the current problem between the old PPCS and the new Eberline server interface that isn't compatible. He asked if this MR corrects the interface communication issue. Mr. desRochers said this issue is being evaluated as part of this modification. We need to ensure the communication interface is working between the new PPCS and the Eberline server.
The MSS recommended approval of the SE revision without comment Mr. Kaminskas approved the document in Mr. Mende's absence.
SE 2000-0045, CR 97-0968, Valve Lineup Changes for Instrument Air and Service Air to Fuel Transfer System.
The MSS recommended approval of the SE without comment. Mr. Kaminskas approved the d et"ri sase . .. .
SE 20000055, MR 99-029*AJ*B, P-38A*B Auxiliary Feedwater Pump Minimum Flow Recirculation line Flow Orifice Replacement.
Mr. Foltynowicz presented the SE for MR 99-029*A and *B thataddresses EWR 99-031 and CR 99-1391. Mr. Herrman asked what constitutes the PMT for this MR. Mr. Foltynowicz said the testing consists of starting the applicable pump and measuring flow through the orifice.
Mr. Herrman questioned the following phrase contained in response to Question 5: "The AF pumps have 9 stages with 0.009" to 0.014" diametrical clearances and a minimum 0.4375" impeller vane path. Since the pumps have multiple stages and small clearances, they will reduce larger particles size contained in the SW to less than 0.015"." He did not feel the statement is entirely nuelecause he felt that particles could go through the vane paths.
Mr. Schroeder said the pump has 9 stages and the particles would have to clear all 9 clearances in order to pass through. This is very unlikely to occur. Mr. Foltynowicz subsequently removed that phrase and included a different discussion contained in the paragraph below.
Mr. Hornak said the discussion that could be included in response to Question 5 is that of plant configuration and the fact that SW is used. Hbalso did not libcf quantitative discussion and recommended discussing the recirculation line flow path. Mr. Foltynowicz removed the quantitative discussion and included the following in the SE: "In addition, the safety related function of the AF pumps P-38A and P-38B is to deliver sufficient flow for accidents th4 are time sensitive to AF system startup (LONF, LOOP), LOL, SGTR and MSLB accidents and provide sufficient flow for long term decay heat removal for accidents such as a SBLOCA. The recirculation line flow path is not required to support this function since the pump discharge valves will automatically open fully in response to the accident and provide a flowpath for the pump. The recirculation line AOV automatically closes approximately 95 gpm and increasing. Failure to pass flow through the recirculation orifice
POINT BEACH NUCLEAR PLANT MSSM 2000-023 MANAGER'S SUPERVISORY May 12, 2000 STAFF MEETING Page 5 during the 45 seconds would be conservative since flow to the SGs would be delivered sooner. The recirculation line AOV is also designed as a failed closed valve to ensure that recirculation flow is not diverted from the SG in the event of a loss of instrument air."
With the change made, the MSS recommended approval of the SE. Mr. O'Grady approved the document in Mr. Mende's absence.
SE 2000-0050, MR 98-002*D/*E, Safety Assessment System (SAS)IPlant Process Computer System (PPCS) Replacement Modification - Control Room Remodeling and Rewiring.
Mr. Flynn presented the SE and associated FCR that addresses MR 98-002 design packages D and E. Design package D addresses the control room modifications to the operator consoles (3 for each unit), new SAS alarm screens, new PCs, new work space and raised floor changes. A humrn f a eiiii was-iperf6riiil ori-thle iiew design layout. Four activities will include the use of volatile organic chemicals (VOCs). Activities are: Touch up painting of file cabinets; silicone adhesives for solid surface countertop installation; pedestal adhesives; and carpet adhesives. Mr. Flynn said use of these VOCs will be observed and usage limited during this MR to assure personnel health is maintained. Mr. Flannagan asked if the area will have monitors to ensure safe levels are present at all times. Mr. Flynn said he is working with Ms. Sipiorski to ensure a safe work environment.
Design package E addresses the electrical portion of the work in the control room. The work includes providing vital power to non-vital circuits. Mr. Flynn said loads will be limited during the time the old and new PPCS are run in parallel. He said an HVAC evaluation was performed to ensure the 750 F temperature limit is not exceeded in the control room because of the load limitations. The temperature will not be exceeded.
