ML030870354
| ML030870354 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/06/2003 |
| From: | Kreil J Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2003-0094 | |
| Download: ML030870354 (186) | |
Text
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 25 12/14/2001 REACTOR TRIP RESPONSE Page I of 22 A. PURPOSE
- 1. This procedure provides directions to stabilize and control plant conditions following a reactor trip without safety injection.
- 2.
This procedure is applicable for specified entry conditions.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. This procedure is entered from the following procedure when SI is neither actuated nor required:
a EOP-O UNIT 1. REACTOR TRIP OR SAFETY INJECTION. Step 4 C. REFERENCES
- 1. Technical Specifications for Point Beach Nuclear Plant
- 2.
Final Safety Analysis Report for Point Beach Nuclear Plant
- 3.
As-built plant drawings
- 4.
Generic Technical Guidelines developed by the Westinghouse Owners Group (WOO).
This consists of the following documents:
- a. Low pressure version of the WOG Optimal Recovery Guidelines. Status Trees, and Functional Restoration Guidelines
- b. Background documents for each low pressure version Optimal Recovery Guideline. Status Tree. and Functional Restoration Guideline
- c.
WOG Emergency Response Guideline Executive Volume
- d. WOG Emergency Response Guideline Maintenance Program Summary
FOLDOUT PAGE FOR EOP-O.l UNIT 1
- 1.
SI ACTUATION CRITERIA IF either condition listed below occurs. THEN actuate SI and go to EOP-O UNIT 1. REACTOR TRIP OR SAFETY INJECTION, Step 4:
"o RCS subcooling based on core exit thermocouples - LESS THAN 351F OR "o PZR level -
CANNOT BE MAINTAINED GREATER THAN 10%
- 2.
EOP-0 UNIT 1. REACTOR TRIP OR SAFETY INJECTION, TRANSITION CRITERIA IF SI actuation occurs during this procedure.
THEN go to EOP-O UNIT I, REACTOR TRIP OR SAFETY INJECTION, Step 4.
- 3.
RCS TEMPERATURE CONTROL CRITERIA IF RCS temperatures are rising OR RCS cold leg temperature is greater than 547*F. THEN adjust feedwater flow and dump steam as necessary to stabilize RCS cold 16g temperature less than or equal to 547 0 F.
- 4.
AFW SUPPLY SWITCHOVER CRITERIA IF CST level lowers to less than 8 feet. THEN switch to alternate AFW suction supply per AOP-23 UNIT 1. ESTABLISHING ALTERNATE AFW SUCTION SUPPLY.
- 5.
AFW MINIMUM FLOW REQUIREMENTS IF any AFW pump mini-recirc valve fails shut. THEN maintain minimum 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-38B minimum flow -
GREATER THAN 50 GPM "o P-29 minimum flow -
GREATER THAN 75 GPM P.
I
Nuclear Poni er Business Unit TEMPORARY CHANGE REVIEW AND APPROVAL Note: Refer to NP 1 2.3, Temporary Procedu-e Changes, for requirements Page I of I - LNITLATION Doc Number EOP-O. 1 Current Rev 25 Unit PB. 1 Temp Chang-: No. 7
,. t, - D.,
DocumentTitle REACTOR TRIP RESPONSE Existing Effective Temporary Changes Brief Description MNODIFY FOP FOR MINIMUM AFWV FLOW TO INCLUDE LOW IA IiB&A PRESSURE (Identify specific e',anges on Form PBF-0026c, Document Review and Approval Conmnuation, and include.,th the package) 0] Initiate PBF-0026h and include with the change.
Other documents required to be effective concurrently with the temporary change:
Changes pre-screened according to NP 10.3.1?
0 NO E] YES (lfYes. h, rceref'ees and eMena on PBF-0.6cXrc to NP 103 1)
Scrccning completed according to NIP 10.3. 1?
[1 NA 0 YES Safety Evaluation Reauired? 0 NO DlYES Oft..
Ye,#ristn av be omsed ( ral rene' and eawro%-as ~lbe out-beroDe einkcmrw&tno Determine if the change constitutes a Change Of Intent to the procedure by evaluating the following questions (If any ans%-crs are YES, a rc ision may be processed or {rinal reviews and approvals shall be obtained before implementing)
Will the proposed change:
YES
- 1. Require a change to, affect or invalidate a requirement, commitment, e% aluation or El description in the Current Licensing Basis (as defined in N'P 10.3.1)?
NO ED
- 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 performance also being deleted?
- 5.
Change Tech Spec or other regulatory acceptance criteria other than for re-baselining purposes?
- 6.
Require a change to th rocedure Purpose or change the proc uc assftfca-n?
Initialed By (printsign)
If El 0
El 0
El ED El 0
Date -I21 /
II - INITIAL APPROVAL This change is correct and complete, can be performed as vritten, and does not Iversely affect personnel or nuclear safety, or Plant operating conditions.
Date Group Supervisor (prantrsign)
V~
V rA j
xr 57~i C '----IZ Dt (Cannot be the lrutiatar)
This change does not adversely affect Plant operating conditions (Safet. Rc.t p.cdures onl%)
Senior Reactor Operator (priwintas)
Ml 6* 0-s)
S] 1 (Cannot be thselrdtiator or Group Su Iper~isor)
A-Date jzo o!
III - PROCEDIOhE OWNER REVIEW ED Permanent El One-time Use
- E Expiration Date, Event or Condition:
0l4old change until procedure completed (final rcview and approval still required within 14 da) s of initial approval)
,Z QR/MSS Review NOT Required (Ad4rzn;.N.-SR only)
(] QR Review Requiredl-- MSS Review Required CRera-.*.N'P
- 6.5)
Procedure COv.ner (print/sign)
& /, 4
,r't.a' I
/i5,1705 Date
/
This Chanee an. suppr,oiting reouiren'enis correctlh co6rDleied and proc.ssed I
IV - FINAL REVIEW AND APPROVAL Musat be completed,ithin 14 dwis ofio~tial approal)
(The Initiator, QR and Approial.Authorit% shall be independent from each other)
QR,. ISS (print'sign)
At' b/,1 7 /A'1/4 I
Date / 7.h/k/6/
Indic2lcs 5o.59n2.4S applicabiliy assessed, ae. necessary screenmgs!.e-a.haat6Tnis pel(ooe d,,eterminawtn made as to,..heth:r addittorn!
cross-disciplinary-cev, rtiquired. and ifrcqwred, perfortred.
MSS Meeting NoDae/'
d Appro~al Authority (printtsign)
-- 2.
ý;Chtty Dat V -REVISION INFORMATION FOR PERMANENTt;CHANGES
/
Post Typingr Re~iC-.v(print/sign-)
e,.
/&
"U
6
/ /[, -
Dat
-~ ', //
Indicates lemporary change(s) mncorpoaictd eyactly as appro\\ ed and no od4.-ichangen
'ud'to.,cunt Incorporated into Revision Number
,9 Effective Date JAN 1 02 0D9,
~~ECD~
0A D.
120 PKr.0026e Revision 12 11 ()t;ZoQ I I
-W I I
I t*
ýJ I
I IkIvIVI
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Pa-ze
- - of I
--- 4--
I Other Comment,.
otc Rco:.,Frg ofStep.NurmKr !,.s nc~t :t.tuircd fer mul;ple oc,'.rer.zcs of iefcnlical tnforrnz.,!
net '--.r:rf.C'31 s¢*O%'L%%C.S I
i I
I Doec Numbe'-r EOP-0.1 Revision 25 Unit Title REACTOR TRIP RESPONSE Temporary Change Numbcr 7C.j joi - o'1 Description of Changes:
Step
- Change/Reason CHANGE: ADDED REFERENCE TO ANNUNCIATOR CO1 A 1-9. IA HEADER LOW PRESSURE FOR THE AFW MINIMUM FLOW REQUIREMENTS.
FOLDOUT REASON: TO ENSURE MINIMUM FLOW IS MAINTAINED THROUGH THE AFW PUNMPS PAGE DURING OPERATION.
This procedure change has been processed as follows: (Manual/Location)
Peo IPerformed EL Copy included in work package for field implementation. (WO No.
)
W]
Copy' filed in Control Room temp change binder (Operations oniv).
i,.
,.-0
[]
Original change package provided to
__to obtain Procedure Owner Review (e g, o,,er reviewmvey c oordnated by In-Group 0AlI. Procedure Writer, Pro -cdure Supervisor. ete) l El Performcd By (print and sign)
'*t..
\\ -
i*i_.-6
,'J-.-
Date f2
,)L "01 II - PROCEDURE OWNER REVIEW ON PBF-0026e (may be. performed by OA U, Procedure Writr, etc.)
I Date This procedure change has been processed as follows: (ivanual/Lccaticn)
Performed
[]
Copy sent to Document Control Distribution Lead for Master File.
I
-c (Not required for one-urne use change)
Se LI Copy filed in Group satellite file. (Not required for one-tine use changes)
El Copy filed in Group one-time use file.
[]
Original Temp Change pro% ided to b S to obtain Final Approvals
]
(c g. final appro',a] may be-coordiu.vt by In-Group OA II, Procedure Writer, Procedure Supcr'visor, etc)
)
II ElI Performed By (pi int and sign)
(r.
/
A "./_
-,e-.
Date
/2-2 PBF-0026h
,,f..
.,i Point Beach Nuclear Plant TEMPORARY CHANGE AFFECTED MLNUAL LOCATION Pave i
of Procedure Number EOP-0.1 Revision 25 Unit PB1 Title REACTOR TRIP RESPONSE Temporary ChangeNumber T".V 1_.0'/
I - IMMEDIATELY AFTER INITIAL APPROVAL ON PBF-0026e (Non-,*ntent chan*gs)
(after Fstual Approval if change oftntent involved)
Point Beach Nuclear Plant 10 CFR 50.59/72.48 SCREENING (NEW RULE)
,,zoo,.2 - Co*.
\\'cnf, SCR numbý en ae pag.s P_2e 1 TI
'roposed Activity:
ECPIARP Actions For AFW Mini-Recirc Flo% Requiremcnt Asso-a.ted Referenco(s) ".
Prcplrcd by:
Bob Wartenberg Date:
Name (Print) iyzature FcNevie'i cd by:
tL 4
1f4S iatr Date:
1 1
Namne (Print)
Sigý(ature PART I (50.59172.4$) - 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 Manuna) and NEI 964-7. 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.
I. 1 Dcscribc the proposed acivity and the scope of the activity being covered by this screening (Thc 10 CFR 50 59 1 72.48 revicw of other pordons of the proposed activity may be documented via the applicabilit% and pre-screenhig process requircencnts i NP 5.1.8.) Appropriate dcscripfive material may be attached.
EOP-0 and EOP-0.1 for both units revised the AFW Minimum Flow Requirements foldout-page criteria to include the instrument air header pressure low annunciator in the alarm state as additional criteria for monitoring ArWV mini-recirc flcw requirements.
AFW Minimum Flow Requirements foldout-page item was added to ECA-0.0. The foldout-page item is identical to the AFW mixumuin flow requirements used in EOP-0 and EOP--0.1.
Step 3.1 x%.as added to ARP CO1 A 1-9. The step states to monitor and maintain AFW mini-recire flow requirements should tie AFW punip mini-recire valve fail shui 1.2 Search the PBNP Current Licensing Basis (CLB) as follows: Final Safety Analysis Report (FSAR). FSAR Ch.angec Requests (rFCRs) with assipned numbers, the Fire Protection Evaluation Report (FPER), the CLB (Regulator.) Commitment Database.
the Technical Sp.zcifications (both Custom and Improved), the Technical Specifications Bases, and the Tczhmrcal Rteouirements Mlanual. Search the ISFSI licensing basis as follows: VSC-24 Safety" Analysis Report. the VSC-24 Certificate of Compliance. the CLB (Regulator.) Commitment Database. and the VSC-24 10 CFR 72.212 Site Evaluation Re.ort Describe the periir.ent design function(s), performance requirements, and methods of e aluation for both die plant and fcr the c.,slqSFSl as appropriate. Idendif" vhere the pertinent infornation is described in the atbove documents (by document section number "id title). (Resource Manual 5.3.1 and NEI 96-07, App. B, B.2)
FSA.R 10.2. Auaxflar.. Feedwater System.
1 3 Does the proposed aztivity involve a change to any Custom or Improv'd Techrucal Spt:cficauon (ITS)'* Ch...ngcs :o "Tccclhni.al Spccifications require a Licen.s Amendmeunt Request (Reszource Manual Sect'on 5 3.1 2).
"T'ccmic iI Si,:;ifizction Change :
[:)Yes [D No 1f a Techlic.I Spc,::ficaiion change is required. explain vihat tie chage should be ;and iih ' :. rzquitcd
Point Beach Nuclear Plant SCR -.
____-/_'_.
10 CFR 50.59172.48 SCREENI-NG (NEW RULE) verify SCRnumber on al page SPage 2
Docs the proposed activity invol'e a change to the terms, conditions or specifications incorpomrted in any VSC-24 cask Ccrrficatc of Compliance (CoC)? Changes to a VSC-24 cask Certificate of Compliance require a CoC amendment request.
0 Yes ED No If a storage cask Certificate of Compliance change is required, explain i hat the change should be and i hy 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 ansver the followving que.sions:
YES NO QUESTION ED El Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?
i]
0]
Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?
C3 0
Does the proposed activity involve SSCs whose failure could initiate a timisient (e.g., reactor trip, loss of fecedwAtcr, ctc.) or accident, OR whose failure could impact SSC(s) credited in the Safety Analyses?
[]
0 Does the proposed activity involve CLB-descnrbed SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications?
0]
Does the activity involve a method of evaluation described in the FSAR?
l 01 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 pioduct barrier (DBLFPBA?
(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 CYR 50.59 screening in the conclusion section (Part IV), then proceed dhiectly to Part V - 10 CFR 72.48 Pre-screening Questions.
If any of the abov.', questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.
FSAR 10.2 states each AFW pump has an AOV controlled recirc line back to the CST to ensure minimum flow to dissipate heat. This changec ensures the minimum AFNV flow requirements will be maintained on any running AF\\V pump in the case of a failed shut AFW mini-rccirc flow ccntrol valve PART ll1 (50.59) - DETERMINE WHETHER TILE ACTIVITY IINVOLVES ADVERSE EFFECTS (Resource Manual 5.3.3)
If ALl, the questions in Part II are ansviered NO, then Part III is El NOT.APPLICABLE.
Answer the following questions to detennine if the activity has an adverse effect on a design function Any YES answ\\er means that a 10 CFR 50 59 E\\,iluation is required; EXCEPT where noted in PaTi I1I.3.
111.1 Cl-IANGES TO THE FACILITY OR PPOCEDUbRES YES NO QUESTION El Z
Does the activity adversely affect the designfunctirwr of an SSC credited in safety analyses?
PBF-151c
ý "
N-11 '
I T
,, o tGCm5 Point Beach Nuclear Plant SCR,A-
- ],,f 10 CFR 50.59/72.48 SCREENING (NEW RULE) vrifys,-- rumbe onall pages
- &&-?
1-4-OA Page 3 I]
[
Does the activity adversely affect the mc-iod of perfcrming 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 anvwers are NO describe the basis for the conclusion (attach additional discussion as necessary):
This change ensures that minimum re-irc flow requiremen:- as stated in FSAR 10.2 are not violated.
111.2 CHANGES TO A METHOD OF EVALUATION (If the activity does not involve a method of evaluation, these questions are Z NOT APPLICABLE.)
YES NO QUESTION El El Does the activity use a revised or different method of evaluation for performing safety analysas than that described in the CLB?
0l
[]
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 answcr 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 EXTER.IMENTS If the activity is not a test or experiment, the questions in Ifl.3.a and I11.3.b are [9 NOT A.PPLICABLE.
- a. Answer these two questions first:
YES NO QUESTION
[I
[I Is the proposed test or ex-peri mert bounded by other tests or experiments that are described in the CLB?
0 M"
Are the SSCs affected by the proposed test or experiment isolated from the facility?
If the answer to B OTH questions in V.3.a is NO, continue to IIL3.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 1l1.3.a above.
If the ans-\\er to either question in 11L3.a is YES, then these three questions are El NOT APPI ICABLE.
YES NO QUESTION El El Does the activity utilize or control an S.C in a manner that is outside the reference bounds cf the design bascs as described in the CI.B?
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 Ml Does the activity place the facility in a condition not previously c, aluated or that could affect the ca.pabihty of an SSC to perform its intended functions?
If any answer in 111.3.b is YES, a 10 CFR 50.59 Exaluaiion is required. If the ansN'ers in 111.3 b are ALL NO, describe the basis for the conclusion (attach additional discussion as necessary):
PBF-151.c
- 1
-1 0
Point Beach Nuclear Plant SCR 10 CFR 50.59172.48 SCREENING (NEW RULE) 4.fY SCR =,=
P on ai p4,
=
A-4 _a Pag e 4 1-.
. - 10 CFR 50.59 SCREENLNG CONCLUSION (Rcsource Manual 5.3.4).
Check all that appl%.
A 10 CFR 50.59 Evaluation is U required or Z NOT required.
A Point Beach FSAR change is E] required or [E NOT required. If an FSAR change is required. then initiate an FSAR ClMneC Request (FCR) pcr NP 5.2.6.
A Regalatory Commitment (CLB Commitment Database) change is U] required or [Z NOT required. If a Regulatory Commitment Changre is required. initiate a commitment change per NP 5.1.7.
A Technical Specification Bases change is El required or Z NOT required If a change to the Technical Specification Bases is required, thln initiate a Technical Specification Bases change per NP 5.2.15.
A Technical Requirements Manual change is [I required or [D NOT required. If a change to the Technical Requirements Manual is requircd, then initiate a Technical Requiremcnts Manual change per NP 5.2.15.
S -------
10 CFR 72.48 SCREENING.
NOTE: NE! 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 "V determines if a full 10 CFR 72.48 screening is required to be completed (Parts VI and V\\1) for the proposed activity.
YLS NO QUESTION U1 ID Does the proposed activity involve IN ANY MANNER the dry fuel storage cask(s), the cask trarnsfer/transpon cwuipment, any ISFSI facility SSC(s), or any iSFSi facility monitoring as follows: Multi-Assembly Sealed Basket (M%1SB), 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 PPCSISFSI Continuous Temperature Monitoring System?
U
[]
Does the proposed activity involve IN ANY MANNER SSC(s) installed in the plant specifically added to support
,ask ioading/unloading activities, as follows: Cask Dcwatering System (CDW), Cask Reflood System (CRE), or HN drogen Monitoring System?
U
[
Does the proposed activity in-,olve 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 (VNIPAB), Drumming Area Ventilation S) stem (VNIDRM),
RE-10.: (SFP Low Range Monitor). RE-135 (SFP High Range Monitor). RE-221 (Dnnuring Area Vent Gas Monitor), RE-325 (Dimurrning Area E.haust Low-Range Gas Monitor), PAB Crane, SFP Platform Bridge.
Truck Access Area. or Decon Area?
U
[
Does the proposed activin in% oh'e a change to Point Beach CLB design criteria for external evcnts such as earthqutkes, tornadoes. hi' ghwinds. f.oodinm eta ?'
U
[
Does the activity invoh e plant h]ea.
load requirements or procedures for areas of the p!ant used to support cask loadtiegtmloading activities'9 U
[
Does the a:tiity inlohle an. picitial for fire or explosion %\\iere casks we loaded. unloaded, transported or stored?
"N of the Pxt V questions arc answered ES. ilicna full 10 C
,R72.-,S screening is required and answcrs to the questiszis in i-.... VI and Par Vli :,re to be provided If ALL the questions in Part V are answered NO. then cheek Pars \\i,'nd VrI as not applicable Complc:c Pan VIll to document the conclusiou tlht no 10 CFR 72.48 evaluation is reqtured.
,.2.oa) -:,- -,5 Point Beach Nuclear Plant SCR 10 CFR 50.59172.48 SCREENING (NEW RULE)
V5 YSCR.=ber on ail pae:
J/--,A-Page5 TT VI (72.48) - DETERMiNE IF TIE CHANGE INVOLVES A ISFSI LICENSING BASIS LESIGN FU7VCTION (L.,L the questions in Part V are NO th-n Part VI is 0 NOT APPLICABLE.)
Compare the proposed activity to the relevant portions of the ISFSI licensing basis and ansi er the following questions:
YES NO QUESTION El
[i Does the proposed activity involve cask/ISFSI Safety Analyses or plant/cask/lSFSI structures, sys'e,"as and components (SSCs) credited in the Safety Analyses?
r-El Does the proposed activity involve plant, cask or ISFSI SSCs that support SSC(s) credited in the Safety Analyses?
El El Does the proposed activity involvh plant, cask or ISFSI SSCs Nihose function is relied upon for prevcntion of a radioactive release, OR whose failure could impact SSC(s) credited in the Safety Analyses?
El El Does the proposed activity involve cask/ISFSI described SSCs or procedural controls that perform functions that are required by, or othenrise necessary to comply with, regulations, license conditions, CoC conditions, or orders?
El El Does the activity involve a method ofjevaluation described in the ISFSI licensing basis?
E I]
Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)
]
El Does thz activity exceed or potentially affect a cask design basis limit for afissioti product barrier (DBLFPB)?
(NOTE: If TIlS questions is ansvercd YES, a 10 CFR 72.48 Evaluation is required.)
If thilic.t.nwers to ALL of these questions are NO, -mark Parts VII as not applicable, and document the 10 CFR 72.48 screening in the conc' -us'on section (Part VIII).
"- ny of the above questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) ved.
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 zie answered NO, then Part VII is Z NOT APPLICABLE.)
Answer the following questions to determine if the activity has an adherse effect on a design function. Any YES answr means that a 10 CFR 72.48 Evaluation is required; EXCEPT where noted in Part VIL3.
VII.1 Changes to the Facility or Procedures YES NO QUESTION
[]
El Does the activity adversely affect the design unction of a plant, cask, or ISFSI SSC credited in safety analyses?
0I El Does the activity adversely affect the method of performing or controlling the desqgn function of a plant.
cask. or ISFSI SSC credited in the safety analyses?
If any answer is YES, a 10 CFR 72 -if, Evalmution is reqaired. If both ansvwc~s are NO, describe t'he basis for the conclusion (attach additional discussion, as nc.ssar'):
?13F.ISlSc
,snO 1;V24'O!
Refcrcn:e N-5 1.3
Point Beach Nuclear Plant 5CR 4
10 CFR 50.59/72.48 SCREENING (N-EW RULE)
V'.s sc r'-=b.- on an p*.c
-.2 Changcs tw a Method of Evaluation (If the acfi:, does not :avolve a method of cvaluation. these questions are Li NOT APPLICA3LE)
YES NO QUESTION
[
Li Does the activity use a revised or different r1:cthod of evaluation for performing safety atZ1N ses than Ltha described in a cask SAR?
fl Li Docs the actirivt use a revised or different method of evaluation for eve'uating SSCs creditede in saJfety anal% ses _han that described in a cask SAR?
If an. ainsx*c is YES. a 10 CFR 72.48 Evalhation is required. if both an-,N ers are NO. describe the basis for tfr. conclusion (attach additional discussion, as necessar ):
VX. 3 Tests or Experiments (If the activity is not a test or experiment, the questions in VIL3.a and \\qL3.b arc Li NOT APPLICABLE.)
a Answer these tv' questions first:
YES NO QUESTION El ED Is the proposed test or ex-periment bounded by othcr tests or expcriLnents that are described in the cask ISFSI licensing basis?
Li Li Are the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISFSI facility?
If the ansscr to both questions is NOL continue to \\r1.3.b. If the ains erto EITHER question is YES. then briefly describe the basis.
- b..Answcr these additional questions ONLY for tests or e.perLmenes xkich do not meeit the criteria givcn in \\I.3.a aLo, e.
If the ans%% er to either question in VII.3.a is YES, then these three quastions arc [3 NOT APPLICABLE:
YES NO QUESTION Li Does the activit utilize or control an SSC in a mantier that is out-=ide the referenme boun-S of the design bases as described in the ISFSI licensing basis?
Li L
Does the activity utilize or control a plant, cask or ISFSI facihr. SSC in a m.'amer that is monsis'ent%;t i-h the analyses or descriptions in the ISFSI licensing basis?
Li F"'1 Does the acuvity place thle cask or ISFSI facility iii a condition no; -,reio,!y c% alu.ted cr !hLt Could affict the capability of a plant. cask. or ISFSI SSC to perform its intende5d functions9 If nn.
i;mns cr in Vi 3 is YES. a 10 CFR 72.4-E% aluation is required If the pSwc.s are,all NO. de-ib asis fort.:
cceiclusin' (attach addit~onal discussion as necessan T
-..:."Z :.. i'.
!0:4' (M
Point Beach Nuclear Plant 10 CFR 50.59/72.48 SCREENUNG (NEW RULE)
VmrNf" SCR number on all pages J-4-C'- Page"7
- RT VIII - DOCUMENT THE CONCLUSION OF THE 10 CFR 7?2.48 SCREENING Check all that apply:
A 10 CFR 7T..48 Evaluation is El requircd or Z NOT required-Obtain a screning number and proaide the original to Records Management regardl-ss of the "onclusion of t&.; 50.59 or 72.48 screening A V'SC-U4 cask Safcty Analysis Report change is tj required or [D NOT rcquired. If a VSC-24 cask SAR change is rcquirca, then contact the Point Beazh Dr, Fuel Storage group upervsor.
A Rcgulatory Commitment (C13 Comr,,itment Database) change is 0 required or [0 NOT required. Lfa Regulatory Commitment Ch-mnge is required, initiate a commitment change per NP 5.1.7.
A change to the VSC-24 10 CFR 72.212 Site EDtauation Report is E] required or [E NOT required. If a VSC-24 10 CFR 72.212 Site Evaluation Report change is required, then contact the Point Beach Dry Fuel Storage group superv-sor.
PB3F-1515c
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 1 of 22 A.
PURPOSE
- 1. This procedure provides directions to stabilize and control plant conditions following a reactor trip without safety injection.
- 2.
This procedure is applicable for specified entry conditions.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. This procedure is entered from the following procedure when SI is neither actuated nor required:
- EOP-O UNIT 1. REACTOR TRIP OR SAFETY INJECTION. Step 4 C. REFERENCES
- 1. Technical Specifications for Point Beach Nuclear Plant
- 2.
Final Safety Analysis Report for Point Beach Nuclear Plant
- 3.
As-built plant drawings
- 4.
Generic Technical Guidelines developed by the Westinghouse Owners Group (WOG).
This consists of the following documents:
- a.
Low pressure version of the WOG Optimal Recovery Guidelines. Status Trees. and Functional Restoration Guidelines
- b. Background documents for each low pressure version Optimal Recovery Guideline. Status Tree. and Functional Restoration Gdideline
- c.
WOG Emergency Response Guideline Executive Volume
- d.
WOG Emergency Response Guideline Maintenance Program Summary
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 2 of 22 ACTION/EXPECTED RESPONSE
-RESPONSE NOT OBTAINED NOTE Foldout page shall be monitored throughout this procedure.
I Verify RCS Temperature Control:
Perform the following:
- a. Check RCS wide range cold leg
- 1. IF RCS cold leg temperature less temperatures:
than 5471F AND RCS temperatures
-are trending lower. THEN
ANDAD a) Stop dumping steam.
a STABLE b) Ensure S/G blowdown isolations
-SHUT c) IF cooldown continues. THEN control feed flow:
- 1) Reduce total feed flow.
- 2) Maintain total feed flow greater than or equal to 200 gpm until level greatex than 29% in at least one SIG.
d) IF cooldown can NOT be stopped by controlling feed flow. THEN isolate steam lines:
- 1) Shut both main steam isolation valves.
M114S-2018 for S/G A
- 1MS-2017 for S/G B
- 2) Ensure main steam isolation bypass valves -'BOTH SHUT lMS-234 for S/G A
- MS-236 for SIG B
- 2.
IF RCS cold leg temperature greater than 5470F OR RCS temperature trending higher. THEN stabilize RCS temperature at or below 547°F as follows:
o Dump steam to condenser.
ORR o Dump steam using atmospheric "steam dumps.
POINT BEACH NUCLEAR PLANT EOP-O.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 3 of 22 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 2
Verify Feedwater Isolation:
- a. Check RCS average temperature LESS THAN 554 0 F
- b. Check feedwater regulating control valves - BOTH SHUT
" JCS-466 for S/G A
"* 1CS-476"for S/G B
- a.
WHEN RCS average temperature is less than 554*F.
THEN do Step 2.b.
OBSERVE CAUTION AND NOTE PRIOR TO STEP 3 and continue with Step 3.
- b. Shift controllers to manual and shut valves.
"* lfC-466. train A reg valve
"* 1RC-476.
train B reg valve OCAUTION FOn nattural circulation. RTD bypass manifold temperatures are~not accurate.
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 4 of 22 III I
I r
....... t I
ACTION/EXPE.CTD
)LA4tiz I
CAUTION If motor-driven auxiliary feedwater pump flow is greater than 240 gpm.
its motor *breaker may trip due to over current.
NOTE If both units require AFW flow. at least one AFW pump must be aligned to each unit.
3 Transfer Feedwater Control To Bypass Regulating Valves:
- a. Check main feedwater pumps - AT LEAST ONE RUNNING
- b. Control feedwater flow using regulating bypass valves.
o ICS-480. train A e ICS-481. train B
- c. Maintain both S/G levels greater than 25%
- d. Reset Loss Trip of Feedwater Turbine
- e. Secure any running AFW pumps IF S/G levels can NOT be maintained using main feed bypass..THEN establish AFW flow:
- 1. Ensure motor-driven AFW pumps BOTH RUNNING
"* P-38A. train A
"* P-38B. train B
- 2. Verify AFW valve alignment:
a) Ensure Unit 1 valves OPEN
"* AF-4023. train A
"* AF-4021. train B b) Ensure Unit 2 valves SHUT
"* AF-4022. train A
"* AF-4020..train B c) Ensure motor-driven AFW pump discharge control valves MAINTAINING DISCHARGE PRESSURE AT 1200 PSIG
"* AF-4012. train A
"* AF-4019. train B
- 3.
IF both S/G levels are less than 25% OR motor-driven AFW pumps are NOT available. THEN ensure steam supply valves to turbine-driven AFW pump are both open.
"* 1MS-2020. train A
"* lMS-2019. train B I
RESPONSE NQ1 L)BIAINED
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 5 of 22 TE ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAIN4ED 4
Stabilize SIG Levels:
1*
- a. Check S/G levels - GREATER THAN
- a. Maintain total feed flow greater 29%
than 200 gpm until level greater than 29% in at least one S/G.
