ML030800062
| ML030800062 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/01/2002 |
| From: | Groehler R, Harper R, Sokol K Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 2002-0852, AOP 10B, Rev 3, FOIA/PA-2003-0094 PBF-0026c, Rev 13 | |
| Download: ML030800062 (51) | |
Text
Unit Nuclear Power Business TEMPORARY CHANGE REVIEW AND APPROVAL Note: Refer to NP 1.2.3, TemporaryProcedureChanges,for requirements. f-Page I o I - INITIATION Doc Number AOP 10B Document Title Safe to Cold Shutdown in local control Existing Effective Temporary Changes n/a Brief Description Add AFW Minimum Flow requirements to foldout page.
Current Rev 3 Unit PBI Temp Change No.
(identify specific changes on Form PBF-0026c, Document Review and Approval Continuation, and include with the package) 0D Initiate PBF-0026h and include with the change.
2002-0852 K
Other documents required to be effective concurrently with the temporary change:
Changes pre-screened according to NP 5.1.8? 0 NO El YES (Novide documentaion acrdmg to NP 5.1.8)
Screening completed according to NP 5.1.8? El NA 0 YES (An*ch coy)
Safety Evaluation Required? [D NO [I YES (if vea revsion ay be pmc or finmlreviews and apprvals shall be _-'ained bWo hirnke..-
Determine ifthe change constitutes a Change Of Intent to the procedure by evaluating the following questions.
S........ 61 (If any answers anrYES, a revision may be processed or final reviews and approvals shall be obtained before implementing)
Will the proposed change: YES NO
- 1. Require a change to, affect or invalidate a requirement, commitment, evaluation or description in the Current or ISFSI Licensing Basis (as defined in NP 5.1.8 and NP 5.1.7)? 0 0]
- 2. Cause an increase in magnitude, significance or impact such that it should be processed as a revision?
- 3. Delete or modify a prerequisite, initial condition, precaution, limitation or other steps that El could have safety significance or affect the procedure's margin of safety?
- 4. Delete QC hold points, Independent Verification or Concurrent Check steps without the related step(s) that require the performance also being deleted?
- 5. Change Tech Spec or other regulatory acceptance criteria other than for re-baselining purposes? 0D
- 6. Require a change to the procedure Purpose or change the procedure classificatio ? 0 0 Initiated By (printsign) Ross Groehler I Q A,.,. , , '.-Date 11/01/2002 HI - INITIAL APPROVAL This change is correct and complete, can be performed as wri', ýd does not adversely affect personnel or nuclear safety, or Plant operating c o ons. d doe_, Date Group Supervisor (print/sign) /C OL)e' ) LY Dat /0 1-\O (Cannot be the Initiator)
This change does not adversely[effect I 9prating co dit -ed procedures only)
Senior Reactor Operator (print/sign) 71D Date 10 (Cannot be the Initiator or Group Supervisor) \j M - PROCEDURE OWNER REVIEW S Permanent El One-time Use El Expiration Date, Event or Condition:
El Hold change until procedure completed (final review and approval stiJLac~quired within 14 days of initial approval)
El QR/MSS Review NOT Required (Admin/NN,§ýy) K QR Review Re* . ] MSsview Required (w, n Yarc16-)
Procedure Owner (print/sign) A/'-'
A,) I IVV- / " Date ////b A This Change and suonortine reourements correctly completed and orocessed. I IF IV - FINAL REVIEW AND APPROVAL
[ dst be comoleted within 14 days tinitial a fl I'he Initiator. OR and Aw oAutoritv shall be Indenendent from uchpther)
S (print/sign) t;,i-I Date1L///La .
Indicates 50.59/72 48 applicability assemssd, any necessary screenings/evaluations perf'"rmed, determination made as to whethof additional cross-disdplinaryreview required, and ifrequired, performed.
MSS Meeting No.
Avoroval Authority (print/sign) J- 'e -- I T').
Dte V- REVISION INFORMATION FOR PERMANENT CHANGES Post Typing Review (print/sign) /
Indicates temporary change(s) incorporated exactly as approved and no other changes made to document.
Incorporated into Revision Number Effective PBF-0026e
References:
NP 1.23 Revision 13 01116/02
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Page ___ of_
Doc Number AOP 10B Revision 3 Unit PBI Title Safe to Cold Shutdown in Local Control Temporary Change Number 2002-0852 Description of Changes:
Step
- Change/Reason Add AFW minimum flow requirement: Monitor and maintain minimum AFW discharge flow or stop the affected AFW pump as necessary to control S/G levels] See SCR 2002-0458, CAP029908, P-38A MD AFWP has inadequate Recirc Flow during IT-10. See Also CAP029952, AFW recire line possible common mode failure.
- Note: Recording of Step Number(s) is not required for multiple occurrences ofidentical information or when not beneficial to reviewers.
PBF-0026c Revision 6 04/1l101 Refercnces: NP 1.1.3, NP 1.2.3
Point Beach Nuclear Plant TEMPORARY CHANGE AFFECTED MANUAL LOCATION Page of Procedure Number AOP 10B Revision 3 Unit PBI Title Safe to Cold Shutdonw in local control Temporary Change Number 2002-0852 I - IMMEDIATELY AFTER INITIAL APPROVAL ON PBF-0026e (Non-intent changes)
(after Final Approval if change of intent involved)
This procedure change has been processed as follows: (Manual/Location) Date Performed I'] Copy included in work package for field implementation. (WO No.
[] Copy filed in Control Room teinp change binder (Operations only). / " 0f-7Z...
[ Original change package provided to t< c, S to obtain Procedure Owner Review (c.g., Owner review may be coordinated by In-Group OA Procedure 1I, Writer, Procedure Supervisor, etd4). 1/"
El El Performed By (print and sign) Carol Schroeder / 6-- z CA & Date c/-[
Z-It - PROCEDURE OWNER REVIEW ON PBF-0026e (may be performed by OA IL,Procedure Writer, etc.)
Date Performed This procedure change has been processed as follows: (Manual/Location)
- ] Copy sent to Document Control Distribution Lead for Master File. / [_/- o*.
(Not required for one-time use change)
El Copy filed in Group satellite file. (Not rquired for one-time use changes.)
El Copy filed in Group one-time use file.
[] Original Temp Change provided to ,-(- If- to obtain Final Approvals / r-62
(&g., final approval may be coordinated by In-Group OA II, Procedure Writer, Procedure Supervisor, etc.)
[] PAB OPS Shop
[* OPS Office I] Simulator (Training OAI) _ _ _
Performed By (print and sign) Carol Schroeder /14*ct5 Date / --- o 2-PBF-0026h Revision 5 06/13/01
Reference:
NP 1.23
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) VerifySCRnumberonallpa Page.
Title of Proposed Activity: AFW minimum flow requirement change to AOP, EOP, CSP, ECA, SEP, 01-62 A/B procedures Associated Reference(s) #: Removal of internals from AF- 117 and upgrade open function of AFW pumps minirecirc vlaves to safety -related (MR 02-029); SCR 2002-005-01 EOP/ARP actions for AFW mini-recirc requirement; 2002-0055, P-38A/B mini recirc flow orifice replacment (MR 99-029 *A, *B);
Flowserve Corporation Pump Division letter dated March 2, 20012; CAP 29908; CAP 29952 Prepared by. Eric A. Schmidt IJohn P. Schroeder ,~, .A Name (Print) (e Reviewed by- ___ _________ 16____A __I____ Date: 4 )
Nam-niPrint) Signature PART I (50.59/72.48) - DESCRIBE THE PROPOSED ACTIVITY AND SEARCH THE PLANT AND ISFSI LICENSING BASIS (Resource Manual 5.3.1)
NOTE: The "NMC 10 CFR 50.59 Resource Manual" (Resource Manual) and NEI 96-07. Appendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 10 CFR 50.59 and 10 CFR 72.48 screenings.