Mr. Flynn noted that an extra operator will be available in the control room during periods when the work precludes a stationed operator from getting to the boards. Mr. O'Grady asked if a fire loading analysis was performed. Mr. Flynn responded that the carpeting, countertops and raised floor were procured as low flame and low smoke composition. Cables are kept in conduit runs. He noted the response to Question 2 addresses fire retardancy requirements.
This was acceptable to the MSS. Mr. Hornak also asked if SQUG considerations were part of the MR planning process. Mr. Flynn responded that this was evaluated and no SQUG impacts were identified. The SE description addresses the SQUG aspects.
Mr. Hornak questioned the VOC usage and whether an acceptance range is known to assure the charcoal filters remain undamaged. Mr. Flynn said the industry references 3 lbs. VOC within a 24-hour period as being acceptable. This was confirmed with Mr. Moseman in RP.
Mr. Flynn said even at this level, no impact would occur to the recirculation mode and TS are not challenged. He said if the recirculation mode is entered, the IWP instructs personnel to close up the VOC containers and halt work with the chemicals. Mr. Hornak asked whether the amount of vapor allowed for this MR is tolerable for all operators. Mr. Flynn said the use of VOCs will be limited throughout this work. A machine is brought in to draw out some of
POINT BEACH NUCLEAR PLANT MSSM 2000-023 MANAGER'S SUPERVISORY May 12,2000 STAFF MEETING Page 6 the vapors from the control room to help minimize the fumes. A monitoring device may also be staged in the control room as validation of actual levels in various areas. Mr. Flynn said the PPCS project team is very sensitive to the VOC issue. Mr. O'Grady also questioned if this work impacts control room habitability and the life of the charcoal filters. Mr. Herrman asked whether spare filters are available. The filter availability was not known at this time.
Mr. Hornak asked whether another AC mode of operation was considered during this MR.
The MSS requested Engineering reevaluate contingencies with respect to the charcoal filters and control room habitability. Mr. Corell also noted that the chemical concentrations used for these installations may have some affect on the surface of the charcoal. He asked if this was evaluated. Mr. O'Grady also questioned the VOC impact on the charcoal filters. He is not comfortable with an engineering statement that 3 lbs. VOC is acceptable without having formal documentation supporting this number. Mr. Homn alk so agreed that more "justifiationis required. He gave an example of ai-roblem that may develop during the MRl work that would require an operability determination (OD). At this time there is insufficient data to support writing an OD. Messrs. Hornak and O'Grady requested additional justification be provided in the SE to support chemical usage without damage to the charcoal filters and affecting the control room habitability study. The MSS agreed with the request Based on this issue, the MSS did not approve the SE. (See MSSM 2000-025 for additional discussions.)
- 5. 1999 Maintenance Rule Annual Report. In accordance with NP 7.7.7, Step 5.3, Ms. Jilek presented the 1999 Maintenance Rule annual report. (See NPM 2000-0325 dated March 30, 2000 for detailed report.)
The Maintenance Rule requires periodic review by management. This year's report contains an executive summary. Ms. Jilek said this report was reviewed by the Overview Expert Panel and discussed during the morning manager's meeting. The MSS asked the composition of the Overview Expert Panel. Ms. Jilek said she is on the panel as well as Maintenance, Operations, System Engineering and PSA representatives. Based on this, Mr. Herrman asked the MSS if they felt this should be reviewed during MSS meetings.
Messrs. O'Grady, Homak and Corell felt the executive summary should be provided to the MSS and discussed during a MSS meeting. They did not feel the entire report was necessary, only the summary. Based on this opinion, the MSS recommended the continuation of this review of the Maintenance Rule annual report.
Mr. Herrman questioned the reporting of the emergency diesel generators (EDGs). He gave an example of having one EDG out of service but having the other EDG aligned to the bus.
Ms. Jilek sai6dthat is counted as downtime-each EDG is counted separately.
The MSS thanked Ms. Jilek for the presentation.
POINT BEACH NUCLEAR PLANT MSSM 2000-023 MANAGER'S SUPERVISORY May 12,2000 STAFF MEETING Page 7
- 6. In accordance with NP 1.6.5, the following set of procedures was reviewed to determine the appropriate review/approval of future revisions.
SEM, Site Engineering Manual.
It was agreed that SEM 7.11.8 and SEM 7.11.9 would have future revisions approved by the MSS because the procedures are classified as IPTE. SEM 7.11.5 and SEM 7.11.6 are reclassified as administrative procedures. Other SEMs will be reviewed by a QR.
