ML030800237
| ML030800237 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/04/2002 |
| From: | Dvzak D, Groehler R, Sokol K Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| AOP 10, Rev 0, FOIA/PA-2003-0094 PBF-0026c, Rev 13 | |
| Download: ML030800237 (31) | |
Text
Nuclear Power Business Unit TEMPORARY CHANGE REVIEW AND APPROVAL Note: Refer to NP 1.2.3, Temporary Procedure Changes, for requirement. ~L I-4-002 I - INITIATION DocNumber AOP 10 Current Rev 0
1 b*
Document Title Control Room Inaccessibility Existing Effective Temporary Changes 2002-0764 Brief Description Add caution to Attachments A and B to address Al (Identify specific changes on Form PBF-0026c, Document Review and Approval Continuatio 0D Initiate PBF-0026h and include with the change.
Other documents required to be effective concurrently with the temporaryi Changes pre-screened according to NP 5.1.8?
0 NO El YES (fvi, Screening completed according to NP 5.1.8?
El NA 0 YES (nacd Safety Evaluation Required?
10 NO [I YES (if n a. reidon my t. pr=ocese f
Determine if the change constitutes a Change Of Intent to the procedure by (If any answers arc YES. a revision my be processed or final reviews and approvals shall beI Will the proposed change:
I. Require a change to, affect or invalidate a requirement, commitme description in the Current or ISFSI Licensing Basis (as defined in
- 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 0
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 El 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?
11
- 6. Require a change to the procedure Purpose or change the procedure classificat jon?.l 0
Initiated By (print/sign)
Ross Groehler
/
-/* - "A 1
Daý-11/04/2002 H - INITIAL APPROVAL This change is correct and complete, can he performed as written, and does not ad vsely affect personnel or nuclear safety, or Plant operating conditions.
7 Group Supervisor (print/sign) /-
tZ*
tr / 7_te t7or (Cannot be theloitiator This change does not adversely aect an operating conditios(Sa ty Relate.rocedures only)
Senior Reactor Operator (printisign)
[(.. *t
.o I
/
(iV Date /_111W)
" (Cannot be the Initiator or Group Supervisor)
IMI - PROCEDURE OWNER REVIEW
)R Permanent El One-time Use
[I Expiration Date, Event or Condition:
El Hold change until procedure completed (final review and approval still required within 14 days of i itial approval)
El QRJMSS Review NOT Required (AdminNNSR only)
QR Review R ired 2d ()
Procedure Owner (print/sign)
.7i/,'4, W, gyr- /
This Change and suvoortini recuirements correctlv comoleted and oroceseda IV.-INAL REVIEW AND AP L
ust be comnleted within 14 days of Initial aqrovyli I'he Initiator. OR and AnItvaAut~rv a all be lndeuendentfrom ach thru 0
SS(print/sign)
Y CI )
I
/_
Date'/1 M /
=
Indicates 5059f72.48 applicability assessed, any necessary screenings/evaluations p-eormed, det n made as to wh r additional cmss-d'isciplinary review required, and if required, performed.
MSS Meeting No.
ADmroval. Authoritv (print/sizn) D*, a 71 V -REVISION INFORM A TKIOj Q-PE MAD C!6ii ANGES Post Typing Review (print/sign) e________Date_________I_____r\\___
Indicates temporary change(s) incorporated exactly as approved and no other changes made to document.
Incorporated into Revision Number Effective Date PBF-0026e
References:
NP 1.2.3 Revision 13 01/16/02
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Page
_ of__
Doc Number AOP 10 Revision 0
Unit PBO Title Control Room Inaccessibility Temporary Change Number 2002-0865 Description of Changes:
Step
- Change/Reason Attachment A Add caution associated with AFW minimum flow requirement: To prevent AFW pump damage monitor pg 6 of 18, and maintain minimum AFW discharge flow or stop the affected AFW pump as necessary to control S/G Attachment B levelsi See SCR 2002-0458, CAP029908, P-38A MD AFWP has inadequate Recirc Flow during IT-10.
pgXof 18 See also CAP029952, possible common mode failure associated with AFW pump recirc line.
Y i
___________________ I
- Note: Recording of Step Number(s) is not required for multiple occurrences of identical information or when not beneficial to reviewers.
PBF-0026c Revision 6 04/19/01
- 1
References:
NP 1.13, NP 1.2.3 j
i i
j I
Point Beach Nuclear Plant TEMPORARY CHANGE AFFECTED MANUAL LOCATION Page of Procedure Number AOP 10 Revision 0
Unit PBo Title Control Room Inaccessabitily Temporary Change Number 2002-0865 I - IMMEDIATELY AFTER INITIAL APPROVAL ON PBF-0026e (Non-intent changes)
(after Final Approval if change of intent involved)
Date This procedure change has been processed as follows: (Manual/Location)
Performed ED Copy included in work package for field implementation. (WO No.
)
[]
Copy filed in Control Room temp change binder (Operations only).
/
5 c Z.
[]
Original change package provided to to obtain Procedure Owner Review (eg, Owner review may be coordinated by In-Group OA II, Procedure Writer, Procedure Supervisor, etc.).