.1 Describe the proposed activity and the scope of the activity being covered by this screening. (The 10 CFR 50.59 / 72.48 review of other portions of the proposed activity may be documented via the applicability and pre-screening process requirements in NP 5.1.8.) Appropriate descriptive material may be attached.
This screening supports procedural uprgrades to address the Auxiliary Feedwater (AFW) System issue as identified in CAP 29908 and CAP 29952. Procedural guidance for operation of AFW System will be changed such that the operator must ensure that discharge flow for P-38 A/B must be greater than 50 gpm and 1/2 P-29 discharge flow must be greater than 75 gpm. If pump flow cannot be maintained within these requirements, the pump must be secured.
1.2 Search the PBNP Current Licensing Basis (CLB) as follows: Final Safety Analysis Report (FSAR), FSAR Change Requests (FCRs) with assigned numbers, the Fire Protection Evaluation Report (FPER), the CLB (Regulatory) Commitment Database, the Technical Specifications, the Technical Specifications Bases, and the Technical Requirements Manual. Search the ISFSI licensing basis as follows: VSC-24 Safety Analysis Report, the VSC-24 Certificate of Compliance, the CLB (Regulatory)
Commitment Database, and the VSC-24 10 CFR 72.212 Site Evaluation Report. Describe the pertinent design function(s),
performance requirements, and methods of evaluation for both the plant and for the caskfISFSI as appropriate. Identify where the pertinent information is described in the above documents (by document section number and title). (Resource Manual 5.3.1 and NEI 96-07, App. B, B.2)
FSAR 10.2 Auxiliary Feedwater System (AF) - The AFW system shall automatically start and deliver adequate AFW flow to maintain adequate steam generator levels during accidents which may result in main steam safety valve opening, such as: Loss of normal feedwater (LONF) and Loss of all AC power to the station auxiliaries (LOAC). AFW system shall also deliver sufficient flow to the steam generators supporting rapid cooldown during such accidents as: steam generator tube rupture (SGTR) and main steam line break (MSLB).
Each pump has an AOV controlled recirculation line back to the condensate storage tanks to ensure minimum flow to prevent hydraulic instabilities and dissipate pump heat.
TS 3.7.5 Auxiliary Feedwater (AFW) System TS Bases B 3.7.5 Auxiliary Feedwater (AFW) System FSAR 7.3.3.4 Manual AFW Flow Control During Plant Shutdown Manual control of steam generator water level using the AF pumps to remove reactor decay and sensible heat.
FPER 6.6.4 Auxiliary Feedwater System The Auxiliary Feedwater Pumps are provided with a mini-recirc line to ensure a minimum amount of flow is established to keep the pumps from dead heading PBF-151Sc
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Venry SCR number on all pal Page FSAR 10.2 Auxiliary Feedwater System (AF)
TS 3.7.5 Auxiliary Feedwater (AFW) System TS Bases B 3.7.5 Auxiliary Feedwater (AFW) System FSAR 7.3.3.4 Manual AFW Flow Control During Plant Shutdown FPER 6.6.4 Auxiliary Feedwater System 1.3 Does the proposed activity involve a change to any Technical Specification? Changes to Technical Specifications require a License Amendment Request (Resource Manual Section 5.3.1.2).
Technical Specification Change: El Yes ED No If a Technical Specification change is required, explain what the change should be and why it is required.
1.4 Does the proposed activity involve a change to the terms, conditions or specifications incorporated in any VSC-24 cask Certificate of Compliance (CoC)? Changes to a VSC-24 cask Certificate of Compliance require a CoC amendment request.
[- Yes [3No 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 1 (50.59) - DETERMINE IF THE CHANGE INVOLVES A DESIGN FUNCTION (Resource Manual 5.3.2)
Compare the proposed activity to the relevant CLB descriptions, and answer the following questions:
YES NO QUESTION 0 El Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?
[E ED Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?
0] El Does the proposed activity involve SSCs whose failure could initiate a transient (e.g., reactor trip, loss of feedwater, etc.) or accident, OR whose failure could impact SSC(s) credited in the Safety Analyses?
0] El Does the proposed activity involve CLB-described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications?
E] 0 Does the activity involve a method of evaluation described in the FSAR?
El 0 Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)
El 0 Does the activity exceed or potentially affect a design basis limitfor afission product barrier(DBLFPB)?
(NOTE: If THIS questions is answered YES, a 10 CFR 50.59 Evaluation is required.)
If the answers to AL.L of these questions are Q, mark Part III as not applicable, document the 10 CFR 50.59 screening m the
- onclusion section (Part IV), then proceed directly to Part V - 10 CFR 72.48 Pre-screening Questions.
If any of the above questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.
PBF-1515c
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59172.48 SCREENING (NEW RULE) Venfy SCR number on all pagp Page 3 MR-02-029 upgraded the open function of the AFW pumps mini-recirc AOV to safety-related. The safety-related boundary includes the recirc orifice and all associated upstream components and piping. It is postulated that a failure of the piping downstream of the recirc orifice will not have any adverse affects on the AFW system. The availability of the recirculation flowpath provides an additional flowpath to support minimum flow requirements. This procedure change will improve the reliability of the AFW pumps by not relying upon the recirc flow path for operability as it has been concluded that the restrictions in the recirc orifice may not be adequate for use. Whereas current guidance mandates that the operator verify the position of the recirc AOV and the status of the Instrument Air system, these procedural changes will only require the operator to monitor pump discharge flow.
PART III (50.59) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (Resource Manual 53.3)
If ALL the questions in Part II are answered NO, then Part III is [] NOT APPLICABLE.
Answer the following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a 10 CFR 50.59 Evaluation is required; EXCEPT where noted in Part 111.3.
1I. I CHANGES TO THE FACILITY OR PROCEDURES YES NO QUESTION El 0 Does the activity adversely affect the designfunction of an SSC credited in safety analyses?
0l 0 Does the activity adversely affect the method of performing or controlling the designfunction of an SSC credited in the safety analyses?
If any answer is YES. a 10 CFR 50.59 Evaluation is required. If both answers are NO describe the basis for the conclusion (attach additional discussion as necessary):
Minimum flow requirements will be maintained within recommendations from the vendor by monitoring pump discharge flow and securing the pump as required. Starting and stopping of the AFW pumps has been previously evaluated in 50.59 Evaluation 2002-005, which addressed procedural changes to reduce the potential of pump damage as a result of the loss of the recirculation flow path.
111.2 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 El E] Does the activity use a revised or different method of evaluation for performing safety analyses than that described in the CLB?
El El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in the CLB?
If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion, as necessary).
111.3 TESTS OR EXPERIMENTS If the activity is not a test or experiment, the questions in III.3.a and III.3.b are 0 NOT APPLICABLE.
- a. Answer these two questions first:
YES NO QUESTION El E] Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?
[E] M Are the SSCs affected by the proposed test or experiment isolated from the facility?
PBF-1515c Rrerrnce" NP i I R
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Venfy SCR numberon all pages Page 4 If the answer to BOTH questions in V.3.a is NO, continue to III.3.b. If the answer to EITHER question is YES, then describe the basis.