M. B. Arnold W.AppHroean (Approved) cc: OSRC c/o B. J. Onesti M. E. Reddemann M. B. Arnold A. B. Beach G. M. Krieser M. B. Sellman D. B. Black J. J. Walsh File
MSSM 2000-023 Attachment INTERNAL Page 1 of 3 CORRESPONDENCE To: Manager's Supervisory Staff From: Duane Schoon, Steve Gucwa Date: April 3, 2000 SubJect: Site Engineering Manual (SEM) Procedures Classification Review Copy To: M. B. Koudelka Attached is a listing of the Site Engineering Manual (SEM) procedures which were classified as "Minor" procedures, as well as identifying those which are designated as an Infrequently Performed Tests or Evolution (IPTE). This listing also includes recommendations to classifications as described in and in accordance with NP 1.1.2, "Procedure and Administrative Controls in NPBU", and NP 11.3, 'r-ce'-uPrel'paration, Review aid Approval".
The following is the decision making process employed in determining the recommended procedure classification:
"* Documents were reviewed to the criteria of NP 1.1.2 to determine the document type; whether a Licensing Document, Policy, Guideline, Controlled Reference Document, Work Plan, or Procedure. The review concluded that the attached SEMs are procedures.
"* Procedures were then reviewed to determine whether they were "Technical Procedures" or "Administrative Procedures". SEM 7.11.5, "RSC Leak Test for Unit 1", and SEM 7.11.6, "RSC Leak Test for Unit 2",were identified as "Administrative Procedures". These procedures are documentation tools (a commitment to the NRC to document findings when performing the Tech Spec required leak test). The leak tests are performed/controlled by IT 230, "Leak Test Of Class 1 Components Following A Refueling Shutdown Unit 1", IT 235,
"¶LeakTest Of Class 1 Components Following A Refueling Shutdown Unit 2", and/or 01 119, "RCS Normal Leak Test". These Administrative types received the recommended classification of Admin.
"* Technical Procedures were subsequently reviewed for Safety Related (SR) or Non-Nuclear Safety Related (NNSR) classifications. The following was employed for recommendations of SR Procedures:
- 1. Procedures that are described in Unit I or 2 Technical Specifications, Section 15.6.8 (Plant Operating Procedures), or
- 2. Activities are performed on equipment designated as SR on the Quality List (Q-List) or designated SR in CHAMPS, or
- 3. Recommendations of typical Safety-Related activities per Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)", Appendix A, Typical ProceduresI. for Pressurized Water Reactors and Boiling Water Reactors.
MSSM 2000-023 Attachment April 3,2000 Page 2 of 3 Page 2 SEMs were then examined to determine a recommended level of "Reviews" as described in and in accordance with NP 1.6.5, "Manager's Supervisory Staff and Qualified Reviewer".
"* SEM 7.11.8, "Removal of Steam Generator Nozzle Dams Unit 2", and SEM 7.11.9, "Installation of Steam Generator Nozzle Dams Unit 2", received a MSS Review Required (MSS) recommendation based on the procedures being classified as IWME.
"* SEM 7.11.10, "Reactor Vessel Interior Inspection", and SEM 7.11.11, "Steam Generator Primary Channeihead Closeout Inspection", received a Qualified Reviewer (QR) recommendation based on the following:
- 1. Procedure is not designated as IPTE
- 2. Performance of procedure does not involve multiple (more than 2) groups
- 3. Performance of the procedure does not directly challenge Nuclear Safety
- 4. Performance of the procedure does not directly challenge Power Production Bascd upon our reviews of the.proceeitres,.the, attached-recommendations are being provitei- fo.r.
MSS consideration and approval.
Respectfully, Duae6rcoon System Engineering Manager Steve Gucwa Procedures Group
I.
