ML030780089
| ML030780089 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/29/2002 |
| From: | Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2003-0094 AOP 10, Rev 0 | |
| Download: ML030780089 (28) | |
Text
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Nuclear Power Business Unit TEMPORARY CHANGE REVIEW AND APPROVAL Note: Refer to NP 1.2.3, Temporary ProcedureChanges,for requirements.
Page I of I - 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 (Identify specific changes on Form PBF-0026c, Document Review and Approval Continuation, and include with the package) 0 Initiate PBF-0026h and include with the change.
Other documents required to be effective concurrently with the temporary change:
WJ (A Changes pre-screened according to NP 5.1.8? 0 NO El YES (Provide docanenution according to NP 5.1.8)
Screening completed according to NP 5.1.8? U NA 0 YES (Anach copy)
Safety Evaluation Required? 0_ NO El YES (ifr. ,re be p rosd-or r, ie" and ** syiv*ts a be obained bore iplemenong)
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 S..dercription in the Currenn.-r ISFSI Licenfig Basis- dfid 1U'Tn1 5.1:8"aia N i' 3. s E. -.
- 2. Causean increase in magnitude, significance or impact such that it should be processed as a El Z 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 performance also being deleted? El z
- 5. Change Tech Spec or other regulatory acceptance criteria other than for re-baselining
/ A purposes? El 2 .
- 6. Require a change to the procedure Purpose or change the procedure classifieption?
El z Initiated By (print/sign) Ross Groehler ./ Date 10/29/2002 II - INITIAL APPROVAL This change is correct and complete, can be performed as written, and does not a nuclear safety, or Plant operating conditions. ely affect personnel or Group Supervisor (printtsign-) IZ ' rt,
-2 i*,'--'AA <A44KZ--t~-I~
(Cannot be theinitiator) ate This change does not adversely affect! Tan"operating conditions.
ty Relatcedures only)
Senior Reactor Operator (print/sign) ratingc ony)Date-(Cannot--------------------------
the Initiator o S soupr tj Ml - PROCEDURE OWNER REVIEW
[ Permanent Rl One-time Use El Expiration Date, Event or Condition:
El Hold change until procedure completed (final review and approval still required within 14 days El QRIMSS Review NOT Requircd-Wmin/NNSR only) 0 QR Review Re uqjd El MSS ý'ý equiof initial approval)
(Refer*-e NP 1.6.5)
Procedure Owner (printlsig) eo_____nt___o__ecy__o____ edan This Chaneeand sucoortina, renutrementscorrectiv completedand nroag~f Date.d-"
IV - FINAL REVIEW AND APPROVAL 5-j)Must be comoleted within 14 days of ini Intiator. The OR and A o A-h- hall beindeoendent from yaehot:crl
'16'MSS'(print/sign) i '!se F n-Io T-/'It " 1 Date nSappli ability assessed, any necessary screenings/cvaluations performed, determination made as to whethe1 addit6na cross-disciplinary review required, and ifrequired, performed.
PISS Meeting No. n 4 Date d Av~rovaI Authority (ornt/si2n) 9. -//x w L Post Typing Review (print/sign) V - REVISION INFORMATION FOR PERMANENT CfiANGES 1
Jfl Da Indicates temporary change(s) incorporated exactly as approved and no other changes made to document.
Incorporated into Revision Number _
Effective Date PBF-0026c Revision 13 01/16102 Refirences: NP 1.2.3
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Page of Doc Number AOP 10 Revision 0 Unit PBO Title Control room inaccessibilitv Temporary Change Number 2002-0764 Description of Changes:
Step
- ChanaelReason l - . I Attachment A pg 5 of 18, Add caution associated with AFW minimum flow requirement: To prevent AFW pump damage monitor Attachment B and maintain minimum AFW discharge flow or stop the affected AFW pump as necessary to control S/G p7ofl18 levels] Se eSiC R 2 00n' ln4lRrA PA ' *oo nR P _2A 1i. r * ^Arllr ,--.-:_ a . . . . , _ . .. .