- b. Answer these additional questions ONLY for tests or experiments which do NOT meet the criteria given in III.3.a above.
If the answer to either question in III.3.a is YES, then these three questions are El NOT APPLICABLE.
YES NO QUESTION 1] El Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the CLB?
El E] Does the activity utilize or control an SSC in a manner that is inconsistent with the analyses or descriptions in the CLB?
I] E] Does the activity place the facility in a condition not previously evaluated or that could affect the capability of an SSC to perform its intended functions?
If any answer in III.3.b is YES, a 10 CFR 50.59 Evaluation is required. If the answers in III.3.b are ALL NO, describe the basis for the conclusion (attach additional discussion as necessary):
Part IV - 10 CFR 50.59 SCREENING CONCLUSION (Resource Manual 5.3.4).
Check all that apply.
A 10 CFR 50.59 Evaluation is [] required or Z NOT required.
A Point Beach FSAR change is Ml required or Z NOT required. If an FSAR change is required, then initiate an FSAR Change Request (FCR) per NP 5.2.6.
A Regulatory Commitment (CLB Commitment Database) change is E] required or 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 [E required or Z NOT required. If a change to the Technical Specification Bases is required, then initiate a Technical Specification Bases change per NP 5.2.15.
A Technical Requirements Manual change is [E required or [0 NOT required. If a change to the Technical Requirements Manual is required, then initiate a Technical Requirements Manual change per NP 5.2.15.
10 CFR 72.48 SCREENING NOTE: NEI 96-07, Appendix B, Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 72.48 screenings.
PART V (72.48) - 10 CFR 72.48 INITIAL SCREENING QUESTIONS Part V determines if a full 10 CFR 72.48 screening is required to be completed (Parts VI and VII) for the proposed activity.
YES NO QUESTION El Z Does the proposed activity involve IN ANY MANNER the dry fuel storage cask(s), the cask transfer/transport equipment, any ISFSI facility SSC(s), or any ISFSI facility monitoring as follows: Multi-Assembly Sealed Basket (MSB), MSB Transfer Cask (MTC), MTC Lifting Yoke, Ventilated Concrete Cask (VCC), Ventilated Storage Cask (VSC), VSC Transporter (VCST), ISFSI Storage Pad Facility, ISFSI Storage Pad Data/Communication Links, or PPCS/ISFSI Continuous Temperature Monitoring System?
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 5 El 0] Does the proposed activity involve IN ANY MANNER SSC(s) installed in the plant specifically added to support cask loading/unloading activities, as follows: Cask Dewatering System (CDW), Cask Reflood System (CRF), or Hydrogen Monitoring System?
El 0] Does the proposed activity involve IN ANY MANNER SSC(s) needed for plant operation which are also used to support cask 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 (Drunmning Area Exhaust Low-Range Gas Monitor), PAB Crane, SFP Platform Bridge, Truck Access Area, or Decon Area?
El ED Does the proposed activity involve a change to Point Beach CLB design criteria for external events such as earthquakes, tornadoes, high winds, flooding, etc.?
E] 0 Does the activity involve plant heavy load requirements or procedures for areas of the plant used to support cask loading/unloading activities?
El 0] Does the activity involve any potential for fire or explosion where casks are loaded, unloaded, transported or stored?
If ANY of the Part V questions are answered YES, then a full 10 CFR 72.48 screening is required and answers to the questions in Part VI and Part VII are to be provided. If ALL the questions in Part V are answered NO then check Parts VI and VII as not applicable. Complete Part VIII to document the conclusion that no 10 CFR 72.48 evaluation is required.
PART VI (72.48) - DETERMINE IF THE CHANGE INVOLVES A ISFSI LICENSING BASIS DESIGNFUNCTION "If ALL the questions in Part V are NO then Part VI is [] NOT APPLICABLE.)
Compare the proposed activity to the relevant portions of the ISFSI licensing basis and answer the following questions:
YES NO QUESTION El 0] Does the proposed activity involve cask/ISFSI Safety Analyses or plant/cask/ISFSI structures, systems and components (SSCs) credited in the Safety Analyses?
El 0] Does the proposed activity involve plant, cask or ISFSI SSCs that support SSC(s) credited in the Safety Analyses?
El 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?
El 0] Does the proposed activity involve cask/ISFSI described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, CoC conditions, or orders?
El ED Does the activity involve a method of evaluation described in the ISFSI licensing basis?
El Z Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)
El 0* Does the activity exceed or potentially affect a cask design basis limitfor afisszon product barrier(DBLFPB)?
(NOTE: If THIS questions is answered YES, a 10 CFR 72.48 Evaluation is required.)
If the answers to ALL of these questions are NO, mark Parts VII as not applicable, and document the 10 CFR 72.48 screening in the conclusion section (Part VIII).
If any of the above questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.
PART VII (72.48) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (NEI 96-07, Appendix B, Section B.4.2.1)
(If ALL the questions in Part V or Part VI are answered NO, then Part VII is 0 NOT APPLICABLE.)
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 6 Answer the following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a 10 CFR 72.48 Evaluation is required; EXCEPT where noted in Part VII.3.
VII. 1 Changes to the Facility or Procedures YES NO QUESTION El E] Does the activity adversely affect the design function of a plant, cask, or ISFSI SSC credited in safety analyses?
El El Does the activity adversely affect the method of performing or controlling the designfunction of a plant, cask, or ISFSI SSC credited in the safety analyses?
If any answer is YES, a 10 CFR 72.48 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion, as necessary):
VII.2 Changes to a Method of Evaluation (If the activity does not involve a method of evaluation, these questions are El NOT APPLICABLE.)
YES NO QUESTION E3 El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in a cask SAR?
El El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in a cask SAR?
If any answer is YES, a 10 CFR 72.48 Evaluation is required. If both answers are NO. describe the basis for the conclusion (attach additional discussion, as necessary):
VII.3 Tests or Experiments (If the activity is not a test or experiment, the questions in VII.3.a and VII.3.b are E] NOT APPLICABLE.)
- a. Answer these two questions first:
YES NO QUESTION El El Is the proposed test or experiment bounded by other tests or experiments that are described in the cask ISFSI licensing basis?
El El Are the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISFSI facility?
If the answer to both questions is NO, continue to VII.3.b. If the answer to EITHER question is ES, then briefly describe the basis.
- b. Answer these additional questions ONLY for tests or experiments which do not meet the criteria given in VII.3.a above.
If the answer to either question in VII.3.a is YES, then these three questions are E] NOT APPLICABLE:
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 7 YES NO QUESTION El E] Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the ISFSI licensing basis?
[1 El Does the activity utilize or control a plant, cask or ISFSI facility SSC in a manner that is inconsistent with the analyses or descriptions in the ISFSI licensing basis?
El [I Does the activity place the cask or ISFSI facility in a condition not previously evaluated or that could affect the capability of a plant, cask, or ISFSI SSC to perform its intended functions?
If any answer in VII.3.b is YES, a 10 CFR 72.48 Evaluation is required. If the answers are all NO describe the basis for the conclusion (attach additional discussion as necessary):
PART VIII - DOCUMENT THE CONCLUSION OF THE 10 CFR 72.48 SCREENING Check all that apply:
A 10 CFR 72.48 Evaluation is El required or 0 NOT required. Obtain a screening number and provide the original to Records Management regardless of the conclusion of the 50.59 or 72.48 screening.