Manual: SEM IIManual Unit: 0 Document Type Procedure i, Old A New Safety Related MSS Eval Classification - Csssffleatlon Reviewer Meeting Complete 1 SEM7.11.5 RCS Leak Test for Unit l Minor - NA NA MSSM2000-023 [
2 SEM7.1l.6 RCS Leak Test forUnit 2 Minor ] ~. NA NA MSSM2000-023 []
2 SEM 7.11.8 Removal of Steam Generator Nozzle Dams Unit 2 Minor [ Sfety Related Managers MSSM 2000-023 [
Supervisory Staff 2 SEM 7.11.9 Installation of Steam Generator Nozzle Dams Unit Minor [ Sfty Related Managers MSSM 2000-023 R 2 1 Supervisory Staff 0 SEM 7.11.10 Reactor Vessel Interior Inspection Minor C] Stfety Related Qualified Reviewer MSSM 2000-023 []
0 SEM 7.11.11 Steam Generator Primary Channelhead Closeout Minor [ Safety Related Qualified Reviewer MSSM 2000-023 Ga Inspection I
¶ I
9I Olo 0Ma
~V~w Friday, April 28. 2000
Point Beach Nuclear Power Plant Organizational Assessment Charter Purpose The purpose of this assessment is to identify processes, procedures, and organizational dynamics that have resulted in continued issues with Operational and Engineering performances at Point Beach Nuclear Power Plant (PBNPP) and barriers to improving plant and personnel performance.
Background
The success and-long-term viability of Point Beach Nuclear'Power Plant is dcpcndmnt oa the ability of the organization to identify issues and resolve them effectively to insure safe, reliable, cost effective operation.
Previous events at PBNPP indicate the organization has not been fully effective in timely resolution of plant issues. Although the station has run well and performance has improved, the pace of improvement is not sufficient to achieve Excellence in the time frame consistent with other NMC plants. Recent examples have occurred in several areas. The trend in maintenance and engineering related backlogs indicate progress has been lagging. Causes of equipment issues in the Spring 2002 Refueling Outage, related to human performance, have been evident in subsequent issues. Overall these are indications of a need to fully understand the processes and interactions that have hindered improvements.
The most recent PBNPP red finding on the Auxiliary Feedwater System identified similar potential problems. Resolution of this old design issue required the station to reevaluate the way previous portions of the system were treated. Subsequently Point Beach identified another potential failure mechanism on this system involving the same recirculation path. The issues associated with these modifications indicate insufficient Engineering rigor and Operational oversight. The processes, procedures and organizational dynamics leading to the recent issue need to be understood and a plan put in place for effective resolution.
The assessment team will accomplish this based on a review of procedures, processes, oversight committees, external communications, interviews with personnel and observation of station activities.
I
Sco&e The scope of this assessment will include a review of a broad spectrum of processes that have the potential for identifying process and organizational weaknesses. In addition, to ensure a thorough understanding of past decisions and their completeness and impact on current plant operation several specific area reviews will be conducted. The following list of items will be included for review/observation.
SPECIFIC ANALYSIS AND REVIEW Overall AFW System Adequacy - Identify and evaluate limiting performance areas Modifications performed on AFW (Safety and Non-Safety) - Past 10 years Modifications on other Safety Systems whether the modifications were safety related or not - Past 5 years Previous 50.59s and 50.59 screenings - Past 5 years Past MSS minutes - Past 5 years Plant and Site Oversight Committee Minutes - Past 5 years CAP/Action Requests- Past 5 years - non-lower level items that are system/problem related Performance Indicators performance - persistent areas of slow improvement Regulatory Submittals - Past 5 years NRC Inspection Reports - Past 5 years Last 3 INPO evaluations Excellence Plan Action Areas and Completeness 2
DECISION MAKING AND PROCESS IMPLEMENTATION
'Organizational Dynamics Work Planning/Scheduling process
- Accountability methods
- Culture Survey Operability Determinations
- Teamwork methods Modification Process CAP Process Temporary Modification Process Root Cause Process MSS Process System Related Health Reporting 50.59 Prooess/50.59 Screening process Deliverable The deliverable for this review will be a document that provides weaknesses and strengths identified during the review and will specifically evaluate the Auxiliary Feedwater System adequacy. From this input, the information will be overall Streamed and the Excellence Plan will be updated to provide the overall guidance for Station improvement Point Beach Assessment and Organizational Assistance Team - General Areas for Review Team Lead and Coordinator - Gary Van Middlesworth Engineering/Regulatory Areas - Jim Taylor Regulatory - Ed Weinkam Operations and Organization - Warren Fujimoto Culture Survey -Synergy (John Guibert)
Organization/Excellence Plan - Doug Cooper Technical & Engineering - Mano Nazar Operations/Work Planning - Jack Purkis Technical and Engineering - Mark Reddemann Organizational Dynamics - Dan and Beth Nilsson (Tall People)
Licensing/Root Cause/Operability Determination- Keith Young/Ed Weinkam Lori Armstrong - Observer Mentor - John kIolden INPO -TBD 3
DECISION MAKING AND PROCESS IMPLEMENTATION Organizational Dynamics - Doug Cooper Warren Fujimoto Nilsson's
- Accountability methods
- Culture Survey - John Guibert
- Teamwork Methods CAP Process - Jim Taylor & Mano Nazar Root Cause Process - Keith Young System related health reporting - Jim Taylor & Mano Nazar MSS Process - Warren Fujimoto & Gary Van Middlesworth 50.59 Process/50/59 Screening process - Ed Weinkiam Modification Process - Mano Nazar Temp Mod Process - Jack Purkis Work Planning/Scheduling process - Jack Purkis 4
SPECIFIC ANALYSIS AND REVIEW AREAS Overall AFW System Adequacy - Mano Nazar & Mark Reddemann Modifications on AFW, Safety and non-safety - Jim Taylor & Mark Reddemann Modifications, Safety Systems - Mano Nazar & Mark Reddemann 50.59's and 50.59 screenings - Ed Weinkam & Keith Young MSS minutes -- Warrerrfujirnoto & Gay-Van Middleswvith ...