S.. . ....
. .. ...... . .. ... .... ... * .*o -,*,* *r ,i* ina dequate tlecirc Fllow curing IT-10.
Other__ome o&ntsc ActL
. I Other Comments
" Note: Recording of Step Numberts) is not required for multiple occurrences of identical information or when not beneficial to reviewers PBF-0026c Revision 6 041181/01 References NP 1 1.3. NP 1 2.3
Point Beach Nuclear Plant TEMPORARY CI-HANGE AFFECTED MANUAL LOCATION Page of Procedure Number AOP-10 Revision 0 Unit PRO Title CONTROL ROOM -NACCESSIBILITY Temporary Change Number 2002-0764 I - LMMEEDIATELY AFTER INTIAL APPROVAL ON PBF-0026e (Non-Irntent change)
(after Final Approval if change of intent involved)
This procedure change has been processed 0 as follows: (Manual/Location) Date Performed El Copy included in work package for field implementation. (WO No. )
[YJ Copy filed in Control Room temp change binder (Operations only). 3o. o a,
". - Original change package provided to "T-, c' to obtain Procedure Owner
'-Review (g2Oýýr iiani9 b i%3i~b_ 11 Procedure WneýrcdrSupervwsr, tc.ý. C> 7 Elf El El El El Performed By (print and sign) C:,;., . ,. Date /o-.o-cZ-II - PROCEDURE OWNER REVIEW ON PBF-0026e (may be performed by OA IL. Procedure Writer- etc)'
Date This procedure change has been processed as follows: (Manual/Location) Performed SCopy sent to Document Control Distribution Lead for Master File.
(Not required for one-time use change)
El Copy filed in Group satellite file. (Not required for one-time use changes.)
El Copy filed in Group one-time use file.
[] Original Temp Change provided to b_
"5 to obtain Final Approvals / o. j'-o z..
(e g . final approval may be coordinated by In-Group OA IILProcedure Wnter, Procedure Supervisor, etc.)
Performed By (print and sign) _ct,( - I6*/.j fe&.<_/1 Date /- -
PBF-0026h Revision 5 06113/01 Rercrence" NP 1.2.3
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Vefy SCR numbe 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. Scy:idt/JohnP. Schroeder , !f Name ( Print)
Reviewed by: iK 1 *,6, 4 __ _____Date: /eA
-Namn-Z'Print) 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 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 flo'" to the steam generators supporting rapid cooldown during such accidents as: steam generator tube rupture (SGTh) 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.
RFAF 51 Si r
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: 0l Yes Z No If a Technical Specification change is required, explain what the change should be and why it is required.
1.4 Does the proposed activity involve a change to the terms, conditions or specifications incorporated in any VSC-24 cask Certificate of Compliance (CoC)? Changes to a VSC-24 cask Certificate of Compliance require a CoC amendment request E]Yes ED No If a storage cask Certificate of Compliance change is required, explain what the change should be and why it is required.
10 CFR 50.59 SCREENING PART 11 (50.59) - DETERMINE IF THE CHANGE INVOLVES A DESIGN FUNCTION (Resource Manual 5.3.2)
Compare the proposed activity to the relevant CLB descriptions, and answer the following questions:
YES NO QUESTION 0] El Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?
El 0] Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?
0 El Does the proposed activity involve SSCs whose failure could initiate a transient (e.g., reactor trip, loss of feedwater, etc.) or accident, OR whose failure could impact SSC(s) credited in the Safety Analyses?
0 El Does the proposed activity involve CLB-described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications?
El 0] Does the activity involve a method ofevaluation described in the FSAR?
El 0] Is the activity a test or experiment? (i.e., a non-passive activity which gathers data) k 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 ALL of these questions are NO mark Part III 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.
PBF-1515c
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages 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 111 (50.59) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (Resource Manual 5.3.3)
If AL.L the questions in Part II are answered NO, then Part Ill 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 11-.3.