A VSC-24 cask Safety Analysis Report change is El required or [D NOT required. If a VSC-24 cask SAR change is required, then contact the Point Beach Dry Fuel Storage group supervisor.
A Regulatory Commitment (CLB Commitment Database) change is E] required or 0 NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.
A change to the VSC-24 10 CFR 72.212 Site Evaluation Report is El required or 0 NOT required. If a VSC-24 10 CFR 72.212 Site Evaluation Report change is required, then contact the Point Beach Dry Fuel Storage group supervisor.
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 1 of 40 A. PURPOSE
- 1. To provide the steps necessary to perform a natural circulation cooldown of the reactor plant using the plant equipment available subsequent to a fire.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. The following is the only entry condition for this procedure:
- a. Following completion of AOP 10A. SAFE SHUTDOWN LOCAL CONTROL when unable to initiate and attain Cold Shutdown from the control room.
C. REFERENCES
- 1. 10 CFR 50 Appendix R Fire Protection Program For Nuclear Power Facilities
- 2. OM 3.27. Control Of Fire Protection & Appendix R Safe Shutdown Equipment
- 3. CR 00-2654. PT Limits During Safe Shutdown Cooldown
- 4. INPO OE 11273. Appendix R and Appendix G Safe Shutdown Analysis
- 5. SE 97-120-01. Revisions to AOP 1OA. Safe Shutdown Local Control. and AOP lOB Safe To Cold Shutdown In Local Control CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 2 of 40
'IiI i I ISTEPI ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I
NOTE If at any time control room access and control of shutdown functions from the control room can be regained. this procedure can be exited once a shutdown strategy is developed.
Direct I&C To Perform.The Following:
- b. ICP 10.1. ESF System And AMSAC System Bypass 2 Ensure Completion Of The Following AOP-10A. SAFE SHUTDOWN - LOCAL CONTROL. Follow-up Actions:
- EPIPs implemented
- Pressurizer level
- Auxiliary feedwater established to "B" steam generator
- Service water valve alignment
- Hydrogen purged from main generator e Battery loads minimized 3 Check Minimum Safe Shutdown Equipment Available
- a. Check both 480 volt safeguards a. Go to Attachment A, COOLDOWN buses free from fire damage USING G-05 AS POWER SOURCE.
"*IB-03
"*IB-04
- b. Turbine-driven auxiliary b. Align motor-driven auxiliary feedwater pump supplying "B" feedwater pump to supply "B" steam generator steam generator.
a IP-29 o P-38A o P-38B CONTINUOUS USE
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL .Page 3 of 40 iSTEP I ACTION/EXPECTED RESPONSE I i I RESPONSE NOT OBTAINED -I-4 Check Both 480 Volt Safeguards Buses Go to Ste* 6.
Energized
"*1B-03
"*IB-04 5 Go To Step 11 6 Reenergize 480 V Safeguards Buses: Go to Attachment A. COOLDOWN USING G-05 AS POWER SOURCE.
- a. Open all breakers on 1B-03 and IB-04
- b. Close G-03 DC control power switch D72-28-01
- c. Start G-03 at C81:
- 1) Place G-03 in - LOCAL
- 2) Depress - SHUTDOWN RESET
- 3) Depress - FAST START
- d. Close G-03 output breaker on 1A-06 1A52-80 1
- f. Close 1B-04 to IX-14 breaker on IB-04
- IB52-17B
- g. Close IB-04 to 1B-03 tie breaker on IB-03 o IB52-16C CONTINUOUS USE
POINT BEACH NUCLEAR PLANT POINT LEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 4 of 40 I STEP I ACTION/EXPECTED RESPONSE I
I T RESPONSE NOT OBTAINED I
7 Establish Normal Power Supply To Charging Pump
- a. Check at least one charging pump a. Locally close 1P-2A or 1P-2C powered from normal power supply charging pump breaker on IB-03 or 1B-04.
"o 1P-2A "o IP-2B o 1B52-13A "o IP-2C OR o IB52-20A
- d. At 1B313B-2B337B perform the following:
- 1) Close breaker Bi
- 2) Open breaker B3
- 3) Open breaker B2
- 4) Close breaker B4
- e. At 2B337B-1B313B perform the following:
- 1) Close breaker B8
- 2) Open breaker B6
- 3) Open breaker B7
- 4) Close breaker B5 CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 5 of 40 STEPiI ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED Iý 8 In Pipeway #2 Isolate Component Cooling Water To RCPs
- a. Shut component cooling supply to containment 1CC-719 1
- b. Shut RCP return valves
"*ICC-759A
"*ICC-759B 9 In Pipeway #2 Entry. Align Component Cooling For Restoration
- a. Shut RHR heat exchanger component cooling supply valves
"*1CC-738A
"*1CC-738B
- b. Shut boric acid evaporator isolation
- c. Shut non-regenerative heat exchanger isolation
- d. Shut seal return heat exchanger isolation
- e. Shut CCW Emergency Make-Up a ICC-815 CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORYAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 6 of 40 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I 10 Restore Component Cooling:
- a. Start one component cooling water a. Perform the following:
pump by local breaker closure on IB-03 or IB-04 1) Cross-connect Unit 1 and Unit 2 component cooling.
"o 1B52-10A a) Ensure at least one Unit 2 OR component cooling water pump running.
"o IB52-23B b) Open the following valves:
"*Suction cross-connect a CC-722A
"*HX-12C Unit 1 inlet 1CC-726C
"* EX-12C Unit 2 inlet
"*EX-12BIC inlet
- 2) Request TSC evaluate possible system repairs.
11i Ensure At Least Two Service Water Go to Attachment A. COOLDOWN USING Pumps Running G-05 AS POWER SOURCE.
"o P-32A "o P-32B "o P-32C "o P-32D "o P-32E "oP-32F CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORM~AL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 7 of 40 STEP IACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINEDL__J 12 Restore Power To Safeguards MCCs
- b. Open all breakers on MCC IB-42
- d. On IB-04. close 1B-42 feeder a 1B52-23C 13 Restore Emergency Lighting And Gai-Tronics
- a. On 1B-42. close XL-10 breaker
- b. On 2B-32. close XL-20 breaker
- 14 Maintain Component Cooling Heat *
- Exchanger Outlet Temperature Between *
- 70*F And 125*F *
- ITI-621D
- 15 Maintain Surge Tank Level Between *
- 25% And 75% *
- ILI-618
- CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 8 of 40 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 16 Maintain RCS Conditions:
- a. Check pressurizer level - GREATER a. Locally operate charging pumps as *
- THAN 20% necessary to maintain level - *
- GREATER THAN 20%.
- 1LI-426-1 on INI0
- b. Check RCS subcooling greater than b. IF pressurizer heaters available
- 351f per FIGURE 2. NATURAL THEN energize heaters to maintain
- CIRCULATION COOLDOWN LIMITS subcooling - GREATER THAN 350 F *
- lPI-420C-1 I on IN11 o IB52-20A 17 Consult TSC To Restore Desirable Electrical Loads Within Capacity Of Power Source
- a. Ventilation Fans
"*Auxiliary feed pump room. 4kV switchgear room, and battery room ventilation
"* W-46
"* W-46A
"* W-1OA
"* W-1OB
"*PAB battery room or inverter room ventilation
"*Cable spreading room
"*Control room ventilation
"*Computer room ventilation
- b. EDG Non-vital loads on B-40
- c. Security battery chargers CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Revision 3 5/3/2001 Page 9 of 40 STEP ACTION/EXPECTED RESPONS RESPONSE NOT OBTAINED CAUTION RCS pressure should be allowed to drift down as temperature is reduced. No attempt to manually control RCS pressure is required at this time.