- b. Control feed flow to maintain S/G
- b. IF level in intact SIG continues levels between 29% and 65%
to rise. THEN stop feed flow to that SIG.
Verify Generator Trip:
- a. Check main-generator and exciter
- a. Open main generator and exciter field breaker - OPEN field breakers.
- b. Check Auto Bus Transfer -
- b. Manually transfer power to COMPLETE transformer IX-04:
- 2) Close IA-O to IA-03 and 1A-02 to IA-04 tie breakers.
e IA52-37. train A
- IA52-55. train B 5
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 6 of 22 SI
~
i i
I L SEPiI ACTION/EXPECTED RESPONSE I
6 Ensure Miscellaneous Electrical Loads Are Energized:
- a. Ensure MCCs ENERGIZED
- 1B-31. IB52-14C. train A a B-43. IB52-21C. train B
- b. Check battery chargers supplying DC buses -
ENERGIZED "o D-07 "o D-09 "o D-108 "o D-109 RESPONSE NOT OBTAINED
- b. Restore battery dhargers:
- 1) Close affected battery charger supply contactor.
- 2) IF contactor does NOT close OR battery charger will NOT operate. THEN restore battery chargers per AOP-0.0. VITAL DC SYSTEM MALFUNCTION. while continuing with this procedure.
- c. Ensure cavity cooling fan -
ONE RUNNING
.o IW'4A. train A o IW-4B. train A
- d.
Check cable spreading room ventilation operating:
- 1) Check cable spreading room recirc fans -
ONE RUNNING "o W-13A1 "o W-13A2
- 2) Check CSR chilled water recirc pumps -
ONE RUNNING "o P-1lIA "o P-111B
- e.
Start additional loads as necessary to meet current plant conditions.
Refer to AOP-22 UNIT 1. EDG LOAD MANAGEMENT
- d. Restore cable spreading room ventilation per 01-90.
CONTROL.
COMPUTER.
AND CABLE SPREADING ROOM VENTILATION SYSTEMS.
I r
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT I EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 7 of 22 1 ACTION/EXPECTED R*ESPONSE Check Control Room Fans Armed:
- a. Check Control Room Charcoal Filter Fan W-14A - WHITE LIGHT OFF RESPONSE NOT OBTAINED
- a. At MCC IB-32. depress Control Circuit Arming pushbutton for Control Room charcoal filter fan W-14A.
- b. Check Control Room Recire Fan W-13B2 -
WHITE LIGHT OFF
- b. At MCC IB-42. depress Control Circuit Arming pushbutton for Control Room recirc fan W-13B2.
Check All AC Buses - ENERGIZED BY OFFSITE POWER a
S a
S 0
0 6
1A-03.
IA-05.
IB-01.
1A-04.
IA-06o IB-02.
train train train train train train train train train train Restore offsite power to all AC buses while continuing with this procedure.
A A
A A
A B
B B
B B
7 I ACTION/EXPECTED
RESPONSE
I I
I I
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 8 of 22 STP ACTION/EXPECTED RESPONSE L
RESPONSE NOT OBTAINED 9
Verify Charging Pump Suction Alignment:
- a.
Check VCT level " GREATER THAN
- a. Perform the following:
17%
SHUT
- 3)
WHEN VCT level is greater than 17%.
THEN do Steps 9.b and 9.c.
Continue with Step
- 10.
ICV-112C
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 9 of 22 I
I L~jI, ACTION/EXPECTED RESPONSE 10 Verify Charging Flow:
I RESPONSE NOT OBTAINED
- a. Ensure RCS Loop A Cold Leg Normal Charging Isolation Valve -
OPEN
- b.
Check charging ONE RUNNING "o IP-2A. train "o 1P-2B. train "o 1P-2C. train pumps.-
AT LEAST A
A B
- b. Perform the following:
- 1) IF component cooling water flow to any RCP thermal barrier is lost. THEN locally shut affected RCP(s) seal injection throttle valve before starting charging pumps.
"o 1CV-300A.
- 2) Start charging pumps as necessary to establish at least one running.
- c. Start additional charging pumps and adjust speed on running charging pumps as necessary to establish desired charging flow
- d. Adjust charging line flow controller as necessary to maintain labyrinth seal AP greater than 20 inches
- IHC-142 I
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 10 of 22 STP ACTION/EXPECTED RESPONSE L
RESPONSE NOT OBTAINED 11 Check All Control Rods FULLY IF two or more control rods NOT INSERTED fully inserted, THEN initiate emergency boration as follows:
- a. Record initial level for in service BAST
- b. Start one boric acid transfer pump.
- c. Establish maximum charging flow:
- 1) Fully open charging flow control valve.
1 IHC-142
- 2) Start additional charging pumps.
- IP-2A. train A 9 IP-2B. train A a IP-2C, train B
- 3) Adjusting charging pump speed as necessary to maintain charging flow less than 140 gpm.
- d. Open emergency borate valve.
- ICV-350
- e. Borate 1200 gallons for each control rod not fully inserted.
Use TLB-5. BORIC ACID STORAGE TANKS, to determine BAST level change.
- f.
IF emergency boration can NOT be established. THEN perform boration per OP-5B. BLENDER OPERATION/DILUTION/BORATION.
while continuing with this procedure.
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 11 of 22 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 12 Check PZR Level GREATER THAN 12%
Perform the following:
- a. Ensure all PZR heaters OFF
- IT-lA e iT-lB e IT-IC a iT-ID a IT-lE
- b. Ensure ill letdown isolation valves -
SHUT
"* 1CV-427 e ICV-200A
"* ICV-200B
"* ICV-200C
- c. Ensure excesr letdown isolation valve -
SHUT I
lCV-1299
- d. Control charging to restore PZR level to greater than 12%.
- e. DO NOT CONTINUE until PZR level is greater than 12%.
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT I EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 12 of 22 STEP ACTION/EXPECTED RESPONSE PRESPONSE NOT OBTAINED 13 Check If Letdown Should Be Established:
- a.
Check normal letdown - ISOLATED
- b. Establish letdown:
- 1) Open letdown line isolation valves containment
"* ICV-371A
"* ICV-371
- 3) Ensure component cooling flow to non-regenerative heat exchanger - ESTABLISHED a 1HC-130
- 4) Ensure charging flow - AT LEAST 20 GPM
- 5) Adjust backpressure as necessary and open letdown isolation valves to establish desired letdown flow "o 1CV-200A "o iCV-200B "o 1CV-200C 14 Check PZR Level -
BETWEEN 20% AND 30%
- a.
Go to Step 14.
- b. Perform the following:
a) Establish excess letdown per OP-5E. ESTABLISHING AND SECURING EXCESS LETDOWN.
b) IF excess letdown can NOT be established. THEN control charging as necessary to maintain PZR level.
Control charging and letdown to maintain PZR level between 20% and 30%.
POINT BEACH NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE REACTOR TRIP RESPONSE EOP-0.1 UNIT 1 EOP-0.1 UNIT 1 SAFETY RELATED Revision 26 1/10/2002 Page 13 of 22 ACTION/EXPECTED RESPONSE Check RCS Pressure -"GREATER THAN 1735 PSIG 16 Check PORVs And PORV Block Valves:
- b. Check PZR PORVs BOTH SHUT
"* IRC-430
"* 1RC-431C
- c.
Check PORV block valves -
OPEN
"* IRC-515 for 1RC-431C
"* 1RC-516 for IRC-430 RESPONSE NOT OBTAINED Perform the following:
- b.
Go to EOP-0 UNIT 1. REACTOR TRIP OR SAFETY INJECTION.
- a. Reduce RCS pressure:
- 1) Ensure at least one "PZR PORV and its associated block valve
- OPEN
- 2)
WHEN RCS pressure is less than 2335 psig. THENN do Steps 16.b and 16.c.
Continue with Step 17.
- b. Perform the following:
- 1) Manually shut affected PORVs.
- 2)
IF any PORV can NOT be shut.
THEN manually shut associated block valve.
o 1RC-515 for IRC-431C o 1RC-516 for 1RC-430
- c.
IF associated PORV is shut. THEN open PORV block valve(s).
15
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 14 of 22
-EP IACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 17 Check PZR Spray Valves:
- a. Check RCS pressure
- LESS THAN 2260 PSIG
- b. Place PZR.pressure controller to manual, adjust setpoint to desired pressure. and then return to auto
- c. Check PZR spray valves -
SHUT
"* IRC-431A. loop A
"* IRC-431B. loop B
- d.
Check auxiliary spray valve SHUT
- ICV-296
- a. Reduce RCS pressure:
- 1) Ensure at least one PZR spray valve -
OPEN
- 2) WHEN RCS pressure is less than 2260 psig. THEN do Steps 17.b.
17.c and 17.d.
Continue with Step 18.
- c. Perform the following:
- 1) Manually shut PZR spray valves.
- 2) IF any spray valve can NOT be shut. THEN place manual override switch to close for failed spray valve(s).
"o lRC-431A-S for IRC-431A "o 1RC-431B-S for 1RC-431B
- 3) IF any spray valve can NOT be shut using manual override.
THEN stop RCP supplying failed spray valve(s).
o For IRC-431A. stop RCP A o For IRC-431B. stop RCP B
- d. Perform the following:
- 1) Manually shut auxiliary spray valve.
- 2) IF auxiliary spray valve can NOT be shut. THEN minimize
.charging and shut charging line flow control valve.
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 15 of 22 I
I L TJI ACTION/EXPECTED RESPONSE 1
RESPONSE NOT OBTAINED 18 Check PZR Pressure Control:
- a. Energize PZR heaters as necessary to saturate PZR 0
0 0
0 0
IT-1A IT-lB IT-IC IT-ID IT-lE
- b. Operate PZR heaters as necessary to stabilize PZR pressure at 2235 psig.
- c. Use the following:
- 1) IF letdown is in service. THEN use auxiliary spray.
I ICV-296
- 2) IF auxiliary spray can NOT be used. THEN use one PORV.
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE REACTOR TRIP RESPONSE EOP-0.1 UNIT 1 SAFETY RELATED Revision 26 1/10/2002 Page 16 of 22 ACTION/ExPEcTED RESPONSE 19 Check Reactor Makeup Control:
- a. Check makeup set for desired boron concentration
"* 1HC-110
"* IHC-Il1
- b. Ensure makeup armed and in auto
- c.
Check VCT level -
GREATER THAN 17%
20 Monitor Turbine Generator Coastdown:
- a. Ensure HP seal oil backup pump RUNNING
- IP-37B
- b. Ensure turning gear lube oil pump RUNNING
- IP-37C
- c.
Ensure HP turbine drains -
OPEN I
I1 RESPONSE NOT OBTAINED
- a. Adjust controls as necessary to establish makeup at desired boron concentration.
- c. Redirect makeup to VCT inlet:
- 1) Open boric acid blender to VCT.
- iCV-IhOC
- 2) Shut boric acid blender to VCT flow control.
a 1CV-11OB
- b.
IF turning gear lube oil pump can NOT be started. THEN ensure DC pump running.
- IP-37D J
i
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 17 of 22 ACTION/EXPECTED RESPONSE 21 Establish SIG Pressure Control:
- a. Check MSIVs - ANY OPEN o lMS-2018 for S/G A o lMS-2017 for SiG B
- b. Ensure condenser available:
a Check circ pumps -
AT LEAST ONE RUNNING
- Check condenser vacuum GREATER THAN 22 INCHES Hg
- d. Place Steam Dump Mode Selector switch in Manual 22 Check RCS Wide Range Hot Leg Temperatures -
STABLE RESPONSE NOT OBTAINED IF condenser steam dump is NOT available.
THEN maintain current S/G pressures using atmospheric steam dumps.
"* lMS-2016 for S/G A
"* lMS-2015 for SIG B Control flow as hot leg steam dump and total feed necessary to stabilize RCS temperatures.
23 Check If An RCP Should Be Started:
- a.
Check RCPs - BOTH STOPPED
- a. IF any RCP is running. THEN go to.
Step 26.
CAUTION If RCP seal cooling had previously been lost. affected RCPs should not be started prior to a status evaluation.
I
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE REACTOR TRIP RESPONSE EOP-0.1 UNIT 1 SAFETY RELATED Revision 26 1/10/2002 Page 18 of 22 ACTION/EXPECTED RESPONSE Jýi 24 I
I
- RESPONSE NOT OBTAINED Establish Conditions For Starting Desired RCP:
"* Ensure power to RCP -
AVAILABLE o IA-01 for RCP A o 1A-02 for RCP B
"* Ensure RCP thermal barrier CCW flow -
NORMAL
"* Ensure number I seal AP -
GREATER THAN 200 PSID
"* Ensure number I seal leakoff flow
- WITHIN NORMAL OPERATING RANGE OF FIGURE 1
"* Ensure RCP labyrinth seal AP GREATER THAN 20 INCHES
"* Ensure VCT pressure GREATER THAN 15 PSIG
"* Start RCP oil lift pump i
J Try to establish starting conditions.
IF starting conditions can NOT be established for any RCP.
THEN perform the following:
- a. Verify natural circulation:
a RCS subcooling based on core exit thermocouples - GREATER THAN 350 F
- Core exit thermocouples STABLE OR SLOWLY TRENDING LOWER
- S/G pressures - STABLE OR SLOWLY TRENDING LOWER
- RCS cold leg temperatures NEAR SATURATION TEMPERATURE FOR S/G PRESSURE
- b.
IF natural circulation can NOT be verified. THEN raise steam flow.
- c. Stop RCP oil lift pumps.
- d.
Go to Step 26.
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 19 of 22
]sTEP I ACTION/EXPECTED RESPONSE 1
25 Establish Forced Circulation:
- a. IF all RCPs are stopped. THEN perform the following:
- 1) Verify natural circulation:
- RCS subcooling based on core exit thermocouples -
GREATER THAN 350 F a RCS hot leg temperatures STABLE OR SLOWLY TRENDING LOWER
- Core exit STABLE OR LOWER thermocouples SLOWLY TRENDING
"* S/G pressures - STABLE OR SLOWLY TRENDING LOWER
"* RCS cold leg temperatures NEAR SATURATION TEMPERATURE FOR S/G PRESSURE
- 2) IF natural circulation can NOT be verified. THEN raise steam flow.
- b. Stop RCP oil lift pumps 26 Check If Source Range Detectors Should Be Energized:
- a. Check intermediate range flux LESS THAN 1.5x10"I' AMPS
" 1N-35. train A
" 1N-36.
train B
- b.
Check source range detectors BOTH ENERGIZED o 1N-31.
train A 0 1N-32. train B
- c. Transfer both pens of NR-45 recorder to source range scale-
- a. WHEN intermediate range flux is less than l.5x10" 10 amps.
THEN do Steps 26.b and 26.c.
Continue with Step 27.
- b. Depress both intermediate range permissive defeat push-buttons.
A B
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 1 EMERGENCY OPERATING.PROCEDURE SAFETY RELATED Revision 26 1/10/2002 REACTOR TRIP RESPONSE Page 20 of 22
.I I
I iSTEP 1 27 28 Shut Down Unnecessary Piant Equipment 29 Check If Secondary Systems Should Be Secured:
- a. Check secondary steam systems NO LONGER REQUIRED Start or stop pumps and align valves as necessary to establish desired configuration.
- a.
Go to Step 30.
- b. Secure secondary steam systems per OP-13B.
SECONDARY SYSTEMS SHUTDOWN.
while continuing with this procedure 30 Maintain Stable Plant Conditions:
"* PZR pressure -
BETWEEN 2200 PSIG AND 2300 PSIG
"* PZR level -
BETWEEN 20% AND 30%
"* S/G levels " BETWEEN 29% AND 65%
"* RCS cold leg temperatures LESS THAN PR EQUAL TO 547OF I
ACTION/EXPECTED RESPONSE I
Stabilize Secondary Plant:
- a. Ensure circulating water pump(s)
AT LEAST ONE RUNNING "o lP-30A. train A "o IP-30B. train B
- b. Ensure S/G feed pumps -
ONE RUNNING "o lP-28A "o IP-28B
- c.
Ensure heater drain pumps -
ONE RUNNING "o IP-27A "o IP-27B "o lP-27C
- d. Ensure condensate pumps ONE RUNNING "o 1P-25A "o IP-25B I
I RESPONSE NOT OBT£AINE:D
POINT BEACH NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE REACTOR TRIP RESPONSE EOP-O.1 UNIT 1 EOP-0.1 UNIT I SAFETY RELATED Revision 26 1/10/2002 Page 21 of 22 ACTION/EXPECTED ESPONSE I
ESPONSE NOT OBTAINED i
31 Notify DCS And STA Of Current Plant Conditions 32 Perform NP 5.3.3. INCIDENT INVESTIGATION AND POST-TRIP REVIEW.
While Continuing With This Procedure 33 Select Recovery Procedure:
- a. Check RCPs -
AT LEAST ONE RUNNING
- a. Go to EOP-0.2 UNIT 1. NATURAL CIRCULATION COOLDOWN.
- b. Go to appropriate plantProcedure
-END
POINT BEACH NUCLEAR PLANT EMERGENCY OPERATING PROCEDURE REACTOR TRIP RESPONSE EOP-0.1 UNIT 1 SAFETY RELATED Revision 26 Page 22 of 22 FIGURE 1 RCP No.1 SEAL LEAKOFF FLOWS NO.1 SEAL DIFFERENTIAL PRESSURE (PSI)
B NO.1 SEAL LEAK RATE (GPM) 0.2
I FOLDOUT PAGE FOR EOP-0.1 UNIT 1
- 1.
SI ACTUATION CRITERIA IF either condition listed below occurs. THEN actuate SI and go to EOP-0 UNIT 1, REACTOR TRIP OR SAFETY INJECTION, Step 4:
o RCS subcooling based on core exit thermocouples - LESS THAN 351F OR o PZR level -
CANNOT BE MAINTAINED GREATER THAN 10%1.
- 2.
EOP-0 UNIT 1. REACTOR TRIP OR SAFETY INJECTION, TRANSITION CRITERIA IF SI actuation occurs during this procedure.
THEN go to EOP-0 UNIT 1.
REACTOR TRIP OR SAFETY INJECTION, Step 4.
- 3.
RCS TEMPERATURE CONTROL CRITERIA IF RCS temperatures are rising OR RCS cold leg temperature is greater than 5470F.
THEN adjust feedwater flow and dump steam as necessary to stabilize RCS cold leg temperature less than or equal to 5470F.
- 4.
AFW SUPPLY SWITCHOVER CRITERIA IF CST level lowers to less than 8 feet. THEN switch to alternate AFW suction supply per AOP-23 UNIT 1. ESTABLISHING ALTERNATE AFW SUCTION SUPPLY.
- 5.
AFW MINIMUM FLOW REQUIREMENTS IF any AFW pump mini-recirc valve fails shut OR annunciator C01 A 1-9.
INSTRUMENT AIR HEADER PRESSURE LOW in alarm. THEN monitor and maintain minimum AFW flow or stop the affected AFW pump as necessary to control SIG levels.
o P-38A minimum flow - GREATER THAN 50 GPM o P-38B minimum flow - GREATER THAN 50 GPM o P-29 minimum flow - GREATER THAN 75 GPM I
T EMPOR %IN"'
CHANGE REVIENN AND APPROVAL I-INITIATION
~
rc-visiu'.?t 3
Delete or rnodil[ a prerequisite initial condition, precaution-himitation or othe-r Steps that could have safety significance or affect the procedure's margin of saifety9 4I Delete QC hold points, Independent Vcrification or Concurrent Check steps v~ithout the related st.-p(s) that require the performance also being deleted?
- 5. Change Tech Spec or other regulatory acceptance criteria other thant for re-basehnting purposes?
6 Require a change to the procedure Purpose or change the procedure classification" Initiated By (pnntsirn)
-.A./
,r D
0 D
Date
.~~-ILI -
If-INITIAL APPROVAL Tnis change is cor.cct and eo!nplcie. can ic, performed as written, and does not ad' ersely affect personnel or r~ucleczi safctm or Plant operating conditions Group Supcr%-isor ri
-.:gn) 7~ Vy
~
j,.~
te (Cannot be the 1ittatur)
Seio Tis chanLce, does not ad;ci sely :iffcct Plant operating condit: *.Ps (Safet~ly Peicd Frucdircs ant) Dt SeirReactor Operator (prontsign) it I
c (C.&nnut be the Initiator or Group Supe iscr)
[it -
PROCEDURE OWNER REVIEW Perrinai'czit 0
One-time Use I
L Expiration Date,., Ent or Cond Won.
SHold change until procedure completed tfinal rev-iex% and uppro--al still req,,.ircd '.%ithin 14 da-,s of initial approval)
I]QRIMISS Rtevw% sa NOT equire-d (A.X NY oni>)
QR l~Iteu.
Riecizfed
/Iv' S RZeviex% Rcquiredl tRci' rcr:c N 4M/
Procedure Oi~ne-r (print/lSign)i~____
Dte" c
IV FINAL REVIEW AND APPROVAL 4A.Ibe cianpteud %-ithi,,:1 aso nii prnI Teii~oQ n
\\in~tAt'rt hl indeptridit from each other)
MSS %lective NO
/-I
.pprma
At.
kn-.in v
- REVisioN iNroiz;ITiolN FOi, npT:R Iticorpor~'ted inte P.-\\ ision Nultber E~ffecti'.c Daite I
U;C,, rDEC 17 ZG i rLI~t Rd:rc::d N.' :.
I F-i),Xc Number EOP-oI CU.'Tent Rev 23-LI'.. 11132 icmnpU42--
. 0U -.(c.
I L-C f_
'Dýzunint'liiz RLEACTOR TRIP RESPONSE 1ineEffcu e Trtnpcrar% 'Chranges
~fDsciptonADDED FOP ITEMT 1 C).\\DDlZLSS AFW MLlNLM1UM FLOWN
~~': '~c1~
~cs r~~o~ zBF(O2~.D t.,s~
~
Ara!a Ca nd trcl.,dc wan Incr.a :c.:)
Zbutiatc FBF-010261h and include withi the hace lOthe,-r docuraicnts rcLcu,,rcd tobecffeýLcti'e con~cur.-entl)\\ %%iiithictempwrarn -hange NONE CILUnges pre-scrcen-ld according to NP 10 3 U'
-1 NO TlXE
~'
Screening con-pleted ac:-ording to NP 10 3 1**
Ej NA El YES Safet; E~aluation Recquircd' ED NO [I YE S a'~
arc.ian Ma' bx ýozeusez z-r~
Imms =&a-'C ea~ss-.L:l br-.',P-Wrfc~
De-termine if the chiange-constitutcs a Chan=e Of intenlt to the procedure byv exaluating thie foliowingc questions
,I -*xi a evr s aie Y ES,.I
%t, or. may bec p.-o s.d r,:
%a i % an d appreo a Is sha 1 be obt a :-.d b efore,I n?1:mi n 2)
WNill the proposed change:
Y'ES NO
- 1. Require a.change to.affect or i.n~ahdatc arcquirenicnt, commilLMent,,, ealuation or Ej dcscnptiAon-in the Current Licensmng Basis (as defined in NP 10 _3
,1)?
2 Cause an inzcrasc in mag.,rit,.idc, smgnificancc cr impact such that it should be processcd as a 3
R,-%týwn 12 11 GN V
Poant BE.c'.ic Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTUINUATION' Pjve
ý'
of
!'ýL-Other Comments NCc 'tCording 0f! ':p iob~rs).. is n~ot rejuired :r :i2r oc~urrmu u'ftci~k.tal irtarnation er %%hc.'i na' bene..31,c tomov r!3F-eo:Gc R..ision6 04'IS.OI fnn.s NP I 3 N5 I 23 Doc Numnbcr LOP-0.1 Re, isicr' 23
- Unit, 2
Thil REACTOR TRWTRESPONSE Tf. mporary Changu Nurnbcr i o o t C* 7' Desriifonof Chances:
Z I %-
L.1.
1.1 4-'
LiLM "TE.MPOkRY CIIANGE AFFECTED MANUAL LOCATION Parec Yj1 of Z" I
II - PROCEDURE OWNER REVIEW ON PBF-0026e (rna. be performed by OA [I, Procedure Wuiter, etc)
This procedure change hzs been processed as follows: (Marnual/Location)
S Copy sent to Document Control Distribution Lead for Master File.
(Not rcquired for one-time use change)
E]
Copy filed in Group satellite file. (Not requtcd for one-11-e us changes)
E]
Copy filed in Group one-time use file.
Original Tcmp Change piovided to __
121-,.
to obtain Final Approvals (C gS fhnal approval ma) be coordtnated b% In-Group OA 1I, Proccdurte \\
Mric, Procedure Superisor. ti:)
PAP.
PDIF-0026h
,Isis;ort 06,1;3 01 R e f-c.-,:n
" ' 2 3
"SA I
'S I
-2 I
Pr.cedure Number EOP-0.1 R evsion 23 Um!t PB2 Tidle REACTOR TRIP RESPONSE Tcmporan Chiage Number,
t
/ /
I -.BLMIEDL-kTELY AFTER INITIAL APPROVAL ON PBF-0026e kNon-Intenteonges)
(afttr rinal Approszl.f change cfmtent invol,,cd)
I Date This procedure change has been processed as follows: ('Manual/Location)
I Perfomed El Copy included in \\*ork package for field implementation WVO No.
)
7 Copy filed in Control Room temp change binder (Operations only).
+
11/.
) 1 SOriginal change package provided to (n -. 6 to obtain Procedure Owuner El*
Re'view ',: S.
- Orerr-_%ic~v rnaytc coorA.rx.aed by~ In-Gro.jp 04 11, P-o~ccdurt Write-r. Procedure Supernis.r. etic El El Performed By (print and sign.)
D c
j
\\
Q /jZ1L; D ate )
-S!
I
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 2 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 24 12/14/2001 REACTOR TRIP RESPONSE Page 1 of 22 A. PURPOSE
- 1. This procedure provides directions to stabilize and control plant conditions following a reactor trip without safety injection.
- 2.
This procedure is applicable for specified entry conditions.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. This procedure is entered from the following procedure when SI is neither actuated nor required:
- EOP-0 UNIT 2.
REACTOR TRIP OR SAFETY INJECTION.
Step 4 C. REFERENCES
- 1. Technical Specifications for Point Beach Nuclear Plant
- 2.
Final Safety Analysis Report for Point Beach Nuclear Plant
- 3.
As-built plant drawings
- 4.
Generic Technical Guidelines developed by the Westinghouse Owners Group (WOG).
This consists of the following documents:
- a. Low pressure version of the WOG Optimal Recovery Guidelines. Status Trees*. and Functional Restoration Guidelines
- b. Background documents for each low pressure version Optimal Recovery Guideline. Status Tree. and Functional Restoration Guideline
- c.
WOG Emergency Response Guideline Executive Volume
- d.
WOG Emergency Response Guideline Maintenance Program Summary
FOLDOUT PAGE FOR EOP-0.1 UNIT 2
- 1.
SI ACTUATION CRITERIA IF either condition listed below occurs. THEN actuate SI and go to EOP-O UNIT 2. REACTOR TRIP OR SAFETY INJECTION. Step 4:
"o RCS subcooling based on core exit thermocouples - LESS THAN 35'F OR "o PZR level - CANNOT BE MAINTAINED GREATER THAN 10%
- 2.
EOP-O UNIT 2. REACTOR TRIP OR SAFETY INJECTION, TRANSITION CRITERIA IF SI actuation occurs during this procedure.
THEN go to EOP-O UNIT 2.
REACTOR TRIP OR SAFETY INJECTION, Step 4.
- 3.
RCS TEMPERATURE CONTROL CRITERIA IF RCS temperatures are rising PR RCS cold leg temperature is greater Than 5470 F.
A adjust feedwater flow and dump steam as necessary to stabilize RCS cold leg temperature less than or equal to 547 0 F.
- 4.
AFW SUPPLY SWITCHOVER CRITERIA IF CST level lowers to less than 8 feet, THE switch to alternate AFW suction supply per AOP-23 UNIT 2. ESTABLISHING ALTERNATE ANW SUCTION SUPPLY.
- 5.
ANW MINIMUM FLOW REQUIREMENTS IF any AEW pump mini-recirc valve fails shut. THE maintain minimum flow or stop the affected ANW pump as necessary to control SIG levels.
"o P-38A minimum flow -
GREATER THAN 50 GPM "o P-38B minimum flow -
GREATER THAN 50 GPM4 "o P-29 minimum flow
-GREATER THAN 75 GPM
4A Page I of I INITLATION Doc Number EOP-O.l Current Rev 24 Unit PB2 Temp Change No.,-IC Document Title REACTOR TRIP RESPONSE Existing Effective Temporary Changes BiicfDcscription MODIFY FOP FOR MINL-IU-M AFWV FLOW TO INCLUDE LOW LA HDR PRESSURE (Identify specific changes on Form PBF-0026c. Document Review and Appro,,al Continuation, and mclude %%tah the package)
[0 Imnate PBF-0026h and include with the change.
Other documents required to be effective concurrently with the tcmporary. change.
Changes pre-screened according to NP i0 3.1?
[
NO El YES (I Yes. Ist re.c-ce-:cs **-
- nt-n*a cn-PBF0:6cX--efr N*. 10 3 1)
Screening completed according to NP 10.3.1?
El NA [
YES Safety E% aluation Required 9 S NO [, YES iryt., *re*u.r- -v bt proc....d o, f=1
.e..
&-. ae,rmil, w..a be eau-d bc*f*-e u-or-entr-z)
Determine if the change constitutes a Change Of Intent to the procedure by evaluating the following questions (If ar) ans%%ers 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, comrunitmem, eN aluation or description in the Current Licensing Basis (a:; defined in NP 10.3. 1)?
- 2. Cause an increase in magnitude, significance or impact such that it should be processed as a El revision?
- 3. Delete or niodify 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 %ithout the El 0
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?
- 6. Require a change to the procedure Purpose or change the procedu'r classiAa'on?
Initiated By (p-insign) 1,-4
- /Zk "..A
/
WI
,*n- -
Datei II - INITIAL APPROVAL This clhange is correct and complete, can be performed as % ritten, and does not adversely affect pcrsonnel or nuclear safety, or Plant operating c~ndlitions, Group Supervisor (print'sign) n ii
'*Z
- s.
c
/
Date
,/
/
This change does not adversely affect Pla operating conditions (saftro Related
)c s nl Senior Reactor Operator (pnnt'sign)
/
.Zfe
_elat,'
-Z/solDat (Cannot be the Initiator or Group Superi isor)
Ill - PROCEDURE OWNER REVIEWV Permanent E] One-time Use I] Expiration Date, Event or Condition:
El Hold change until procedure completed (final review and approv-al stil required itd un 14 days of initial approN al)
,0 "QRIIMSS Review NOT Required (AdmirnNNSR only)
[*
QR Review Required.