,TO FACTJIT, OR PR E , .. .... ,
YES NO' QUESTION
[1 0 Does the activity adversely affect the designfunction of an SSC credited in safety analyses?
El 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 El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in the CLB?
El - E] 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 EXPER(MENTS If the activity is not a test or experiment, the questions in III.3.a and llI.3.b are 0 NOT APPLICABLE.
- a. Answer these two questions first:
YES NO QUESTION El El Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?
El EL Are the SSCs affected by the proposed test or experiment isolated from the facility?
PBF-1515c
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Venfy SCR number an all 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 IfI.3.a above.
If the answer to either question in II.3.a is YES, then these three questions are [] NOT APPLICABLE.
YES NO QUESTION
[E E] Does the activity utilize or control an SSC in a manner that is outside the reference bounds bf 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?
" '-D~e ia-- *1i 'iitZ "i61-ot previoustiy dvaluated orthat could-affect the-rapability---
of an SSC to perform its intended functions?
If any answer in fLI.3.b is YS a 10 CFR 50.59 Evaluation is required. If the answers in mL3.b are ALL NO describe the basis for the conclusion (attach additional discussion as necessary):
Part IV - 10 CFR 50.59 SCREENING CONCLUSION (Resource Manual 5.3.4).
Check all that apply:
A 10 CFR 50.59 Evaluation is El required or Z NOT required.
A Point Beach FSAR change is El 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 El required or Z NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.
Specification Bases is A Technical Specification Bases change is El required or Z NOT required. If a change to the Technical 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 to determine the NOTE: NEI 96-07, Appendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance proper responses for 72.48 screenings.
PART V (72.48) - 10 CFR12.48 INITIAL SCREENING QUESTIONS proposed activity.
Part V determines if a full 10 CFR 72.48 screening is required to be completed (Parts VI and VII) for the VES NO QUESTION tC] 0 Does the proposed activity involve IN ANY MANNER the dry fuel storage cask(s), the cask transfer/transport Basket equipment, any ISFSI facility SSC(s), or any ISFSI facility monitoring as follows: Multi-Assembly Sealed (MSB), MSB Transfer Cask (MTC), MTC Lifting Yoke, Ventilated Concrete Cask (VCC), Ventilated Storage Links, Cask (VSC), VSC Transporter (VCST), ISFSI Storage Pad Facility, ISFSI Storage Pad Data/Communication or PPCSJlSFSI Continuous Temperature Monitoring System?
PrIP-1 IqI ý"
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCRnumber on all pages Page 5 I] 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?
5l z 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?
5 0] Does the proposed activity involve a change to Point Beach CLB design criteria for external events such as earthquakes, tornadoes, high vrinds, flooding, etc.?
5 0] --
Does the activity involve plant heavy load requirements or procedures for areas of the plant used to support cask jainloa* kuod'i activities? ....... ..... - . ......
5l z Does the activity involve any potential for fire or explosion where casks are loaded, unloaded, transported or stored?
If ANY of the Part V questions are answered YES, then a full 10 CFR 72.48 screening is required and answers to the questions in Part VI and Part VeI 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 I] NOT APPLICABLE.)
.ompare the proposed activity to the relevant portions of the ISFSI licensing basis and answer the following questions:
YES NO QUESTION El] Z Does the proposed activity involve cask/ISFSI Safety Analyses or plant/cask/ISFSI structures, systems and components (SSCs) credited in the Safety Analyses?
[E Z Does the proposed activity involve plant, cask or ISFSI SSCs that support SSC(s) credited in the Safety Analyses?
l 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?
5] 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 Z Does the activity involve a method ofevaluation described in the ISFSI licensing basis?
[I] 0 Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)
S 0] Does the activity exceed or potentially affect a cask design basis limitfor a fission productbarrier(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 Z NOT APPLICABLE.)
PRF-1 SIsc
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 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. descr*be the basis for the conclusion
~ ..... _j (attach additiouaLdis'a.saion,.s.nJ~c~v,~)=. - .*....-.... ............. .. .......... ...