18 Initiate Natural Circulation I
Cooldown To 5001F
- b. Dump steam from "B" steam generator 1MS-2015 l
d.
- c. Maintain "B" steam generator level within range of FIGURE 1.
STEAM GENERATOR LEVEL
- ILI-470A-1 on 1C-205
- d. Maintain RCS hot leg and cold leg temperature within acceptable region of FIGURE 2. NATURAL CIRCULATION COOLDOWN LIMITS 19 Check RCS Hot Leg Temperature Return to Step 18.
BETWEEN 500OF AND 510OF e 1TI-451B-1 on 1C-205 CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit I ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 9 of 40 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED CAUTION RCS pressure should be allowed to drift down as temperature is reduced. No attempt to manually control RCS pressure is required at this time.
18 Initiate Natural Circulation Cooldown To 5000F
- a. Maintain cooldown rate in RCS cold leg - LESS THAN 25 OF/HR
- lTI-451B-I on 1C-205
- b. Dump steam from "B" steam generator lMS-2015 I
- c. Maintain "B" steam generator level within range of FIGURE 1.
STEAM GENERATOR LEVEL 7 61 e 1LI-470A-1 on 1C-205
- d. Maintain RCS hot leg and cold leg temperature within acceptable region of FIGURE 2. NATURAL CIRCULATION COOLDOWN LIMITS 19 Check RCS Hot Leg Temperature Return to Step 18.
BETWEEN 500OF AND 510OF
- ITI-451B-1 on 1C-205 CONTINUOUSITg T
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP OB Unit I ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 10 of 40 STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE The follouiing steps may require a containmer nt entry. Coordinate entry with-the TSC and ensure the guidelines of EPIP-10.1. Emergency Reentry are followed.
20 Depressurize RCS To 1400 PSIG:
- a. Shut normal charging isolation
- 2) Activate close relay inside 2) Locally shut valve in breaker cubicle for 1CV-1298 Regenerative heat exchanger room.
a 1CV-1298
- b. Stop all but one charging pump
- c. Open charging pump flow control valve bypass
- ICV-323B
- d. Throttle shut auxiliary charging isolation to reduce RCS pressure
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 11 of 40 I STEP. I ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED
-F-CAUTIONS
- RCS cooldown must be carefully coordinated to ensure subcooling and PTS limits, as shown by FIGURE 2. NATURAL CIRCULATION COOLDOWN LIMITS. are not violated.
- RCS pressure will drift down as temperature is reduced. It may not be necessary to manually reduce pressure in the following step.
22 Commence Natural Circulation Cooldown To 400°F
- a. Maintain cooldown rate - LESS THAN 25 aF/HR
° ITI-451C-l on 1C-205
- b. Maintain "B" steam generator level within the range of FIGURE
- 1. STEAM GENERATOR LEVEL o ILI-470A-1 on IC-205
- c. Dump steam from "B" steam generator e IMS-2015
- d. Throttle auxiliary charging d. Perform the following:
isolation as necessary to maintain RCS pressure - BETWEEN o Locally operate pressurizer 1000 PSIG AND 1400 PSIG heaters.
@ ICV-323A OR o Allow pressurizer to go solid and operate charging pumps as necessary.
23 Check B Loop RCS Cold Leg Return to Step 22.
Temperature - LESS THAN 400°F
- ITI-451C-I (IC-205)
CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 12 of 40 STEP fl ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED
' I 24 Maintain RCS Stable For At Least 30 Hours
- a. Maintain Thot - BETWEEN 390°F AND 400°F a 1TI-451B-1
- b. Check RCS pressure - BETWEEN 1000 b. Perform the following:
PSIG AND 1400 PSIG o Locally operate pressurizer
- IPI-420C-1 heaters.
OR o Allow pressurizer to go solid and operate charging pumps as necessary.
- 25 Slowly Raise Steam, Generator Level IF desired AFW flow can NOT be
- -BETWEEN 500 INCHES AND 510 INCHES obtained. THEN perform the 4
- following:
- 1LI-470A-1 (1C-205) *
- a. Reduce turbine-driven AFW pump
- recirc flow.
- a IAF-4002 *
- b. Monitor turbine-driven AFW pump *
- temperature to ensure adequate *
- flow for cooling.
- r'A1JrTVTUT1nTTC ITCV
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE AOP-10B Unit 1 SAFETY RELATED SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Revision 3 5/3/2001 Page 13 of 40 LSI I ACTION/EXPECTED RESPONSE
!J RESPONSE NOT OBTAINED 26 Align One RHR Train For Operation:
"*ISI-851A
"*ISI-851B
- b. Open RHR pump suction valves
"*IRH-704A
"*lRE-704B
- c. Open RBR pump discharge valves
"* 1RH-709A
"*1RHl7O9B
- d. Open RER inlet cross-connect valves
- f. Open RER heat exchanger outlet cross-connect valves
- IRH-716C
- IRH-716D
- g. Perform the following: g. Perform the following:
- 1) Throttle RHR heat exchanger a) Throttle RHR heat exchanger discharge valve to 1/4 turn discharge valve to 1/4 turn OPEN OPEN
- IRH-716A
- IRH-716B
- IRH-716B e IRH-716A
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 14 of 40 I
SITEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I NOTE The following steps may require a containment entry. Coordinate entry with the TSC and ensure the guidelines of EPIP-10.1. Emergency Reentry are followed.
27 Depressurize RCS To 375 PSIG
- a. Check pressurizer level - LESS a. Go to Step 29.
THAN 90%
e ILI-426-1 CINI1)
- b. Check pressurizer level - GREATER b. Operate charging pump to THAN 20% establish pressurizer level GREATER THAN 20%.
lLI-426-1 I.
- c. Stop all but one charging pump
- d. Open charging pump flow control valve bypass
- f. Check RCS pressure LESS THAN f. Return to Step 27.e 375 PSIG 1PI-420C-1 1
28 Verify Pressurizer Solid: Perform the following
- Pressurizer level - GREATER a. WHEN pressurizer level greater THAN 90% than 90%. THEN go to Step 29.
ILI-426-1 I (lN11) b. Continue depressurization per Step 27.
29 Secure Charging And Float RCS Pressure On SI Accumulators CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 15 of 40 STEP1I ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I
- 30 Monitor RCS Pressure And Level
- a 1LI-426-1 ,
31 Check RCS Pressure - LESS THAN Perform the following:
375 PSIG
32 Reestablish Charging Flow:
- b. Open auxiliary charging isolation a 1CV-323A
- c. Shut charging flow control valve bypass 1CV-323B I
- 33 Maintain RCS pressure - BETWEEN ,
- IPI-420C-1 CONTINUOUS USE
NUCLEAR PLANT POINT BEACH POINT BEACH 1NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 16 of 40 ISTEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 34 Align Component Cooling System For RHR
- a. Align component cooling for RHR train aligned in Step 26
- b. Adjust RHR heat exchanger component cooling water outlet throttle valves to approximately 30 degrees open
"*1CC-824A
"*1CC-824B
- d. Check component cooling water d. Perform the following:
pump discharge pressure - GREATER THAN 85 PSIG 1) IF within the capacity of the power source THEN locally
"* 1B52-10A
"* IB52-23B
- 2) IF an additional component cooling water pump can NOT be started. THEN perform the followng:
"o Secure additional component cooling water.loads.