El 5
El Performed By (print and sign)
Carol Schroeder Date VWY2 II - PROCEDURE OWNER REVIEW ON PBF-0026e (may be performed by OARI, Procedure Writer, etc.)
This procedure change has been processed as follows: (Manual/Location)
Date Performed
(] Copy sent to Document Control Distribution Lead for Master File.
y (Not required for one-time use change)
E]
Copy filed in Group satellite file. (Not required for one-time use changes.)
El Copy filed in Group one-time use file.
S Original Temp Change provided to I>) G+/- 3>
to obtain Final Approvals 2-.
(ecg., final approval may be coordinated by In-Group OA AI Procedure Writer, Procedure Supervisor, etc.)
SPAB SOPs OFFICE Performed By (print and sign)
Carol Schroeder Date 1~fl04t2O02 coPy PBF-O026h Revision 5 06/13/01
Reference:
NP 1.23
Nuclear Power Business Unit TEMPORARY CHANGE REVIEW AND APPROVAL Note: Refer to NP 1.2.3, Temporary Procedure Changes. for requirements.
Page-i of IP-INITIATION Doc Number AOP 10 Current Rev 0
Unit PBO Temp Change No.
2002-0764 Document Title Control Room Inaccessibility Existing Effective Temporary Changes N/A Brief Description Add caution to Attachments A and B to address AFW Minimum Flow requirements (identif specific changes on Form PBF-0026c, Document Review and Approval Cominuaiton, and include with the package)
ED Initiate PBF-0026h and include with the change.
Other documents required to be effective concurrently with the temporary change:
A) (A Changes pre-screened according to NP 5.1.8?
O NO 0 YES (Providc docmcton a-cor,,n to wP S.IJ)
Screening completed according to NP 5.1.8?
EN NA 0 YES (Attah vy)
Safety Evaluation Required?
( NO E] YES (
,iar, somay bepces*dorraw
'-*an_*ppvshan be*o b;* d
,mr p,
Determine if the change constitutes a Change Of Intent to the procedure by evaluating the following questions.
(If any answers are YES. 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)?
- 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 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 performince also being deleted?
- 5. Change Tech Spec or other regulatory acceptance criteria other than for re-baselining purposes?
0l 0
z E0 00 El
[
- 6. Require a change to the procedure Purpose or change the procedure classifiction?
0 0
Initiated By (print/sign)
Ross Groehler I
Ddf7=**
Date 10/29t2002 II - INITIAL APPROVAL This change is correct and complete, caa be performed as written, and does not nuclear safety, or Plant operating conditions.
Group Supervisor (print/sign)
A,-#,,
/-ZI
"(Cannott This change does not adversely affect,'an* operating Senior Reactor Operator (print/sign) _
I-,,
)
)z
)
(Cannot be the Initiator erCrrnn III - PROCEDURE OWNER REVIEW
[Z Permanent 0l One-time Use 0l Expiration Date, Event or Condition:
0 Hold change until procedure completed (final review and approval still required within 14 days of initial approval) 0 QR/MSS Review NOT Requi minINNSR only)
[
QRRevicwRe u' d 0] MSS
,equld(Re,,=eceHt6.5)
Procedure Owner (print/sign) uo2 ruient Date 47 This Chante and supooruing reouirements correctly cornoleted and orce9m f ust be comoleted within 14 days orinji !anooI'6 rThe lnltiator. OR and v I Aut hon*]rPhall be lndeoendezt from areb.
-hot er)
SS (printsign) 6I'.()
1k I V/
I f
6-L I' "
Date 6 c
Indicates 50.59M.48 applicability assessed, any necessary sercenings/cvauations performed, determination made as to whethr additi6nal cross-disciplinary review required, and if required, performed.
MSS Meeting No.
A'!,
Approval Authority (orint/sien)
.9,54d-A z'-
Date /41"*-' 'L V-REVISION INFORMATION FOR PERMANENT ANGES Post Typing Review (print/sign)
I Date Indicates temporary change(s) incorporated exactly as approved and no other changes made to document Incorporated into Revision Number Effective Date PBF-0026e Revision 13 01/16/02 coPY
References:
NP 1.23 IV -FINAL REVIEW AND APP RVA, I
(Cannot be the Initiator or Grogin !
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Page of
- Note: Recording orStep Number(s) is not required for multiple occurrences oridentical information or when not beneficial to revice.ers.
PRBF6-00R6c NP
=W.
Reviion6 0/1801
References:
NP 1 13. NP 1.2.3 I
,m
°
Point Beach Nuclear Plant TEMPORARY CHANGE AFFECTED MANUAL LOCATION Page of Procedure lNumber AOP-10 Revision
__0 Unit PB0 Title CON'TROL ROO&M Di'ACCESSIBILITY Temporary Change Number 2002-0764 I - IMMEDIATELY AFTER LNITIAL APPROVAL ON PBF-0026e (Non-intrit changes)
(after Final Approval if change of intent involved)
This procedure change has been processed as follows: (Manual/Location)
Date
-- 0
-Performed
[E Copy included in work package'for field implementation. (WO No.