Plant and Site Oversite Committee Minutes - Jim Taylor CAP/Action Requests - Warren Fujimoto, INPO Performance Indicators - Doug Cooper & Gary Van Middlesworth Regulatory Submittals - Ed Weinkam & Keith Young NRC Inspection reports - Jack Purkis Last 3 INPO Evals - Warren Fujimoto Excellence Plan - Doug Cooper & Mano Nazar 5
NM INTERNAL CORRESPONDENCE NPM 2002-0642 To: A. J. Cayia From: Ken Peveler Date: December 5, 2002
Subject:
UPDATE ON POTENTIAL AFW COMMON MODE FAILURE EVENT RESOLUTION TEAM ACTIONS Ref: CAP029952 Copy To: File The pprpose of this message is to provide an update on the subject team actions, to recommend adjustments at tlis time due to progress made and to address organizational changes. Overall, my recominendation is that this team be dissolved and the remaining actions remain with the responsible organizations and processes.
First, an update on the specific group tasks:
&AsIncident Investigation - Stu Thomas - Task completed.
&.dOn-line Work Risk Management - John Anderson - This task is completed relative to AFW in-plant work activities impacting the schedule beyond the impacts of the potential common mode plugging issue (i.e., condensate storage tank inspection, service water system flush, AFW and recirculation orifice inspections completed, and AFW returned to service with compensatory actions in place). This places the station in a base level of risk of low "yellow" due to the AFW recirculation line remaining in a degraded condition due to the orifice design. Orifice design activities are in progress for replacement. In the mean time, plant risk assessment continues to manage station work while in this yellow condition.
Resolution of this issue is expected to be extended into January, when the newly designed orifices will be available for installation and return the station to a "green" base line.
fInterim Corrective Actions - Duane Schoon - Task completed.
&Adssue Resolution Team and Root Cause Evaluation - Jim Freels - Issues remain open in this area. Rich Flessner continues to work on the root cause under the management sponsorship of Jim Freels. Input from PH and PSA is required to complete the root cause. At this time, most if not all of the facts of the situation have been gathered and the first drafts of the root cause have been submitted for comment. Design resolution and risk assessment activities remain open (See attached).
At this time, I am recommending that I be relieved of my team leadership responsibilities for the AFW Common Mode Failure Event Resolution Team and the remaining activities be pursued in accordance with our normal processes. This will support my return to Nuclear Oversight and support preparations for my transition to Engineering Programs at Point Beach in January of 2003. Risk management and design resolution activities require close attention.
NPM 2002-0642 December 5, 2002 Page 2 Attached is an update of the Charter and an action matrix to reflect status and remaining activities. Specific areas for attention going forward include:
- 1. Coordination of design and installation activities for the new orifices, including coordination
'of other AFW work to minimize unavailability time.
- 3. Completion of risk analysis work at PHI
- 4. A decision on whether to perform testing on an orifice
- 5. Completion of the root caus..
- 6. Response to regulatory issues
/bjo Attachments:
- 1. Team Charter, with status
- 2. Action matrix
Point Beach Auxiliary Feedwater Review Team Charter
Purpose:
The purpose of the Point Beach Auxiliary Feedwater Review Team is to systematically investigate the design and licensing bases of the Auxiliary Feedwater System and to verify the as-built and tested conditions of the system satisfy those bases.
Objectives: The objectives of this review team are to revalidate the design bases by:.