OR "o Throttle 1CC-824A and ICC-824B CrnwT T1,TJ1 TT1C TTQ 'V
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-1OB Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 17 of 40 ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I 35 Cooldown To Less Than 350*F
- a. Ensure I&C has calibrated and insialled temperature indicators in RHR inlet temperature RTD well a 1TE-630
- b. Maintain cooldown rate less than 25OF/HR
- c. Cooldown using atmospheric steam dump
- IMS-2015
- d. Maintain steam generator level. d. Throttle shut turbine driven AFW BETWEEN 500 INCHES AND 510 INCHES mini-recirc valve as necessary to obtain desired flow.
e ILI-470A-1 (IC-205) 1AF-4002 I
- e. Maintain cooldown rate between 200F/HR and 250F/HR when RCS temperature is between 375 0 F and 350°F CONTINUOUS USE
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORIAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 18 of 40 III i I I STEPl IIACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I I
CAUTION The following steps may require a containment entry. Coordinate entry with the TSC and ensure the guidelines of EPIP-10.1. Emergency Reentry. are followed.
36 Establish RHR:
- a. Check RCS temperature between a. WHEN temperature is between 340OF 340OF and 350 0 F and 350*F. THEN continue with*
Step 36.b.
- b. Check RCS pressure - LESS THAN
- c. Pressurize RHR system
- 1) Close the following breakers:
"*IB52-324M
"*1B52-424M
"*1B52-325M
- 2) Activate open relay inside 2) Locally open valve.
breaker cubicle for IRH-700 IRH-700
- 3) Activate open relay inside 3) Locally open valve.
breaker cubicle for IRH-701 IRH-701 I
- 4) Activate open relay inside 4) Locally open valve.
breaker cubicle for IRH-720
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY-RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 19 of 40
[STEP ACTION/EXPECTED RESPONSE IRESPONSE NOT OBTAINED Step 36. (continued from previous page)
- d. Isolate accumulators from RCS: d. Manually close accumulator outlets in containment.
- 1) Close breakers for accumulator outlets: "*ISI-841A
"*1SI-841B
"*1B52-324F (SI-841A)
"*Activate close relay inside breaker cubicle
- 2) Activate close relay inside breaker cubicle
"*ISI-841A
"* ISI-841B
- e. Check Step 35".c complete e. DO NOT CONTINUE until RER system has been pressurized.
- f. Locally close one RHR pump
- f. Close one RFR pump alternate breaker supply breaker.
o IB52-12A e B52-55B (B-08)
OR o IB52-21A
- g. Ensure one RUR pump running g. IF one RBR pump can NOT be started. THE* perform the o IP-10A following:
OR 1) Contact maintenance for repairs.
o IP-10B
- 2) WHEN repairs complete. THEN start one RHR pump and complete Step 36.h.
- h. Operate RRR pump on recirc for at least 5 minutes
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 20 of 40 STEP I ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 37 Check Component Cooling Water Heat Adjust service water flow to Exchanger Outlet Temperature Less establish - LESS THAN 1250F.
Than 125*F 1TI-621D NOTE When on RHR. loop temperatures will not be indicative of actual core temperature.
38 Commence Cooldown To Less Than 200OF
- a. Maintain cooldown rate less than 25 OF/hr as determined from RHR inlet temperature
- ITE-630
- c. Check RHR temperatures trending c. Return to Step 38.b lower a ITE-630 39 Check RHR Inlet Temperature Less Return to Step 38.
Than 200'F e 1TE-630 40 Secure RCS Cooldown
- a. Throttle RHR HX outlet valve for aligned train to stabilize temperature o 1RH-716A OR o IRH-716B rA'k7rrT1KT77nTVZ 114ZV
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-1OB Unit I ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 21 of 40 ISTEP fl ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED I 41 Check Pressurizer Level - GREATER WHEN pressurizer level greater than THAN 90% 90%. THEN stop all charging pumps.
a 1LI-426-1 42 Float RWST On RCS
- b. Ensure RCS pressure between 10 psig and 30 psig lPI-420-1 I
- 43 Maintain Stable Plant Conditions ,
- Until A Recovery Plan Is Established ,
- a. RHR inlet temperature - LESS *
- THAN 200FF
- AND30 PSIG ,
- c. Pressurizer solid
-END-
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 22 of 40
[STP ACTION/EXPECTED RESPONSE I SPON E NOT OBTAINED ATTACHMENT A (Page 1 of 14)
COOLDOWN USING G-05 AS POWER SOURCE Al Ensure Minimum Safe Shutdown Equipment Available:
- a. G-05 gas turbine supplying a. Return to Main Body, Step 2.
alternate shutdown bus
"*B-08
"*B-09
- b. Turbine-driven auxiliary feedwater pump supplying "BN steam generator
- 1P129
- c. One charging pump providing RCS makeup
- d. RBR pump alternate power available
- 01-112. ALIGNING\EQUIPMENT TO APPENDIX R POWER SUPPLY A2 Ensure At Least Two Service Water Return to Main Body, Step 2.
Pumps Running "oP-32B "oP-32C "oP-32E "oP-32F
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 23 of 40 I -- I I SACTION/EXPECTED RESPONSE I I RESPON 3E NOT OBTAINED I
ATTACHMENT A (Page 2 of 14)
COOLDOWN USING G-05 AS POWER SOURCE A3 Perform Tagout To Allow Maintenance To Install Alternate Shutdown Cables
- a. Battery and inverter room fan
- 1) Select fan to be operated and tag open normal supply breaker o W-85. breaker 2B52-3213C OR o W-86. breaker 2B52-426J
- 2) Tag open alternate power supply breaker on B81
- B52-811K
- b. Component cooling water pump
.1) Select pump to be operated and tag normal supply breaker RACKED OUT o IP-liA, IB52-1OA OR o.lP-11B. 1B52-23B
- 2) Tag alternate power disconnect switch - OPEN a B855C A4 Contact Maintenance To Connect Alternate Power Cabling Per Attachment B. INSTALLATION OF Attachment B. INSTALLATION OF i .....
ALTERNATE POWER CABLES f* NW7 LrT1 /*T7TnT T0 T 7TOV
POINT BEACH NUCLEAR PLANT AOP-1OB Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 24 of 40 I II - - - I I LST II ACTION/EXPECTED RESPONSE I I E
RESPONSE NOT OBTAINED II ATTACHmENT A (Page 3 of 14)
.COOLDOWN USING G-05 AS POWER SOURCE A5 Check Seal Injection Supplied By Locally isolate component cooling Charging to RCPs.
- a. Shut component cooling supply to containment
- 1CC-719
- b. Shut RCP seal return valves ICC-759A 1
ICC-759B I
A6 Align Component Cooling For Restoration
- a. In pipeway #2 shut RER heat exchanger component cooling supply valves
"*1CC-738A
"*1CC-738B
- b. In Pipeway #2 entry shut boric acid evaporator, isolation e ICC-744B
- c. In Pipeway #2 entry shut non-regenerative heat exchanger isolation e 1CC-740B
- d. In Pipeway #2 entry shut seal return heat exchanger isolation
- ICC-750B
('IhT V77TTMIMTT TTLIV
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/312001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 25 of 40 ISTEP ACTION/EXPECTED RESPONSE I I I
RESPONSE NOT OBTAINED ATTACHMENT A (Page 4 of 14)
COOLDOWN USING G-05 AS POWER SOURCE A7 Restore Component Cooling
- a. Check maintenance has completed a. Cross-connect Unit 1 and Unit 2 installing alternate shutdown component cooling:
cables to Unit 1-component cooling water pump 1) Check at least one Unit 2 component cooling water pump running.