)
[]
Copy filed in Control Room temp change binder (Operations only).
10 R I*
Original change package provided to rtc V to obtain Procedure Owner Review (e.g., Owner review may be coordinated by In-Group OA II. Procedure Writer, Procedure Supervisor, etc.
(o "*'-"
10 El El LI Performed By (print and sign) Cl, Ie___*/.ztr Date /o-.3o-cz-II - PROCEDURE OWNERREVIEW ON PBF-0026e (may be performed by OA 11 Procedure Writer, etc.)
This procedure change has been processed as follows: (ManuaL/Location)
Copy sent to Document Control Distribution Lead for Master File.
(Not required for one-time use change)
Eli Copy filed in Group satellite file. (Not required for one-tire use changes.)
r-]
Copy filed in Group one-time use file.
[]
Original Temp Change provided to __
'1)
- to obtain Final Approvals (e.g, final approval may be coordinated by In-Group OA IIL Procedure Writer, Procedure Supervisor. etc.)
I LLA.4 eta._
O (w '
~
0
ýS" Performed By (print and sign)
- c.
'6~~4x tit, cfJ Date /P, 5 L ~~
PDF-OO26h Revision 5 06113ADl COY'l, f Reference-NP 1 2 3 I
'h "lit" I
I
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE)
Verify SCR number on all pages Page 1 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-1 17 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 / John P. Schroeder
,f(
,Pai/
te:
/0.
Name (Print)
Reviewed by-/
Date:
,/
Nam-Print)
Signature PART I (50.59172.48) - DESCRIBE THE PROPOSED ACTIVITY AND SEARCH THE PLANT AND ISFSI LICENSING BASIS (Resource Manual 53.1)
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 72A8 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. Pr6cedural 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 cask/ISFSI as appropriate. Identify where the pertinent information is described in the above documents (by document section number and title). (Resource Manual 5.3.1 and NEI 96-07, App. B, B.2)
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 7.3.3A 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.
I)
I
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE)
Verify SCR number on all pages Page 2 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:
[] 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 0 No If a storage cask Certificate of Compliance change is required, explain what the change should be and why it is required.
S....
10 CFR 50.59 SCREENING PART 11 (50.59) - DETERMINE IF THE CHANGE INVOLVES A DESIGN FUNCTION (Resource Manual 5.3.2)
Compare the proposed activity to the relevant CLE descriptions, and answer the following questions:
YES NO QUESTION 0]
[J Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?
[1 0]
Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?
0Z
[I 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 1]
Does the pro'posed 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?
O]
0]
Does the activity involve a method of evaluation described in the FSAR?
O 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 limit for afission product barrier (DBLFPB)?
(NOTE: If THIS questions is answered YES, a 10 CFR 50.59 Evaluation is required.)
If the answers to ALL of these questions are NO, mark Part MI as not applicable, document the 10 CFR 50.59 screening in 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.
COPY
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE)
Verify SCR number on all pagps 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 wiUl 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 m (50.59) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (Resource Manual 5.3.3)
If ALL the questions in Part H are answered NO then Part III is [1 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 11I3.
mI.I CHANGES TO THE FACILITY OR PROCEDURES YES NO QUESTION 0] "
[
Does the activity adversely affect the design function of an SSC credited in safety analyses?
1]
0]
Does the activity adversely affect the method of performing or controlling the design function of an SSC credited in the safety analyses?
If any answer is YES. a 10 CFR 50.59 Evaluation is required. If both answers are NO. describe the basis for the conclusion (attach additional discussion as necessary):
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 El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in the CLB?
El 0]
Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in the CLB?
If any answer is YES a 10 CFR 50.59 Evaluation is required. If both answers are NO describe the basis for the conclusion (attach additional discussion, as necessary).
111.3 TESTS OR EXPERIMENTS If the activity is not a test or experiment, the questions in I.3.a and I.L3.b are 0 NOT APPLICABLE.
- a. Answer these two questions first:
YES NO QUESTION E]
D]
Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?
El El Are the SSCs affected by the proposed test or experiment isolated from the facility?
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59172.48 SCREENING (NEW RULE)
Veify SCRnumber o l pages Page 4 If the answer to BOTH questions in V.3.a is NO continue to III3.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 re] NOT APPLICABLE.
YES NO QUESTION El El Does the activity utilize or control an SSC in a manner that is outside the reference bounds Uf the design bases as described in the CLB?
El El Does the activity utilize or control an SSC in a manner that is inconsistent with the analyses or descriptions in the CLB?
El El Does the activity place the facility in a condition not previously evaluatqd or that could affect the capability of an SSC to perform its intended functions?
If any answer in 111.3.b is _
a 10 CFR 50.59 Evaluation is required. If the answers in IL3.b are ALL desc'be 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 [I required or 0 NOT required.
A Point Beach FSAR change is El required or 0 NOT required. If an FSAR change is required, then initiate an FSAR Change Request (FCR) per NP 5.2.6.
A Regulatory Commitment (CLB Commitment Database) change is El required or 0D NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.