VII.2 Changes to a Method of Evaluation (If the activity does not involve a method of evaluation, these questions are [E NOT APPLICABLE.)
YES NO QUESTION El El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in a cask SAR?
El El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in a cask SAR?
If any answer is YES, a 10 CFR 72.48 Evaluation is required. If both answers are NO describe the basis for the conclusion (attach additional discussion, as necessary):
VII.3 Tests or Experiments (If the activity is not a test or experiment, the questions in VII.3.a and VII.3.b are 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?
E] E] AVe 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 E] NOT APPLICABLE:
PBF-1515c
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) ven fy SCR number on all pages Page 7 YES NO QUESTION
[E E] Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the ISFSI licensing basis?
[] El 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 E] required or Z NOT required. Obtain a screening number and provide the original to Records Management regardless of the conclusion of the 50.59 or 72.48 screening.
A VSC-24 cask Safety Analysis Report change is [] required or Z NOT required. If a VSC-24 cask SAR change is required, then contact the Point Beach Dry Fuel Storage group supervisor.
A Regulatory Commitment (CLB Commitment Database) change is [] required or Z NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.
A change to the VSC-24 10 CFR 72.212 Site Evaluation Report is [I required or 0 NOT required. Ifa VSC-24 10 CFR 72.212 Site Evaluation Report change is required, then contact the Point Beach Dry Fuel Storage group supervisor.
PBF-1515c
POINT BEACH NUCLEAR PLANT AOP-10 SAFETY RELATED ABNORMAL OPERATING PROCEDURE Revision 0 2/18/2002 Page 1 of 18 CONTROL ROOM INACCESSIBILITY 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:
- Offsite power is available.
to
- All controls ar*e operational and no failures are expected to occur the control board which precludes the safe operation of equipment from outside the Control Room.
No other accident condition exist requiring use of the Emergency Operta.tiin~g' Prýoc;aUedurs or a -r -e fi
- Both Units are it power. Mode 1.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. The following are eztry 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
- 4. DBD-O1. Auxiliary Feedwater
- 5. DBD-04. Chemical And Volume Cpntrol System
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 Page 2 of 18 CONTROL ROOM INACCESSIBILITY ISTEP II ACTION/EXPECTED RESPONSE I I l *RESPONSE NOT OBTAINED lily-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 SUnit 1 Reactor - TRIPPED
- Unit 2 Reactor .- TRIPPED 2 Ensure Both Units Turbine - TRIPPED
"*Unit I Turbine - TRIPPED
"*Unit 2 Turbine - TRIPPED 3 Shut Main Steam Isolation Valves
"*IMS-2018
"*IMS-2017
"*2MS-2017 4 Adjust Atmospheric Steam Dump Controllers To 1005 PSIG"
"*1HC-468
"*1HC-478
"*2HC-468
"* 2HC-478" 5 Align Charging Pump Suctions To RWST:.
".LCV-112B
"*1CV-112C
"*2CV-112C
POINT BEACH NUCLEAR PLANT AOP1O POINT BEACH NUCLEAR PLANT AOP-10 SAFETY RELATED ABNORMAL OPERATING PROCEDURE Revision 0 2/18/2002 Page 3 of 18 CONTROL ROOM INACCESSIBILITY RESPONSE NOT OBTAINED STEP ACTION/EXPECTED RESPONSE !
z 6 Start Turbine-Driven AFW Pumps
- AF-4021 for S/G 1B
- AF-4022 for S/G 2A auto start of.
Placing Main Feed-Pump control switches in pull-out will defeat the Motor Driven AFW pumps.