ME
.SS Revit.w Required (R,,'enCe.%V 1 6 5)
Procedure Owner (print/sign)
ZT'.m,,/IY.,
I
/,
Date
/
Th_; Ch--.ne and surmortine reo-a.lemcnts correctly corr.etcd an-d troces,ed rV - FINAL REVIEW AND APPROVAL
_(Mut be completed tiithjn 14 dais ofinhlal approtal)
(The Iutiator, QR and Appro% At Authorit% shall be independent from ch olhr)
'QRIMSS (printsign) t{d' J
4,747'*
I s
i, Date /,.-
Indiates 50 59,72 48 applicahility asscs,%,A& any necessary s.eer--ngs:e. a.il.ad'ns perforne. d-t.r.r-raton made as to hz.he:r. add.tio,',l cLross-disciplhn-ry review required, and 3frequired, performed, MSS Meeting No.
Approval Autlhont. (prnL's~g,*gn)
/
/*/is---//
-D ate /"
V-REVISIONINFORMATION FOR PERAMANE-N7 CiANGES
/ /
Post% TyingRctiC% (prinLsitn) 1)',
/*.l,
I 1 /-1., ;
Date
"/.J-lie" Infi.atcs temporary char'gc(s) irc,4rporated cxatl, as aprrc.cdx and no ot.r..hangesn n....,
Incorporated into Revision Number Effective Datejr&4 JI
,F
._.D.jjT~f i-*O r n 11131'-0026c Re% tbior 12 1.V08,)99
' -Cot it j -e I o
Nuclear Power Business Unit TEMPORARY CHANGE REVIEW AND APPROVAL Note: Refer to,VP 1.2.3. Temporary Procedure Chanmes, for requirements sYfJ'Al I
R-.:rrne.s NP 1 2
Pagej-of Doc Numbcr EOP-0.1 Re-ision 24 Unit 2
Title REACTOR TRIP RESPONSE Tcmpr,:.ay Change Number 7a" 2c 9 /3 Description of Changes:
Step
- Change/Reason CHANGE: ADDED REFERENCE TO AN.INUNCIATOR COI A 1-9, IA HEADER LOW PRESSURE FOR THE AFW MINIMUM FLOW REQUIREMENTS REASON: TO ENSURE MINIMUM FLOW IS MAINTAINED THROUGH THE AFW PUM.-N.PS DURING OPERATION Other Comnicnts Now i.':d;n of Swp Xu:ibr:, ir nor ::a i*J fŽ mL'r!. c Lrcr.cs of &niaI ifur: a:l. C:
n z:
- o ri.:rs PBl'-002(.c Rcision 6 (04 IS 1 Point Beach Nuclear Plan, DOCUMENT REVIEW AND APPROVAL CONTINUATION Change/Reason t
%~*ct".r.,-'
I I I N2-I : 3
I -r A r
ci Proccdurc Number E&OP-0.1 Revision 24 Unit PB2 Title REACTOR TRIP RESPONSE Tcmporar. Change Number "re!,.t,.
(_3 I - IMMEDIATELY AFTi-ER INITIAL APPROVAL ON PBF-0026e (Non-I.t.tchrt gs)
(after Fina! Appo-va if change of intent in% olved)
Date This procedure change has been processed as follows: (ManuaLIocation)
Performed LI Copy includcd in work package for field implementation. (WO No.
) J
[]
Copy filed in Control Room temp change binder (Operations onh).
1 /2
-C-. I.
Z)
Original change package p:ovided to
- Y to obtain Prccedure Owner j /
Rex Jew (( g, Ownecr rc'ic%, ma. be coordinated by In-Crom.p OA 11, Procdurc V, tiler., Prccedure Supcr"isor, dc)
U LI I_
[]
I__
I- ___________________
Perfcnr.cd By (print and sign)
(.
I
\\.
j t
,L"Date
/.2 ;)o -c II - PROCEDURE OWNER REVIEW ON PBF-0026e (riuiy N*; rrformed by OA II, Procedure Writer, etc)
Date This procedure change has been processed as folloxss: (ManuablLocation)
Performed F
Copy sent to Document Control Distribution Lead for Master FHle.
/..-.--.,
(Not requ-red for one-timc use change)
LI Copy filed in Group satellite file. (Notrequired for one-time use changes)
LI Copy filed in Group one-time use file.
[
Original Tcino Chance provided to S
to obtaLn Final Approi al (e E. tinr! acproJ ma'. be cocrdanat.z Ir. in-Grrup OA II, Pr*cedurc Writer, Proced.re.- pSrvi'sor, ctc)
~~
~
L Z
~
I jot El lPcrf-rn1d By (print and sign) oC
( S,,
,-.c
/ (_'.
I!
- Dat, Rer.I 1
.',. 13 0, Point Beach Nuclear Plant TEMPORARY CHANGE AFFECTED MLANUAL LOCATION Page
\\
2
ottD~- *:*
Point Beach Nuclear Plant
- o."
10 CM 50.59n2.48 SCREENING (NEW RULE) wafy SCR5~bon-al pagG
.*/..*Page I
f t Proposed Activity:
EOP/ARP Actions For AFW Mini-Recirc Flow Requirement Associated Reference(s) #:
Prepared by:
Bob Wartenberj
~.~T Date:
P Name (Print)
- Reviewed by:
Date:
Nalne (Print)
PART I (50.59fl2.48) - DESCRIBE THE PROPOSED ACTIITY AND SEARCH THE PLANTr AND ISFSI LICENSING BASIS (Resource Manual 5.3.1)
NOTE: The "NMC 10 CFR 50.59 Resource Manual" (Resource Manual) and NEI 96-07, Appendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 10 CFR 50.59 and 10 CFR 72.48 screenings.
1.1 Describe the proposed activity and the scope cf 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-screning process requirements in NP 5.1.8.) Appropriate descriptive material may be attached.
EOP-0 and EOP-0.1 for both uwits revised the AFW Minimum Flow Requirements foldout-page criteria to include the instrument air header pressure low annunciator in the alarm state as additional criteria for monitoring AFW mini-retcirc flow requirements.
AFW Minimum Flow Requirements foldout-page item was added to ECA-0.0. The foldout-page item is identical to the AFW minimum flow requirements used in EOP-0 and EOP-0.1.
Step 3.1 Ias added to ARP CO1 A 1-9. The step states to monitor and maintain AFW mini-recirc flow require;.:ents should the AFW pump mini-recirc valve fail shut.
L2 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 ('PER), the CLB (Regulatory) Commitment Database, the Technical Specifications (both Custom and Improved), the Technical Specifications Bases, and the Technical Requirements Manual Search theISFSI 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 finction(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, Au:xiliary Feedater System.
1.3 Does the proposed activity involve a change to any Custom or Improved Technical Specification (ITS)? Changes to Technical Specifications require a License Amendment Request (Resource Manual Section 5.3.1.2).
Technical Specification Change:
0l Yes E No If a Technical Specification change is required, explain what the change should be and why it is required-P ceferee " NP 5 1.g P?-F-1515c
Point Beach Nuclear Plant SCR 10 CFR 50.59/72.48 SCREENI'NG (NEW RULE)
Ve'S SCR :
on all pagcs Pacee 2 Does the proposed activity involve a change to the terms, conditions or specificationrs in:o.porited in an'v VSC-2I cask Certificate of Compliance (CoC)? Changes to a VSC-24 cask Certificate of Compliancc requLre a CoC an.nicn,,ent request.
Di Yes Li No If a storage cask Certificate of Compliance change is required. explain what the change sho'ud be and wh% n is required.
10 CFR 50.59 SCREENING PART !I (50.59) - DETERMINE IF THE CHANGE INVOLVES A DESIGN FUNCTION (Resourcc M.anual 5.3 2)
Compare the proposed activity to the relevant CLB descriptions, and anrwer the following questions:
YES NO QUESTION ED D
Does the proposed activity involve Safcty Anmlyses or structures, sYstems and components (SSCs) credcted in the Safet" Analyses?
5]
[]
Does the proposed activity' involve SSCs that support SSC(s) credited in the Safety Analyses?
D
[D Does the proposed activity involve SSCs whose failure could initiate a transient (e.g, revactor trip. loss of fecciwatcr, etc.) or accident, OR whose failure could impact SSC(s) credited in the Safet.v Analyses?
F 0
Does the proposed activity involve CLB-described SSCs or procedural controls that perform functions that arc required by, or otherwise necessary to comply with, regulations, license cond'tions, orders or mechnical specifications?
0]
Does the activity involve a method of evaluation described in the FSAR?
Li 0
Is the activity a test or experiment? (i.e., a non-passive activity which gathers data) 5 0
Does the activity exceed or potentially affect a de.Fign bascls liritlfor afission o
roduct barrier (DBLFB)?
(NOTE: 11T HIS questions is answered YES, a 10 CFR 50.59 Evaluation is required.)
If the answers to ALL, of these questions are NO., mark Part 11 as not applicable, document the 10 CFR 50.59 screening in the conclusion section (Pan IND, then proceed direc.ly to Part V - 10 CFR 72.4S Pre-screeping Questions.
If any of the above questions are marked YES identify below the specific design fuinction(s). method of eN aluation(s) or DBLFPB(s) involved FSAR 10.2 states each AFW pump has an AOV controlled reciic line back to the CST to ensure minimum flow to dissipate heat This charnze ensures the mirumum AFW flow requirements will be maintained on any running A.WA piunip in the case of a failed shut AFWV nuni-rccmrc flow contrcl valve PART Im (50.59) - DETER.MINE WHETHER T71iE ACTIVITY INVOLVES ADVERSE EFFECTS (Resource Manual 5.3.3)
Ift A.LL the questions in Part II are answered NO. then Pan III is U] NOT APPLIC.A3LE.
Answer the follmi\\ irng questions to determine if the activity has an,c"erse effect on a design funcuan Any 3EnS ans".r nta.: int a 10 CFR 50.59 Evaluation is required; EXCEPT m here noted in Part 1113.
i11 1 CHIANGES TO ThE FACILITY OR PROCEDURES YES NO QUESTION S
[]
Docs tihe activity ad\\erscly affect the deszgnfiuncfion of a3, SSC.....
in s" fN andiŽ P1W-15i5 11*
F Rc':*,vr.C 102401Ol
N Point Beach Nuclear Plant 10 CFR 50.59/72.48 SCREEN`INMG (NEW RULE)
C=4OCZ,, -
cx:
SCR _
66_1,_
Verif' SCR rramber on all rates J.-* O*Pace 3 D
[
Does the activity adversely affcct the method of performing or controlling the desgnfunction of an SSC credited in the safet; anal scs?
If any ans.er is YES, a 10 CFR 50.59 Evaluation is required. If both answers 21e NO. describe the basis for the conclusion (attach additional discussion as necessary):
This change ensures that minimum recirc flow requirements as stated in FSAR 10.2 are not violated.
111.2 CHANGES TO A METHOD OF EVALUATION (If the activity does not involve a method of evaluation, these questions arc F NOT APPLICABLE.)
YES NO El El QUESTION Does the activity use a reMised or different method of evaluation for performing safety analyses than that described in the CLB?
El El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safctv analyses than that described in the CLB?
If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers arc NO, describe fie basis for the conclusion (attach additional discussion, as necessary).
111.3 TESTS OR EXPERIMENTS If the activity is not a test or experiment, the questions in III.3.a and m.3.b are [
NOT APPLICABLE.
- a. Answer these two questions first:
YES NO D
El D
M QUESTION Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?
Arc the SSCs affected by the proposed test or experiment isolated from the facility?
If the answer to BOTH questions in V.3.a is NO, continue to IIL3.b. If the answer to EITHER question is YES, then describe the basis.
- b. Answecr these additional questions ONLY for tests or experiments iwhich do NOT meet the criteria given in UII.3.a above.
If the answer to either question in 11.3.a is YES, then these three questions are E] NOT APPLICABLE.
YES NO El El QUESTION Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the CLB?
El El Does the activity utnlize or control an SSC in a manner that is inconsistent i ith the analyses or descripuons in the CLB?
E]
El Does the activity place the facility in a condition not pr.,
lously cvaluated or J-.at could affect the capabilitv of an SSC to perform its intended functions?
If am answer in lII.3.b is YES. a 10 CFR 50.59 Evaluation is required. If the ans\\e crs in III 3.b arc ALL..NO. describe t'ic basis for the conclusion (attach addtional discussion as necess*ry).
PBF.-151ic
- C.-. - :C* NP ý 1 9 k
Point Ecach Nuclear Plant SCR _-___-,_____"_
10 CFR 50.59."72.48 SCREENING (NEW RULE)
Vef SCR rrmc, on.il paes S+ '*
1-_,4-oaPage 4 S1-10 CFR 50.59 SCREENING CONCLUSION (Resource Manual 5.3.4)
Check all dial app]'.
A 10 CFR 50 59 En aluation is El required or -Z NOT required.
A Point Beach FSAR change is El required or 0, NOT required. If an FSAR change is required. then initiate an FSAR Clange Request (FCR) per NP 5 2.6.
A Regulator" Commitment (CLB Commitment Database) change is El required or [
NOT required. Ifa Regulato.-%
Commitment Change is required, inntiate a comminitmencflt change per NP 5.1.7.
A Technical Specification Bases change is D. rcquired or ID NOT required. If a change to the Technical Specification Bases is required thcn initinte a Technical Specificauon Bases change per WNP 5.2.15.
A Technical Requirements Manual change is Ml rcquied or [ NOT required Ifa change to the Technical Reqirements Manual is required. then initiate a Technical Requirements Ma.nual change per NP 5.2.15.
10 CFR 72.48 SCREENING NOTE: NEI 96-07. Appendix B, Guidelines for 1]( 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 v determines ifa full 10 CFR 72.48 screening is required to be completed (Parts Vi and VII) for the proposed activity.
"Y r.
NO QUESTION Does the proposed acti.,ity inolve IN ANY ]MTAIN.NER the dry fuel storage cask(s), the cask transfer/transport equipment, any ISFSI facility SSC(s), or any ISFSI facility monitoring as follorts: Multi-Assembly Sealed Basket (MSB), MSB Transfer Cask (MFTC). MrTC Lifting Yoke, Ventilated Concrete Cask (VCC). Ventilated Storage Cask (VSC). VSC Transporter (VCST), ISFSI Storage Pad Facility, ISFSI Storage Pad Data/Communica*ion Links, or PPCSrISFSI Continuous Temperature Monitoring System?
El f]
Docs the proposed activity in ol' c IN ANY MANNER SSC(s) installed in the plant spccifically added to support cask loading/unloading actihitics, as follaws: Cask Dewaering System (CD.V), Cask Reflood System (CRF), or H% drogen Monitoring System?
El Does the proposed activity in\\ olvc IN ANY MANNER SSC(s) needed for plant operation which are also used to support cask loadinglunloadine a:Li% ities. as follov s" Spent Fuel Pool (SFP). SFP Cooling and Filtration (SF),
Prim'anr Auxiliarv Building Venilation System (\\VYPAB), Drumrning Area Ventilation System (\\VNDPJM).
RE-10_ (SFP Lov Range Monheor). RE-135 (SFP Htgh Range Monitor). RE-221 (Drumning.Area Vent Gas Monitor), RE-325 (Dnmuninv Ajea E.daust Low-Range Gas Moiitor). PAB Crane, SFP Platform Bridge.
Tru:!k Access Area. or Decon Area?
Dzoes die proposed actiN ity in% oh c a change to Point Beach CLB design c.riteria for eyxternal events such as earthquakes. tornadoes. !high N%
,:,ds. floodine. etc.?
I--,
[]
Dees the aci.ity inN oh e plant v':,
- load requirements or procedures for areas of the plant used to support cask Ioaidnr_,unloading aciiviues" E"- []
Does th. actnr, i rinoohe am V-:.i':al for fire or explosion m\\here casks are !oad'-d. unloaded, tur_*-spord or stor.ed?
"Y of the P.:rut V.jtrcstiens are answered YFý; the"n a frill 10 CFP. 7_ 410, screcnin-g is req*a:e-a'
""sers to the 0uzstioas in l
,'a:d P.am VII re to be pro% ided IfALI
.. ",'1,*csiron_ in Pan V are aDareted NO. thea cheek Pars \\I and VII as not
- rpplhcnblc Conipic:e Part \\1Il o do:iicimern t, :i:,:lusion tihlt no 10 CFR 72.45 exarua'uon is reqturei
]-'e....,'n It 1
(t
Point Beach Nuclear Plant SCR 10 CFR 50.59172.48 SCREENING (NEW RULE) vci*r scR mrbr on &!I pa*es J~~)
,-Page 5
RT VI (72.48) - DETERMINE IF THE CHALNGE INVOLVES A ISFSI LICENSING BASIS IESIGN FUNCTION (ij ALI. the questions in Part V are NO, then Part VI is [D NOT APPLICABLE.)
Compare the proposed activity to the relevant portions of the ISFSI licensing basis and arsver the following questions:
IVES NO QUESTION
[]
fl Does the proposed activity involhe cask/ISFSI Safety Analyses or plant/cask/ISFSI structures, systems and components (SSCs) credited in the Safety Analyses?
[]
E]
Does the proposed activity involve plant, cask or ISFSI SSCs tlat support SSC(s) credited in -he Safety A-alyscs?
[]
E3 Does the proposed activity involve plant, cask or ISFSI SSCs whose function is relied upon for prevention of a radioactive release, O._R whose failure could impact SSC(s) credited in the Safety Analyses?
E]
El Does the proposed activity involve cask/ISFSI described SSCs or procedural controls that perform functions that are rcquired by, or othenvise necessary to comply with, regulations, license conditions, CoC conditions, or orders?
El EL Does the activity involve a method of evaluation described in the ISFSI licensing basis?
[]
El Is the activity a test or experiment? (i.e., a non-passive activity which gathcrs data)
M E]
Does the activity exceed or potentially affect a cask design basis limit for afission product barrier (DBLFPB)?
(NOTE: If T[IS questions is answered YES a 10 CFR 72.48 Evaluation is required.)
Ifthe 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).
- -y of the above questions are marked YES identify belo, 'he specific design function(s), method of evaluation(s) or DBLFPB(s)
%ed.
PART NrII (72.48) - DETERMINE WVHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (-Ei 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 2 NOT APPLICABLE.)
Answer the followisng questions to determine if the activity has an ad%,erse effect on a design function. Any YES answer means that a 10 CFR 72.48 Evaluation is required; EXCEPT where noted in Part VIL3.
VII. 1 Changes to the Facility or Procedures YES NO QUESTION D]
fl Does the activity adversely affect tie design funcnton of a plant, cask, or ISFSI SSC creited in safet.
analyses?
[]
LI Does the activity adversely affect the method of performing or controlling the designflunction of a plant cask, or ISFSI SSC credited in the safety analyses?
!f any ansi'cr is YES. a 10 CFR 72.48 Evaluation is required. If both ansuers are NO. describe the basis for the con-h.sion (atta.ch additional discussion, as nctessary)
PDF.1~515
.. 2.Ooo--.
- oo*D Point Beach Nuclear Plant SCR 0
10 CFR 50.59/72.48 SCREENR11G (NEW RULE)
-i6' SCR numberon all P 1-4 --
Page 6 "2 Changes to a Method of Evaluation (If the activity does n-t involve a method of evaluation, these questions are [] NOT APPLICABLE.)
YES NO QUESTION
[J El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in a cask SAR?
[I 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 VII3.a and VII.3.b are [] NOT APPLICABLE.)
- a. Answer these two questions first:
YES NO QUESTION El El, Is the proposed test or experiment bounded by other tests or experiments that are described in the cask ISFSI licensing basis?
El
[:
Are the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISFSI facility?
f the anser to both questions is NO continue to VIL3.b. If the answer to EITHER question is YES, then briefly describe the basis.
- b. Ansrer 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 El NOT APPLICABLE:
YES NO QUESTION El El Does the activity utilize or control an SSC in a manner that is outside the reference bounds of the design bases as described in the ISFSI licensing basis?
[E El Does the activity utilize or control a plant, cask or ISFSI facility SSC in a -narier that is inconsi-stent 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 functions?
If any answer in VII.3.b is YES, a 10 CFR 72.48 Evaluation is required. If the ansN% ers are all NO, describe the basis for the conclusion (attach Wdditional discussion as necessary):
PBF-1515c Rcvision0 10/24/01 Pefe-ence VP 5 1 8
,:2CC0,Z - oc_.*_
Point Beach Nuclear Plant SCR,_________-__-_
10 CFR 50.59/72.48 SCREENING (NEW RULE) vei sCR numbe or. ll M-s
____1-4--o0 Page 7
'%T VIII - DOCUI-ENT THE CONCLUSION OF THE 10 CER 72.48 SCREENING Check all that apply:
A 10 CFR 72.48 Evaluation is E] required or 0 NOT required. Obtain a screening nmLnber and provide the original to Records Management regardless of the conclusion of the 50.59 or 72.48 sc-eening A VSC-24 cask Safety Analysis Report change is I] required or Z NOT required. If a VSC-24 cask SAR change is rcquircd, then contact the Point Beach Dry Fuel Storage group supervisor.
A Rcgulatory Conmmitment (CLB Commitment Database) change is [0 required or E NOT required. If a Regulatory Comnritment Change is required, initiatc a commitment change per NP 5.1.7.
A change to the VSC-24 10 CFR 72.212 Site Evaluation Report is El rcquired or Z, NOT required. If a \\'SC-24 10 CFR 72.212 Site Evaluation Report change is required, th-n contact the Point Beach Dry Fuel Storage group supervisor.
PDF.ISIc
POINT BEACH NUCLEAR PLANT EOP-0.1 UNIT 2 EMERGENCY OPERATING PROCEDURE SAFETY RELATED Revision 25 1/10/2002 REACTOR TRIP RESPONSE Page 1 of 22 A. PURPOSE
- 1. This procedure provides directions to stabilize and control plant conditions following a reactor trip without safety injection.
- 2.
This procedure is applicable for specified entry conditions.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. This procedure is entered from the following procedure when SI is neither actuated nor required:
& EOP-O UNIT 2. REACTOR TRIP OR SAFETY INJECTION. Step 4 C. REFERENCES
- 1. Technical Specifications for Point Beach Nuclear Plant
- 2.
Final Safety Analysis Report for Point Beach Nuclear Plant
- 3.
As-built plant drawings
- 4.
Generic Technical Guidelines developed by the Westinghouse Owners Group (WOG).
This consists of the following documents:
- a. Low pressure version of the WOG Optimal Recovery Guidelines. Status Trees. and Functional Restoration Guidelines
- b. Background documents for each low pressure version Optimal Recovery
.Guideline.
Status Tree. and Functional Restoration Guideline
- c.
WOG Emergency Response Guideline Executive Volume
- d. WOG Emergency Response Guideline Maintenance Program Summary
1 4.
FOLDOUT PAGE FOR EOP-O.1 UNIT 2
- 1.
SI ACTUATION CRITERIA IF either condition listed below occurs. THEN actuate SI and go to EOP-0 UNIT 2. REACTOR TRIP OR SAFETY INJECTION, Step 4:
o RCS subcooling based on core exit thermocouples - LESS THAN 350 F OR o PZR level -
CANNOT BE MAINTAINED GREATER THAN 10%
- 2.
EOP-0 UNIT 2. REACTOR TRIP OR SAFETY INJECTION, TRANSITION CRITERIA IF SI actuation occurs during this procedure.
THEN go to EOP-O UNIT 2, REACTOR TRIP OR SAFETY INJECTION, Step 4.
- 3.
RCS TEMPERATURE CONTROL CRITERIA IF RCS temperatures are rising PR RCS cold leg temperature is greater Than 547°F. THEN adjust feedwater flow and dump steam as necessary to stabilize RCS cold leg temperature less than or equal to 5470 F.
- 4.
AFW SUPPLY SWITCHOVER CRITERIA IF CST level lowers tb less than 8 feet, THEN switch to alternate AFW suction supply per AOP-23 UNIT 2.
ESTABLISHING ALTERNATE AFW SUCTION SUPPLY.
- 5.
AFW MINIMUM FLOW REQUIREMENTS IF any AFW pump mini-recirc valve fails shut OR annunciator C01 A 1-9.
INSTRUMENT AIR HEADER PRESSURE LOW in alarm. THEN monitor and maintain minimum AFW 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-38B minimum flow -
GREATER THAN 50 GPM "o P-29 minimum flow - GREATER THAN 75 GPM I
Nuclear Power Business Unit
]' I' TEMPO R.-RY CHANGE REVIEW AND.-APPROVAL Note: Refer to VP 1.23. Temporary Procedure Changes, for requirements.
Page 1 of I - INIT*ATION Doc Number ECA-0.0 Current Rev 29 Unit PB I Temp Change No.Ld.
_ 0','7 Document Title LOSS OF ALL AC POWER Existing Effective Temporary Changes Brief Description ADD FOP FOR.MINIMUM AFW FLOW (IdentNf) specific changes on Form PBF.0026c. Dccument Review and Approval Continuation, ran inelut.v.;,h,he package) 0 Initiate PBF-0026h and include with the change.
Other documents required to be effective concurrently with the temporary change:
Changes pre-screened according to NP 10.3.1?
0 NO El YES (if Yes. list ritfierces and c--e*.-en PF-O026:X-efr so N'P 103 1)
Screening completed according to NP 10.3.1?
[] NA 15ý YES Safety Evaluation Required?
[ NO D-YES f Y.. ac,,,er, rn--
2:.. z.-,o,6,,>,
-e,-!-r,,un0 Deterrrlic hifthc 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 remiews and approvals shall be obtzined before implementing)
Will the proposed change:
YES NO
- 1. Require a change to, affect or invalidate a requirement, commitment, evaluation or 0
description in the Current Licensing Basis (as defined in NP 10.3.1)?
- 2. Cause an increase in magnitude, significance or impact such that it should be processed as a 0
revision?
- 3. Delete or modify a prerequisite, initial condition, precaution, limitation or other steps that C
0 ED 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 0
related step(s) that require the performance also being deleted?
- 5.
Change Tech Spec or other regulatory acceptance criteria other than for re-baselining E]
S purposes?
- 6.
Require a change to the pzpcedure Purpose or change the proCed cla..
aU59%
Initiated By (pint/sign)
IDate 1l - INITIAL APPROVAL This change is correct and complete, can be performed as written, and does not ady v affect rsor nel or nuclear safety, or Plantq.perating conditions.
Date Group Supervisor (printhign) 7-Date
"(Cannot be the lihtiator) v This change does not adversely affect Plant operating conditions (Safety Relate Pro dures only)
Senior Reactor Operator (print/sign)
/
C- ý. 0.
/
U4 Date /Z-..,
c0( '
(Cannot be flnitiator or Croup Super-
"or)
Ill - PROCEDURE OWNER REVIEW0 0
Permanent
[] One-time Use
[] Expirztion Date, Event or Condition:
]-Iold change until procedure completed (final review and approval still required within N4 da'Ns of initial approval)
QR/MSS Review NOT Required (Adrni-'%SR only)
[
QR Review Required El MSS Re% ie%% Required (,cf
... NP' 16 5)
Procedure Owner (print/sign)
/,
1
./
/
Date /5A _._/_/_
T,% Change and supportine recuirements correctW compnleted and vrocessed IV - FINAL REVIEW AND A PPROVAL
-.*,NIust be completed -Aithin 14 days of Initial approial)
(The Initiator, QR and Approu al Au*horit. shall be independent firom each other)
QRI SS (pnnt/sign) 15'd W
lAt~~I/.
~{/,C ae____
Indicates 50 59(7248 apphicabilhtyasses,,ed anycessa.-y screeningstcvaluations perfo-m e..*cermrinaon made as to %shesheh adddtional crosa-disciplinasy review required, and if required, performed.A MSS Meeting No.
/
-1 Approval Au=hey 0/
.(---
L elDate
_c ApcvlAu....... y (print'sign)
ý0 Da
/ic' #
V - REVISION. lNFOiRMATION FOR PERMAfýN-ENft 1ANGES Post T% pin-Review (print sign) 1'r, t".--
/
, 74, -
'.Date
/Z.
" Indicates temporary chang*(s) ancorporaitcd c',cacyasi appro cd an no other ch-nge made o docurren.
Incorporated into Revision Number Effective Date SJ-d4 1 (' 2'n'
-_CD JA, 11 ZOE
- c. -,
3 PBF-0026e 11R4fcr"cHs.,1 23 r....:......-
I. I one
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTlNUATION Pagee 4
of.._
Doc Number ECA-0.0 Re\\Jislon 29 Unit 1
Title LOSS OF ALL AC POWER Tempomrar Change Number 7"e h.l -y l
-2 0
1 Description of Changes:
Step Change/Reason CHANGE: ADDED FOP CRITERIA FOR THE AFW` MIN12MI.JM FLOW REQUIREMEIN'I It FOLDOUT REASON: TO ENSURE MNLNNIIMM FLOW IS IMtMNTAJN"ED THROUGH TH AFW PUMPS PAGE DURING OPERATION.I II Other Comments
- Note: Recording or'step Nunbc=r's) is not requ---ed for rr'ul-pic ozccrrce--es of identncz!
caar:
- c" hm not '=6cfim] to rt% W.AMr PaF-00o6e
a
,0 I.
Procedure Number ECA-0.0 Re~ision 29 Unit PB1 Tide LOSS OF ALL AC POWER TemporaryChangeNumber E-m,ev 201o
- 0q/97 I - IMMEDIATELY AFTER INITIAL APPROVAL ON PBF-0026e (Non-Intent changes)
(after Final Approval if change of intent inolh ed)
Date This procedure change has been processed as follows: (Manual!Location)
Performed 0
Copy included in work package for field implementation. (WO No.
)
I[I Copy filed in Control Room temp change binder (Operations only).
I o
[
Original change package provided to PLSTLO to obtain Procedure Owner I
Review (e g.. Owner r* *-ew may be cordinaod by In-Group OA I, Procedure Writer, Procedure Supervisor. e'c.)
I ;--.* * "
Performed By (print and sign)
Cc,,ý 7
,~
Date /-.oo 11-. PROCEDURE OWNER REVIEW ON PBF-0026e (may be performed by OA H, Procedure Writer, etc.)
Date This procedure change has been processed as follows: (Manual/Locat~nn)
Pefrd 0J Copy sent to Document Control Distribution Lead for Master File.
[] Cp ie nGopstliefl.(Not required for one-time use changes))_____
CpfieinGopstliefl.(Not required for one-time use chznge)
EL Copy filed in Group one-time use file.
S Original Temp, Change provided to
-)
&to obtain Final Approvals 1,
f.-_2 el 0~
(c g. fina approval may be coordinated by In-Group OA 11, Procedure Writer, Procedure Superisor. eci..)