A Technical Specification Bases changý is [] required or [0 NOT required. If a change to the Technical Specification Bases is required, then initiate a Technical Specification Bases change per NP 5.2.15.
A Technical Requirements Manual change is El 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: NET 96-07, Appendix B, Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 72.48 screenings.
PART V (72.48) - 10 CFR 72.48 INITIAL SCREENING QUESTIONS Part V determines if a full 10 CFR 72.48 screening is required to be completed (Parts VI and VII) for the proposed activity.
.ES NO QUESTION El 09 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 50.59172.48 SCREENING (NEW RULE)
VeifySCR niu*beron 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?
I]
0]
Does the proposed activity involve IN ANY MANNER SSC(s) needed for plant operation which are also used to support cask loading/unloading activities, as follows: Spent Fuel Pool (SFP), SFP Cooling and Filtration (SF),
Primary Auxiliary Building Ventilation System (VNPAB), Drumming Area Ventilation System (VNDRM),
RE-105 (SFP Low Range Monitor), RE-135 (SFP High Range Monitor), RE-221 (Drumming Area Vent Gas Monitor), RE-325 (Drumming Area Exhaust Low-Range Gas Monitor), PAB Crane, SFP Platform Bridge, Truck Access Area, or Decon Area?
0 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.?
El 0]
Does the activity involve plant heavy load requirements or procedures for areas of the plant used to support cask loadingfunloading activities?
El []
Does the activity involve any potential for fire or explosion where casks are loaded, unloaded, transported or stored?
If ANY of the Part V questions are answered YES, then a full 10 CER 72.48 screening is required and answers to the questions in Part VI and Part VII are to be provided. If ALL the questions in Part V are answered NO then check Parts VI and VII as not applicable. Complete Part VIII to document the conclusion that no 10 CFR 72.48 evaluation is required.
PART VI (72.48) - DETERMINE IF THE CHANGE INVOLVES A ISFSI LICENSING BASIS DESIGN FUNCTION (If ALL the questions in Part V are _NO then Part VI is El 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?
E]
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?
E]
[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 0
Does the activity involve a method of evaluation described in the ISFSI licensing basis?
[]
0 Z
Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)
E]
[0 Does the activity exceed or potentially affect a cask design basis limit for afission product bari'er (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 El Does the activity adversely affect the designfunction of a plant, cask, or ISFSI SSC credited in safety analyses?
El El Does the activity adversely affect the method of performing or controlling the design function 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):
VIL2 Changes to a Method of Evaluation (If the activity does not involve a method of evaluation, these questions are El NOT APPLICABLE.)
YES NO QUESTION El 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):
VI.3 Tests or Experiments (If the activity is not a test or experiment, the questions in VIL3.a and VII3.b are El NOT APPLICABLE.)
- a. Answer these two questions first:
YES NO QUESTION El
[]
Is the proposed test or experiment bounded by other tests or experiments that are described in the cask ISFSI licensing basis?
El El Are the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISFSI facility?
If the answer to both questions is NO continue to VII.3.b. If the answer to EITHER question is YES, then briefly describe the basis.
- b. Answer these additional questions ONLY for tests or experiments which do not meet the criteria given in VII.3.a above.
If the answer to either question in VII.3.a is YES then these three questions are [] NOT APPLICABLE:
Point Beach Nuclear Plant 10 CFR 50.59/72.48 SCREENING (NEW RULE)
YES NO SCR 2002-0458 Verify SCR number on all pages Page 7 QUESTION El El Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the ISFSI licensing basis?
E]
El Does the activity utilize or control a plant, cask or ISFSI facility SSC in a manner that is inconsistent with the analyses or descriptions in the ISFSI licensing basis?
El El Does the activity place the cask or ISFSI facility in a condition not previously evaluated or that could affect the capability of a plant, cask, or ISFSI SSC to perform its intended finctions?
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 [I required or ED 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 0 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 El 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 ED 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-1O ABNORMAL OPERATING PROCEDURE SAFETY RELATED ReVision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 1 of 18 A.
PURPOSE
- 1. To provide the operators with the instructions required to maintain Hot Shutdown in the event conditions in the Control Room require evacuation as deemed necessary by the Control Room staff.
The following assumptions apply to this procedure:
a Offsite power is available.
- All controls are operational and no failures are expected to occur to the control board which precludes the safe operation of equipment from outside the Control Room.
e No other accident condition exist requiring use of the Emergency Operating Procedures or'any other Abhormal Operating Procedure.
a Both Units are it power. Mode 1.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. The following are entry conditions for this procedure:
- a. Toxic gas in the Control Room, requiring evacuation.
- b. Confirmed bomb -threat in or adjacent to the Control Room requiring evacuation.
- c. Other life threatening conditions, as determined by the DSS or his designee, that cause the Control Room to be uninhabitable.
C. REFERENCES
- 1.
General Design Criteria #19
- 2.
Technical Specifications for Point Beach Nuclear Plant
- 3.
FSAR for Point Beach Nuclear Plant
- 5.