8 Stop Main Feedwater Pumps And Place Control Switches In - AUTO a IP-28A
"... ,t*t IP-28B S 2P-28A a 2P-28B 9 Stop Heater Drain Tank Pumps 6 IP-27A a IP-27B a IP-27C a 2P-27A 2P-27B S 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
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 4 of 18 I I I LiL ACTION/EXPECTED RESPONSE I l I I
RESPONSE NOT OBTAINED I 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 I CHARGING PUMP OPERATOR PUMP OPERATOR 14 Dispatch Operators To Perform Local Actions:
- a. Attachment E. TURBINE HALL OPERATOR
- b. Attachment F. 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
- I
- 17 Check Control Room - HABITABLE WHEN Cpntrol Room can be re-entered.
- THE____NN go to EOP-0. REACTOR TRIP OR
- "SAFETY INJECTION.
-END-t
POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 5 of 18 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINE ATTACHMENT A (Page I of 2)
Unit I 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.
Al Check Turbine-Driven AFW Pump - Direct PAB Operator to locally align RUNNING steam supply to turbine-driven AFW pump:
1P-29 I
o Locally open B SIG steam supply valve.
e lMS-2019 o Locally open A S/G steam supply valve.
1MS-2020 l
- A2 Manually Control A S/G Level: *
- a. Engage clutch and throttle *
- 1AF-4001 to maintain S/G level *
- -BETWEEN 300 INCHES AND *
- 330 INCHES *
- S* IRK-38 ILI-460-AA on *
- e 1LI-460-BA on 1RK-38 *
- A3 Manually Control B S/G Level: *
- a. Engage clutch and throttle *
- 1AF-4000 to maintain S/G level *
- -BETWEEN 300 INCHES AND *
- 330 INCHES *
- 1LI-470-AA on 1RK-38
- I *
- 1LI-470-BA on 1RK-38
- POINT BEACH NUCLEAR PLANT, AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOK INACCESSIBILITY Page 6 of 18 I ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I ATTACHMENT A (Page 2 of 2)
Unit 1 AFW PUMP OPERATOR A4 Maintain S/G Levels Between IF level can NOT be maintained using 300 Inches And 330 Inches Using turbine-driven AFW pump. THEN Turbine-Driven AFW Pump maintain SIG "lA" level using motor-driven AEW pump as follows:
- a. At N-01. place P-38A in - LOCAL.
- b. At N-01. depress ktart
- c. Open AF-4023.
- e. Inform DOS that "IB* SIG Atmospheric Steam Dump should be isolated.
lMS-2015 l
A5 Inform DOS That S/G Levels Are BETWEEN 300 INCHES AND 330 INCHES
-END-i
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 7 of 18 I I I I LsYJ II ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT 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 Turbine-Driven AFW Pump -
I B1 Check Turbine-Driven AFW Pump Direct PAB Operator to locally align V RUNNING steam supply to turbine-driven AFW pump:
e 2P-29 "oLocally open B SIG steam supply valve.
- 2MS-2019 "o Locally open A S/G steam supply valve.
- B2 Manually Control A S/G Level:
- a. Engage clutch and throttle *
- 2AF-4001 to maintain S/G level *
-BETWEEN 300 INCHES AND
- 330 INCHES
- e 2LI-460-AA on 2RK-38,
- a 2LI-460-BA on 2RK-38
- B3 Manually Control B S/G Level: *
- a. Engage clutch and throttle 2AF-4000 to maintain S/G level
-BETWEEN 300 INCHES AND 330 INCHES
- - 2LI-470-AA on 2RK-38 ,
- e 2LI-470-BA on 2RK-38 I
- POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 8 of 18 ISTEP It ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINEDI I ATTACHMENT B (Page 2 of 2)
Unit 2 AFW PUMP OPERATOR B4 Maintain S/G Levels Between IF level can NOT be maintained using turbine-driven AFW pump. THEN 300 Inches And 330 Inches Using maintain S/G "2B" level using Turbine-Driven AFW Pump motor-driven AFW pump as follows:
- a. At N-01. place P-38B in - LOCAL.
- b. At N-01. depress start
- - -- e--... - --- --
-'pushbdtt6h'.
- c. Open AF-4020.