L "tI L _
- -7
- - F, L ) I SLACLO 6
1Az El Perform-ed By (print and sign)
C1 (Xc.A Date /}-d)t-o(
[Cc PBF-0026h Retisinn 5 06,13 C!
Reference. NP 1 2 3 Point Beach Nuclear Plant TEMPORARY CIHA-NGE AFFECTED MIANITAL LOCATION Page
_1 r
U I
Point Beach Nucl"-e; Plant 10 CFR 50.59/72.48 SCREENING (NEW RULE)
SCR aO50-I' 0;t Vcri?, SCR.-umrnb on all pages
..rage.
of Proposed Activity:
EOP/ARP Actions For AFW Mini-Recirc Flow Requirement Asscciated Refcrcnce(s) #:
Prepared by:
Bob Wartenberg
./
Date:
Name (Print)
S~na"ure Reviewed by:
Date:
11416 Namne (Print)
Sioature PART 1 (50.59172.48) - DESCRIBE THE PROPOSED ACTIVITY AND SEARCH THE PLANT AND ISFSI LICENSING BASIS (Resouarce Manual 5.3.1)
NOTE: The "NM'SC 10 C-R 50.59 Resource MNanual" (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.
I. 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 porti 3ns 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.
EOP-0 and EOP-0.1 for both units revised the AFW Minimum Flow Requirements foldout-page criteria to include the instrument air header pressure low annunciator in the alarm state as additional criteria for monitoring AFV mini-recirc flo%ý requirements.
AFW Minimum Flow Requirements foldout-page item was added to ECA-0.0. The foldout-page item is identical to the AFW minimum flow requirements used in EOP-0 and EOP-0.1.
Step 3.1 was added to ARP C01 A 1-9. The step states to monitor and maintain AFW mini-recirc flow requirements should the AFW pump mini-recirc valve fail shut 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 Technica! Specifications (both Custom and Improved), 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 th:
cask/ISFSI as appropriate. Identify where the pertinent information is cescribed 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.
1.3 Does the proposed activity involve a change to any Custom or Improved Technical Specification (ITS)? Changes to Technical Specifications require a License. Amendment Request (Resource Manual Section 5.3 1.2).
Technical Specificafion Change:
EJYes OJNo If a Technical Specification change is required, explain what the change should be and % hy it is required.
P13F-1515c Pvc'isionO 104,01
-V.
Re!frcnc NP 5 1 3
Point Beach Nuclear Plant SCR RO61-IG2N 10 CFR 50.59/72.48 SCREEIXNýG (NEW IRULE) wV.,- SCR nu*bc. on--il r-ges Page 2 Does the proposed activity involve a change to the terms, conditions or specificationrs incor-poa,-ed in any VSC-24 cask Czruficate of Compliance (CoC)? Changes to a VSC-24 cask Certificate of Compliance require a CoC amendment request.
EYes 2 No If a szorage cask Certificate of Compliance change is required, explain what the change should be and %%h it is required.
10 CFR 50.59 SCREEN'ING----------
PART I (50.59) - DETERMWNE IF THE CHANGE INVOLVES A DESIGN FTjNCTIO.N (Resource Manual 5.3.2)
Compare the proposed activity' to the relevant CLB descriptions, and answer the foUowing qu:stions" YES NO QUESTION 0
[
Does the proposed activity involve Safety Analyses or structures, s) stems and components (SSCs) credited in the Safety Analyses?
El
]
Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Anal ses?
C]
0D Does the proposed activity involve SSCs whose failure could initiat. a transient (e.g., reactor trip, loss of feedwaer, etc.) or accident, OR whose failure could impact SSC(s) credited in the Safery" Analyses?
F]
[]
Does the proposed activity involve CLB-describcd SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications?
1 0
Does the activity involve a method ofevoluation described in the FSAR?
El]
0 Is the activity a test or experiment? (ie., a non-passive activity which gathers data) 1]
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 III as not applicable, document the 10 CFR 50.59 screening in the con. ".sion 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.
FSAR 10.2 states each AFW pump has an AOV controlled recirc line back to the CST, :o ensure minimum flow to dissipate heat. Th change ensures the minimum AFWV flow requirements will be maintained on any rurning AFW pump in the case of a failed shut AFV imSm-recirc flow control valve PART IH (50.59) - DETERMINE WHETHER THE ACTIVITY iNVOLVES ADV'ERSL EFFECTS (Resource Man'al 5.3.3)
If ALL the questions in Part II are answered NO., then Pail Il is E] NOT APPLICABLE.
Ansver the following questions to determine if the activity has an adverse effect on a design function Any YES ansN er means that, 10 CFR 50.59 Evaluation is required; EXCEPT where noted in Part III 3.
111.1 CHANGES TO THE FACILITY OR PROCEDURES YES NO QUESTION I]
0 Does the activity adecrsely affect the desn func'lon of an SSC crcdited in safety anoihses?
PBF-iSISc t,
-:-A.
sf/sII
Point Beach Nuclear Plant SCR
?00[-6,,,L*
10 CFR 50.59172.48 SCREENgNG (*NEW RULE) v.sc.R.-.=b'-ron all pgs Page 3 El 2
Does the activity adversely affect the method of performing or controlling the design function of an SSC credited in the safety analyses?
Ifany ans'.,er is Y'ES, a 10 CFR 50.59 Evaluation is required. If both answers are NO. descnbe the basis for the conclusion
(-attch additional discussion as necessary):
"This chan,,e ensures that minimum recirt flow requirements as stated in FSAR 10.2 are not violated.
111.2 CHANGES TO A METHOD OF EVALUATION (If the activity does not involve a method of evaluation, these questions are Z NOT APPLICABLE.)
YES NO QUESTION
[]
El Does the activity use a revised or different method of evaluation for performing safety an2alyses than that described in the CLB?
El
[]
Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in the CLB?
If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers are NO. describe the basis for the conclusion (attach additional discussion, as necessary).
111.3 TESTS OR EXPERIMENTS if the activity is not a test or experiment, the questions in 11L3.a and fll.3.b are [] NOT APPLICABLE.
- a. Ansver these taro questions first:
YES NO QUESTION El C)
Is the proposed test or experiment bounded by other tests or c 'periments that arc described in the CLB?
El El Are the SSCs affected by the proposed test or experiment isolated from the facility?
If the answer to BOTH questions in V.3.a is NO, continue to I1L3.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 gi en in II1.3.a above.
If the ans-wer to either question in I1.3 a is YES, then these three questions are [] NOT APPLICABLE.
YES NO QUESTION
[I El Does the activity utilize or control.,i SSC in a manner that is outside the reference bounds of 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?
[E
[E Does the activity place the facility in a condition not pre,-iousl. e,. aluated or that could affect the capability of an SSC to perform its intended functions?
If am, an,-% er in 111.3 b is YES, a 10 CFR 50 59 Evaluation is required. If the anst.% ers in II.3 b are ALL No, describe the basis for the conclusion (attach additional discussion as necessary):
PBF-1 0 I 5c
-R,"i,'i cn 0 10,2-4 01R
- .-e.*"Y 5 1 3 IN-
Point Beach Nuclear Plant SCR 1 0 CFR 50.59M72.48 SCREENING (NEW RULE) sc.l.-SC.* onaIg Page "
-hrV -10 CFR 50.59 SCREEND.G CONCLUSION (Resource Manual 5.3.4).
Cneck all that apply:
A 10 CFR 50.59 Evaluation is E] required or, NOT required.
A Point Beach FSAR change is [J 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 [I required or [
NOT required. Ifa Regulatory Commitment Change is required, initiate a comnmnitment change per NP 5.1.7.
A Technical Specification Bases change is El required or [D 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 [] required or Z 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,¢CREE,*'iG 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.
"S NO QUESTYYJ El
[
Does the proposed activity involve IN ANY WMANER the dry fuel storage cask(s), the cask transfer/tansport equipment, any ISFSI facility SSC(s), or any ISFSI facility monitoring as follows: Mlulti-Assembly Sealed Basket (MSB), MSB Transfer Cask (MTC),.TC Lifting Yoke, Ventilated Concrete Cask (VCC), Ventilated Storage Cask (VSC), VSC Transporter (VCST), ISFSI Storage Pad Facility, ISFSI Storage Pad Data/Comnmunication Links, or PPCS/ISFSI Continuous Tempernnire to-i-g System?
[]
1 Does the proposed activity involve IN AN-Y MANINER SSC(s) installed in the plant specifically added to support cask loading/unloading activities, as follows: Cask D)ewateming System (CD*,N%, Cask Reflood System (CRF), or Hydrogen Monitoring System?
l []
Does the proposed activity involve IN ANY KMANINNER SSC(s) needed for plant operation which are also used to support cask loading/unloading activ'ities, as follows: Spent Fuel Pool (SFP), SFP Cooling and Filtration (SF),
Primary Auxiliary Building Vent'~on System (VNAB), Drumming Area Ventilation System (VNDPRM),
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?
[]
[
Does the propose* activity involve a change to Point Beach CLB design criteria for external events such as earthquakes, tornadoes, high winds, flooding, etc.?
l
[
Does the activity involve plant heavy load tequirements or procedures for areas of the plant used to support cask loading/unloading activities?
l
[]
Does the activity in olve any potential for fire or explosion i*here casks are loaded, unloaded, transported or stored')
ANY of the Part V questions are answered YES, the-n a full 10 CFR 72.48 screening is required and answers to the questions in rt VI and Part VIl are io be provided. If ALL the qu.stions in Part V are answered NO. then check Parts VI and 1VI as not tppl.cab!e. Complete Part VIII to document the conclusion that no 10 CFR 72.4S e~aluation is required.
PBF-1513c Pcv"sionO 1*4tO1 Refe:ence-NP 5 18
Point Beach Nuclear Plant SCR
- --_____"_1_6__Y 10 CFR 50.59/72.48 SCREENLNG (NEW RULE)
VC5.*, ScR
- -.:C.
r
-T VI (72.48) - DETERMINE IF THE CHANGE INVOLVES A ISFSI LICENSING BASIS DESIGN
.U.VCTIO.V (1
ihe quCsua.ns Ln F,.rt V are N*. then Part VI is Z.NOT APPLICA.ELE.)
Co.a-re the proposcd activity to the relevant portior-of the ISFSI licensing basis and ans, er the following questions:
YES NO QUESTiON F
U Does the proposed activity involve caskfLSFSI Safety Analyses or p!ant'cask'1SF5! structures.
sCiistems znd ccmponents (SSCs) credited in the Safety Analyses?
F--
[]
Does the proposed activity involve plant, cask or ISFSI SSCs that support SSC(s) credited i* the S-afety.Anal' ses"
['-
U]
Does the proposed activity involve plant, cask or ISFSI SSCs %%hcse function is relie* upon for pre ention of a radioactive release, OR whose failure could impact SSC(s) credited in the Safetv Anal' ses?
F, Does the proposed activity involve caslkJISFSI described SSCs or procedural controls that perform functions that arc required by, or othenrise necessary to comply wnith, rcgulaions, license conditions, CoC conditions, or orders" U
Does the activity involve a method of evaluation described in the ISFSI licensing basis9 U-1 Is the activity a test or experiment? (i.e., a non-passive activity %%Dich gthers data)
E
[]
Dc-s the activity exceed or potentially affect a cask design basis limit for afission product ba' r:er (DBLFPBj?
(NOTE: If THIS questions is answ,*ered YES, a 10 CFR 72.48 Evaluatkcn is required.)
If the answers to ALL of these questions are NO, mark Parts VII as not applicable, and document the 10 CFR 72.A4 s:retning in the conclusion section (Part VIII).
,,.v of the above questions are marked YES, identify below the specific design function(s), mrdiod of e. aluation(s) or DBLFPB(s) ed PART V1I (12.4S) - DETERMINE WVHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (N"E196-07.
Appendix B. Section B.4.2. 1)
(1f ALL the questions in Part V or Part VI are answered NO, then Part VII is Z NOT APPLICABLE.)
Ans%% er the follo;'. ing questions to determine if the activity has an adverse effect on a design fiunction. Any YES an.ser means ihat a 10 CFR 72.48 EN' auation is required, EXCEPT where no' t in Part VI1.3.
VII. !
Changes to the Facility or Procedures YES NO QUESTION U
U Does the acrivity adversely affect the design function of a plaint. cask, or ISFSI SSC credi-ed in sat-e:'
anal)ses?
1]
-]
Does the activity adversely affect the method of performing or controilling the designfanzt'n of a pr,-.L cask, or ISFSI SSC credited in the safet' anal, ses?
If any ann., er is YES, a 10 CFR 72.43 Evalu,,aton is required if both answ.vers are NO. d*_sUri the basis f#r- "- cop ci:Sfr (attIch addc.t.onl cd*scussion, as necessary)
Ip.-
l.
Poi.t Be-cl Nuclear Plant S CR eC?6 b20 10 CFR 50.59/72.43 SCREE.NL'G (N-EW RULE)
'o Pa2e 6 C!.anges to a.,,-thod cf Eva:uation (ih-he octivi', cteas not inolhe a method of evalunton these questions are [I NOTAPPPLICABLE)
"YES NO QUESTION Ur-U]
Does the acut'irv use a revised or different method of evaluadion *cr nef.fo.ra-ingsa,", anal; ses than that described in a cask SAR?
U U
Does the activity use a reviscd or different method of evaluation for ev aluating SSCs crei::ted -n safe" anal,% ses than that described in a cask SAR?
If, ansncr is YES, a 10 CFR 72.48 Evaluation is required. Ifloth answers _-c NO descrbe Lhi basis for tie copzlusion (aatach add.tional discussion, as necessary)
VII.3 Tests or E\\-periments (If dhe activity is not a tcst or experiment, the questions in VIt3.a and VII.3.b are U NOT APPLICABLE.)
- a. Answer these t%%o questions first:
YES NO QUESTION U
U Is the proposed test or experiment bounded by othcr tests or expriments that are descnbed in the cask ISFSI licensing basis?
U]
U]
Are the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISFSI facility?
If the answ cr to both questions is N continue to VIH.3.b. If the answer to EITHER question is YES then briefly describe the basis
- b. AnswNer these additional questions ONLY for tests or experiments w*hich do not meet the criteria civen in \\L%.3.a abo e.
If the ainsw. er to either question in VII.3.a is YES, then these t.hree questions are U -NOT APPLICABLE:
YES NO QUESTION U
U Does the activity utilize ot control an SSC in a manner that is omside xhl reference bounds of the desizrn bases as described in the ISFSI licensing basis?
Ul UD Does the activity utilize or control a planwt cask or ISFS1 faciliby SSC in a manner that is LnconsisTent ith the analyses or descriptions in t0= ISFSI licensing basis?
Ul U
Does the activity place the cask cr ISFSI fa.ilibt in a condition not -revioasly evalujaed or g!n' could affcct the capablhity of a plant. cask, or ISFSI SSC to perform its intended f',un:'ions?
If ay answer in VII.3.b is YES, a 10 CFR 72..13 EDaluaticn is required. If dhe a*._7,Ners zie all NO. describe the bcss for te conclusion (anach additional discussion as nccessac-a,)"
PbF.:IiC
Point Ecach Nuclear Planr 10 CFR 50.59172.48 SCREENLNG (N.EW RULE)
P':c r agei Pge "
PART Viii - DOCU IMENT THE CONCLUSION OF THE I0 CFR,42.-8 SCREENLNG Check al' dcat apply:
A 10 CFR ";2 48 Evaluation is 1] requ;red or [] NOT required. Obtain a s:iec-.ng rumtcr wid provide the ori.-;-t. to
,,crds., Mxnacinmen, regardless ofthe :onclusion of the 50.59 or 72.48 scree.-'ru A VSC-24 cask Safety A.nalysis Report change i6 D] required or [D NOT requir-Cc If a VSC-24 cask SA.:. chae.e ;s required, then contact the Point Beach Dry Fuel Storage goup supervisor.
A Reg-ulatory Corrmmitment (CLB Commitmcnt Database) change is [- required or Z NOT required. ira Repz-+/-or-v Commitment Change is reouired, initiate a commitnent change per NP 5.1 7.
A change to the VSC-24 10 CFR 72.212 Siic Evaluation Report is El required or 3 NOT requied. Ira 'SC-4 10 CFR 72.212 Site Evaiuanion Report change is required, then contact the Point Beach Dry-.. Fuel Stomrag group supervisor.
P2 F-151 "c Rc'.:.1,10:1 0
"0"':"
?*
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page I of 63 A. PURPOSE
- 1. This procedure provides directions to respond to a loss of all 480 Vac' power for plant conditions greater than or equal to 350'F and RHR not in service.
- 2.
This procedure is applicable for all plant conditions where RCS hot leg temperature is greater than or equal to 3501F with accumulators in service, and assumes the RER system is not in service for decay heat removal and all SI system components are aligned for normal power operation.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. The symptom of a loss of all AC power is the indication that both 4160 Vac safeguards buses or both 480 Vac safeguards buses are deenergized.
- 2.
This procedure is entered directly-on indication that safeguards buses are deenergized or from the following procedures on the indication that safeguards buses are deenergized:
"o EOP-O UNIT 1. REACTOR TRIP OR SAFETY INJECTION. Step 3 "o EOP-0.0 UNIT.1.
REDIAGNOSIS.
Step 2 "o CSP-S.l UNIT 1. RESPONSE TO NUCLEAR POWER GENERATION/ATWS.
Step Al "o CSP-H.l UNIT 1. RESPONSE TO LOSS OF SECONDARY HEAT SINK. Step Cl C. REFERENCES
- 1. Technical Specifications for Point Beach Nuclear Plant
- 2.
Final Safety Analysis Report for Point Beach Nuclear Plant
- 3.
As-built plant drawings
- 4.
Generic Technical Guidelines developed by the*Westinghouse Owniers Group (WOG).
This consists of the following documents:
- a. Low pressure version of the WOG Optimal Recovery Guidelines. Status Trees. and Functional Restoration Guidelines
- b. Background documents for each low pressure version Optimal Recovery Guideline. Status Tree. and Functional Restoration Guideline
- c.
WOG Emergency Response Guideline Executive Volume
- d.
WOG Emergency Response Guideline Maintenance Program Summary
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 2 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTES
"* Steps 1 and 2 are immediate acti6n steps.
"* Critical safety function status trees should be monitored for information only.
Critical safety procedures (CSPs) shall not be implemented.
Verify Reactor Trip:
"* Check reactor trip breakers and bypass breakers -
OPEN
"* i-52/RTA
"* i-52/RTB
"* 1-52/BYA
"* 1-52/BYB
"* Check neutron flux -
LOWERING
& 1N-35
"-Verify Turbine Trip:
- a. Check turbine stop valves - BOTH SHUT:
"o SL and SR -
SHUT OR "o Annunciator IC03 1E1 4-3.
TURBINE STOP VALVES TWO CLOSED LIT OR o Turbine Valves Closed bistable lights - LIT Manually trip reactor.
- a. Shutdown turbine as follows:
- 1) Depress turbine trip pushbutton.
- 2) IF turbine will NOT trip. THEN perform the following:
a) Manually run back turbine.
b) Stop both EH oil pumps and place in pull-out.
c) IF turbine still has NOT tripped. THEN shut main steam isolation valves.
- IMS-2018 for S/G A a IMS-2017 for S/G B 0
0
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1110/2002 LOSS OF ALL AC POWER Page 3 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Foldout page shall be monitored throughout the remainder of this procedure.
3 Maintain RCS Inventory:
- a. Check PORVs BOTH SHUT
- IRC-430
- IRC-431C
- b.
Check letdown orifice outlet valves -
SHUT
- a. IF PZR pressure less than 2335 psig. THEN manually shut PORVs.
- b. Manually shut valves.
"* ICV-200A
"* ICV-200B
"* 1CV-200C
- c.
Check letdown containment isolation valves -
SHUT a 1CV-371A
- d. Check RCP seal return isolation valve -
SHUT
- e.
Check RCS sample valves -
SHUT
- ISC-966A. PZR steam space sample containment isolation valve I
ISC-966B.
PZR liquid space sample containment isolation valve
RCS hot leg sample containment isolation valve
- f. Check head vent system ENERGIZED
- g. Check reactor vessel head vent solenoids SHUT
- c. Manually shut valves.
- d. Manually shut valve.
- e. Manually shut valves.
- f.
OBSERVE CAUTION PRIOR TO STEP 4 and go to Step 4.
- g. Manually shut valves.
"* IRC-570A. train A
"* IRC-570B. train B
- h.
PZR vent valves -
SHUT
- h. Manually shut valves.
" IRC-580A. train A
"* IRC-580B. train B
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 4 of 63 LOSS OF ALL AC POWER 4
Verify Turbine-Driven AFW Pump Operations:
"* IMS-2020. train A
"* 1MS-2019. train B
- b. Check turbine-driven AFW pump discharge flow -
GREATER THAN OR EQUAL TO 200 GPM a IFI-4002
- b. Establish motor-driven AFW flow from Unit 2 as follows:
i) Ensure Unit 2 motor-driven AFW pump - RUNNING
.P-38B
- 2) Locally open "B" steam generator discharge valve.
e AF-4021. train B
- c. Control pumps and align valves as necessary to maintain S/G level between [51%]
29% and 65%
5 Verify Cooling To AFW Pump Bearings:
- a. Check service water pumps AT LEAST ONE RUNNING "o P-32F. train A "o P-32D. train B "o P-32E. train B
- a. Ensure diesel-driven fire pump RUNNING
- P-35B CAUTION If motor-driven auxiliary feedwater pump flow is greater than 240 gpm.
its motor breaker may trip due to over current.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 5 of 63 i
i p
I ACTION/EXPECTED RESPONSE I -I Check Emergency Diesels -
ALL RUNNING
- G-01. train A
- G-02. train A
- G-03. train B
- G-04. train B 7
Check Diesel Status:
- a. Ensure service water header pressure - GREATER THAN OR EQUAL TO 40 PSIG
- b. Ensure frequency on running diesels - BETWEEN 59.5 HZ AND 60.5 HZ
- c. Ensure voltage on running diesels
- BETWEEN 4050 VAC AND 4300 VAC RESPONSE NOT OBTAINED 6
b) Pull fuel supply operator.
cut-off valve 8
Check Bus IA NOT ENERGIZED Go-"to Step 11.
I CINEPCTDRSOS I
Try to start non-running diesels as follows:
- a. Ensure diesel mode selector switch in AUTO.
- b. Place control switch to START.
- c. Ensure generator field flash occurs automatically.
- d. Ensure green READY TO LOAD light is energized.
IF diesel temperatures, voltage, or frequency can NOT be maintained.
THEN shutdown affected diesel(s):
- 1. Trip affected diesel(s) and place output breaker in pull-out.
o G-O1.
IA52-60 o G-02.
IA52-66 o G-03.
- 2. Place local/rembte switch for affected diesels to local.
- 3. IF any diesel can NOT be shutdown from Control Room.
THEN locally shutdown affected diesel(s):
a) Push both engine stop push-buttons.
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER I
ACTION/EXPECTED ESPONSE I
9 Energize Bus 1A-05 From Normal Diesel Supply G-01:
- a. Check G RUNNING
- b. Ensure IA-03 to 1A-05 bus tie breaker - OPEN e IA52-57
- c.
Check G-01 to bus IA-05 breaker CLOSED e 1A52-60 RESPONSE NOT OBTAINED
- a.
Go to Step 10.
- b.
Go to Step 11.
- c.
Perform the following:
- 1) Try to auto-close breaker by placing control switch to trip position then release.
- 2)
IF breaker will NOT auto-close. THEN perform the following:
a) Place mode selector switch in EXERCISE.
b) Turn synch switch on.
c) At C-02. manually close breaker control switch.
- 3) IF 1A-05 is NOT powered from normal diesel supply G-01.
THE___N go to Step 10.
- d.
Go to Step 11 ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 6 of 63 i
lA I
t
POINT BEACH NUCLEAR PLANT.
ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 7 of 63 I
ACTION/EXPECTED RESPONtE 10
- e.
Check G-02 CLOSED
- IA52-66 to bus 1A-05 breaker -
I I RESPONSE NOT OBTAINED
- a. Go to Step 11.
- b.
Go to Step 11.
Energize Bus 1A-05 From Alternate Diesel Supply G-02:
- a. Check G RUNNING
OPEN 1
- c. Ensure G-O1 to bus IA-05 breaker OPEN AND IN PULL-OUT 1
- d. Unlock and place G-02 to bus 1A-05 breaker control switch in AUTO
- e. Perform the following:
- 1) Try to auto-close breaker by placing control switch to trip position then release.
- 2) IF breaker will NOT auto-close.. THEN perform the following:
a) Place mode selector switch in EXERCISE.
b) Turn synch switch on.
c)At C-02. manually close breaker control switch.
11 Check Bus IA NOT ENERGIZED
- c.
Go to Step 11.
l I
I Go to Step 14.
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 8 of 63 SEI ACTION/EXPECTED RESPONSE I
12 Energize Bus 1A-06 From Normal Diesel Supply G-03:
RESPONSE NOT OBTAINED
- a. Check G-03
- RUNNING
- c. Check G-03 CLOSED
- a. Go to Step 13.
- b. Go to Step 14.
- c. Perform the following:
- 1) Try to.auto-close breaker by placing control switch to trip position then release.
- 2) IF breaker will NOT auto-close. THEN perform the following:
a) Place mode selector switch in EXERCISE.
b) Turn synch switch on.
c) At C-02. manually close breaker control switch.
- 3) IF 1A-06 is NOT powered from normal diesel supply G-03.
THEN go to Step 13.
- d.
Go to Step 14 I
i
POINT BEACH NUCLEAR PLANT POI1NT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 9 of 63 LsEPiI ACTION/EXPECTED RESPONSE I
13 Energize Bus 1A-06 From Alternate Diesel Supply G-04:
- a. Check G RUNNING
- c. Ensure G-03 to 1A-06 bus tie breaker -
OPEN AND IN PULL-OUT e 1A52-80
- d. Unlock and place G-04 to bus 1A-06 breaker control switch in AUTO 1
- e. Check G-04 to bus 1A-06 breaker CLOSED 1
1A52-86 I
RESPONSE NOT OBTAINED
- a.
Go to Step 14.
- b.
Go to SteD 14.
- c.
Go to Step 14.
- e. Perform the following:
- 1) Try to auto-close breaker by placing control switch to trip position then release.
- 2) IF breaker will NOT auto-close. THEN perform the following:'
a) Place mode selector switch in EXERCISE.
b) Turn synch switch on.
c) At C-02. manually close breaker control switch.
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 10 of 63
[s PI ACTION/EXPECTED RESPONSE I
14 Check 480 Vac Safeguards Buses - AT LEAST ONE ENERGIZED o 1B-03. train A o 1B-04. train B RESPONSE NOT OBTAINED Try to restore power to either bus as follows:
- 2) Close bus 1B-03 normal feed.
- . 1B52-16B. train A
- b. IF IA-06 energized.
THEN energize bus 1B-04:
- IA52-84. train B
- 2) Close bus 1B-04 normal feed.
- IB52-17B. train B
- c. IF no 480 Vac safeguards bus can be energized. THEN go to Step 17.
" IB-03. train A
"* 1B-04. train B 15 Verify Service Water System Operation:
- a. Ensure service water header pressure GREATER THAN OR EQUAL TO 40 PSIG 16 Return To Procedure And Step In Effect
- a. Manually start pump(s) and align valves as necessary to establLsh service water header pressure greater than or equal to 40 psig.
Go to EOP-O UNIT
- 1. REACTOR TRIP OR SAFETY INJECTION. Step 4.
II I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 11 of 63 STEP I
ACTION/EXPECTED RESPONSE 17 Check Bus H ENERGIZED Check Bus 1A NOT ENERGIZED Restore Power To Bus IA-03:
- a. Check bus E-02 ENERGIZED
- b. Reset and close bus H-02 feed to IX-04 a E52-22
OPEN' 1
lA52-337
- d. Reset and close bus IA-03 normal feed e IA52-36 I
RESPONSE NOT OBTAINED IF bus H-01 is energized OR bus H-03 energized.
THEN perform the following:
- a. Ensure bus E-02 normal feed OPEN
- H52-20
- b. Ensure H-03 to H-01 bus tie breaker -
CLOSED E
H52-31
- c. Close E-02 to H-01 bus tie breaker.
9 H52-21 Go to Step 20.
IF bus 2A-03 is powered from 2X-04..
THEN cross-feed IA-03 from 2A-03 as follows:
- 1. Ensure bus IA-03 normal feed OPEN I
IA52-36
- 2. Ensure lA-01 to IA-03 bus tie breaker -
OPEN 1
- 3.
Turn on synchronizing switch for.
IA-03 to 2A-03 bus tie breaker.
- 4. Close IA-03 to 2A-03 bus tie breaker.
I IA52-40 18 19
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER L
II ACTION/EXPECTED RESPONSE I
20 Energize Bus lA-05 From 1A-03:
"a.
Check bus 1A-03 ENERGIZED
- b. Ensure G-01 to bus 1A-05 breaker OPEN I
IA52-60
- c. Ensure G-02 to bus IA-05 breaker OPEN I
- d. Turn on synchronizing switch for 1A-03 to IA-05 bus tie breaker I
- e. Trip and close IA-03 to 1A-05 bus tie breaker a 1A52-57 21 Check Bus IA NOT ENERGIZED 22 Restore Power To Bus IA-04:
- a. Check bus H ENERGIZED "b. Reset and close bus E-02 feed to ix-o0 H
152-22
- c. Ensure 1A-02 to IA-04 bus tie breaker -
OPEN 1
- d. Reset and close bus 1A-04 normal feed a 1A52-56 ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 12 of 63 RESPONSE NOT OBTAINED
- a.
Go to Step 21.
Go to Step 23.
IF bus 2A-04 is powered from 2X-04.
THEN cross-feed 1A-04 from 2A-04 as follows:
- 1. Ensure bus lA-04 normal feed OPEN 1
IA52-56
- 2. Ensure 1A-02 to IA-04 bus tie breaker -
OPEN
& IA52-55
I 1A52-52
- 1A52-52 I
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 13 of 63 JI ACTION/EXPECTED RESPONSE I
I RESPONSE NOT OBTAINED 23
- a.
Go to Step 24.