DBD-04. Chemical And Volume Cpntrol System oapy
POINT BEACH NUCLEAR PLANT AOP-1O ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 2 of 18 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES Steps in this procedure may be performed out of order as deemed necessary by the DSS or his designee.
- Entry into EOP-O.
REACTOR TRIP OR SAFETY INJECTION. is not required.
1 Initiate Manual Reactor Trip For Both Units
"- Unit 1 Reactor - TRIPPED a Unit 2 Reactor.- TRIPPED 2
Ensure Both Units.Turbine - TRIPPED'
- Unit 1 Turb Ine - TRIPPED
- Unit 2 Turbine - TRIPPED 3
Shut Main Steam Isolation Valves e lMS-2018 e 1MS-2017 a 2MS-2018
- 2MS-2017 4
Adjust Atmospheric Steam Dump Controllers To 1005 PSIG"
5 Align Charging Pump Suctions To RWST:.
- 2CV-112B
POINT BEACH NUCLEAR PLANT AOP-l0 POINT BEACH NUCLEAR PLAINT ABNORMAL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY AOP-10 SAFETY RELATED Revision 0 2/18/2002 Page 3 of 18
[~~I ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED 6
Start Turbine-Driven AFW Pumps a 1P-29 a 2P-29 7
Place Motor-Driven AEW Discharge Valves In - MANUAL PULLOUT AND CLOSE AF-4021 for S/G lB AF-4022 for S/G 2A SStop Main Feedwater Pumps And Place Control Switches In - AUTO 6
0 0
0 IP-28A 1P-28B 2P-28A 2P-28B 9
Stop Heater Drain Taýk Pumps a
U 0
a 0
a 1P-27A IP-27B IP-27C 2P-27A 2P-27B 2P-27C 10 Ensure Only One Condensate Pump Running Per Unit o P-25A OR o P-25B 11 Evacuate Control Room And Obtain Copies Of This Procedure From The Work.Control Center 12 Notify CAS Of Control Room Evacuation CAUTION
- Placing Main Feed-Pump control switches in pull-out will defeat auto start of:
the Motor Driven A!W pumps.
I I
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POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 4 of 18
[I~
ACTION/EXPECTED RESPONSE REPOS NO BAM 13 Dispatch Four Licensed Operators To Perform Local Actions:
- a. Attachment A. Unit I AFW PUMP OPERATOR
- b. Attachment B. Unit 2 AFW PUMP OPERATOR
- c. Attachment C. UNIT 1 CHARGING PUMP OPERATOR
- d. Attachment D. UNIT 2 CHARGIN9 PUMP OPERATOR 14 Dispatch Operators To Perform Local Actions:
- a. Attachment E. TURBINE HALL OPERATOR
- b. Attachment *.
PAB OPERATOR 15 Direct STA To Report To TSC And Implement Emergency Plan
"* 16 Locally Monitor Operating Equipment Until Control Room Can Be Re-Entered
- 17.
Check Control Room -
HABITABLE WHEN Cpntrol Room can be re-entered.
- THEN go to EOP-O, REACTOR TRIP OR SAFETY INJECTION.
-END-
POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 5 of 18 I I ACTION/EXPECTED RESPONSE J
I RESPONSE NOT OBTAINED ATTACHMENT A (Page 1 of 2)
Unit 1 AFW PUMP OPERATOR Al Check Turbine-Driven AFW Pump RUNNING 1P-29 Direct PAB Operator to locally align steam supply to turbine-driven AFW pump:
o Locally open B SIG steam supply.
valve.
- IMS-2019 o Locally open A SIG steam supply valve.
l IMS-2020 A2 Manually Control A SIG Level:
a.'Engage clutch and throttle IAF-4001 to maintain S/G level
-BETWEEN 300 INCHES AND 330 INCHES a 1 LI-460-AA an IRK-38 1
1LI-460-BA on lRK-38
- A3 Manually Control B S/G Level:
- a. Engage clutch and throttle 1AF-4000 to maintain S/G ievel
-BETWEEN 300 INCHES AND 330 INCHES
- ILI-470-AA on IRK-38 ILI-470-BA on 1RK-38 CAUTION To prevent AFW pump damage, monitor and maintain minimum AFW discharge flow greater than 75 gpm or stop the affected AFW Pump as necessary to control SIG levels.
I
POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 6 of 18
[ TE II ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT A (Page 2 of 2)
Unit I AFW PUMP OPERATOR A4 Maintain S/G Levels Between 300 Inches And 330 Inches Using Turbine-Driven AFW Pump IF level can NOT be maintained using turbine-driven AFW pump. THEN maintain A S/G level using motor-driven AFW pump as follows:
- a. At N-01.
place P-38A in
- LOCAL.
- b. At N-O1, depress start pushbutton.
- c. Open AF-4023.
- d. Throttle 1AF-31 to maintain A S/G level -
BETWEEN 300 INCHES AND 330 INCHES.
1 1LI-460-AA on IRK-38 1
1LI-460-BA on 1RK-38
- e.