- d. Throttle AF-45 to maintain "2B" S/G level - BETWEEN 300 INCHES AND 330 INCHES
"*2LI-460AA
"*2LI-460BA I
- e. Inform DOS that "2A" S/G Atmospheric Steam Dump should be isolated.
.B5 Inform DOS That S/G Levels Are BETWEEN 300 INCHES AND 330 INCHES
-END-L
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10 SAFETY RELATED ABNORMAL OPERATING PROCEDURE Revision 0 2/18/2002 Page 9 of 18 CONTROL ROOM INACCESSIBILITY RESPONSE NOT OBTAINED ii I ACTION/EXPECTED RESPONSE I I ATTACHMENT C (Page 1 of 2)
UNIT 1 CHARGING PUMP OPERkTOR C1 Ensure Charging Pump Suction Align charging pump suction to RWST:
ALIGNED TO RWST:
- OPEN I
1CV-358 e ICV-112B
- b. Shut VCT to charging pump suction
-suction MOV - SHUT
". 1CV-112C 11CV-112C CAUTION Placing pressurizer heaters in local defeats heater low level c*utout.
- C2 Check PZR Pressure - BETWEEN Locally operate back-up heaters to *
- - PI-449B
- o Bank C
. o Bank D
- AOP-lO POINT BEACH NUCLEAR PLANT AOP-10 SAFETY RELATED ABNORMAL OPERATING PROCEDURE Revision 0 2/18/2002 Page 10 of 18 CONTROL ROOM INACCESSIBILITY I I ISTE PII ACTION/EXPECTED RESPONSE I 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 a. WHEN PZR level greater than 16%.
THEN do Step C3.b.
16% AND RISING o LI-433C
- b. Establish !etdown: , r th :
. I
- 1) Inside 1B52-426M. locally open a) Operate charging pumps as necessary to maintain PZR letdown line isolation level - BETWEEN 20% AND 45%
lRC-427 I
01-15, CHARGING PUMP LOCAL CONTROL STATION OPERATION
- 2) Locally open letdown isolations as necessary b) Go to Step C5.
"o 1CV-200A "o 1CV-200B o 1CV-200C NOTES
- PZR level may require time to stabilize after letdown restoration.
- Charging pumps should remain running in auto/remote if possible.
Operate charging pumps as necessary *
- C4 Check PZR Level - BETWEEN 20% AND
- 25% to maintain PZR level- BETWEEN
, 20% AND 45%
@ TT-btir 0i 15. CHARGING PUMP LOCAL CONTROL
, STATION OPERATION C5 Inform DOS Of The Following:
"* PZR level
"*PZR pressure
-END-
POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 11 of 18 1ij I ACTION/EXPECTED RSESPONSE I I RESPONSE NOT OBTAINED I I
ATTACHMENT D (Page 1 of 2)
UNIT 2 CHARGING PUMP OPERATOR D1 Ensure Charging Pump Suction Align charging pump suction to RWST:
ALIGNED TO RWST:
- OPEN
-. ~b. Shut VCT to charginp PUmp suction VCT outlet to charging pump MOV.
'suction MOV - SHUT a 2CV-112C
- 2CV-112C CAUTION Placing pressurizer heaters in local defeats heater low level cutout.
- D2 Check PZR Pressure - BETWEEN Locally operate back-up heaters to *
- PI-449B -A
- "oBank C
- "oBank D
POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 12 of 18 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 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 PZR level greater than 16%.
THEN do Step D3.b.
16% AND RISING e LI-433C
- t. 1&tdon? b.-PerffCart1:e following:
- 1) Inside 2B52-427J. locally open a) Operate charging pumps as necessary to maintain Pzr letdown line isolation level - BETWEEN 20% AND 45%
- 2RC-427
- OI-15..CHARGING PUMP LOCAL CONTROL.STATION OPERATION
- 2) Locally open letdown isolations as necessary b) Go to SteD DS.