Energize Bus 1A-06 From 1A-04:
- a. Check bus IA ENERGIZED
- b. Ensure G-03 to bus 1A-06 breaker OPEN 1
- c. Ensure G-04 to bus 1A-06 breaker OPEN I
- e. Turn on synchronizing switch for IA-06 to IA-04 bus tie breaker 1
- f. Trip and close IA-06 to 1A-04 bus tie breaker
- IA52-77 I
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 14 of 63 LOSS OF ALL AC POWER I
ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED 24 Check 480 Vac Safeguards Buses - AT LEAST ONE-ENERGIZED o 1B-03. train A o 1B-04. train B 25 Verify Service Water System Operation:
- a. Ensure service water header pressure.-
GREATER THAN OR EQUAL TO 40 PSIG 26 Return To Procedure And Step In Effect Try to restore power to either bus as follows:
- 1) Close bus IA-05 feed to 1X-13.
I 1A52-58. train A
- 2) Close bus 1B-03 normal feed.
- IB52-16B. train A
- b. IF IA-06 energized. THEN energize bus IB-04:
- 1) Close bus IA-06 feed to 1X-14.
I 1A52-84. train B
- 2) Close bus IB-04 normal feed.
o IB52-17B. train B
- c. IF no 480 Vac safeguards bus can be energized. THEN perform the following:
- 1) WHEN power is restored to at least one 480 Vac safeguards bus. THEN go to Step 48.
- 2) OBSERVE NOTE PRIOR TO STEP 27 and continue with Step 27.
- a. Manually start pump(s) and align valves as necessary to establish service water header pressure greater than or equal to 40 psig.
Go to EOP-O UNIT 1. REACTOR TRIP OR SAFETY INJECTION, Step 4.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED "Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 15 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE If SI actuates after placing pumps in pull-out, DO NOT GO TO EOP-0 UNIT 1.
REACTOR TRIP OR SAFETY INJECTION.
Recovery must be made using either ECA-0.1 UNIT 1. LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED.
or ECA-0.2 UNIT 1. LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED.
27 Place The Following Equipment Switches In Pull-Out:
- a. Reactor coolant pump
"* IP-lA. train A
"* IP-lB. train B
- b. Charging pumps
"* 1P-2A. train A
"* IP-2B. train A
"* 1P-2C. train B
- c. RHR pumps
- IP-1OA. train A
- IP-10B. train B
- d. Component cooling pumps
"* BP-IA. train A
"* IP-1IB. train B
- e. Main feed pumps
- 1P-28A. train A e IP-28B. train B
- f. Heater drain tank pumps
- IP-27A. train A
- IP-27C. train A
- IP-27B. tiain B
- g. Motor-driven AFW pump
- P-38A. train'A (Step 27. continued on next Dage)
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 16 of 63 LIL ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED Step 27.
(continued from previous page)
- h. Containment accident fans
"* IW-lAl.. train A
"* IW-IBI. train A
"* IW-iCI. train B
"* IW-IDI. train B
- i.
Containment spray pumps
"* IP-14A. train A
"* lP-14B. train B J.
SI pumps
- a. Condensate pumps
"* 1P-25A. train A
"* IP-25B. train B
- b. Circulating water pumps
"* IP-30A. train A
"* 1P-30B. train B 29 Try To Restore Power To Any 480 Vac Safeguards Bus While Continuing With This Procedure 0
0 0
0 0
0 Locally start G01 per ATTACHMENT A Locally start G02 per ATTACHMENT B Locally start G03 per ATTACHMENT C Locally start G04 per ATTACHMENT D Start gas turbine per ATTACHMENT E Backfeed 480 Vac buses per ATTACHMENT F Place any non-operating pumps in pull-out.
Align equipment to alternate power source per ATTACHMENT G. ALIGNING EQUIPMENT TO ALTERNATE POWER SOURCE.
while continuing with this procedure.
NOTE Refer to ATTACHMENT J for a list of equipment that should be available during a loss of all AC power.
I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 17 of 63 I
ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED Isolate RCP Seals:
- a. Locally shut RCP seal injection throttle valves
"* ICV-300A. RCP A
"* 1CV-300B.
RCP B
- b. Locally shut RCP component cooling return isolation valves
- lCC-759A. RCP A I
ICC-759B. RCP B c.
Check at least one reactor coolant pump seal return containment isolation valve shut "o 1CV-313 "o ICV-313A 31 Verify Condensate Storage Tank Isolated From Condenser Hotwell:
- a. Ensure condenser hotwell low flow
- a.
make-up valve -
SHUT a 1CS-2125
- c. Locally shut RCP seal water return isolation valve.
0 ICV-313 IF valve can NOT be manually shut. THEN locally shut upstream isolation valve.
- b. Ensure condenser manual fill valve -
SHUT e 1CS-86 30 I
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 18 of 63 SAC TI ON /E X PE C TED R ESPON SEI RESPONSE NOT OBTAINED 32 Check S/G Status:
- a. Main steam isolation valves SHUT
- a. Locally shut MSIVs per ATTACHMENT I.
LOCAL SHUTTING OF MSIV.
"* IMS'2018 for S/G A
"* lMS-2017 for S/G B
- b.
MSIV bypass valves -
SHUT
- b. Locally shut valves.
- lMS-234 for S/G A a lMS-236 for S/G B
- c. Feed regulating valves -
SHUT
- c. Manually or locally shut valves.
- d. Feed regulating bypass valves SHUT
- d. Manually or locally shut valves.
"* ICS-480 for S/G A
"* ICS-481 for SIG B
- e. Blowdown isolation valves -
SHUT
- 1MS-5958 for'S/G A
& IMS-5959 for S/G B
- e. Shut blowdown header isolation valves.
- IMS-2042 for S/G A
- IMS-2045 for SIG B I
7 I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 19 of 63
- TJ I
ACTION/EXPECTED RESPONSE 33 Check If Secondary System Is Intact:
"* No S/G pressure dropping in an uncontrolled manner AND
"* No S/G completely depressurized I
I PRESPONSE NOT OBTAINED Isolate faulted S/G(s):
a.. Reset loss of feedwater turbine trip.
- b. Locally shut motor-driven AFW pump discharge valve.
o AF-4023 for S/G A o AF-4021 for S/G B
- c. Shut turbine-driven AFW pump discharge valve.
"o IAF-4001 for S/G A "o 1AF-4000 for S/G B
- d. Shut steam supply valve to turbine-driven AFW pump.
"o IMS-2020 for S/G A "o lMS-2019 for SIG B
- e. Ensure atmospheric steam dump SHUT o lMS-2016 for S/G A o IMS-2015 for SIG B CAUTIONS
"* A faulted or ruptured S/G that -is isolated should remain isolated.
"* Steam supply to the turbine-driven AFW pump must be maintained from at least one SIG.
I
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT I SAFETY RELATED Revision 30 1/10/2002 Page 20 of 63 ACTION/EXPECTED R.ESPONSE I
I RESPONSE NOT OBTAINED 34
- c. Main steam line e IRE-231 for S/G A e 1RE-232 for S/G B
- d. Activity samples J
- e. Local surveys
"* Request local surveys of main steam lines
"* Request Chemistry to prepare for periodic activity samples of both SIGs I
Check If S/G Tubes Are Intact:
Check secondary system radiation levels -
NORMAL
- a. Condenser air ejector
"* 1RE-215
"* RE-225
- b. S/G blowdown
- IRE-219
-FT Perform the following:
- a. Identify ruptured S/G(s) while continuing with this procedure.
- b.
WHEN ruptured S/G identified, THEN isolate ruptured S/G(s) as follows:
- 1) Locally shut motor-driven AFW pump discharge valve.
o AF-4023 for S/G A o AF-4021 for S/G B
- 2) Shut turbine-driven AFW pump discharge valve.
"o IAF-4001 for SIG A "o 1AF-4000 for S/G B
- 3) Shut steam supply valve to turbine-driven AFW pump.
"o lMS72020 for S/G A "o IMS-2019 for S/G B
- 4) WHEN ruptured S/G pressure less than 1050 psig, THEN ensure atmospheric steam dump shut.
o lMS-2016 for S/G A o IMS-2015 for S/G B
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 21 of 63 I STEP II ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED
- 35 Stabilize Intact S/G Level:
- a. Intact SIG level -
GREATER THAN
[51%] 29%
- b. Control feed flow to maintain intact S/G level between
[51%] 29% and 65%
- a. Maintain total feed flow greater than 200 gpm until level greater than [51%] 29% in at least one SIG.
- b. IF level in intact S/G continues to rise in an uncontrolled manner. THEN isolate ruptured S/G as follows:
- 1) Locally shut motor-driven AFW pump discharge valve.
o AF-4023 for S/G A o AF-4021 for S/G B
- 2) Shut turbine-driven AFW pump discharge valve.
o IAF-4001 for SIG A o.IAF-4000 for SIG B
- 3) Shut steam supply valve to turbine-driven AFW pump.
o lMS-2020 for S/G A o IMS-2019 for S/G B
- 4) WHEN ruptured SIG pressure less than 1050 psig. THEN ensure atmospheric steam dump shut.
o IMS-2016 for S/G A o lMS-2015 for S/G B I
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 22 of 63 STP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE Batteries have the following ratings:
D-05 and D-06 225 AMPS for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D-105 and D-106 187 AMPS for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1D-205 and 2D-205 233 AMPS for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> D-305 225 AMPS for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Verify Adequate DC Capacity:
- a.
Check battery bus D-01 voltage GREATER THAN 105 VDC
- a. Perform:one of the following:
o Restore battery chargers.*
OR o Shift to alternate battery per 0-SOP-DC-001.
BUS D-O1 &
COMPONENTS.
37 Check CST Level -
GREATER THAN 8 FEET 38 Check Intact S/G Levels -
GREATER THAN [51%] 29% IN AT LEAST ONE S/G Switch to alternate AFW suction supply per AOP-23 UNIT 1.
ESTABLISHING ALTERNATE AFW SUCTION SUPPLY. while continuing with this procedure.
Perform the following:
- a. Control atmospheric steam dump as necessary to stabilize RCS temperatures.
- b. Establish and maintain maximum feed flow until level greater than [51%] 29% in at least one intact S/G.
- c.
WHEN level greater than [51%]
29%
in at least one intact S/G. THEN do Step 39.
Continue with Step 40.
- 36
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 23 of 63 I
ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED CAUTION S/G pressures should be maintained greater than 200 psig to prevent injection of accumulator nitrogen into the RCS.
NOTES
"* Depressurization of S/Gs will result in SI actuation from low steam lane pressure.
"* The S/Gs should be depressurized at the maximum rate achievable while maintaining intact SIG levels greater than [51%] 29% to minimize RCS inventory loss.
Cooldown may exceed 100*F in one hour.
e PZR level may be lost and reactor vessel upper head voiding may occur due to depressurization of S/Gs.
Depressurization should not be stopped to prevent.
these occurrences.
39 Depressurize Intact S/Gs To 230 PSIG To Maintain RCS Inventory:
- a. Manually dump steam at maximum rate using atmospheric steam dump
- a. Locally dump steam using atmospheric steam dump.
"* lMS-2016 for S/G A
" 1MS-'2015 for S/G B
- b. Check RCS cold leg temperatures GREATER THAN 345OF
- c.
Check S/G pressures
- LESS THAN 230 PSIG
- d. Control atmospheric steam dumps as necessary to maintain S/G pressures between 200 psig and 230 psig
- b. Control atmospheric steam dump as necessary to stop S/G depressurization.
Go to Step 40.
- c.
WHEN S/G pressures are 230 psig. THEN do Step Continue with Step 40.'
less than 39.d.
- d. Manually or locally control atmospheric steam dumps to maintain S/G pressures between 200 psig and 230 psig.
- IMS-2016 for S/G A
& IMS-2015 for S/G B I
I i
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 24 of 63 SEP]l ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 40 Check Reactor Subcritical:
Control atmospheric steam dump as necessary to stop S/G
- a. Check intermediate range startup depressurization and allow RCS to rate -
ZERO OR NEGATIVE heat up.
- 1N-35. train A I
1MS-2016 for S/G A e IN-36. train B l
IMS-2015 for SIG B
- b. Check source range startup rate -
ZERO OR NEGATIVE
- IN-31. train A
HAS ACTUATED WHEN SI actuated. THEN do Steps 42.
43 and 44.
Continue with Step 45.
42 Reset SI 43 Verify Containment Isolation:
- a. Check containment isolation panels "A" and "B" ALL LIGHTS LIT
- b. Check other valves -
SHUT
"* RS-SA-9. Unit 1 steam supply to rad waste system Any valve which may be open under administrative control 44 Check Containment Pressure Recorder HAS REMAINED LESS THAN 25 PSIG 1
1PR-968 1
IPR-969 45 Check Containment Radiation - LESS THAN 4 R/HR
- a. Perform the following:
- 1) Manually actuate Containment Isolation.
- 2) IF any valve open AND flow path NOT required. THEN shut valve(s).
Refer to ATTACHMENT H.
- b. Manually shut valve(s).
Reset containment spray signal.
IF any system in ATTACHMENT H'NOT needed. THEN shut associated containment isolation valve(s).
POINT BEACH NUCLEAR PLANT ECA-0.O UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 25 of 63 lI, I
I 2
I ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED 46 Check Core Exit Thermocouple Temperatures -
LESS THAN 1200'F 1
lTR-00001A I
1TR-O0001B 47 Check 480 Vac Safeguards Buses - AT LEAST ONE ENERGIZED "o IB-03. train A "o 1B-04. train B 48 Verify Service Water System Operation:
- a. Ensure service water header pressure -
GREATER THAN OR EQUAL TO 40 PSIG 49 Stabilize S/G Pressure:
- a. Control atmospheric steam dumps as necessary to maintain S/G
".pressures stable
"* IMS-2016 for S/G A
"* lMS-2015 for S/G B IF core exit temperatures are greater than 1200'F AND trending higher. THEN go to SACRG-1, SEVERE ACCIDENT CONTROL ROOM GUIDANCE INITIAL RESPONSE.
Continue to control RCS conditions and monitor plant status:
- a. Check status of local actions:
- AC power restoration
- RCP seal isolation
- DC control power
- b. OBSERVE NOTE PRIOR TO STEP 29 and return to Step 29.
- a. Manually.start pump(s) and align valves as necqessary to establish service water header pressure greater than or equal to 40 psig.
- a. Manually or locally control atmospheric steam dumps to maintain SIG pressures stable.
Li I
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER I I ACTION/EXPECTED RESPONSE I
50 Verify Necessary Equipment Loaded On 480 Vac Buses:
- a. Emergency AC lighting and plant alarms
- b. Check B-33 and B AT LEAST ONE ENERGIZED
- c. Communications
- 1) Gai-tronics
- 2) Radio communications ECA-0.0 UNIT I SAFETY RELATED Revision 30 1/10/2002 Page 26 of 63 RESPONSE NOT OBTAINED
- a. Energize 1B-42 or IB-32.
- b.
Observe operator aid on C02R and reenergize B-33 or B-43.
- c. Perform one of the following:
a).IF B-33 is energized. THEN position transfer switch to B-33.
- B-50 b) IF B-33 is NOT energized. THEN position.transfer switch to B-43.
a B-50 7
CAUTION Placing loads on energized AC safeguards buses in excess of the power source's capacity could result in loss of the power source.
Refer to AOP-22 UNIT 1. EDG LOAD MANAGEMENT, for KW ratings.
I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2602 LOSS OF ALL AC POWER Page 27 of 63 I
II ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED 51 Select Recovery Procedure:
- a.
Check RCS subcooling based on core exit thermocouples GREATER THAN [800F) 35°F
- b. Check PZR level GREATER THAN
[34%) 10%
- c. Check SI -
HAS NOT ACTUATED UPON AC POWER RESTORATION
- a.
Go to ECA-0.2 UNIT 1, LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED.
- b.
Go to ECA-0.2 UNIT 1. LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED.
- c.
Perform the following:
- 1) Reset SI.
- 2)
Go to ECA-0.2 UNIT 1. LOSS OF ALL AC POWER RECOVERY WITH SI REQUIRED.
- d.
Go to ECA-0.1 UNIT 1. LOSS OF ALL AC POWER RECOVERY WITHOUT SI REQUIRED*
-END-l I
I II I
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 28 of 63 I S, ECA-O.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 28 of 63 TEP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I
ATTACHMENT A (Page 1 of 5)
G-01 LOCAL MANUAL START Al Check Green "Power On" Light -
IF. green light is NOT lit, THEN ENERGIZED transfer control power to alternate source:
" Panel C-64A "a
Panel C-34
- a. At PAB 8' elevation South of Unit 2 charging pumps, direct PAB operator to shut switch D31-01.
- b. At C-78.
shift to alternate power by swapping paired breakers:
- For annunciators, open breaker I and close breaker 2.
a For start circuit
- 1. open breaker 3 and close breaker 4.
e For control power.
open breaker 5 and close breaker 6.
- For field flash, open breaker 7 and close breaker 8.
A2 Check Overspeed Trip Alarms CLEAR Reset mechanical overspeed trip and alarms as follows:
"* Panel C-64A
"* Panel C-34
- a. Place mode selector switch in Local position.
- b. Reset mechanical overspeed trip.
- c. Place mode selector switch in Auto position.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 29 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT A (Page 2 of 5)
G-01 LOCAL MANUAL START A3 Emergency Start G-01:
- a. At C-34A. place local/remote transfer switches to local
"* Transfer switch No.
1
"* Transfer switch No.
2
- b. At C-34A. start G-01 by
- b. IF-G-01 will N0T emergency start.
depressing "EMERGENCY START" THEN manually start G-01:
push-button
- 1) At C-64. place mode selector switch in "LOCAL START".
- 2) At C-64. depress and hold "ENGINE START" push-button until engine speed rises to idle.
- 3) At C-64. raise engine speed to 900 rpm by depressing idle release push-buttofn.
- 4) IF G-01 can NOT be started.
THEN do not continue and inform Control Room of G-01
- status.
A4 At C-64. Check G-01 Speed -
GREATER Perform the following:
- a. At C-64. place governor mode switch to "HYD" position.
- b. IF diesel speed NOT greater than or equal to 900 RPM, THEN at C-64. raise speed using hydraulic governor control switch.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 30 of 63 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT A (Page 3 of 5)
G-01 LOCAL MANUAL START A5 Contact Control Room To Check G-01 Frequency -
BETWEEN 59.5 HZ AND 60.5 HZ Perform the following:
- a. IF field is NOT flashed.
THEN in C-34. firmly raise and hold bottom plunger of relay FFC for 3 seconds.
- b.
IF hydraulic governor control switch NOT functional. THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
- c.
IF frequency can NOT be maintained.
THEN locally shutdown G-01:
- 1) Push both engine stop push-buttons.
- 2) Pull fuel supply cut-off valve operator.
- 3)
At C-34A. place output breakers in pull-out.
- 4) Return to procedure and step in effect.
Ls=jI CAUTION Rubber gloves with leather protectors and personnel safety equipment are required to locally operate FFC relay plunger.
l II
POINT BEACH NUCLEAR PLANT ECA-0.O UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2062 LOSS OF ALL AC POWER Page 31 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT A (Page 4 of 5)
G-O1 LOCAL MANUAL START A6 Contact Control Room To Check G-01 Perform the following:
Voltage BETWEEN 4050 VAC AND 4300 VAC
- a. IF field is NOT flashed. THEN in C-34'. firmly raise and hold bottom plunger'of relay FFC for 3 seconds.
- b. Contact Control Room to maintain voltage between 4050 Vac and 4300 Vac by adjusting diesel loading.
- c.
IF voltage can NOT be maintained.
THEN locally shutdown G-01:
- 1) Push both engine stop push-buttons.
- 2) Pull fuel supply cut-off valve operator.
- 3) At C-34A. place output breakers in pull-out.
o IA52-60 o 2A52-73
- 4) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 32 of 63 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT A (Page 5 of 5)
G-01 LOCAL MANUAL START A7 Energize Bus 1A-05 From Normal Diesel Supply G-01:
- a. Check G-O1 - RUNNING
OPEN
- a. Return to effect.
- b. Return to effect.
procedure and step in procedure and step in e IA52-57
- c. Locally ensure G-02 to lA-05 bus tie breaker control switch OPEN AND IN PULL-OUT
- c. Return to effect.
vrocedure and sten in
- IA52-66
- d. Locally ensure G-01 to bus 1A-05 breaker control switch - IN AUTO e IA52-60
- e. Locally check G-01 to bus IA-05 breaker - CLOSED a 1A52-60 A8 Check Bus IB ENERGIZED
- d. Place G-01 to bus 1A-05 breaker control switch in AUTO.
- e. Locally perform the following:
- 1) Try to auto-cldse breaker by placing control switch to trip position then release.
- 2)
IF breaker will NOT auto-close. THEN manually close breaker control switch.
IF bus 1A-05 energized, THEN energize bus 1B-03:
- a. Close bus lA-05 feed to IX-13.
1 IA52-58. train A
- b. Close bus 1B-03 normal feed.
e IB52-16B. train A A9 Return To Procedure And Step In Effect
-END-
POINT BEACH NJCLEAR PLANT POINT BEACH NTUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 33 of 63 SE I
ACTION/EXPECTED RESPONSE I
I RESPONSE NOT OBTAINED ATTACHMENT B (Page 1 of 5)
G-02 LOCAL MANUAL START 81 Dispatch Operator With Key #43 To G-02 B2 Check Green "Power On' Light -
IF grel ENERGIZED
- transf, source
"* Panel C-65A
"* Panel C-35
- a. At Uni wat4
- b. At by
- F b
- F b
eF b
- F a
B3 Check Overspeed Trip Alarms -
CLEAR Reset alarms
- Panel C-65A
.Panel C-35
- a. Pla Loc
- b. Res
- c. Pla Aut en light is NOT lit.
THEN er control power to alternate PAB 8' elevation South of t 2 charging pumps, direct PAB ch to shut switch D41-01.
C-79. shift to alternate power swapping paired breakers:
or annunciators, open reaker 1 and close breaker 2-.
or start circuit 1. open reaker 3 and close breaker 4.
or control power. open reaker 5 and close breaker 6.
or field flash, open breaker 7 nd close breaker 8.
mechanical overspeed trip and as follows:
ce mode selector switch in al position.
et mechanical overspeed trip.
ce mode selector switch in o position.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 34 of 63 L iI ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED ATTACHMENT B (Page 2 of 5)
G-02 LOCAL MANUAL START B4 Emergency Start G-02:
- a. At C-35A. place local/remote transfer switches to local
"* Transfer switch No. 1
"* Transfer switch No.
2
- b. At C-35A. start G-02 by depressing "EMERGENCY START" push-button B5 At C-65. Check-G-02 Speed - GREATER THAN OR EQUAL TO 900 RPM
- b. IF G-02 will NOT emergency start.
THEN manually start G-02:
- 1) At C-65. place mode selector switch in "LOCAL START".
- 2) At C-65, depress and hold "ENGINE START" push-button until engine speed rises to idle.
- 3) At C-65. raise engine speed to 900 rpm by depressing idle release push-button.
- 4) IF G-02 can NOT be started.
THEN do not continue and inform Control Room of G-02 status.
Perform the following:
- a. At C-65. place governor mode switch to "HYD" position.
- b.
IF diesel speed NOT greater than or equal to 900 RPM.
THEN at C-65. raise speed using hydraulic governor control switch.
I I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 35 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT B (Page 3 of 5)
G-02 LOCAL MANUAL START "CAUTION Rubber gloves with leather protectors and personnel safety equipment are required to locally operate FFC relay plunger.
B6 Check.G-02 Frequency -
BETWEEN Perform the following:
59.5 HZ AND 60.5 HZ
- a. IF field is NOT flashed. THEN in C-35. firmly raise and hold bottom plunger of relay FFC for 3 seconds.
- b. IF hydraulic governor control switch NOT functional. THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
- c. IF frequency can NOT be maintained.
THEN locally shutdown G-02:
- 1) Push both engine stop push-buttons.
- 2) Pull-fuel supply cut-off valve operator.
- 3) At C-35A. place output breakers in pull-out.
- 4) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 36 of 63 ACTION/EXPECTED RESPONSE I
I RESPONSE NOT OBTINED I
ATTACHMENT B (Page 4 of 5)
G-02 LOCAL MANUAL START B7 Check G-02 Voltage -
BETWEEN Perform the following:
4050 VAC AND 4300 VAC
- a.
IF field is NOT flashed. THEN in.
C-35.
firmly rai~e and hold bottom plunger of relay FFC for
.3 seconds.
- b. Contact Control Room to maintain voltage between 4050 Vac and 4300 Vac by adjusting diesel loading.
- c. IF voltage can NOT be maintained.
THEN locally shutdown G-02:
- 1) Push both engine stop push-buttons.
- 2) Pull fuel supply cut-off valve operator.
- 3) At C-35A. place output breakers in pull-.out.
- 4) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 37 of 63 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT B (Page 5 of 5)
G-02 LOCAL MANUAL START B8 Energize Bus 1A-05 From Alternate Supply G-02:
- a. Check G RUNNING
- b. Locally ensure 1A-03 to IA-05 bus tie breaker - OPEN a.
Return to effect.
- b. Return to effect.
procedure and step in procedure and step in a IA52-57
- c. Locally ensure G-01 to 1A-05 bus tie breaker control switch OPEN AND IN PULL-OUT
- IA52-60
- d. Locally unlock and place G-02 to bus 1A-05 breaker control switch in AUTO I
IA52-66
- e. Locally check G-02 to bus 1A-05 breaker -
CLOSED e IA52-66 B9 Check Bus 1B ENERGIZED
- c. Return to effect.
procedure and step in
- e. Locally perform the following:
- 1) Try to auto-close breaker by placing control switch to trip position then release.
- 2) IF breaker will NOT auto-close. THEN manually close breaker control switch.
IF bus lA-05 energized. THEN energize bus 1B-03:
- a. Close bus IA-05 feed to 1X-13.
l 1A52-58. train A
- b. Close bus 1B-03 normal feed.
a 1B52-16B. train A B10 Return To Procedure And Step In Effect
-END-I ACTION/EXPECTED RESPONSE I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 38 of 63 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT C (Page 1 of 5)
G-03 LOCAL MANUAL START C1 At D-28. Check 125 Vdc Control Power GREATER THAN 105 VDC C2 At Panel C-101.
Alarms -
CLEAR Check Overspeed Trip Switch to alternate 125 Vdc control power
- a. IF power to D-40 NOT available.
THEN return -o procedure and step in effect.
- b. At D-40, place fused disconnect to "ON".
- D72-40-13
- c. At D-28, place main power transfer switch to "OFF".
- D72-28-M
- d. At D-28. place alternate power transfer switch to "ON".
- D72-28-A
- e. At D-28. check voltage -
GREATER THAN 105 VDC
- f. IF 125 Vdc control power is NOT available. THEN return to procedure and step in effect.
Reset overspeed trip and alarm:
- a. Pull down on.gray reset lever to
-the latched position.
- b. At C-101. reset alarm.
Li CAUTION Flashing fields on both G-03 and G-04 at the same time will overload the 125 vdc power supply.
II
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 39 of 63 L~EiI ACTION/EXPECTED RESPONSE I
RESPONSE NOT OBTAINED ATTACHMENT C (Page 2 of 5)
G-03 LOCAL MANUAL START C3 Auto Start G-03:
- a. At C-81. place local-remote switch to "LOCAL"
- b. Ensure governor mode switch in "AUTO"
- c. At C-81. depress "SHUTDOWN RESET" push-button
- d. Check G-03 RUNNING
- d. Fast start G-03:
- 1) At C-81. depress "FAST START" push-button.
- 2) IF G-03 did NOT fast start.
THEN manually start G-03:
a) At C-81. depress "SHUTDOWN RESET" push-button.
b) At C-81. depress "VOLTAGE SHUTDOWN RESET" push-button.
c) At C-81. depress "ALARM RESET" push-button.
d) At C-81. place governor mode switch to "HYD" position.
e) IF G-03 is NOT running.
THEN depress "FAST START" push-button.
- 3) IF G-03 can NOT be started.
THEN return to procedure and steD in effect.
C4 At C-81. Depress "ALARM RESET" Push-Button
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER
'Page 40 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT C (Page 3 of 5)
G-03 LOCAL MANUAL START C5 Check G-03 Frequency BETWEEN Perform the following:
59.5 HZ AND 60.5 HZ
- a.
IF field is NOT flashed, THEN at C-81. depress "VOLTAGE SHUTDOWN RESET" push-button.
- b.
IF hydraulic governor control switch NOT functional. THEN adjust frequency using SPEED control knob on faceplate.of Woodward governor.
c..IF frequency can NOT be maintained. THEN locally shutdown G-03:
I)
Push both engine stop push-buttons.
a Black engine stop e'Red emergency stop
- 2) At C-81. place output breaker in pull-out.
- 3) Return to prbcedure and step in effect.
I I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 41 of 63 SE ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT C (Page 4 of 5)
G-03 LOCAL MANUAL START C6 Check G-03 Voltage -
BETWEEN Perform the following:
4050 VAC AND 4300 VAC
- a.
IF field is NOT flashed. THEN at C-81. depress "VOLTAGE SHUTDOWN RESET" push-button.
- b.
IF voltage can NOT be maintained.
THEN locally shutdown G-03:
- 1) Push both engine stop push-buttons.
a Black engine stop a Red emergency stop
- 2) At C-81. place output breaker in pull-out.
o IA52-80 o 2A52-87
- 3) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 42 of 63 I ACTION/EXPECTED RESP.ONSE I I RESPONSE NOT OBTAINED ATTACHMENT C (Page 5 of 5)
G-03 LOCAL MANUAL START C7 Energize Bus 1A-06 From Normal Diesel Supply G-03:
- a. Check G-03 RUNNING
- b. Locally ensure IA-04 to 1A-06 bus tie breaker -
OPEN
- a. Return to effect.
- b. Return to effect.
procedure and step in procedure and step in
- c. Locally ensure G-04 to IA-06 bus tie breaker control switch OPEN AND IN PULL-OUT
- c. Return to effect.
procedure and step in
- IA52-86
- d. Locally ensure G-03 to bus 1A-06 breaker control switch - IN AUTO
- d. Place G-03 to bus 1A-06 breaker control switch in AUTO.
- e. Locally check G-03 to bus 1A-06 breaker -
CLOSED a IA52-80 C8 Check Bus 1B ENERGIZED
- e. Locally perform the following:
i) Try to auto-close breaker by placing control switch to trip position then release.
- 2) IF breaker will NOT auto-close.
THEN place synch switch to "ON" and manually close breaker control switch.
IF 1A-06 energized. THEN energize bus IB-04:
- a. Close bus 1A-06 feed to IX-14.
1 1A52-84. train B
- b. Close bus 1B-04 normal feed.
- iB52-17B. train B C9 Return To Procedure And Step In Effect
-END-ISTj I
CINEPCTDRSOS I
I I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 43 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT D (Page 1 of 5)
G-04 LOCAL MANUAL START D1 Dispatch Operator With Key #43 To G-04 CAUTION Flashing fields on both G-03 and 0-04 at the same time will overload the 125 vdc power supply.