- 2) Open the following valves:
"*Suction cross-connect e CC-722A
"*HX-12C Unit 1 inlet e 1CC-726C
"* HX-12C Unit 2 inlet
"*BX-12B/C inlet
- b. Start one component cooling pump b. Perform the following:
from C-45
- 1) Request TSC evaluate possible system repairs
- 2) WHEN component cooling has been established. THEN perform Steps A8 and A9.
- 3) Go to Step A1O.
- A8 Check Component Cooling Heat Locally adjust service water outlet *
- Exchanger Outlet Temperature Between valves. *
- 70°F And 125*F *
- rONTT T1*101T1F1E
POINT BEACH NUCLEAR PLANT AOP-1OB Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 26 of 40 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT A (Page 5 of 14)
COOLDOWN USING G-05 AS POWER SOURCE
- A9 Maintain Component Cooling Surge *
- Tank Level Between 25% And 75% *
- ILI-618
- NOTE Solid plant conditions may be required to maintain RCS subcooling. Subcooling limits are provided by FIGURE 2. NATURAL CIRCULATION COOLDOWN LIMITS
- AIO Maintain RCS Conditions Locally operate charging as required
- to maintain RCS parameters..
- a. Pressurizer level - GREATER *
- THAN 20%
- LI-426-1 (IN11)
- b. Subcooling- GREATER THAN 35*F
- - TI-451B-1 (IC-205) *
- PI-420C-1 (INIl) p,.i rfnMi" TN lnl'l 11,"-1
POINT BEACH NUCLEAR PLANT AOP-10B Unit I ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 27 of 40 ISP ACTION/EXPECTED RESPONSE i1 RESPONSE NOT OBTAINED ATTACHMENT A (Page 6 of 14)
COOLDOWN USING G-05 AS POWER SOURCE CAUTION RCS pressure should be allowed to drift down as temperature is reduced. No attempt to manually control RCS pressure is required at this time.
All Initiate Natural Circulation Cooldown To 500'F
- a. Maintain cooldown rate in RCS cold leg - LESS THAN 25 *F/HR
- 1TI-451B-1 (lC-2O5)
- b. Dump steam from "B" steam generator IMS-2015
- c. Maintain "B" steam generator level within the range of FIGURE
- 1. STEAM GENERATOR LEVEL
- 1LI-470-1 (lc-205)
- d. Maintain RCS hot leg and cold leg temperature within acceptable region of FIGURE 2. NATURAL CIRCULATION COOLDOWN LIMITS A12 Check RCS Cold Leg Temperature Return to Step All.
BETWEEN 500OF AND 510°F
- TI-451C-l (IC-205) r('JTTUTTfT1T TTP
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 28 of 40 I STP I ACTION/EXPE4CTED RESPONSE I I RESPON3E NOT OBTAINED -I ATTACHMENT A I
(Page 7 of 14)
COOLDOWN USING G-05 AS POWER SOURCE NOTE The following steps may require a containment entry. Coordinate entry with the TSC and ensure the guidelines of EPIP-10.1. Emergency Reentry. are followed.
A13 Depressurize RCS To 1400 PSIG
- a. Shut normal charging isolation
- 1) At IB-42 shut breaker for 1CV-1298 normal charging isolation
- 2) Activate close relay inside 2) Locally shut valve in breaker cubicle for 1CV-1298 regenerative heat exchanger room.
ICV-1298 I
- b. Stop all but one charging pump
- c. Open charging pump flow control valve bypass e 1CV-323B
- d. Throttle shut auxiliary charging isolation to reduce RCS pressure
- ICV-323A
("3),TTT1ATTT TT*Z
POINT BEACH NUCLEAR PLANT AOP-1OB Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Revision 3 5/3/2001 Page 29 of 40 T ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT A (Page 8 of 14)
COOLDOWN USING G-05 AS POWER SOURCE CAUTIONS
"*RCS cooldown must be carefully coordinated to ensure subcooling and PTS limits, as shown by FIGURE 2. NATURAL CIRCULATION COOLDOWN LIMITS, are not violated.
"*RCS pressure will drift down as temperature is reduced. It may not be necessary to manually reduce pressure in the following step.
A15 Commence Natural Circulation Cooldown To 4000F:
- a. Maintain cooldown rate - LESS THAN 25 OF/HR
- TI-451C-1 on IC-205
- b. Maintain "BE steam generator level within the range of FIGURE
- c. Dump steam from "B" steam generator 1MS-2015 l
- d. Throttle auxiliary charging isolation to maintain RCS pressure - BETWEEN 1000 PSIG AND 1400 PSIG e 1CV-323A r'n TT 1TTT AT 1(TC' TTC-'t
POINT BEACH NUCLEAR PLANT AOP-lOB Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 30 of 40 ISTP II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT I A (Page 9 of 14)
COOLDOWN USING G-05 AS POWER SOURCE A16 Maintain RCS Stable For At Least 30 Hours
- a. Maintain Thot - BETWEEN 3900F AND 4000 F e ITI-451B-1
- b. Check RCS pressure - BETWEEN b. Perform the following:
1000 PSIG AND 1400 PSIG o Energize pressurizer heaters.
e IPI-420C-1 OR o Allow pressurizer to go solid and operate charging pumps as necessary.
- A17 Slowly Raise Steam Generator Level - IF desired AFW flow can NOT be
- BETWEEN 500 INCHES AND 510 INCHES obtained. THEN perform the
- . following:
o 1LI-470A-1 *
- a. Reduce turbine-driven AFW pump
- recirc flow
, IAF-4002 *
"w
- b. Monitor turbine-driven AFW pump
- temperature to ensure adequate *
- flow for cooling.
- 1^1- IT-
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE AOP-10B Unit 1 SAFETY RELATED SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Revision 3 5/3/2001 Page 31 of 40 STEP ACTION/EXPECTED RESPONSE J RESPONSE NOT OBTATNE ATTACHMENT A (Page 10 of 14)
COOLDOWN USING G-05 AS POWER SOURCE A18 Locally Align One RHR Train For Operation:
"*1SI-851A
"* ISI-851B
- b. Open RHR pump suction valves
"* lRB-704A
"* IRH-704B
- c. Open RHR pump discharge valves
- IRH-709A a IRH-709B
- d. Open RHR inlet cross-connect valves
- IRH-713B
- e. Open RflR heat exchanger inlet valves
"*1RH-715A
"*IRH-715B
- f. Open RUR heat exchanger outlet cross-connect valves
"*1RH-716C
"*1RH-716D
- g. Perform the following: g. Perform the following
- 1) Throttle RHR heat exchanger a) Throttle RHR heat exchanger discharge valve to 1/4 turn discharge valve to 1/4 turn OPEN OPEN e 1RH-716A - 1RH-716B
- 2) Close RPR heat exchanger b) Close RHR heat exchanger discharge valve,.... discharge valve.
- ITRW-71R
- 1RH-716B - In,1IDA
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 32 of 40
[STEP II ACTION/EXPECTED RESPONSE I I l RESPONSE NOT OBTAINED 1 g ATTACHMENT A (Page 11 of 14)
COOLDOWN USING G-05 AS POWER SOURCE NOTE The following steps may require a containment entry. Coordinate entry with the TSC and ensure the guidelines of EPIP-IO.1. Emergency Reentry are followed.