Inform DOS that B S/G atmospheric steam dump should be isolated.
e IMS-2015 A5 Inform DOS That S/G Levels Are BETWEEN 300 INCHES AND 330 INCHES
-END-CAUTION To prevent AFW pump damage, monitor and maintain minimum AFW discharge flow or stop the affected AFW pump as necessary to control S/G levels.
"o P-38A minimum flow -
GREATER THAN 50 GPM "o P-29 minimum flow -
GREATER THAN 75 GPM I
I
POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 7 of 18 sTEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACMIENT B (Page 1 of 2)
Unit 2 AFW PUMP OPERATOR CAUTION To prevent AFW pump damage, monitor and maintain minimum AFW discharge flow greater than 75 gpm or stop the affected AFW Pump as necessary to control S/G levels.
BI Check Turbline-Driven AFW Pump -
Direct PAB Operator to locally align RUNNING steam supply to turbine-driven AFW pump:
- 2P-29 o Locally open B SIG steam supply valve.
- 2MS-2019 o Locally Open A SIG steam supply valve.
- 2MS-2020 B2 Manually Control A SIG Level:
- a. Engage clutch and throttle 2AF-4001 to maintain S/G'level
-BETWEEN 300 INCHES AND 330 INCHES S
e 2LI-460-AA on 2RK-38 e 2LI-460-BA on 2RK-38 B
B3 Manually Control B S/G Level:
- a. Engage clutch and throttle 2AF-4000 to maintain SIG level
-BETWEEN 300 INCHES AND 330 INCHES a 2LI-470-AA on 2RK-38
- 2LI-470-BA on 2RK-38
POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 8 of 18 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT B (Page 2 of 2)
Unit 2 AFW PUMP OPERATOR B4 Maintain S/G Levels Between 300 Inches And 330 Inches Using Turbine-Driven AFW Pump IF level can NOT be maintained using turbine-driven AFW pump, THEN maintain B S/G level using motor-driven AFW pump as follows:
- a. At N-02. place P-38B in LOCAL.
- b. At N-02. depress start pushbutton.
- c. Open AF-4020.
- d. Throttle AF-45 to maintain B S/G level -
BETWEEN 300 INCHES AND 330 INCHES.
- 2LI-470-AA on 2RK-38
- 2LI-470-BA on 2RK-38
- e. Inform DOS that A S/G atmospheric steam dump should be isolated.
a 2MS-2016 B5 Inform DOS That S/G Levels Are BETWEEN 300 INCHES AND 330 INCHES
-END-L~JII C
CAUTION To prevent AFW pump damage, monitor and maintain minimum AFW discharge flow or stop the affected AFW pump as necessary to control S/G levels.
"o P-38B minimum flow - GREATER THAN 50 GPM "o P-29 minimum flow -
GREATER THAN 75 GPM 9
k I
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY AOP-10 SAFETY RELATED Revision 0 2/18/2002 Page 9 of 18 ACTION/EXPECTED RESPONSE p
I I I RESPONSE NOT OBTAINED ATTACHMENT C (Page I of 2)
UNIT I CHARGING PUMP OPERATOR C1 Ensure Charging Pump Suction ALIGNED TO RWST:
"* RWST to charging pump suction 10V OPEN 1CV-112B VCT outlet to charging pump suction NOV - SHUT Align charging pump suction to RWST:
- a. Open RWST to charging pump suction.
I 1CV-358
- 1CV-li2C CAUTION Placing pressurizer heaters in local defeats heater low level cutout.
C2 Check PZR Pressure -
BETWEEN Locally operate back-up heaters to.
2200 PSIG AND 2250 PSIG maintain PZR pressure - BETWEEN 2200 PSIG AND 2250 PSIG So Bank C "o
Bank D I
S TE P W
I l
"IK PT-*qR
POINT BEACH NUCLEAR PLANT ABNORKAL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY AOP-10 SAFETY RELATED Revision 0 2/18/2002 Page 10 of 18 I
I I
ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT C (Page 2 of 2)
UNIT 1 CHARGING PUMP OPERATOR C3 Check If Letdown Should Be Established:
- a. Check PZR level -
GREATER THAN 16% AND RISING
- LI-433C
- b. Establish letdown:
- 1) Inside IB52-426M. locally open letdown line isolation I
IRC-427
- 2) Locally open letdown isolations as necessary
- a. WHEN PZR level greater than 16%.
THEN do Step C3.b.
- b. Perform the following:
a) Operate charging pumps as necessary to maintain PZR level - BETWEEN 20% AND 45%
a 01-15.
CHARGING PUMP LOCAL CONTROL STATION OPERATION b) Go to Step C5.
"o 1CV-200A "o 1CV-200B "o 1CV-200C-C4 Check PZR Level - BETWEEN 20% AND Operate charging pumps as necessary, S 25%
to maintain PZR level-BETWEEN 20% AND 45%
- 4.
T./
-A 01 15.
CHARGING PUMP LOCAL CONTROL I
STATION OPERATION C5 Inform DOS Of The Following:
- PZR level
- PZR pressure
-END-T TEP II NOTES
- PZR level may require time to stabilize after letdown restoration.
- Charging pumps should remain running in auto/remote if possible.