?? Determining if the Auxiliary Feedwater System is currently designed, constructed, operated, maintained and tested to meet the requirements of the design and licensing bases
?? Documenting any identified discrepancies in the corrective action program
-?Facilitating the *rclutioptp a;xy idaitiled discrepancies
,?? Ensuring the design bases documents accurately reflect their conclusions.
Scope: The scope of this review is to include, but not limited to, the Point Beach Operating License and Attachments, Updated Final Safety Analysis Report, applicable regulatory commitments and correspondence, applicable plant drawings, design bases documents, installed and pending system modifications, Operability Determinations, applicable normal operating and emergency procedures, component setpoints and bases, preventive/predictive maintenance task scope and frequency, system parameter monitoring and trending activities and bases, applicable maintenance and testing procedures, quality assurance requirements, applicable training lesson plans and simulator scenarios.
Deliverables: Deliverables from this team to include:
?? Identification and documentation of discrepancies via the corrective action program 7? Generation of Operability Determinations as applicable 7? Identification of Limiting system margin components and scenarios
?? Updated design bases documents
?? Revision packages to correct document weaknesses
?? Appropriateness of safety analysis assumptions and identification of margin in these analyses
?? Comprehensive report detailing team activities, findings and recommendations
Team Composition:
TBD based on final charter Schedule:
Charter-Approval: - I---. ... I.-
AJ. Cayia Date Site Vice President
Point Beach Auxiliary Feedwater Plan Action:
Review NUREG 1022 for reportability potential Webb Review licensing and design basis information for previous communication with NRC about Auxiliary Feedwater System Webb/Kendall Independently verify power supplies and instrumentation failure matrix Miller Verify Control Room tags are reviewed and revised as necessary Schoon Review and revise operational guidance as necessary Schoon Provide plan and basis for cross-connecting G03/GO4 should one EDG become inoperable Wood Review PRA model to determine if impact of loss ofDOI and D02 correct Wood Construct Auxiliary Feedwater assessment plan Freels Verify actuation time of 30 minutes is sat with Westinghouse Kendall What impact of manual actuation for SW from HEP perspective Masterlark
OPEN ISSUES WHO WHEN COMMENTS Incident Investgaton S. Thomas 11/21/2002 CA026962 -completed 11/20/02 On-Una Work Risk Management J. Anderson Ongoing CA026958 - Charter task NP 10.3.7, Unplanned Yellow Actions R. Wood Ongoing Ref CA27023 and CAP30052/Mod required for green Interim Corrective Actions Charter Tracking Item D. Schoon 11/27/2002 CA026908 - dosed Ref CAP029999/CA026986- Tmg Needs Analysis is Training Needs Analysis and Related Actions P. Smith 2/2/2003 completed Condition Evaluation/further enhancements M. Schug 12/6/2002 Ref CAP029999/CE10848 Independent Evaluation of Procedure changes D. Schoon 11/27/2002 Ref CA026909 - completed Independent Evaluation of briefings &training D. Schoon 11/27/2002 Ref CA026910 - completed Engineering Resolution Team and RCE __,,, , _,_,,
Conduct Root Cause Evaluation - Draft Report R. Flessner 11/18/2002 Ref RCE0000191 - first draft Issued Conduct Root Cause Evaluation - Final Report R. Flessner 12/4/2002 Final requires PSA inputs PRA/drsk significance evaluation J. Masterlark 1/29/2003 Ref CA026900/Also requires PII Inputs Hydraulic system repsponse T. Kendall 12/16/2002 CA02691 1- PII final report expected 12/20/02 CA026912 - PII/PSA/Engr considerations - Not Develop a test plan T. Kendall TBD assigned Analyze corrosion products B. Zipp 1/20/2003 CA026913- PII final report expected 12/20/02 AFW sources and quantity of corrosion products B. Z1pp 1/20/2003 CA026914 - PHI final report expected 12/20/02 Modifications to orifice design J. McNamara 12/13/2002 Ref CA026918 Fabrication J. McNamara 12/17/2002 5 weeks from 11/12 is December 17 Testing: J. McNamara TBD Coordination Installation J. McNamara TBD Planning for Coordination of work schedule with station - J. McNamara TBD 13 week schedule process - Input OPR000031 Part II/Schedule for correction due - J. McNamara 11/27/2002 Complete Part II of OD 31 NEW - Electrical design modifications M. Rosseau TBD Scope, schedule, etc (Clint Drescher as designer)
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