D2 At D-40. Check 125 Vdc Control Power Switch to alternate 125 Vdc control GREATER THAN 105 VDC power:
- a. IF power to D-28 NOT available.
THEN return to procedure and step in effect.
- b. At D-28. place fused disconnect to "ON".
- D72-28-13
- c. At D-40. place main power transfer switch to "OFF'.
- D72-40-M
- d. At D-40. place alternate power transfer switch to "ON".
- D72-40-A
- e.
At D-40. check voltage GREATER THAN 105 VDC
- f.
IF 125 Vdc control power is NOT available.
THEN return to procedure and step in effect.
D3 At Panel C-102.
Check Overspeed Trip Reset overspeed trip and alarm:
Alarms - CLEAR
- a. Pull down on gray reset lever to the latched position.
- b. At C-102. reset alarm.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 44 of 63 STE 1
ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT D (Page 2 of 5)
G-04 LOCAL MANUAL START D4 Auto Start G-04:
- a. At C-82. place local-remote switch to "LOCAL"
- b. Ensure governor mode switch in "AUTO"
- c.
At C-82. depress "SHUTDOWN RESET" push-button
- d. Check G-04 RUNNING
- d. Fast start G-04:
- 1) At C-82. depress "FAST START" push-button.
- 2)
IF G-04 did NOT fast start.
THEN manually start G-04:
a) At C-82. depress "SHUTDOWN RESET" push-button.
b) At C-82. depress "VOLTAGE SHUTDOWN RESET" push-button.
c) At C-82. depress "ALARM RESET" push-button.
d) At C-82, place governor mode switch to "HYD" position.
e) jF G-04 is NOT running.
THEN depress "FAST START" push-button.
- 3)
IF G-04 can NOT be started.
THEN return To procedure and step in effect.
D5 At C-82. Depress "ALARM RESET" Push-Button I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY'RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 45 of 63 IsEPI ACTION/EXPECTED R*ESPONSE RESPONSE NOT OBTAINED ATTACHMENT D (Page 3 of 5)
G-04 LOCAL MANUAL START D6 Check G-04 Frequency
- BETWEEN Perform the following:
59.5 HZ AND 60.5 HZ
- a. IF field is NOT flashed. THEN at C-82. depress "VOLTAGE SHUTDOWN RESET" push-bucton.
- b. IF hydraulic governor control switch NOT functional. THEN adjust frequency using SPEED control knob on faceplate of Woodward governor.
- c. IF frequency can NOT be maintained. THEN locally shutdown G-04:
- 1) Push both engine stop push-buttons.
a Black engine stop a Red emergency stop
- 2) At C-82. place output breaker in pull-out.
- 3) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 46 of 63 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT D (Page 4 of 5)
G-04-LOCAL MANUAL START D7 Check G-04 Voltage -
BETWEEN Perform the following:
4050 VAC AND 4300 VAC
- a. IF field is NOT flashed, THEN at C-82. depress "VOLTAGE SHUTDOWN RESET" push-button.
- b. IF voltage can NOT be maintained.
THEN locally shutdown G-03:
- 1) Push both engine stop push-buttons.
"* Black engine stop Red emergency stop
- 2) At C-82. place output breaker in pull-out.
o IA52-86 o 2A52-93
- 3) Return to procedure and step in effect.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 47 of 63 I
ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT D (Page 5 of 5)
G-04 LOCAL MANUAL START D8 Energize Bus 1A-06 From Alternate Supply G-04:
- a. Check G RUNNING
- b. Locally ensure IA-04 to IA-06 bus tie breaker -
OPEN
- a. Return to effect.
- b. Return to effect.
procedure and step in procedure and step in
- c. Locally ensure G-03 to 1A-06 bus tie breaker control switch OPEN AND IN PULL-OUT a IA52-80
- d. Locally unlock and place G-04 to bus IA-06 breaker control switch in AUTO
- IA52-86
- e. Locally check G-04 to bus 1A-06 breaker
- CLOSED e 1A52-86 D9 Check Bus 1B ENERGIZED
- c. Return to effect.
Drocedure and steD in
- e. Locally perform the following:
I) Try to auto-close breaker by placing control switch to trip position then release.
- 2) IF breaker will NOT auto-close. THEN place synch switch to "ON" and manually close breaker control switch.
IF IA-06 energized.
THEN energize bus IB-04:
- a. Close bus IA-06 feed to 1X-14.
e IA52-84. train B
- b. Close bus 1B-04 normal feed.
- 1B52-17B. train B D10 Return To Procedure And Step In Effect
-END-Li I
0
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 48 of 63 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT E (Page 1 of 5)
POWER RESTORATION USING GAS TURBINE El Start Gas Turbine:
- a. Check G-05 SHUTDOWN
- b. Check G-501 OPERATING
- c.
On C02R. check G-05 remote system operable light - LIT
- d.
On C02R. check "ready to start" light - LIT
- e. Place rate selector switch to "FAST"
- f. Depress "START" push-button
- g. Depress "MINIMUM LOAD" push-button
- h. Check G-05 "READY TO SYNCH" light LIT
- 152-10
- i. Close power to/from H-01 bus.
o H52-10 E2 Continue Operation Of Gas Turbine per 01-110. GAS TURBINE OPERATION
- a.
Go to Step E2.
- b. Locally start G-501.
- 1) Place Auto-Off-Manual switch to MANUAL position.
- c. Align G-501 output to supply G-05 auxiliaries:
- 1) Locally open TSC emergency supply breaker.
- 52T
- 2) Locally close G-05 auxiliaries back-up supply breaker.
- 52E
- d. Manually align gas turbine auxiliaries per OI-I0..GAS TURBINE OPERATION.
- h.
DO NOT CONTINUE until G-05 "READY TO SYNCH" light is lit.
l ill CAUTION Local manual closure of breakers bypasses all lockouts and interlocks and may result in equipment damage.
I I
I
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~iS PI ACTION/EXPECTED RESPONSE i
RESPONSE NOT OBTAINED ATTACHMENT E (Page 2 of 5)
POWER RESTORATION USING GAS TURBINE E3 Ensure The Following Electrical Breakers Open:
"* 1A52-01. bus 1A-01 normal feed
"* 1A52-17. bus 1A-02 normal feed
"* 1A52-40.
1A-03 to 2A-03 bus tie breaker
"* 1A52-52.
1A-04 to 2A-04 bus tie breaker
"* 1A52-36. bus 1A-03 normal feed
"* 1A52-56. bus 1A-04 normal feed
"* 1A52737.
1A-03 to IA-01 bus tie breaker
"* 1A52-55.
1A-04 to IA-02 bus tie breaker E4 Isolate 13.8 kV Bus From Switchyard By Opening Bus H-02 Normal Feed
- H52-20 E5 Energize Bus E-01 And Bus H-02 From G-05:
- a. Close G-05 main breaker a H52-G05
- b. Close E-02 to E-01 bus tie breaker
- H52-21 E6 Check Bus H ENERGIZED IF breakers will not close. THEN perform the following:
- 1. Check status of the following lockouts.
" IX-04 lockout
"* H-01 bus lockout H-02 bus lockout
- 2. IF lockouts are NOT tripped. THEN locally shut stored energy breakers.
IF
_H-02 can NOT be energized.
THEN perform the following:
- a. Shut down gas turbine.
- b. Operate G-501 to power the TSC per 01-35. ELECTRICAL EQUIPMENT OPERATION.
- c. Return to procedure and steD in effect.
I I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 50 of 63 SV E NOT OBTAINED I
I EP ACTION/EXPECTED RESPONSE I
I SPO ATTACHMENT E (Page 3 of 5)
POWER RESTORATION USING GAS TURBINE E7 Restore Power To Bus IA-03 And Bus 1A-.04-:
- a. Reset and close bus H-02 feed to IX-04 H
E52-22 b..Reset and close bus 1A-03 normal feed
- c. Reset and close bus IA-04 normal feed I
IA52-56 E8 Check Bus IA-03 And Bus IA AT Return to pr LEAST ONE ENERGIZED effect.
E9 Energize Bus 1A-05 From IA-03:
- a. Check bus 1A-03 ENERGIZED
- a.
Go to Ste
- b. Ensure G-01 to bus 1A-05 breaker OPEN I
- c. Ensure G-02 to bus IA-05 breaker OPEN I
- d. Turn on synchronizing switch for lA-03 to 1A-05 bus tie breaker
- e. Trip and close 1A-03 to IA-05 bus tie breaker
- IA52-57 p El0.
ocedure and step in
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 51 of 63 LSTPI ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT E (Page 4 of 5)
POWER RESTORATION USING GAS TURBINE E10 Energize Bus IA-06 From IA-04:
- a. Check bus IA ENERGIZED
- a. Go to Step Ell.
- b. Ensure G-03 to bus 1A-06 breaker OPEN 1
- c. Ensure G-04 to bus 1A-06 breaker OPEN I
IA52-86
- d. Trip and close IA-04 to 1A-06 bus tie breaker 1
IA52-54
IA52-77 Eli Check Bus 1A-05 And Bus 1A-06
- AT Return to procedure and step in LEAST ONE ENERGIZED effect.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 52 of 63 ACTION/EXPECTED RESPONSE 1
RESPONSE NOT OBTAINED ATTACHMENT E (Page 5 of 5)
POWER RESTORATION USING GAS TURBINE E12 Check 480 Vac Safeguards Buses AT Try to restore power to either bus LEAST ONE ENERGIZED as follows:
"o 1B-03.
train A
- a.
IF bus IA-05 energized.
THEN "o IB-04.
train B
energize bus 1B-03:
- 1) Close bus 1A-05 feed to IX-13.
1 1A52-58.
train A
- 2)
Close bus IB-03 normal feed.
train A
- b. IF 1A-06 energized.
THEN energize bus IB-04:
- IA52-84.
train B
- 2)
Close bus 1B-04 normal feed.
- IB52-17B.
train B
E13 Return To Procedure And Step In Effect
-END-
POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.O UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 53 of 63 NOT OBTAINED ACTION/EXPECTED RESPONSE I
RESPONS ATTACHMENT F (Page 1 of 3)
BACKFEED TO 480 VAC SAFEGUARDS BUSES F1 Energize Bus IA-01 From Bus 1A-03 And Bus IA-02 From Bus 1A-04:
- a. Close 1A-03 to IA-01 bus tie breaker
- IA52-37. train A
- b. Close 1A-04 to IA-02 bus tie breaker 1
1A52-55. train B F2 Energize Bus IB-01 From Bus 1A-01 And Bus 1B-02 From Bus IA-02:
- a. Close bus IA-O0 feed to IX-1I 1
1A52-02. train A
- b. Close bus IB-O normal feed 1
1B52-04B. train A
- c.
Close bus IA-02 feed to 1X-12
- 1A52-15. train B
'd. Close bus IB-02 normal feed
- IB52-05B. train B I
ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 53 of 63
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 54 of 63 ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED ATTACHMENT F (Page 2 of 3)
BACKFEED TO 480 VAC SAFEGUARDS BUSES F3 Backfeed Bus IB-03 From Bus IB-01 And Bus 1B-04 From.Bus 1B-02:
- a. Place Unit 1 service water pumps in pull-out 0
S U
P-32A. train A P-32B. train A P-32C. train B
- b. Place DC control power fuse block for each tie breaker to off
1B52-15C. train A
- 2)
In 1B-04. place DC control power fuse block for 1B-04 to 1B-02 bus tie breaker to off
- 1B52-18C, train B
- d. Locally close tie breakers:
-. 1B52-15C. IB-03 to 1B-O1 bus tie breaker "a IB52-18C. 1B-04 to IB-02 bus tie breaker NOTE The breaker trip for transformers IX-1I and 1X-12 is 180 amps each.
Refer to AOP-22 UNIT 1. EDG LOAD MANAGEMENT.
for equipment load ratings.
I I
I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 55 of 63 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT F (Page 3 of 3)
BACKFEED TO 480 VAC SAFEGUARDS BUSES F4 Return To Procedure And Step In Effect
-END-
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 56 of 63 ATTACHMENT G (Page 1 of 3)
ALIGNING EQUIPMENT TO ALTERNATE POWER SOURCE G1 IF power is NOT established to H-01.
THEN return to procedure and step in effect.
NOTE It may be necessary to raise charging pump speed to the speed at which the pump had tripped to engage vari-drive belts.
G2 IF placing charging pump 1P-2A in operation from alternate power supply. THEN perform the following:
- a. Locally ensure 1P-2A normal feeder breaker open.
1 IB52-13A
- b. Outside charging pump cubicles, align 1P-2A charging pump normal/alternate transfer switch to alternate power source:
- At IB313A.B854B.
place transfer switch in A2 ON - A4 OFF position.
- c. Place IP-2A charging pump controller in manual at minimum speed.
- IBC-428A
- d. At 1C04. place IP-2A charging pump switch to start.
- e. Adjust charging line flow controller to full open.
1 iBC-142
- f. At C-45. Alternate Shutdown Control Panel. close IP-'2A alternate feeder breaker.
- B52-54B
- g. Place "In Alternate Control" placard next to 1P-2A charging pump control switch.
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 57 of 63 ATTACHMENT G (Page 2 of 3)
ALIGNING EQUIPMENT TO ALTERNATE POWER SOURCE CAUTION If an undervoltage occurs on B-08 or B-09, local*manual restart of associated service water pump is required.
G3 IF placing a service water pump in operation from bus B-08. THEN perform the following:
- a. Select one service water pump to be powered from alternate supply.
"o P-32B "o P-32F
- b. Locally ensure normal feeder breaker for selected service water pump
- c. In Room G-01. place alternate power selector switch to desired service water pump.
e B854D
- d. In Room G-01. place normal/alternate transfer switch for selected service water pump to B-08 power supply.
o lB311C-B854E! for P-32B o 2B334B-B854D for P-32F
- e. At C-45. Alternate Shutdown Control Panel. check selected service water pump switches properly aligned to B-08 power supply using indicating lights.
- f. At C-45, close P-32B/F alternate feeder breaker.
- B52-54D
- g. In Control Room. indicate which pump is in operation from alternate shutdown panel on service water pump control switches.
POINT BEACH NUCLEAR PLANT ECA-0.O UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 58 of 63 ATTACH.MENT G (Page 3 of 3)
ALIGNING EQUIPMENT TO ALTERNATE POWER SOURCE G4 IF placing a service water pump in operation from bus B-09.
THEN perform the following:
- a. Select one service water pump to be powered from alternate supply.
"o P-32C "o P-32E
- b. Locally ensure normal feeder breaker for selected service water pump Sopen.
o IB52-20C for P-32C o 2B52-27C for P-32E
- c. In Room G-02. place alternate power selector switch to desired service water pump.
- B957D
- d. In Room G-02. place normal/alternate transfer switch for selected service water pump to B-09 power supply.
o IB420C-Bg57D for P-32C o 2B427C-B957D for P-32E
- e. At C-45. Alternate Shutdown Control Panel, check selected service water pump switches properly aligned to B-09 power supply using indicating lights.
- f. At C-45. close P-32C/E alternate feeder breaker.
a B52-57D
- g. In Control Room. indicate which pump is in operation from alternate shutdown panel on service water pump control switches.
"G5 Return to procedure and step in-effect.
-END-
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1110/2002 LOSS OF ALL AC POWER Page 59 of 63 ATTACHMENT H (Page 1 of 2)
CONTAINMENT ISOLATION VALVES PANEL A COMPONENT DESCRIPTION TRAIN ICV-1296 Auxiliary charging line A
1RC-538 Pressurizer relief tank to gas analyzer A
1WG-1788 Reactor coolant drain tank to gas analyzer A
1WL-1698 Reactor coolant drain tank to -19 ft sump A
1WL-1003A Reactor'coolant drain tank pump suction A
1WL-1003B Reactor coolant drain tank pump suction A
1RC-508 Reactor makeup water to containment A or B 1RC-539 Pressurizer relief tank to gas analyzer B
IWG-1789 Reactor coolant drain tank to gas analyzer B
1SI-846 Accumulator nitrogen supply A or B IWL-1721 Reactor coolant drain tank pumps suction B
1VNPSE-3244 Containment purge supply A
1VNPSE-3212 Containment purge exhaust A
IWL-1723 Sump A drain A
ISC-951 Pressurizer steam sample A
1SC-953 Pressurizer liquid sample A
1VNPSE-3245 Containment purge supply B
1VNPSE-3213 Containment purge exhaust B
ISC-966A Pressurizer steam sample A or B 1SC-966B Pressurizer liquid sample A or B
POINT BEACH NUCLEAR PLANT POINT BEACH NUCLEAR PLANT EMERGENCY CONTINGENCY ACTION LOSS OF ALL AC POWER ECA-0.0 UNIT 1 SAFETY RELATED Revision 30 1/10/2002 Page 60 of 63 ATTACHMENT H (Page 2 of 2)
CONTAINMENT *ISOLATION VALVES PANEL B COMPONENT DESCRIPTION TRAIN 1CC-769 Component cooling water outlet from excess A or B
.letdown heat exchanger 1CV-313 Reactor coolant pump seal return A
ICV-371 Letdown line A
IMS-5958 Steam generator blowdown A or B 1MS-5959 Steam generator blowdown A or B IWG-1786 Reactor coolant drain tank vent A
ICV-313A Reactor coolant pump seal return B
ICV-371A Letdown line B
1WG-1787 Reactor coolant drain tank vent B
IRM-3200C
-RE-211/212 supply A
IRM-3200A RE-211/212 return A or B lMS-2083 Steam generator A sample A or B IMS-2084 Steam generator B sample A or B ISC-955 Reactor coolant hot leg sample A
lIA-3047 Instrument air line A or B 1RM-3200B RE-21i/212 supply B
1SC-966C Reactor coolant hot leg sample A or B lIA-3048 Instrument air line A or B
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 61 of 63 ISTP IACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED ATTACHMENT I (Page I of 2)
LOCAL SHUTTING OF MSIV NOTE Throughout this attachment. "affected" refers to the MSIV being shut.
IF Control Room annunciator C02D 4-5. UNIT 1 SAFEGUARDS DC CONTROL POWER FAILURE, is lit.
THEN ensure MSIV solenoid power supplies are on.
"* D72-16-2. train A
"* D72-21-2.
train B 12 Shut affected MSIV using local pushbutton.
- a. IF shutting IMS-2018.
THEN press both pushbuttons in 1RK-33.
"* IMS PB-2018A. train A
"* IMS PB-2018B, train B
- b. IF shutting lMS-2017.
THEN press both pushbuttons in IRK-34.
e IMS PB-2017A. train A a IMS PB-2017B, train B 13 IF affected MSIV will NOT shut using local pushbuttons, THEN vent off air from affected MSIV as follows:
- a. Obtain two adjustable crescent wrenches.
- b. Shut instrument air supply valve to affected MSIV.
o IA-638 for lMS-2018 o IA-636 for IMS-2017
"o IMS-331A for S/G A "o IMS-331B for S/G B
o lMS-331A for S/G A o lMS-331B for SIG B
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT I EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 62 of 63 I
II ACTION/EXPECTED RESPONSE I I RESPONSE NOT TAEDI ATTACHMENT I (Page 2 of 2)
LOCAL SHUTTING OF MSIV NOTE The Direction Position Switch on the ratchet is color coded and unit designated for unit applicability.
14 IF affected MSIV will NOT shdt by venting off air, THEN perform the following:
- a. On 66' PAB near waste distillate tanks, open the M&TE box using the M&TE key and obtain the following equipment:
"* One ratcheting torque wrench
"* One 2 3/4" socket
- b. Apply wrench to the nut on the end of the valve shaft and shut affected MSIV.
15 Notify Control Room of affected MSIV status.
-END-q I
IL
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 1 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 30 1/10/2002 LOSS OF ALL AC POWER Page 63 of 63 ATTACHMENT J (Page 1 of 1)
EQUIPMENT AVAILABLE DURING LOSS OF ALL AC POWER NAWAS
- Gai-tronics
"* Outside line in.TSC a PBX telephone system
" ATCo telephone e Radio channel F
"* Operation's radios C. D. and E
- Guard radio System B
"* Interplant trouble cicuit e NRC telephone
"* Motorola radios Channel 4
- Motorola Radios Channel 6 (point-to-point)
(fire department at the scene coordinator)
LIGHTING Station Battery Fixed (EB)
DC Lights
"* Vital switchgear room
"* Diesel room (doorway only)
"a Control Room
"* Cable spreading room Fixed Emergency DC Sealed Beam Lanterns (EL)
"a These lanterns are located along all entry and egress routes to both safe and alternate shutdown equipment and total approximately 100.
"* Each individual lantern battery pack is designed for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of operation allowing for restoration of normal AC power.
Portable Emergency Lanterns
"* C59
"* TSC"
"* Auxiliary feed tunnel
"* Unit 2 non-nuclear room
"* Brigade ready
"* Fire cart. Unit 2 turbine hall. Elevation 8 ft.
-END-COMMUNICATIONS
FOLDOUT PAGE FOR ECA-0.0 UNIT 1
- 1.
POWER RESTORATION CRITERIA o IF power is restored to a 480 Vac safeguards bus prior to placing ECCS components in pull-out. THEN go to ECA-0.0 UNIT 1. LOSS OF ALL AC POWER, Step 24.
o IF power is restored to a 480 Vac safeguards bus after ECCS components have been placed in pull-out. THEN go to ECA-0.0 UNIT 1.
LOSS OF ALL AC POWER, Step 47.
- 2.
AFW SUPPLY SWITCHOVER CRITERIA IF CST level lowers to less than 8 feet. THEN switch to alternate AFW suction supply per AOP-23 UNIT 1. ESTABLISHING ALTERNATE AFW SUCTION SUPPLY.
- 3.
ADVERSE CONTAINMENT CONDITIONS IF any condition listed below occurs. THEN adverse containment setpoint values in brackets.
[I. shall be used:
"o Containment pressure - GREATER THAN 10 PSIG OR "o Containment radiation level -
GREATER THAN OR EQUAL TO 10' R/HR OR "o Integrated dose to containment - GREATER THAN 106 R 4..
AFW MINIMUM FLOW REQUIREMENTS IF any AFW pump mini-recirc valve fails shut OR annunciator C01 A 1-9.
INSTRUMENT AIR HEADER PRESSURE LOW in alarm. THEN monitor and maintain minimum AFW 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-38B minimum flow -
GREATER THAN 50 GPM "o P-29 minimum flow -
GREATER THAN 75 GPM
)'
I-
1w MOU paae 1. Cf
-ý F-
! -INITIATION Do: Nunmber ECA-0.0, Current Rz% 30)
Unit P132 Temip Cli=Se No 7-1 t, q f D-o:umn, -Tite LOSS OF ALL AC POWVER E%.,suinc Effectn - Teniporaz3 Chan.t________________________________________
Bncfdc1-s.ription ADD 10OP FOR MINIMUM AFW FLOW s!~tf.-fi.
g.as or. Formn rBF-cQ:.
R icAc' IM d -k,!C. Z1C,,ntn.matior., znd t-.cl~d,:
-n-pa~k" brutimnte P-'EF-0026h and includc v. nh the chancze Other document-s required to be cffecu' c conicurrent!: -with tne-iziporarn change__________________
Changes, pre-screene-ld according to.\\-? W 3 3 1?
NO 0 l YES ofyes. 1-st te". recsr!e Z-:
~~r-rN Screcninp con-Tpleted according ito NP 10 3.1?
LINA L7, YES Safetv Exalwitic-iRecouircd" )
NO [LI Y ES esr..
a
-a ý-:sed o- ~
ire-ci za.- Tz.-
'ý2.rc'd!-c rc7:ct-.
Dctcrnune,, if the; chance constitutes a Cicintwe Of Lnicnt to the procedure bN e%
aluaiing the. follov.ing questions (hIfrn% a zu:s YE5, a rc~ision ma% b,:
e' final c~ic%%s and 2ip~o~als shalIl be otitaired becf-e impemzrtng)
Will the proposed change:
YES NO0 1
acquirz a change to, afficc or inm ahdwct a requi rcment. c.rnmitme nt, evaluation or dcszc-iT~uon in the Current Licensing Basis (as defined in N-P 10.3 1)'
Ca-use e increase in mag-nitudc. sigruificanecc or impact such that it should be processed as a TeX ision?
-. Deletc or modify, a prerequisite. iniua! condition, precaution, limitation or othcr stcps that could hax c saifety significance or affect dhe procedure's margin of safety?
4 Delete QC hold points, indepe-ndent Verification or, Concurrent Check steps w~ithout theEl D
related step(s) that requirz the performance also bcing deleted?
- 5. Change Tech Spec or other regulatory acceptanice criteria other than for re--base! inring0 purposes"'
6 Require a chiange to the procedure Purpa.e or change the proce ciictv~?L jInitiated By prnn'sten1K 6,2,>s~
Date~~2o~
11-INIT~IAL APPROVAL This cha-n ge is correct and complete, can be performed as %%ritten, and does not adversely affect personnel or nuclear safety, or Plant operating conditions Group Supe-r% isor (prinidsign)
- r2 r f-' o r1 AJiý
-c~-i-/
(Cannot bie the Initiator)
This chanmce does not zidverselh affect P!
operating conditions Srt d * -d-resoly)
Senior Reactor Oprartor (print~si) fV p
Date/
_____l (Cannot be t'be Initiator or Gro~up Super% ibor)
III - PROCEdURE. OWNER REVIEW v
Pcrrnancnt 0] One-time Use
[I E~xpiration Date, Event or Condition.
H)old chanige unitil procedure completed (finial revic'.N and appro~al still requiicd within 14 days of initial zppro~ al)
~ QJNIS c,,e%% NOT Required (,.!n.irN>$SR onl>)
EY QR Re% ie-wReqtircd E] \\4SS Rev:cv. Required 1R~-e:~
-~55 Procedure O;.ýnzr (priiit'sign) grg !,
i.
Dt III % 0-11`11-C :"d %Linn~rtt'1Z~r ML-Ci c-tis..'..i z
Wrn'ezl%.--J.s d
WV-FIN'AL REV'IENN AND APPROVAL
- -(Mu-st bec cumpttcd t.itbin 14 divs ofiessijii
.pprul.
(Th licn~ualor, Qfl.and ApprO%2i AuthontA shalt be indcptendent from esdi o:i'cr)
QR/MS (r:,t-.z) i-'Ue.~,'
~Date
~ s~in
.r rL'C~ ri~Lira4 nJ ec..jircd p.ile.!'-ed NISS Mceuwi No Appro% al AUL'ICertv (rrvit, bit).
~
~~
Date 4L ~
L N' - RK NI'-i 0N I NFOR 10 1TI10x r o P P ER %LNNE
'c i! N(;Fs Post T--rv:. ugR~e~r~ec L Z-
~
A--
WIC hicorporate d i:n.o Re-, isin Numnber Efi~ccti'c D.it-11
- I.
1`11 -U:.
R,%
isi., ri 2 1 L
R7,i~ Ni P TEMlPORARY CHANGE R'E"ENAND APPRO'VAL X.~c 'r:;
'.2 S Tcmp~rv-..
~.
- ~'C'z..:z...
I; c
- .. !:ei Vs
'I I
=
-v DoccUMEr-N r I.,LVIE\\N AND APPROVAL CONTiNULATION.\\
P-32c
- 2-C Do~c Nuii*b~r ECA-0.0 Rcvisizn 30 Unit 2_
T:,
LOSS OF ALL AC POWVER Tc:prnChange ircr7;'2o)-c:5-I2 P::,criiition of Changes:
Step
- rOLD OUT PAGE Ch angelReason CHýANGE: ADDED FOP CRITERIA FOR THE AEWV MIUNIMUM'.
FLOW REQUIREM"vENTS REASON': TO ENSURE MINL\\NIM FLOW IS MAIN'TAINED THROUGH Tf-E. AFW PUMýPS DURING OPERATION Other (orinients 1%
c~n M e is not,:
ar.~,r.1 i-ci.ýs*..rt:a r
- t
-U(,26.
!lclcrLnces NP 1 13. NP 1 :3
4 Po:n Beazh Nu :leh.. P.:?,*
TE.MPORARY CHANGE AFFECTED MANU.ALL LOCATION Pge 2
_ of Pr3:xc.re Number ECA-0.0 P.e isLon 30 Urut PB2 "Tiije LOSS OF ALL AC POWER TcmpornL-Chalc' Numnoer
,9 1 - Ci4 I - IMMEDIATELY AFTER INITIAL APPROVAL ON PBF-0026e (Nen-L.-.:a changes)
(a,9 cr Final Approvl if change of intent.i n\\oi Date SDate This procedure change has been processed as follows: (ManuaUL'ocation)
Performed Li Copy included in iwork package for ficld implcmentation. (WO No.
)
Copy filed in Control Room temp change binder (Operations only)
Original chang',c package provided to l-e3-t.u to obain Procedure ONN-.er Review (e g. o-,%.er :rcic\\, may be coordmited by In-Group OA ii, ProCe-.-re Wrn:er. Procedure Supernisor. tdc)
['
[I_
LIL__
LI____
El___
El Performcd By (print and sign)
C o
/.-,.,e./
Date /7..2c.,."
II - PROCEDURE OWNER REVIEW ON PBF-0026e (may be pcrformed by OA 11, Prcce-Aure Writer, etc.)
This procedure change has been processed as follows: (Manual/Location)
,7 Copy sent to Document Control Distribution Lead for Master File
(\\oa rcqaired far one-time use change)
El Copy filed in Group satellite file. (Not required fcr one-time use charges)
El Cop)y filed in Group one-time use file.
Original Temp Change provided to "i/")
c to obtain Final Approvals (cg, fc,. al ap;-.\\aI niaa be coordmnatc-h b In-Group OA 11, l'roccdtrc Wn'tr. Pre--cduc Surc.risor.,C t)
.m t{'-.*
E1g bAs S___________________
El 0't S
Ix____________________
Performed Bv (print and sign) o,,.
,-c, Date
,2-2r..
FBF-o026h1 Fc."is,,.n.5 06j13 01 Refcrence ',? 1 2 3
-arf-,-Y
.1 I
POINT BEACH NUCLEAR PLANT ECA-0.0 UNIT 2 EMERGENCY CONTINGENCY ACTION SAFETY RELATED Revision 31 1/10/2002 LOSS OF ALL AC POWER Page 1 of 63 A. PURPOSE
- 1. This procedure provides directions to respond to a loss of all 480 Vac power for plant conditions greater than or equal to 350°F and RHR not in service.
- 2.