I o 2CV-200A o 2CV-200B o 2CV-200C NOTES
- PZR level may require time to stabilize after letdown restoration.
- Charging pumps should remain running in auto if possible.
- 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******************A D5 Inform DOS Of The Following:
"*PZR level
"*PZR pressure
-END-
POINT BEACH NUCLEAR PLANT AOP-10 POINT BEACH NUCLEAR PLANT SAFETY RELATED ABNORMAL OPERATING PROCEDURE Revision 0 2/18/2002 Page 13 of 18 CONTROL ROOM INACCESSIBILITY i j RESPONSE NOT OBTAINED I L TJI ACTION/EXPECTED RESPONSE I I ATTACHMENT E (Page I of 3)
TURBINE HALL OPERATOR NOTE All or some of the actions performed in this attachment may have been performed in the control room prior to evacuation.
- -4 Perfornrthe'follow.Lng:*....
Evacuating Control Room
- a. Open reactor trip breakers TRIPPED for BOTH UNITS.
- Unit I Reactor -
- Unit 2 Reactor - TRIPPED
- 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:
E2 Check Both Turbines Tripped Prior To Evacuating Control Room
- a. At Unit 1 Front Standard. rot*ate
- Unit 1 Turbine - TRIPPED trip~lever to trip position.
- Unit 2 Turbine - TRIPPED
- b. At Unit 2 Front Standard. rotate trip lever to trip position.
E3 Inform DOS Both Reactors And Both Turbines Are Tripped Locilly open breakers as necessary E4 At 1A01. Check The Following Pumps StoppLd Prior To Evacuating The per Attachment H. LOCAL BREAKER Control Room:
OPERATION,
"* IA52-05
- 1P-28A Steam.Generator Feed Pump
"* 1A52-08 SIP-27A Heater Drain Tank Pump
"*1A52-09 1P-27C Heater Drain Tank Pump 1
POINT BEACH NUCLEAR PLANT AOP-IO ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 14 of 18 RESPONSE NOT OBTAINED Lm I ACTION/EXPECTED RESPONSE I I ATTACHMENT E (Page 2 of 3)
TURBINE HALL OPERATOR E5 At IA02. Check The Following Pumps Locally open breakers as necessary Stopped Prior To Evacuating The per Attachment H. LOCAL BREAKER Control Room: OPERATION.
- IP-28B Steam Generator Feed Pump "*IA52-13 1lP-27B Heater Drain Tank Pump "*1A52-10 E6 At 2A01. Check The Following Pumps Locally open breakers as-necessary per iftachineri-H:LOCAI-BR=-R. ..
Control Room: OPERATION.
- 2P-28A Steam Generator Feed Pump "*2A52-23
- 2P-27A Heater Drain Tank Pump "*2A52-20
- 2P-27C Heater Drain.Tank Pump "* 2A52-19 E7 At 2A02. Check The Following Pumps Locally open breakers as necessary Stopped Prior To Evacuating The per Attachment H. LOCAL BREAKER Control Room: OPERATION.
- 2P-28B Steam Generator Feed Pump e 2A52-30
- 2P-27B Heater Drain Tank Pump - 2A52-33 E8 At 1A01 or IA02. Check One Unit 1 Locally open breaker per Attachment Condensate Pump Breaker - OPEN H. LOCAL BREAKER OPERATION.
o 1A52-07 for lP-25A OR o 1A52-11 for IP-25B E9 At 2A01 or 2A02. Check One Unit 2 Locally open breaker per Attachment Condensate Pump Breaker - OPEN H. LOCAL BREAKER OPERATION.
o 2A52-21 for 2P-25A OR o 2A52132 for 2P-25B EIO Inform DOS Of The Following:
- Main Feedwater pumps - STOPPED
- HeaTer Drain Tank pumps - STOPPED
- One Condensate pump - RUNNING
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10 ABNORMAL OPERATING PROCEDURE SAFETY RELATED' Revision 0 2/18/2002 CONTROL ROOM INACCESSIBILITY Page 15 of 18 is iiI _ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED l ATTACHMENT E (Page 3 of 3)
TURBINE HALL OPERATOR Eli Check CST Level - GREATER THAN 15 FT Direct Water Treatment Operator to commence filling CST.