A19 Depressurize RCS To 375 PSIG
- a. Check pressurizer level - LESS a. Go to Step A21.
THAN 90%
- ILI-426-1 (INI1)
- b. Check pressurizer level - GREATER b. Operate charging pump to THAN 20% establish pressurizer level GREATER THAN 20%.
lLI-426-1 1
- c. Stop all but one charging pump
- d. Open charging pump flow control valve bypass 1lCV-323B
- e. Throttle shut auxiliary charging isolation to reduce RCS pressure
A20 Check Pressurizer Solid Perform the following:
Pressurizer level - GREATER a. WHEN pressurizer level is greater THAN 90% than 90%. THEN go to Step A21.
lLI-426-1 1 b. Continue depressurization per Step A20.
A21 Secure Charging And Float RCS Pressure On SI Accumulator rrM'rT'kTT1r%1TC T7CV
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 33 of 40 II ACTION/EXPECTED RESPONSE IRESPONSE ATTACHMENT A NOT OBTAINED 1I (Page 12 of 14)
COOLDOWN USING G-05 AS POWER SOURCE A22 Check RCS Pressure - LESS THAN Perform the following:
375 PSIG
A23 Reestablish Charging Flow
- a. Start one charging pump
- b. Open auxiliary charging isolation
- ICV-323A
- c. Shut charging flow control valve bypass e ICV-323B
- d. Maintain RCS pressure - BETWEEN 350 PSIG AND 375 PSIG e 2PI-420C-1
\
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 34 of 40 LSEJI ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT A (Page 13 of 14)
COOLDOWN USING G-05 AS POWER SOURCE A24 Align Component Cooling For RHR:
- a. Align component cooling for the RUR train aligned in Step Ai8
- b. Adjust both RHR heat exchanger component cooling water outlet throttle valves to approximately 30 degrees open 9 1CC-824A
- c.Locally open one RER heat exchanger component cooling water supply valve "o 1CC-738A OR "o 1CC-738B
- d. Check component cooling water d. Raise discharge pressure:
pump discharge pressure - GREATER THAN 85 PSIG o Secure additional component cooling water loads
"* 1PI-617A
"* IPI-617B OR o Throttle both of the following:
"* CC-824A
"* CC-824B CONTINUOUS USE
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 35 of 40 STE ACTION/EXPECTED RESPON4SEI I RESPONSE NOT OBTAINED ATTACHMENT A (Page 14 of 14)
COOLDOWN USING G-05 AS POWER SOURCE A25 Cooldown To - LESS THAN 350°F
- b. Maintain cooldown rate - LESS THAN 250F/HR
- c. Cooldown using one of the following methods:
o "B" SIG Atmospheric (IMS-2015) o Aux. Feed (UP-29)
- d. Maintain steam generator level d. Throttle shut mini recirc valve BETWEEN 500 AND 510 INCHES as necessary to maintain steam generator level - GREATER THAN 2LI-470A-1 500 INCHES.
- e. Maintain cooldown rate between 20*F/HR and 25*F/HR when RCS temperature is between 375OF and 350°F A26 Return To Main Body. Step 36
-END-
.l ~rI r* %1 J% l*TTd"
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORKAL OPERATING PROCEDURE SAFETY RELATED SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Revision 3 5/3/2001 Page 36 of 40 I
[STEP f ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT B (Page I of 3)
INSTALLATION OF ALTERNATE POWER CABLES 81 Ensure Required Danger Tagging Complete In Accordance With Attachment A. COOLDOWN USING G-05 AS POWER SOURCE B2 Select Battery Room Fan o W-85 "OR o W-86 B3 Move Designated Alternate Shutdown Cables From #3 Warehouse To Alternate Shutdown Switchgear Room NOTES
- Penetration M-2005-6-19-N6 is in the north wall above the yellow inverter room. and penetrates into the alternate shutdown switchgear room about 12 feet above MCC BS1
- Fire zone 217 is the area of the battery room fans.
B4 Direct Electrical Maintenance To Route Alternate Shutdown Cables Through Penetration H-2005-6-19-N6 Above MCC B8l. Into Fire Zone 217 To Selected Battery Room Fan "oW-85 OR "o W-86 NOTES "aSave all hardware when disconnecting cables
"*The slow speed winding leads are labeled Ti. T2. and T3.
85 Disconnect Normal Power Cables For Slow Speed Windings Of Selected Fan Motor B6 Connect Alternate Shutdown Cables To Slow Speed Winding Leads Labeled Ti, T2, And T3 Of Selected Fan Using Hardware Removed In Previous Sten rA"3TT13TT0TTT TTCq.
POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Revision 3 5/3/2001 Page 37 of 40 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT B (Page 2 of 3)
INSTALLATION OF ALTERNATE POWER CABLES 87 Connect Other End Of Alternate Shutdown Cable To Load Side Of Breaker B52-811K In MCC-B81 NOTE W-85 rotates clockwise as viewed from the drive pulley.
W-86 rotates counterclockwise as viewed from the drive pulley.
88 Test Fan Rotation:
I
- a. While observing'fan. close then open breaker B52-811K
- b. Check rotation correct
- c. IF fan rotation is correct. THE go to step BlO.
89 Change Fan Rotation:
- a. Direct operations to danger tag open breaker B52-811K
- c. Remove danger tag from B52-811K and close breaker B10 Determine Component Cooling Pump To Be Run "o IP-1iA OR "o IP-liB 811 Move Alternate Power Supply Cables B55CD And B55CE From Warehouse #3 To Area Of Disconnect Switch B855C 812 Direct Electrical Maintenance To Route Alternate Power Supply Cables B55CD And B55CE Between Disconnect Switch And Selected Component Cooling Water Pump o IP-liA OR o IP-11B rM"TIMMtITO IT~V
POINT BEACH NUCLEAR PLANT AOP-10B Unit I ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 38 of 40 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I ATTACHMENT B (Page 3 of 3)
INSTALLATION OF ALTERNATE POWER CABLES NOTE Save all hardware when disconnecting cables.
B13 Disconnect Normal Power Supply Cable From Selected Component Cooling Water Pump 814 Connect Alternate Power Supply Cables To Selected Component Cooling Water Pump B15 Connect Opposite End Of Alternate Power Supply Cable To Load Side Of Disconnect Switch B855C B16 Remove Danger Tag And Close Disconnect Switch B855C B17 Test Rotation Of Component Cooling Pump:
- a. While observing pump. close then open breaker B52-55C
- b. Check rotation correct
- c. IF pump rotation is correct. THEN go to Step Big.
B18 Change Pump Rotation:
- a. Danger tag open disconnect switch B855C
- b. Switch any two power leads at disconnect switch B855C
- c. Remove danger tag from disconnect switch B855C.
B19 Notify DSS Alternate Power Supply Cables Installed
-END-r( ,,TTrP T TTTf C TTQ '
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT ABNORMAt OPERATING PROCEDURE AOP-10B Unit 1 SAFETY RELATED SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Revision 3 5/3/2001 Page 39 of 40 FIGURE 1 STEAM GENERATOR LEVEL
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10B Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 3 5/3/2001 SAFE TO COLD SHUTDOWN IN LOCAL CONTROL Page 40 of 40 FIGURE 2 NATURAL CIRCULATION COOLDOWN LIMITS 0
0 0
0 01 w
In 4
0 a
.j.
0L CL E 40 13 0.
E 1!
0o I0 Ul) an CL 0
0D 04 (6!sd) SSBCd CONTINUOUS USE