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POINT BEACH NUCLEAR PLANT POINT BEACH NUJCLEAR PLANT ABNORMAL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY AOP-10 SAFETY RELATED Revision 0 2/18/2002 Page 11 of 18 ISTEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT D (Page 1 of 2)
UNIT 2 CHARGING PUMP OPERATOR Dl Ensure Charging Pump Suction ALIGNED TO RWST:
RWST to charging pump suction NOV
- OPEN
- a. Open RWST to charging pump suction.
e 2CV-358
- 2CV-112C CAUTION Placing pressurizer heaters in local defeats heater low level cutout.
D2 Check PZR Pressure-BETWEEN Locally operate back-up heaters to 2200 PSIG AND 2250 PSIG.
maintain PZR pressure - BETWEEN 2200 PSIG AND 2250 PSIG
- PI-44gB oBank C o Bank D i
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- t POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY AOP-10 SAFETY RELATED Revision 0 2/18/2002 Page 12 of 18 s
ONSE NOT OBTAINED
- . D4 Check PZR Level -
BETWEEN Operate'charging pumps as necessary 20% AND 25%
to maintain PZR level-BETWEEN 20% AND 45%
- LI-433C 01 15.
CHARGING PUMP LOCAL CONTROL STATION OPERATION D5 Inform DOS Of The Following:
, PZR level PZR pressure
-END-T ll ACTION/EXPECTED RESPONSE I I ATTACHMENT D (Page 2 of 2)
UNIT 2 CHARGING PUMP OPERATOR D3 Check If Letdown Should Be Established:
- a. Check PZR level -
GREATER THAN
- a.
WHEN P:
16% AND RISING THEN d,
- LI-433C
- b. Establish letdown:
- b. Perfor,
- 1) Inside 2B52-427J. locally open a) Ope:
letdown line isolation' nec' levi m 2RC-427 a 0:
- .1*
NOTES PZR level may require time to stabilize after letdown restoration.
Charging pumps should remain running in auto if possible.
ZR level greater than 16%.
o Step D3.b.
m the following:
rate charging pumps as essary to maintain Pzr el -- BETWEEN 20% AND 45%
1-15..CHARGING PUMP LOCAL ONTROL.STATION OPERATION to Step D5.
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POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY ACTION/EXPECTED RESPONSE AOP-10 SAFETY RELATED Revision 0 2/18/2002 Page 13 of 18
..i I
I ATTACHMENT E (Page 1 of 3)
TURBINE HALL OPERATOR El Check Both Reactors Tripped Prior To Evacuating Control Room
- Unit 1 Reactor - TRIPPED
- Unit 2 Reactor r TRIPPED E2 Check Both Turbines Tripped Prior To Evacuating Control Room
"* Unit 1 Turbine'- TRIPPED
"* Unit 2 Turbine - TRIPPED E3 Inform DOS Both Reactors And Both Turbines Are Tripped E4 At 1AO1, Check The Following Pumps Stopped Prior To Evacuating The Control Room:
1 1P-28A Steam.Gen~erator Feed Pump
- -IP-27A Heater-Drain Tank Pump
, IP-27C Heater Drain Tank Pump Perform the fcllowing:,
- a. Open reactor trip breakers for BOTH UNITS.
- Unit 1 reactor trip breakers OPEN
- Unit 2 reactor trip breakers OPEN
- b. Open reactor trip bypass breakers.
for BOTH UNITS.
"* Unit 1 reactor trip bypass breakers -
OPEN
"* Unit 2 reactor trip bypass breakers -
OPEN Perform the following:
- a. At UOrit 1 Front Standard. rot-ate trip lever to trip position.
- b. At Unit 2 front Standard. rotate trip lever to trip position.
Locklly open breakers as necessary per Attachment H*. LOCAL.BREAKER OPERATION,.
"* 1A52-05 IA52-0B
"* 1A52-09 p.
L&J NOTE All or some of the actions performed in this attachment may have been performed in the control room prior to evacuation.
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t*
I II RESPONSE VqOT OBTAINED I
i
/
1
-o POINT BEACH NUCLEAR PLANT AOP-1O ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0
.2/18/2002 C6NTROL ROOM I14ACCESSIBILITY Page 14 of 18 Fs
ýfI ACTION/EXPECTED RESPONSE I
RSPNE OT BAND ATTACHMENT E (Page 2 of 3)
TURBINE HALL OPERATOR E5 At IA02. Check The Following Pumps Stopped Prior To Evacuating The Control Room:
a 1P-28B Steam Generator Feed Pump e IP-27B Heater Drain Tank Pump E6 At 2A01. Check The Following Pumps Stopped Prior To Evacuating The Control Room:
"* 2P-28A Steam Generator Feed Pump
"* 2P-27A Heater Drain Tank Pump 2P-27C Heater Drain.Tank Pump E7 At 2A02. Check The Following Pumps Stopped Prior To Evacuating The Control Room:
- 2P-28B Steam Generator Feed Pump
- 2P-27B Heater Drain Tank Pump E8 At lAOl or IA02. Check One Unit 1 Condensate Pump Breaker - OPEN Locally open breakers as necessary per Attachment H. LOCAL BREAKER OPERATION.