- LI-4025 for T-24A
- LI-4031 for T-24B
-END-
- - t.4 . - r - -.-..- *.. - -.---.
I
POINT BEACH NUCLEAR PLANT AOP-10 POINT BEACH NUCLEAR PLANT SAFETY RELATED ABNORMAL OPERATING PROCEDURE Revision 0 2/18/2002 Page 16 of 18 CONTROL ROOM INACCESSIBILITY p p RESPONSE NOT OBTAINED LiT I ACTION/EXPECTED RESPONSE I I ATTACHMENT F (Page 1 of 1)
PAB OPERATOR F1 Check Unit 1 MSIVs Were Shut Prior Shut Unit I MSIVs as follows:
To Control Room Evacuation
- a. At IRK-33. depress both 1MS-2018 l for SIG 1A pushbuttons.
1MS-2017 for SIG 1B l
- INS PB-2017A. train A
- IMS. PB-2017B. train B F2 Check Unit 2 MSIVs Were Shut Prior Shut Unit 2 MSIVs as follows:
To Control Room Evacuation
- a. At 2RK-33, depress both
- 2MS-2018 for S/G 2A pushbuttons.
- 2MS-2017 for S/G 2B
"* 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 HOV's
- Open Bkr. 1B52-328F for AF-4023
- Open Bkr. IB52-428C for AF-4021 F4 De-Energize Unit 2 Motor-Driven AFW Pump Discharge HOV's
- Open Bkr. 2B52-328C for AF-4022
- Opeit Bkr. 2B52-428F for AF-4020 F5 Inform DOS Unit 1 And Unit 2 Motor-Driven AFW Pump Discharge MOV's - DE-ENERGIZED
-END-
AOP -10 POINT BEACH NUCLEAR PLANT SAFETY RELATED ABNOR1AL OPERATING PROCEDURE Revision 0 2/1812002 CONTROL ROOM INACCESSIBILITY Page 17 of 18 RESPONSE NOT OBTAINED I LE PJI ACTION/EXPECTED RESPONSE I II ATTACHMENT G (Page 1 of 1)
DOS CHECKLIST G1 Check the following Unit 1 actions completed:
COM) ?LETED Unit I ACTION PERFORMED
( Reactor tripped
( ) Motor-Driven AFW pump discharge valves breakers open
( ) PZR level - 20% to 45%
( ) PZR pressure - 2200 PSIG to 2250 PSIG
-,j..... ---*rb~u--rppta-, ..... ... .
- - .1,1. 11 . - .
C.
) Both Main Feedwater Pumps stopped
( ) All Heater Drain Pumps stopped *
( ) One Condensate Pump running
( 330 inches
) All S/G levels between 300 inches and G2 Check the following Unit 2 actiong completed:
COMPLETED Unit 2 ACTION PERFORMED
( ) Reactor tripped
- ':i ( ) Motor-Drlven AFW pump discharge valves breakers open
( ) PZR level - 20% to .45%
C ) PZR pressure - 2200 PSIG to 2250 PSIG
( ) Turbine tripped
( ) Both Main Feedwater Pumps stopped
( ) All Heater Drain Pumps stopped
( ) One Condensate Pump running C *) All S/G levels between 300 inches and 330 inches.
-END-
- POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT AOP-10 SAFETY RELATED ABNORMAL OPERATING PROCEDURE Revision 0 2/18/2002 Page 18 of 18 CONTROL ROOM INACCESSIBILITY ACTION/EXPECTED RESPONSE RESPONSE N*OT OBTAINED 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 "oDepress 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 0of A 480 Vac Breaker:
- a. Check mechanical indicator indicates closed.
- b. Depress square trip push button.
- c. Check mechanical indicator indicates open.
-END-