This procedure is applicable for all plant conditions where RCS hot leg temperature is greater than or equal to 350'F with accumulators in service, and assumes the RHR system is not in service for decay heat removal and all SI system components are aligned for normal power operation.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. The symptom of a loss of all AC power is the indication that both 4160 Vac safeguards buses or both 480 Vac safeguards buses are deenergized.
- 2.
This procedure is entered directly on indication that safeguards buses are deenergized or from the following procedures on the indication that safeguards buses are deenergized:
"o EOP-0 UNIT 2. REACTOR TRIP OR SAFETY INJECTION. Step 3 "o EOP-0.0 UNIT 2.
REDIAGNOSIS. Step 2 o CSP-S.l UNIT 2. RESPONSE TO NUCLEAR POWER GENERATION/ATWS.
Step Al o CSP-H.1 UNIT 2. RESPONSE TO LOSS OF SECONDARY HEAT SINK. Step C1 C. REFERENCES
- 1. Technical Specifications for Point Beach Nuclear Plant
- 2.
Final Safety Analysis Report for Point Beach Nuclear Plant 3..
As-built plant drawings
- 4.
Generic Technical Guidelines developed by the Westinghouse Owners Group (WOG).
This consists of the following documents:
- a. Low pressure version of the WOG Optimal Recovery Guidelines. Status Trees. and Functional Restoration Guidelines
- b. Background documents for each low pressure version Optimal Recovery Guideline. Status Tree. and Functional Restoration Guideline
- c. WOG Emergency Response Guideline Executive Volume
- d.
WOG Emergency Response Guideline. Maintenance Program Summary
FOLDOUT PAGE FOR ECA-O.O UNIT 2
- 1.
POWER RESTORATION CRITERIA o IF power is restored to a 480 Vac safeguards bus prior to placing ECCS components in pull-out. THEN go to ECA-O.O UNIT 2. LOSS OF ALL AC POWER.
Step 24.
o IF power is restored to a 480 Vac safeguards bus after ECCS components have been placed in pull-out. THEN go to ECA-0.0 UNIT 2.
LOSS OF ALL AC POWER, Step 47.
- 2.
AFW SUPPLY SWITCHOVER CRITERIA IF CST level lowers to less than 8 feet. THEN switch to alternate AFW suction supply per AOP-23 UNIT 2, ESTABLISHING ALTERNATE AFW SUCTION SUPPLY.
- 3.
ADVERSE CONTAINMENT CONDITIONS I
any condition listed below occurs. THEN adverse containment setpoint values in brackets. [].
shall be used:
o Containment pressure - GREATER THAN 10 PSIG OR o Containment radiation level -
GREATER THAN OR EQUAL TO 10' R/HR OR o'Integrated dose to containment - GREATER THAN 106 R
- 4.
AFW MINIMUM FLOW REQUIREMENTS IF any AFW pump mini-recirc valve fails shut OR annunciator C01 A 1-9.
INSTRUMENT AIR HEADER PRESSURE LOW in alarm. THEN monitor and maintain minimum AFW 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-38B minimum flow -
GREATER THAN 50 GPM o P-29 minimum flow -
GREATER THAN 75 GPM I
Nuclear Management Company S,-.
CHANGE HISTORY Initiate by RICHARD FLESSNER Complete and Close by MARYBETH ARNOLD Assign Work 2/5/2002 22958 PM Owner RICK WOOD Done 5'7/2002 10 42 45 AM Owner (None)
Assign by RICK WOOD Conduct Work 2/6/2002 2 52 44 PM Owner RICK WOOD -
Work Complete by RICK WOOD Review &
Approval 5/6,'2002 2 37.07 PM Owner RICK WOOD Approved by RICK WOOD Quality Check 5/6/2002 2 37.59 PM Owner PBNP CAP Admin SECTION 1 Activity Request Id:
Activity Type:
Site/Unit:
Activity Requested:
CA003691 Corrective Action Point Beach Common Submit Date:
2/5/2002 2.29.58 PM Assist Operations in determining what initiating events should be included in the EOP validation process by formally providing information on which initiating events considered risk significant for each EOP.
0 CATPR:
Initiator Department:
Y EX Engineering Processes PB Responsible Department: Engineering Initiator:
Responsible Group Code:
Activity Supervisor:
FLESSNER, RICHARD EPP Engineering Progtams PRA PB E RICK WOOD 6)
Activity Performer:
RICK WOOD SECTION 2 Priority:
2 Mode Change Restraint:
(None) 0 QA/Nuclear Oversight?:
N NRC Commitment?:
N Due Date:
Management Exception From PI?:
" Licensing Review?:
" NRC Commitment Date:
5/6/2002 N
N SECTION 3 Activity Completed:
1/18/2002 12.52PM - LARRY PETERSON Due date extended as requested and approved by F. Cayia in prior update. Retruned to R.
flessner for completion 1/18/2002 12:54PM - LARRY PETERSON Reassigned to R. Flessner for completion following extension.
2/6/2002 3 07PM - RICK WOOD:
The risk informed validation of EOPs considers initiating events with a frequency greater than 5e-3/yr. This includes a Steam Generator tube rupture, Reactor trip without Power Conversion System, Loss of Offsite Power, Loss of CCW and a Steamline Break. I will create a table of those EOPs that should be validated against the above initiating events and forward that to Operations procedures for their review.
https://nmc.ttrackonli ne.comltmtrackltmtrack.dll ?IssuePage&TableId= 1000&RecordId= 11 :... 9/18/2002 Page I of 3
Nuclear Management Company 2/25/2002 4.52PM - RICK WOOD:
I sent the draft matrix and procedure feedback form to Terry Vandenbosch for his comment When his comments are received, I will incorporate them into the draft and finalize my procedure feedback to OM 4.3.3 EOP Validation 5/6/2002 2:37:07 PM - RICK WOOD:
The matrix has been incorpQrated into OM 4.3 3, EOP validation on May 3, 2002 as a temporary change to the procedure. The matrix will also be added to the writers' guide and verification procedures in the near future.
517/2002 10.42:45 AM - MARYBETH ARNOLD:
A permanent temporary change notice is posted on the index on 04/26/02 to OM 4 3.3. TCN 2002-0293 is noted as author Jim Green. CLOSED SECTION 4 QA Supervisor:
SECTION 5 0 Project:
(None)
Licensing Supervisor:
CAP Activities &
Actions :
0 State:
0 Owner:
0 Submitter:
" Last Modified Date:
" Last State Change Date:
0 Close Date:
0 One Line
Description:
NUTRK ID:
Child Number:
References:
Update:
Import Memo Field:
CAP Admin:
Done (None)
RICHARD FLESSNER 5/7/2002 10:42:45 AM 517/2002 10:42:45 AM
" Activelinactive:
AR Type:
Assigned Date:
" Last Modifier:
0 Last State Changer:
Inactive Daughter 2/6/2002 MARYBETH ARNOLD MARYBETH ARNOLD 517/2002 10.42:45 AM Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW CR 01-3595 1
CR 01-2278 RCE 01-069 GOOD CATCH OM 4.3 3 TCN 2002-0293 PBNP CAP Admin Site:
Point Beach OLDACTIONNUM:
Cartridge and Frame:
ATTACHMENTS AND PARENT/CHILD LINKS https://nmc.ttrackonline.com/tmtrackltmtrack.dll?IssuePage&Tableld= 1000&Recordld= 11 '... 9/1812002 (None)
Page 2 of 3
Page 3 of 3 Nuclear Management Company ACE000314" Probabilistic Risk Assessment PRA For Auxiliary Feedwater System AFW CAP001415_ Probabilistic Risk Assessment PRA For Auxiliary FeedwateLSystem_AFW https:H/nmc.ttrackonline.conmtmtrackltmtrack.di]?IssuePage&Tableld=1000&Recordld=
I1'... 9/18/2002 T
~
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o M
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I E1.1POR\\RX Cl-A.N.C-.E RZN&'V E A ND A PPROVA.L Ny e-Rek'ýr to NP 1 2S3. TeporanP.-CLV f 'a Chýz:nges,:frreqrdiremiet:.s P
of:2
=:tn feitTemporar)
Brief 'Des-ription Add PRA Core Dimaut! Ri~k Matrix 1 Initiate PBF-0026h rind inclu.de..%.,
the chan:e.
Othr docu-ments recuiiedto be effective concurr~ritly wýith t'ieterniporary chan-ze.
none Chan:ncs pre-screei'cd according,o NP 5.1.S?
07O 1 YES t~rdeco~u-ýurnn:
3C.OfUllgo M'P S I S)
Screening! :ompleted according :0o N? 5.J.8!
I NA 0l ITS (AcU.:' py)
~~Safety Evaluatio.,Re.,uircd?
0s NO D1 YES ';tl.
~
t~l~or.U~
Determine if:he chjnee constitunes a Char.ga or intent !o the procedur-: by evahzdtnng the follov'ing que.stions.
(if ar) n cr~s wC o: YaES. -- re*i,ion mady re prc:sse'd.re fi~r1 ve-ews an~d approvals i1.al be obrai red bce'oc imr'emcn ing)
Will the proposed change:
YES INO0 1I. Require a chine: to. affect or in% alicate a requirement, commitment, e-a!ua-ticn or0 description in the Cu.-rent or ISFSI Licensing Basis (as defined in NP 5 1.8 ajrd N-P 5.1t '7 ?
- 2. Cause an increase in maviitude. sig:nifilmancc or impact such that it Should be processed as a re~ ision?
- 3. Delete or moýdify a prerequisite. initial condition. precaution, limitation or other steps that Icovld have ; ift-ty significance or affect the procedure's margin of safety?
4 Delete QC hold points. Independent Verification or Concurrent Check iteps U~ithout the related step(.,) -hat require the performance also being deleted-?
- 5. Change Tech Spec or other reglulatory acceptance criteria other than for re-baselining purposes?
6 Require a change :o the procedure Purpose or chanige the Pre'zx' t classificatioii Initiated By tpnnttsign~)
James G Gre~en
~
-NDate 04126/12002 11I-INITIAL APPROVAL This chainge is correct and complete. can be per rrned
- ite, and does nioj txslya personnel or nuclear izifety. or Plai aprtn cniir Group Supervisor lpnntis/
Dee 1nOL (Cennot bc te In:ti2
-116 chzinse does not adversely.%'ffct PIe*,verating conditions. IS c&t'R-Ie~d pr~' due.or:ii Senior Reactor Operator (pnntlsign)
Date
~-
~
/IL (Citnnot be the Isrutuor or G;roup Super, sar)
IIIl PROCEDURE OW'VNER REVIEW SPermanent One-time Use J
Expiration Date. Event or Co.ndition-___
Hold change until procedure completed (final revtew and approval still requi re %' rEir 4ýays or initi' approval Req5ire I
drRin/NNSR criv)
Q.-f4.SS Re'view NO r' Reuie (AmmS cr1y Requr&-vw
_c;M4, I JS Procedure Ou' ner i prit/i)
IAAI*
t7e T',,; Ch.n~ew od ~uvnn!
at IV.- FINAL REVIEW AND 'APPRQVALi i(tust be omoleted within 14dayi of injilti joaril iThe Imnriaor. OR.ind AooruN21 Aut'roftv shill be indc~endrnt from eath oih..ri QR/SS'~a/Iri I
__________D..tc
- n_1caics ;o sqn743 3pparcartLr:o gsedsc1j ancccsýe::
,er c~n r? ec-a~atu31on%
penormnd. dele.runat-i.onzl 2
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AMr~r'a,:! A -thoritv i rrnurszi,
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.<JI'Dat 4
o-01 V-REVISION IN;Fi)Rl
\\TIO)N FOR PER\\!AkEN" 'A.CIts/
P0%tTping Review eprnnu/sigrp/,$-gAi
~
D.te A~t :I_
Incor-oritt-d itReson,\\umber El 'ecti,. e Di*_______
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CVIRC OF I L M ErDRCne Reiswn 13 0i16'&Ca2MY3 C
RECU'D 1AY.i~ 6 c-L MA uN' 2
.1 j
lo-.r..t BeaLh Nuclear Plan:
DOCC.MENT REVIEW AND APP4"OVAL COSTINL'ATI(Y' p'.
-t T:ELOP V'ALID.\\TON Tc-np'r:.rv-Criange Number 920 3 Description of Chan~ges:
Step ChangeiReasan Ade10linn h seo h e PRA CORE DAMAGE RISKMk1 MAR 1Xlhz r
A:n~chment A Added A~achm.rtA. PRA CORE DAMAGE RISK IMATRIX. I'rhis is pr.sc, o Cnt.-Tra 42 Adm~nisL-avic PrcN:-dure._____
_________~A~l'O
,)-1
~I ~?
IS T4 A~-V7
~
Azi...5-4'I c1'C LJ't-______
__________ell, OteCmets
_i Rcr inf
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__irdc niui cu~ce ficuc, o-,,L r-vhtnn,'
?B_________
006
__ýon f-131~
I
?Caint BLaci N~uc~zar Planr TE-MPORARY CHANGE AFFECTED MANUAL LOCATION Page I
of Procedure Numbcr OM..3.3 Re'ision 0
Unit PB Title FOP VALIDAfT!O Temporary Change Number 2')020293 I. MIEDIATELY AFTER INITL-AL APPROVAL ON PBF.00)26e (.on-Intent changes) l Ate re l.pp roval if change of i-itnt in.olh. d)
This procedure change has been processed as follo*,s: (.ManuaViLocation) 1 Dt
[] Copy included in work, package for field implementation. (WO No.)
Peord
[
Copy filed in Control Room temp change binder (Operations only).
Original change package provided to to obtain Procedure Owner 0
Review (e g. O,,ncrr-vir rr:.-
be coordinate; by tn-Gro.:, O*, II. Prredure Writr. Proc-dur= Supefvlso:,.
)-c)
El This procedure cha ngehsbe rcse sflos Mna/oai 7i Originalrep changhabeen proceid ed asto ll ows (Mn/obatain ialAprvls
[]
1 Performed By serint and sign) Control i_.iut_
Lead fo Ma..seDatr F*e'.
(Not b pefforred y OAIs, Proedure ee---
[]
Copy filed in Group..,.tel!:te file. (Not required for one-tirne use eh..mve, )
[]
_diCopy filed in Grout) one-time use file.
SOriginal Temp, Change provided to G
3to obtain Fin~a. Approvals o.
(C. g, final
_)poa z
c-cr.=:e rated oy L,.Grouo OA IL m;.!*.:eurt Wrte.-P:'e. edums:e
,--".t r
J
?en-ormned By kpnint xtnd sign)
D, a tt" r--F bct-W 26In
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'*.. " P I 2 ic is on
Point Beach Nuclear Plant 10 CFR 50.59/72.48 APPLICABILITY FORM P.Iwe Brief Activity Title Revision to OMI 4.3.3, EOP VALIDATION or
Description:
This form is required to be completed and attached to the applicable activity change forms to document all or portions; of an activity that are covered by another regulation other than 10 CFR 50.59 and 10 CFR 72.48 (pre-screening criteria 2). See NP 5.1.8, 10 CFR 50.59/72.48 ApplIcabilty, Screening and Evaluation (New Rule).
NOTE: Guidance for searching the FSAR, Technical Specifications, Regulatory. Commitments (CLB Commitment Database) and other licensing basis documents can be found in NP 5.1.8.
Attachment G.
NOTE: Although 10 CFR 50.59 and 72.48 may not be applicable to the processes listed below, change activities conducted under these piocesses may require changes to the FSAR. If so, initiate FSAR changes per NP 5.2.6, FSAR Revisions.
Regulatory or Plant Process YES NO Does the activity require a change to the Facility Operating License, License Conditions or Technical Specifications? (If the answer is YF.S, process the applicable changes per
[]
[
NP 5.2.7, License Amendment Request Preparation, Review and Approval.)
- 2.
NOTE: The Quality Assurance Plan is described in FSAR Section 1.4.
Does the activity require a change to the Quality Assurance Program? If the arnswer is 0
YES, process the applicable changes per NP 11.1.3, QA Program Revisions.
- 3.
NOTE: Implementation of Security Plan changes that require physical changes to the plant, or changes to operator access to the plant require a screening.
NOTE: Security is described in FSAR Section 12.7.
Does the activity require a change to the PBNP Security Plan, a safeguards contingency plan, or security training and qualification plan? If the answer is YES, assess the acceptability of the change per 10 CFR 50.54(p) using Security procedures.
- 4.
NOTE: The Emergency Plan is described in FSAR Section 12.6.
Does the activity require a change to the Emergency Plan? If the answer is YES, process the applicable changes per NP 1.8.1, Emergency Preparedness Procedurcs.
- 5.
NOTE: The Radiation Protection Program is described in FSAR Section 11.4.
Does the activity require a change to the PBNP Radiation Protection Program described in NP 4.2.9, Radiation Protection, OR is the activity within the scope of NP 4.2.9 and 10 CFR 20, Standards for Prote,-don Against Radiation?
- 6.
NOTE: Changes to the plant or method of evaluation that result in re-analysis or the FSAR loss-of-coolant accident (LOCA) analysis require a screening.
Does the activity require a change to the FSAR LOCA analysis results subject to 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors? If the answer is YES, process the applicable changes per NP 5.2.12, 10 CFR 50.46 Reporting Requirements, and NP 5.2.6 FSAR Revisionw.
- 7.
NOTE:
Regulatory commitments are found in the CLB Commitment Database.
Does the activity involke a change to, Regulatory Comnitmeit
? h zh,: answer is YES.
[?I [
process the applicable changes per NP 5.1.7. Regulator), Commitment Changes.
- 8.
Does the activity invoke a change to the Environmental Manual (EM), Radiolocical Effluent Control Program Manual (RECM). Offsite Dose Calcuiation Manual (ODC.M).
or Process Control Program (PCP), ANT does NOT insolve chmngp, in use of exp!us'se fl
[
gases in ;,vaste treatment systems? If the 'nswe; is YES, document the applicable changes per the requirements of TS 15.7.8.7.B (ITS 5.5.1 1.
1I I'I'. I 15a 1,;'1%1n 0 10/24/01 Rtkfl.,
NMV 5 1 N ftpdr"*-Mba"^ý P,12C 1
Point Bcach Nuclcar Plant 10 CFR 50.59/72.48 APPLICABILITY FORM
-Pa*cz 2 Regulatory or Plant Process YES NO NOTE: For purposes of determining 10 CFR 50.59 / 72.48 applicability, the determination of an administrative procedure below takes precedence over definitions or classifications in other plant procedures or guidelines.
- 9.
Does the activity require a change to an administrative procedure or controlled document ONLY?
F-1 ALL of the following statements shall be true for the procedure or controlled document to be considered administrative.
- a.
DOES NOT dihect how plant structures, systems, or components are operated, maintained, tested or repaired either specifically OR generically.
- b. DOES NOT specify acceptance criteria or operating limits for plant structures, systems, or components.
- c.
DOES NOT specify parts, materials, chemicals, lub'icants, etc. to be used in plant structures, systems, or components.
- d. ROES NOT specify compensatory action(s) to address plant structures, systems, or components ot of service, or to address non-conforming condition.s.
- e.
DOP.S NOT affect operator access to operating areas of the plant.
10 CFR 50.59172.48 APPLICABILITY CONCLUSION NOTE: If ANY portion of the activity is NOT. controlled by one or more of the processes above, further 10 CFR 50.59 /72.48 review is required (i.e., portions not covcred by the above processes shall be prescreened to other criteria or screened).
ALL aspects of the activity,re controlled by one or more of the processes above, therefore NO YES NO additional 10 CFR 50.59 and 72.48 review is required.
If the above question is answered NO, briefly describe the portions of the activity NOT covered by one or more of the above processes:
Performed By James G Green I
Datc 04/26&'2002 Name (Print) w"it'nature ReiieS 6 Wy).
,~i~jj i
JJI Date '-.L.
I-;
Namc (Print)
Signature RPIas-or.1 I RL6l.int NP6 I W
. PO!.NT BEACH NUCLEAR PLANT OM 4 3.3 OPERATIONS MANUAL NNSR RcviJion 0 EOP VALIDATION July 30. 1993 1.0 PURPOSE The objective of EOP validation is to guide a systematic evaluation of procedure changcs to determine their useability and operational correctness. Useabili'., is the characteristic of providing sufficient, understandable information to operators. Operational correctness is defined as compatible with plant responses, plant equipment, equipmen, identification, available manpower and human factors criteria.
The emphasis of validation is on performance. Techniques to be used are in-plant walkthroughs, special testing, computer simulation, and performance.
o The matrix in Attachment A was developed based on initiating events with a frequency of core damage greater than 1E-6 and an initiating event frequency of greater that IE-3. The selected scenarios were then compared to the procedures that the operator would most likely use to prevent core damage. It is expected that procedure validation would consider those scenarios where an X is marked. This matrix is risk based only and should not be used as the sole consideration for determining scenarios for procedural validation.
20 PERSONNEL RESPONSIBILITIES 2.1 Verification Team - Determine the validation type required and perform the verification process on procedure changes which are made due to validation activities.
"2.2 EOP Writer - Shall be a participant in all validation; performed.
2.3 Validation Assessment Team - Shall consist of at least a DSS and the EOP writer.
Addi.ional members may be added depending on the magnitude of the validation activities and the technical expertise needed. A DSS shalB act as the leader of the validation team.
2.4 Operations Manager - Shall have overall responsibility for the EOP validation process.
He shall designate the personnel wh-io will comprise the validation team.
2.5 Procedure Reviewer - Shall evicw and approve the validation activities planned by the validation team te ensure the)y adequately assess the procedure change. Hc shall te' iev and approve the resoluticn of any discrepancies found by the validation team befoic procedure changes arc made.
Page 1 of 12
POINT BEACH NUCLEAR PLANT 01 4.3.3 OPERATIONS MANUAL NNSR Revision 0 EOP VALIDATION Julv 30, 1993 3.0 VALIDATION PROCESS 3.1 Scopihig Phase The verification team shall meet after the preparation. assessment, and resolution phases of verification are complete (i.e., all discrepancies resolved and incorporated). They will determine, based upon the final draft of the procedure, the validation activities that need to be performed before the procedure is approved for use. The required -ctivities shall be documented on EOP Verification Form I. Validation can include, but is not iirited to the following:
3.1.1 Physical walkthrough of the plant - to check for accessibility, labeling, timing, etc.
3.1.2 Actual performance using a special test - to check feasibility when deemed necessary for unusual applications.
3.1.3 Technical Review - to check calculations, interfacing with other procedures.
3.1.4 Simulator performance - to check usability of change as a %%
hole including computer modeling.
3.1.5 None necessary - for those cases where it is evaluated that the procedure could not be enhanced by a validation. In this case the validation process ends with this entry.
3.2 Preparation Phase 3.2.1 Personnel Designation. The Operations manager shall designate the personnel to comprise the validation team. A DSS shall be designated as leader. The magnitude and technical requirements shall be considered 0i determining the number of people and level of technical expertise needed to make up an acceptable validation team.
3.2.2 Organization. The EOP WMiter is a general observer and will not be gi'ben specific responsibilities. The leader is responsible for the organization, activity assignments, and preparation by ramn members. The leader is respons;ble for logistic,, such as arranging.,imulator time via training coordi;:ating personnel schedules and nce,ing completion nilestones. In addition, the leader is responsible for the co,:cnt and performance of validation activities, assesments, and conclu Iqion/recoinrendatno.,.
3.2.3 Source Documents and Validation Form,. Ceitain source documents neced to be reviewed, such as the pioposed EOP chanue Other,odurce documents, r,:y be useful, such as the generic (WOG) EOP documents. PB3NP FSAR. INPO validation procedure and plant specific document,,/procedurcs such a.% P&eID!)s.
logic drawings and normal operating proceduc'e,.
Page 2 of 12 MUM ISOMER
FOiN'T BEACH NUI OPERATIONS MA, EOP VALIDATiON
ifiufljU'
CLEAR PL.ANT QIUAL OM 4.3.3 NNSR Revision 0 Juiv 30. 1993
- a. The preparation sections of EOP Validation Foi m 1, must be completed.
- b. A scenario must be developed and scenario form 2. completed. The instructions on this form should,dentify the purpose of each assessment, describe the activities which make up the assessment, --n expected/
anticipated result, and acceptance criteria. Table 1, Evaluation Critera, should be included.
Page 3 of 12
4.0 ASSESSMENT
PHASE This is the actual conduct of the validation exercise. The following activities should be performed by the validation team.
4.1 Brief participating personnel on the scope of validation and how the assessment will be conducted.
4.2 Follow the developed scenario by first giving the initial plant conditions and then give the changing plant parameters via walk (talk) through. Some team members may act as "players" while other members function only as independent observers who evaluate tile actions being performed.
4.3 Review/observe the personnel performing the procedure by using the evaluation criteria applicable to the validation method chosen.
4.4 Stop the assessment scenario for discussion of any important perceived discrepancies.
Good must be used. The gains of any discussion must be considered when interrupting a scenario. For example, if timing of activities is being done, the scenario should not be stopped.
4.5 Conduct a V:efing with the personnel as soon as possible after each assessment using the following sequence:
4.5.1 Brief the participants on the purpose and objectives for debriefing.
4.5.2 Have peisonnel present problems and discrepancies %%
hich they identified during assessment.
4.5.3 Have personnel provide possible reasons for problems.
4.5.4 Suminatize the findings of the debriefing.
4.6 Complete the EOP Validation form.
4.7 Complete the Scenario form.
L
POINT BEACH NUCLEAR PLANT OiM 4.3.3 OPERATIONS MANUAL NNSR Rex ision 0 EOP VALIDATION July 30. 1993 4.8 Indicate on a discrepancy sheet (EOP Verification Procedure Form 3) each discrepancy observed during the assessment phase.
5.0 RESOLUTION PHASE The resolution phase, consists of the following steps. During this phase F*,rm 3 is completed.
5.1 The is to determine a solution for each discrepancy and indicate this as the resolution on the discrepancy sheet.
5.2 The will then review and comment on the proposed resolutions.
5.3 After review and comment by the. the discrepancy sheet with the EOP and any applicable source documents ave sent to the for approval of the resolution.
5.4 If the resolution is not approved, the is to determine a revised solution. A new discrepancy sheet is to be initiated and the resolution process :-estarted.
5.5 In the event the and cannot agree on a resolution :o a discrepancy, the shall determine the resolution to be incorporated in t procedure changeý.
5.6 A resolution acceptable to the, and shall be converted to a draft procedure change by the The draft change shall be given to the for verification.
5.7 Following approval, the change shall be prepared for MSS review and approval.
5.8 Documentation will exist for the following:
5.8.1 Scope of the validation 5.8.2 Validation method used 5.8.3 Participants 5.8.4 Scenario description 5.8.5 Evaluation criteria 5.8.6 Observer worksheet 5 8.7 Identifed discrepancies P:we 4 of 12 4C, 7 7----':ý
POUNT BEACH NUCLEAR PLANT OPERATIONS MANUAL EOP VALIDATION OM 4.3.3 NNSR Revision 0 July 30. 1993 5.8.8 Discrepancy resolutions 5.8.9 Re icw and approval of resolutions All validation fomi shall be maintained per PBNP Administrative Procedures.
Page 5 of 12
POINT BEACH NUCLEAR PLANT OPERAI IONS MANUAL EOP VALIDATION OM 4.3.3 NNSR Revision 0 Ju1 I; 30. 1993 FORM I EOP VALIDATION FORM EOP TITLE:
EOP NUMBER:
SCOPE OF VALIDATION:
REVISION:
VALIDATION METHOD (01R METHODS TO BE USED:_____________
Designated Obscrvers/Reviewers Preparation Completed By Assessment Completed By Operators Involved:
Date:
Date:
Qualification: (SRO, RO. Other)
Resolution Completed By Documentation Package Forwarded By Date:
Pa2e 6 of 12 CIL
POINT BEACH NUCLEAR PLANT OPERATIONS MANUAL EOP VALIDATION OM 4.3.3 NNSR Revision 0 July 30, 1993 FORM 2 SCENARIO FORM SCENARIO NO.:
DATE:
TITLE:
SCOPE:
SCENARIO DESCRiPTION AND SPECIAL INSTRUCTIONS:
INITIAL PI NNT CONDITIONS:
PROCEDURE NO., STEP NO.. & DESCRIPTION PLANT PARAMETER/
SYMPTOM TRANSITION TO LPROCEDURE NO.)
Pagc 7 of 12
"Y S
OM 4.3.3 NNSR Revision 0 July 30. 1993 FORM 3 SCENARIO EOP:
REV.:
PAGE NUMBER:
STEP NUMBER:
DISCREPANCY:
RESOLUTION:
VERIFICATION TEAM REVIEW:
LOP WRITER REVIEW AND COMMENTS:
BY:
RESOLUTION APPROVED: YES NO (circle one)
PROCEDURE REVIEWER:
RESOLUTION LNCORPORATED BY:
DATE:
DATE:
DATE:
DATE:
Page 8 of 12 p I POINT BEACH NUCLEAR PLANT OF"RATIONS MANUAL EOP VALIDATION
'mm
A POINT BEACH NUCLEAR PLANT OM 4.3.3 OPERATIONS MANUAL NNSR Revisir, 0 EOP VALIDATION July 30, 1993 USABILITY A.
LEVEL OF DETAIL
- 1.
Is there sufficient infor-mation to perform the specified actions at each step?
- 2.
Arc the alternatives adequately described at each decision point?
- 3.
Are the labeling, abbreviations, and location infornation as provided in the EOP sufficient to enable the operator to find the needed equipment?
- 4.
Is the EOP mising information needed to manage the emergency condition?
- 5.
Are the contingency actions sufficient to address the symptoms?
- 6.
Are the titles and numbers sufficiently descriptive to enable the operator to find referenced and branched procedures?
- 7.
Are the action steps structured to minimize personnel movements around the control room.
- 8.
Are action steps structured to avoid unintentional duplication of tasks?
B.
UNDERSTANDABILITY I.
Is the EOP easy to read?
- 2.
Are the figures and tables easily and accurately read?
- 3.
Is interpolation of values on figures and charts difficult?
- 4.
Are caution and note -,tatements readily understandable?
- 5.
Are the EOP steps readily understandab!e?
- 6.
Are th, em, :iasized items noti:ed?
Page 9 of 12 pill
PO','NT BEACH NUCLEAR PLANT OM 4.3.3 OPERATIONS MANUAL NNSR Revision 0 EOP VALIDATION July30,. 993 II.
OPERATIONAL CORRECTNESS A.
PLANT COMPATIBILITY
- 1.
Can the actions specified in the procedure be performed in the designated sequence?
- 2.
Are there alternative success paths that are not included in the EOP9
- 3.
Can the information from the plant instrumentation be obtained, as specified by the EOP?
- 4.
Are the plant symptoms specified by the EOP adequate to enable the operator to select the applicable EOP?