- IA52-10 Locally open breakers as..necessary per Attachment H. LOCAL BREAKER OPERATION.
"* 2A52-23
"* 2A52-20
"* 2A52-19 Locally open breakers as necessary per Attachment H. LQCAL BREAKER OPERATION.
"* 2A52-30
"* 2A52-33 Locally open breaker per Attachment H. LOCAL BREAKER OPERATION.
o 1A52-07 for-IP-25A OR o 1A52-11 for 1P 25B E9" At 2A01 or 2A02.
Check One Unit 2 Condensate Pump Breaker - OPEN Locally open breaker per Attachment H. LOCAL*BREAKER OPERATION.'-
o 2A52-21 for 2P-25A OR a 2A52-32 for 2P-25B EIO Inform DOS Of The Following:
- Main Feedwater ppmps - STOPPED
- Heater Drain Tank pumps - STOPPED
- One Condensate pump - RUNNING b.
d
/
1.
POINT BEACH NUCLEAR PLANT ABNORMAL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY AOP -10 SAFETY RELATED' Revision 0 211812002 Page 15 of 18 p
[STEP I ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED ATTACHMENT E (Page 3 of 3)
TURBINE HALL OPERATOR Level - GREATER THAN 15 FT Direct Water Treatment Operator to commence filling CST.
"* LI-4025 for T-24A
"* LI-4031 for T-24B "END-Ell Check CST J
II I
POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 16 of 18 SSTEP I ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED ATTACHMENT F (Page 1 of 1)
PAB OPERATOR F1 Check Unit 1 MSIVs Were Shut Prior To Control Room Evacuation
"* IMS-2018 for SIG 1A
"* lMS-2017 for SIG 1B F2 Check Unit 2 MSIVs Were Shut Prior To Control Room Evacuation
"* 2MS-2018 for SIG 2A
"* 2MS-2017 foe SIG 2B Shut Unit 1 MSIVs as follows:
- a. At IRK-33. depress both pushbuttons.
"* IMS PB-2018A. train A
"- IMS PB-2018B. train B
- b. At PRK'34. depress both pushbuttons.
- a. At 2RK-33, depress both pushbuttons.
- 2MS PB-2018A. train A
- 2MS PB-2018B. train B
- b. At 2RK-34. depress both pushbuttons.
- 2MS PB-2017A. train A 2MS PB-2017B. train B F3 De-Energize Unit 1 Motor-Driven AFW "Pump Discharge MOV's O
Open Bkr. 1B5"2-328F for AF-4023
- Open Bkr. lB52-428C for AF-4021 F4 De-Energiie Unit 2 Motor-Driven AEW Pump Discharge HOV's
- Open Bkr. 2B52-328C for AF-4022
- Open Bkr. 2B52-428F for AF-4020 F5 Inform DOS Unit 1 And Motor-Driven AFW Pump MOV's - DE-ENERGIZED Unit 2 Discharge
-END-I I
I
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT ABNOP14AL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY AOP-10 SAFETY RELATED Revision 0 2/18/2002 Page 17 of 18 STEP I ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED ATTACHMENT G (Page I of 1)
DOS CHECKLIST GI Check the following Unit I actions completed:
Unit-1 ACTION PERFORMED Reactor tripped Motor-Driven AFW pump discharge valves breakers open PZR level - 20% to 45%
PZR pressure - 2200 PSIG to 2250 PSIG Turbine tripped Both Main Feedwater Pumps stopped All Heater Drain Pumps stopped One Condensate Pump.running All SIG levels between 300 Inches and 330 inches G2 Check the following Unit 2 actiong completed:
Unit 2 ACTION PERFORMED "Reactor tripped Motor-Driven AFW pump discharge valves breakers open.
PZR level - 20% to.45%
PZR pressure - 2200 PSIG to 2250 PSIG Turbine tripped Both Main Feedwater Pumps stopped All Heater Drain Pumps stopped One Condensate Pump running All SIG levels between 300 inches and 330 inches.
-END-.
COMPLETED
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(
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(
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(
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(
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POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT
- ABNORMAL OPERATING PROCEDURE CONTROL ROOM INACCESSIBILITY AOP-10 SAFETY RELATED Revision 0 2/18/2002 Page 18 of 18 S
CTION/EXPECTED RESPONSE I I SPoN ATTACHMENT H (Page 1 of 1)
LOCAL BREAKER OPERATION HI Local Opening Of A 4160 Vac Breaker:
- a. Open breaker cubicle door.
- b. Check mechanical indicator indicates closed.
- c. Open breaker by performing one of the following:
o Depress trip/open push plate.
OR o Depress trip/open button.
OR o Lift red tab on open coil.
- d. Check mechanical indicator indicates open.
- e. Close breaker cubicle door.
H2 Local Opening *f A 480 Vac Ireaker:
- a. Check mechanical indicator'indicates closed.
- b. Depress square trip pu'sh button.
- c. Check mechanical indicator indicates open.
-END-SE NOT OBTAINED
£'
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