- 5.
Is information or equipment not specified in the EOP required to accomplish the task?
- 6.
Are the instrument readings and tolerances stated in the EOP consistent with the instrument values displayed on the instruments?
- 7.
Is the EOP physically compatible with the work situation (too bulky to hold, binding would not allow them to lay flat in work space, no place to lay the EOPs down to use)?
- 8.
Are the instrument readings and tolerances specified by the EOP for remotely located instruments accurate?
B.
OPERATOR COMPATI3ILITY
- 1.
If time intervals are specified, can the procedure action steps be performed on the plant within or at the designated time intervals?
- 2.
Can the procedure action steps be performed by the operating shift considering, for example, manning, communication capabilities, conflicting tasks, etc.?
- 3.
If specific actions are assigned to individual shift personnel, does the EOP adequately aid in the coordination of actions among shift personnel where necessary?
- 4.
Can the operating shift follow the designated action step sequences?
I
- 5.
Can the particular steps or sets of steps be readily located wv'hcn required?
Page 10 of i2
.-A
.]
POINT BEACH NUCLEAR PLANT OM 4.3.3 OPERATIONS MANUAL NNSR Revision 0 EOP VALIDATION July 30, 1993
- 6.
Can the procedure exit poin, be returned to without omitting steps when required?
- 7.
Can the branched procedure be entered at the correct point?
- 8.
Are tOP exit points specified adequately?
Page i1 or 12
POiNT BEACH NUCLEAR PLANT OPERATIONS MANUAL EOP VALIDATION OM 4.3.3 NNSR Revision 0 July 30, 1993 ATTACHMENT A (Page i of 1)
PRA CORE DAMAGE RISK MATRLX Pagc 12 of 12 0
0 Z-lý
Nuclear Powecr Bu-,iness Unit DOCUM ENT REVIEW AND APPROVAL Note. Refer io NP 1. LJ for reqtitellzreais P.t'"e I of I - INITIATION Doc Numiber OMI 4.3.3 Unit PBO Usage Lesel Inforniation Proposed Re% No ~
Title LOP Validation
_______Cla~sification NA E] Re vision E Cancellation E]Ncw Document EjOther (e g. pý nueic re,.ie4. adminr hold)
List Temsporary Clianges/Feedbacks Incorporatcd-_________________________________
Dcsription of A] tcr,,non/.Reason (If nCes~4Iy. continue description of changes on PBF.0026c and attachi OMI 4.3.3 information m as incorporated in OM 4.3.2, EOP/AOP Verificatiom/Validation Process (Re,,. 2)
List othcr docuei~nts required to he effective concurrently with the rcvisicnt (c g. other procedures. forms drz%.ingfs etc OMI 4.3 2. EOPIAOP VerificationlValidation Process (Rev. iL~
Dout-tncrt Preparer (prnn'sigen)
J,-rr;zut G Green
/
C-A.9._-4 Datce
.~~
Indieatecs d raft prepared accordins! to N~P 1 1 3. any commitmrcnits/bascs deh ýhavc been documented and rtsuked IT.- TECFHNiCAL REVIEW Cannot be the Preparcr or A pprosa ayt.ihoritt)
Technical Reviewer (print~sign) je
~
Dt Lidicates draft technically correct. conststc-it 'Aith rcekrnrces/base-luppcir tier requiements. rcuircmrntf`of NP I I 3 comp~cied
[if - DOCUMENT OWNER REVIEFW Required Revicuers'Oreanizations:
__________/___________________5________
V3lidation Required?
tZ NO 0YES 5WAIVED (Group Head Appro% at and Rexton Required)
Reaso Valdatin Waved:Continue on PBF.00216c if neccssary, Validation \\Viiver Approval:_____________________________________________
Ciroup tiead Signature Changesc pre-screened according to NP 5.1.? 03NO [3 YES (Pmside documwn~ation accord'i', io NP S E)
Screenting completed according to NP 5. 1.8? 0 NA 5 YES (Attach copy) Safety evaluation required? [: NO [3 YES Training or briefing required? ;9 NO [: YES If YES, training or briefing! required before issue?
F-NO El YES I rai ni I I assi sitace desired?
[6NO E-YES IftYES. Trai ring Coordinator contacted/date:I QRJMISS Review NOT Required (Admin orNNSR only) LI QR Reviess Required
[] MSS Resew Required treferencee NP) 65)
Documenrt Owner (prnnslsign) /&-
ý1
/<
/1).
Indicates docutnent.:s tcchni-ailv correct, can be performed ixs written, does not ~vcenely affect Ixr%(onnLI orniecicsr safe.) armpropnatc re~i' iehase beei1 Verforniul tic t
.echnical, cross-dJissciplinary. vulidiation and 50 59nl2 48), comments hase been resol.ed and incorporatied as pp-opritci. affecie.)
ducunieniJ trainingt/bieflng ha~c been idLituifed tiid %turd processin.- completed. Document Control notified a iemer.criet issuance required te g. may be iess than 2 da-ys for procedure iscuance)
TV -APPROVAL (The Preparer. QuAlifrid Reicv erk (R),
and Approial Autiharti-sh.ll tic difTercni indis iduaI,,)
QR/MNSS tpnnrlsigr) r'i //Zj Date_______
lndiciime' 5o 59nl2 48 applieahilitý assessed any necssary scro~ningslcvatmuations performed de:eminnaiion mide as, to sshether addtticnul cr. N disciplinary r'.Nim' rnqiired. and if required performedi MS Meeting No C
Apprvsal Authority (prnnilsign
,}
Date V '- RELE4SkE FOR DISTRIBUTION I1NA [] YES Pre-implemcniitiois requirements comnplete (e P I trainin.,lbricfngs, 3lfeLled dss~uncrits Ninrd proces~.ifg. etc SSpecific effective date not required. Issue per Document Control schedule.
El Required effective date-
_____________(Coordinaie dalt.
11iih tDocuim-tE irf-l)
Document Owncr/Designce tprnif/aign)
&4 I
I.
Daiic ilectine Date (to beN ni~rcd bs Dniunienit Cinirol, MA 1L 3 M
PEC'D M.1AY 16 2002 MAY 3 1 2002 I~re:n.,- %1 I1 \\Pl I ".III 16IC is,,
is-i
". 1\\11 a -u-11,+/-
K-.%1'1')1123 IJI6Jo-'
The following procedure, OM 4.3.3, Revision 0, was canceled on May 13, 2002
NLuclcar Powecr Business Unit C.
DOCUM',ENT REVIEW NV~D APPROVAL.
-e-1
~
N\\ore: Ref'r lo N'sP 1.i 3for requirements.
t aelo
- t.
I - INITIATION Doc Nur~c O\\1 4-3.2 Unit P130 Usace Le'~e1 lrformation Proposedklcv No 2
FTitle EOPJAOP Vez-ifination.IValidation Process Classification NA
~ Re'iion ElC.rncelato n lNcw Docmcnt F] Other (e g. periodic review. idmirn hold)
MSS Meeting No.
Date Approval Authority (prnt/sign)
IQ1.
C1 V
-;RELEASE FOR DISTRIBUTION N NA El]YES Pre-implementation requirements cor-pltc (e.g.. trainrngi/brtzintngs. :ittecteouý'slmns%-
Fdrcseec l0 Specific effective date not required. Issue per Document Control schedule.
El Required ef.kctise date Coordi-iate date shD OoCLtn~nt ontre)
Dtcument Owner/De-si-ence (onrtlsign)
/
V1' Da5te '5 Fffei-ci~e Da-te tto be enrtered by Documcnt Control)
MAI_______I REC'D MAY0-1I ROFILM I'5F CNJ:6a Rrfer'nics. NP I I I NP I 1 5. N",1 LAAV
'N 2ri2% \\1'12 1 -'PI :'s List Temporx-Y Chan~ges/FeedbaCks Incorporatcd: _______________________________
Descriptioin of AlleratiordRcasofl (if neceesar). continue description of changes on PBF-0026e and attach)
Towa resi rite, revised format per Pro~cedure WVriters' Guide, incorporated ONI 4-3.3. EOP Validation (Rev~. 0) information and PRA Core Darrsaze Risk dat2. See PBF-001.6c tar details.
List other documents required to be effectise concurrently with :i'te revision (c g. atberprocedurea., forms. drawings.ectC1 PBF-2102a (Rev. 0). 2102b (Rev. 0). 2103a (Rev. 0). 2103b (Rev. 0). -2103c (Rev. 0)
Document Preparer (pnnrltsin)
James G Green I
Date 1~I
(
trd'c-atcs jiaft prepared accordine to NP) 1 3. any commiuncentsJt'acis5 c m+/- have been docunvnrtted and rcioI'.ed I! - TECHNICAL REVIEW (Cannot be the Irtpartr or Approval Authoritj,ý'
D t 6?.
Technical Reviev. er (pnnt'sagf)
~
~
/%~
Indicuics draft teehne-allv correzt. consisetwt ceecshSa~rte uircrriertts. requirmemnfts of N*P 1 1.3 completed III - DOCUMENT OWVNER REVIEW Required Rc~itewersJOrganizations:
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PA Validation Rcquirc..7 IZ NO El YES El WAIVED (Group H-ead Approsal and Rason Require)
'POTE: 9L4q 94IC',i Reason Validatiun Wa~ived-
___________________/_________________
Continue on PIDF-0026c if necessaty.
v Validation Waiver Apý-7val:
Group H-ead Signature Chasnes pre-scrcene-d according to NP 5.1.8? [:]NO [3 YES (Provide documentat'on according to '1P 't 1.8)
Scree.cing :omplered accordin-to NP 5..8?,ONA El]YES (Attach cop)) Safety evaluation required? [-]NO [] YES Training or b~tc.ing required*ý9SO O YES If YES, training or brie fing required be fore issue?."
NO ElYES Traininie assistance desired?y~p-iNO 9
YES If YES. T.alijng Coordinlator conuctcd/date:
'?
S-/- A a
N QMN/ŽISS Pcesicsv NOT Required (Adnin or NNSR only)
[] QR Review Required El MISS Review Required (reference NP 1 65)
Document Oxvn, injin 6.3~U & u (73'
ý4 Date & ' '
In.licaics documrent i. iechnxcally correct, can be performed as writ' does not adverscly affec pcrsonnel or u esar sfty. appropnat rev-iews have been perforated tic.. technical. cross-disciplinary. %alidation and 50-59/72.43). cotireents have been resol%cd and irated us appropriate. affected documnrsrra.~stinirg/bncefing base been identificd arid % ord.rwcessing completed Document Control notified if emecrgent issuance requiredc Zeg.
mnaybe IV -APPROVAL (The PreparerQualiraed Re~iieer(Q(
- 1. 2nd Appro% ai Authority 0h2ll bed difft ent indiisidutlSb)
QMISS Ipitiin reI Date _____
indicaues 5O.9fl2 -S applicability asisessed, any nccesis-ry scricsiings/c~aluations pcrfort'ed, determination mnad as; to shether additiona ýcross dttcipbn-%rN resie' required. and if required. peifonncd
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DOCUMENT REVIEW AND APPROVAL CONTINUATION Pae
?
__a__
FDoc 7:nubcr ON 14.3.2 Rcvision I
Unit PB0 ritic EOP/AOP VeriricationIN'alidation Process Temporary Chance Nu.imber Description Of ChangCs:
Step
- Change/Reason Ccscr Sheet Added co% er sheet per Procedures Writers' Guide. I This is pre-screencd to Criteria #I -EditoraW.
1.0 Simplified PURPOSE statement and incorporated information from OM 4.3.3. EOP Validation. / This is pre-screened to Criteria #2-Administrati\\e Procedure.
Added new section IDISCUSSION j per Procedures Writers' Guide. / This is pre-screened to Criteria # 1 Editoral.
2.1 Incorporated information from the old PURPOSE sections of OM 4.3.2 and OM a.3.3 / This is pre throurh wrcened to Criteria # 1-Editoral.
22 Added steep 2.3 in reference to the PRA Risk matrix in Attachment D. IThis I. p,e-screened to Criteria
- 2-Administratise Procedure.
24 ~
Added steps. I This information is clarifying in nature. This is pre-scrcencd to Criteria 42-Administrative throuch Procedure.
2.7 3.1 Incorporated information from the old RESPONSIBILITIES sections of OM 4.3.2 and OM 4.3.3. /This throuh is pre-screened to Criteria #1-Editoral.
Added 'Nuclear Engineering) responsibility. /This is pre-screened to Criteria #2-Administrative ProcedLre.
Added {Reactor Engineering) responsibility. I This is pre-screened to Criteria #2-Administra.ive Procedure.
Added (General) step to Section 4.0./This allows a place for non-specific information to be located together. I This is prc-screcnad to Criteria #2-Administrative Procedure.
4.1J1 Incorporated information from the old PROCEDURE sections of OM 4.32 and OM 4.3.3. /This is pre thro4gh st cened to Criteria # I-Editoral.
4.1.2 4.1.3 Added step to cross reference other procedures in the '%,-'.t standard steps are revised lThis is vre screened to Criteria #2-Administrative Procedurl..
4.1.4 Added step to reference the EOPSTPT for applicable changes I This is prc-screened to Criteria #2 Administrative Procedure.
4.1.5 Incorporated information from tme old PROCEDURE section of OM 4.3.2. /'This is pre-screened to Criteria 0 1 -Editoral.
4.1.6 Added step referencing the De'iarion Documcni. /'I his is pre-screcned to Criteria #2-Administratie Procedure Added NOTE. /This inlormafion is clariOing in nature. "I is is pre-.crcened to Criteria 42 Adniinistrati c Procedure.
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Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION MAt I ' 2222 Page,_.J.of 5 Doc Number OM 4.3.2 Revision 1
Title EOP/AOP Verifi=tionrValidation Process Temporary Change Number Description of Changes:
Step Cbange/Rea--on Added new step to diff'renfiate between Technical changes and Editorial changes to EOPsMAOPs. / This 4.2.1 is pre-screened to Criteria #2-Administrative Procedure 4.2.2 Added new step to reference new Attachment A for Tech. Evaluation Guidelines. I ThIs is pre-screened to Criteria p2-Administrative Procedure.
4.2.3 Incorporated information from the old PROCEDURE section of OM 4.3.2. /This is pre-screened to Criteria #1-Editoral.
Incorporated information from the old PROCEDURE section of OM 4.3.2. Revised IVernfication Teamn 4.2.4 requirements. Added PRA Group reference. / Expanded Team member requiremer.; and PRA Group
- nvolvement allows for more accurate evaluation. This is pre-screened to Crite;ia #2-Administrative Procedure.
4.2.5 Incorporated information from the old PROCEDURE section of OM 4.3.2. / This is prz-screcned to through Criteria #' -Editoral.
4.2.8 Added NOTE. I This information is clarifying i:i nature. This is pre-screened to Cnteria #2 Administrative Procedure.
4.2.9 Added step referencing a safety review. (This is pre-screened to Criteria #2-Ad.mnistrative Procedure.
4.2.10 Note Added this NOTE and steps to ensure specific groups review/evaluate procedure changes that effect their through areas of responsibility. /This is pre-screened to Criteria #2-Administrati'c Procedure.
4.2.13 Incorporated information from the old PROCEDURE section of OM 4.3.2, and revised and reformatted Section 4.3 the steps within the section to be consistent with the Validation steps. /This is pre-scrcencd to Criteria
- 1-Editoral.
Section 4.4 Incorporated information from the old PROCEDURE section of OM 4.3.3. IThis is prc-screencd to Criteria #_ -Editoral.
4.4.1.a Incorporated information from the old PROCEDURE section of OM 4.3.3. / This is pre-screened to through Criteria #l-Editoral.
4.4. i.b 4 4.Added new step referencing Table-top validation method. ( This is pre-screened to Criteria #2 Admir.istratic Procedure.
Added new step defining the Validation Team Leader qualifications. (This is pre-screened to Criteria #2 Administrati,,c Procedure.
Incorporated informatior, from the old PROCEDURE section of 0M 4.3 3./This is pre-screcne-t to Criteria,1-Editoral.
No;c Rccerding of Step Nurr&"nasv; ri nol rquinfd for nu!tiple oc-urrcn.cs cf :dkntcj inItorrniinn or uthen not tvnficn io 10 rc% le, er PBF.006c P,\\isscn6 (W/IS101 RIf.r'Lnct NP I I ". NP 1 23 Unit PB0 Other Comments
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUAT!ON Page.j+ of 5
Doc Numbcr ON 4.3.2 Revision I
Unit PB0 Title EOP/AOP VerificationfValidation Process Temporary Change Number Description of Changes:
Step
- ChangefReason 4.4.3.b Added new sub-steps dcf.ning the requirements of the Validation Team. I This is pre-screuned to Cnrteria through
- 2-Administrative Procedure.
4.4.3.d 4 4.3.e Incorporated information from the old PROCEDURE section of OM 4.3.3./(This is pre-screened to Criteria #W -Editoral.
Added new step directing the Validation Team to review the Verification Teams work. I This is pre 4.4.5 screened to Criteria #2-Administrative Procedure.
Incorporated information from the old PROCEDURE section of OM 4.3.3./ This is pre-screened to through Criteria # 1-Ed itoral.
4.4 6 4.5 I'a Incorporated information from the old PROCEDURE section of GM 4.3.3. /This is pre-screened to through Criteria #I -Editoral.
4.5.1.d Added new step for the evaluation of the Simulator response. / This is pre-screened to Criteria #2 Administrative Procedure.
4.5.!.f Incorporated information from the old PROCEDURE section of OM 4.3.3. /This is pre-screened to Criteria # 1-Editoral.
4-5.2.a Added new step to define the course of the simulator scenario performance. !This is pre-screened to I Criteria #2-Administrative Procedure.
4.5.2.b 4.5.-2.b Incorporated information from the old PROCEDURE section of OM 4.3.3./This is pre-screened to throuah Criteria #l-Editoral.
4.5.2.C Added new step to direct use of alternative methods of validation for parts of the procedure that are not evaluated by the simulator. /This is pre-screened to Criteria #2-Administrative Procedure.
Section 4.6 Added new section to define the steps to be followed during a Walkthrough Validation. I This is pre screened to Criteria #2-Administrative Procedure.
Added new section to define the steps to be followed during a Table-top Validation. / This is pre Section 4.7 screened to Criteria #2-Administrative Procedure.
Section 4.8 Added NOTE to direct the re-performance of portions of the verification or validation processes. I This is NOTE pre-s:-recned to Criteria #2-Administrative Procedure.
4 8.1 Incorporated information from the old PROCEDURE section of OM 4 3.3. /This is pre-screened to through Criteria #W -Edtioral.
4 8.2 4.S.3 Added new step for the evaluation of the Simulator response. I This is pre-screened to Criteria #2 Administrati'.c Proiedure.
Other Comments
- Noie Fccwrdig of Slcp Nt,:bcri is net rjuitcd for multiple occurrnce, of tdcnatc:l,fikrin:ion or %%hen not tbnefiiaal to t:\\icyrs.
1PBF-0026e Kcximon 6
/1.1 Rc~frcnc,, NP i 1. NP 1 2 3 I I
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Pa.e._. of 5
)oc Number OM 4.3.2 Revision 1
Unit PB0 Title EOP/AOP VerificationlValidation Process Temporary Chaige Number Description of Changes:
Step Change/Reason 4.8.4 Incorporated information from the old PROCEDURE section of OM 4.3.3. / This is pre-screened to through Criteria
- 1-Editoral.
4.8.5 4.8.6 Added new step describing the post-validation responsibilities of the Team Leader. /This is pre-screened to Criteria #2-Administrative Procedure.
1NOTE Added NOTE. I Ihis information is clarifying in nature. This is pre-screened to Criteria #2 Administrative Procedure.
4.9.1 Added new step describing the final approval process. / This is pre-screened to Criteria #2 Administrative Procedure.
4.9.2 Incorporated information from the old PROCEDURE section of OM 4.3.3. /This is pre-screened to Criteria #1-Editoral.
Added new step describing the Operations Manager responsibilities. /This is pre-screened to Criteria #2 Administrative Procedure.
5.1 Incorporated information from the old PROCEDURE section of OM 4.3.2. Deleted references to (NOM through EOP), {NP 1.2.2) and {PBNPEOP)./{NOM EOP) and {NP 1.2.2) have been canceled. {PBNPEOP) 5.9 is a redundant reference. This is pre-screened to Criteria # !-Editoral.
5.10 Added references to new forms developed from the forms in the old OM 4.3.3./This is pre-screened to through Criteria #2-Administrative Procedure.
5.12 6.0 Added new sec.tion [BASIS) per Procedures Writers7 Guide. / This is pre-screened to Critera #1 Editoral.
Added new Attachment A to provide guidance for Technical Evaluations. / This is pre-screened to Criteria #2-Administrative Procedure.
Attachment B Added new Attachment B to provide guidance for Status Tree Evaluations. /This is pre-screened ;o Criteria #2-Administrative Procedure.
Attachment C Incorporated infoimation from the old TABLE I of OM 4.3.3. / This is pre-scrcened to Criteria # I Editoral Attachment C Added new step for guidance in the validation of actions taken outside the Control Room. / This is pre step 2.3 screened to Criteria #2-Administrative Procedure.
Added new Attachment D to provide guidance for PRA Core Damage Risk Assessment. / This is pre Attachment D screened to Criteria #2-Administrative Procedure.
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.'1V I 1 1, M, 1 2 3
OM 4.3.2 EOP/AOP VERIFICATION/VALIDATION PROCESS DOCUMENT TYPE:
REVISION:
EFFECTIVE DATE:
APPROVAL AUTHORITY:
PROCEDURE OWNER (title):
OWNER GROUP:
Administrative 2
May 13, 2002 Department Manager Group Head Operations
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 1.0 PURPOSE The purpose of this procedure is to establish the requirements for the verification and validation processes for the Emergency Operating Procedures (EOP) and Abnormal Operating Procedures (AOP).
The verification and validation processes are applicable to procedures designated with EOP, ECA, SEP, CSP, ST, and AOP.
2.0 DISCUSSION 2.1 Verification of EOPs and AOPs is the process of independently checking that the procedures are technically correct, that any deviations from the corresponding ERG/ARG guidance are justified, that the procedures are compatible with plant hardware, and that the procedures adhere to the guidance in OM 4.3. 1, AOP and EOP Writers' Guide.
2.2 Validation of EOPs and AOPs is the process of exercising procedures to ensure that they are usable, that the language and level of information is appropriate, and that the procedures will function as intended. The validation requirements of this procedure are not applicable to revisions made for the correction of typographical errors.
2.3 The matrix in Attachment D was developed based on initiating events with a frequency of core damage greater than 1E-6 and an initiating event frequency of greater that 1E-3. The selected scenarios were then compared to the procedures that the operator would most likely use to prevent core damage. It is expected that procedure validation would consider those scenarios where an X is marked. This matrix is risk based only and should not be used as the sole consideration for determining scenarios for procedural validation.
2.4 EOPs, AOPs, and supporting documentation are revised for the following reasons:
Plant design changes Operator comments or change requests Industry or plant operating experience ERG or ARG revisions Corrective action program Tech Spec changes Revisions to other related program instructions INFORMATION USE Page 2 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 2.5 EOP revisions associated with design changes, Tech Spec changes, or other related procedure changes should normally be implemented concurrently with the change. EOP revisions required to correct technical deficiencies in the EOPs shall be completed in a timely manner.
2.6 Operator requalification training on EOPs provides a means of periodically verifying the technical adequacy of emergency procedures. Operators and training personnel are responsible for ensuring that problems or discrepancies discovered in EOPs during training are documented. Proposed enhancements and suggestions for improvement of the EOPs should also be encouraged.
2.7 Temporary changes to the EOPs and AOPs will be processed and controlled by NP 1.2.3, Temporary Procedure Changes. These changes are usually limited to emergent technical changes and do not require verification or validation per this procedure.
3.0 RESPONSIBILITIES 3.1 Manager's Supervisory Staff (MSS)
The MSS shall have the responsibility of reviewing and approving revisions to the EOPs and AOPs.
3.2 Operations Manager The Operations Manager shall have the overall responsibility for the EOP Verification and Validation processes.
The Operations Manager shall designate the personnel who will comprise the Verification Team.
3.3 EOP Writer The EOP writer shall determine the need for revision of the EOP supporting documents and develop revisions for those documents as necessary.
3.4 Nuclear Engineering Nuclear Engineering should coordinate the necessary changes if a revision to the EOPSTPT is required.
3.5 Reactor Enzineering Reactor Engineering should initiate revisions to the Safety Parameter Display System (SPDS) if revision to CSP-ST.0, Critical Safety Function Status Trees, are required.
These revisions shall not be implemented until approval of the CSP-ST.0 revision.
INFORMATION USE Page 3 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.0 PROCEDURE 4.1 General 4.1.1 The Emergency Response Guideline (ERG) or Abnormal Response Guideline (ARG) documents shall be reviewed to evaluate the intent of the corresponding ERG/ARG steps and whether the proposed change constitutes a deviation from the WOG guidelines.
4.1.2 The applicable EOP Deviation Document shall be reviewed to ensure that previous commitments are properly evaluated and to assess the justification for the present version of the step.
4.1.3 Similar or related steps/actions contained in other emergency procedures shall be evaluated for potential impact.
4.1.4 When setpoints are involved, the EOP Setpoint Document (EOPSTPT) shall be reviewed to ensure that setpoints are correctly implemented and to determine if revision of the EOPSTPT is required.
4.1.5 Review the applicable portions of OM 4.3.1, AOP and EOP Writers' Guide, to ensure compliance with the writers guide.
4.1.6 All safety related deviations from the WOG guidelines shall be documented and justified in the associated Deviation Document.
4.2 Verification Process NOTE:
Technical changes involve any of the following:
"* Changing the method of performing a step or the sequential order of steps
"* Changing the intent of any step, note, or caution
"* Adding, deleting, or changing numerical values, limits, bands, or setpoints
"* Changing instrumentation or controls used in the procedure
"* Changing entry/exit conditions or symptoms
"* Addition or deletion of steps, notes, cautions, graphs, tables, etc.
"* Any change which deviates from the WOG guidelines 4.2.1 Technical changes to EOPs should be verified by a multi-discipline team (at least three members) to maximize effectiveness of the verification process.
Non-technical (editorial) changes to EOPs and changes to AOPs may be verified by a single individual provided that the individual is a licensed operator and a qualified reviewer.
INFORMATION USE Pace 4 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.2.2 Technical changes to EOPs and changes to AOPs should be evaluated using Attachment A, Technical Evaluation Guidelines. Changes to Critical Safety Function Status Trees should be evaluated using Attachment B, Status Tree Evaluation Guidelines.
4.2.3 To ensure an independent verification process, personnel who have been involved in the development of the procedures(s) being verified should not be selected as verifier or appointed to the Verification Team.
4.2.4 The Verification Team members shall consist of, as a minimum, a Chairman, a licensed operator (SRO or RO), and a Training representative. Other members should be selected based on the type of change(s) being made to the procedure. For technical changes, a member of the PRA Group should review the changes but does not have to be a part of the Verification Team meeting.
4.2.5 The Verification Team members shall be listed on PBF-2102a, EOP Verification Team Meeting Form.
4.2.6 Verification Team members should obtain source documents as necessary, such as WOG guidelines, Deviation Documents, and Background Documents.
Other documents such as Tech Specs, FSAR, and other supporting procedures may also be applicable.
4.2.7 Review applicable portion(s) of the revised procedure. Depending upon the scope of the revision, it may be necessary to review the entire procedure and other interfacing procedures to adequately verify the revision. If step numbering or sequencing is affected by the revision, then the entire procedure shall be verified for internal step number referencing.
NOTE:
Minor discrepancies may be resolved by the Verification Team without the use of PBF-2102b, EOP Verification Discrepancy Form.
4.2.8 Identify and document discrepancies on PBF-2102b, EOP Verification Discrepancy Form.
4.2.9 A safety evaluation, in addition to the screening review, should be prepared for changes which involve new deviations from the WOG guidelines.
INFORMATION USE Page 5 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE NOTE:
The required reviews contained in the following steps may be performed concurrently with the verification process if the appropriate personnel are part of the verification team. If performed separately, the review should be identified as a Cross-Discipline Review. The Operations procedure writer is responsible for ensuring that assigned reviewers understand the scope of the review required.
4.2.10 Engineering shall review EOP/AOP revisions which involve any of the following:
- a. New deviations from WOG guidelines or changes in the method or scope of deviations from the ERG or ARG.
- b. Addition, deletion, or changes in setpoints or setpoint usage.
- c. Changes to status trees or other changes affecting SPDS displays.
- d. Additions or changes to actions outside the control room which could impact radiation dose estimates.
- e. Changes in instrumentation used in EOPs which could affect compliance with Reg Guide 1.97, Post Accident Monitoring Instrumentation.
- f. Proposed revisions to AOPs that affect Technical Specifications surveillance requirements.
4.2.11 Reactor Engineering should review proposed revisions to EOPs or AOPs which may affect Reactivity Management.
4.2.12 The PRA Group should review any proposed major revisions to EOPs or AOPs.
4.2.13 Organizations other than Operations (such as Chemistry, Radiation Protection, or Maintenance) should review proposed revisions to EOPs and AOPs which affect actions by the affected organization.
4.3 Resolution of Verification Discrepancies 4.3.1 Verification discrepancies are documented using PBF-2102b, EOP Verification Discrepancy Form, so that future revisions will not undo corrections or improvements made as a result of the verification process.
4.3.2 The Validation Chairperson shall assign personnel (preferably those responsible for writing the procedures) to prepare a resolution for each discrepancy.
INFORMATION USE Page 6 of 28
POINT BEACH NUCLEAR PLANT OPERATIONS MANUAL EOP/AOP VERIFICATION/VALIDATION PROCESS OM 4.3.2 Revision 2 May 13, 2002 TOTAL REWRITE INFORMATION USE Page 7 of 28 4.3.3 The personnel assigned to resolve the discrepancy shall:
- a. Propose a resolution to correct the discrepancy on PBF-2102b, EOP Verification Discrepancy Form.
- b. Obtain concurrence from the Verification Chairperson, as applicable.
- c. If the Verification Chairperson does not concur with the resolution, coordinate efforts to assess and resolve the discrepancy.
- d. Document the final resolution on PBF-2102b, EOP Verification Discrepancy Form.
4.3.4 If the discrepancy cannot be resolved between the personnel assigned to resolve the discrepancy and the Verification Chairperson, then the Verification Chairperson shall recommend a corrective action and obtain approval from the Operations Manager or designee.
4.3.5 After resolution of the discrepancy has been determined, the Verification Chairperson shall:
- a. Ensure the procedure is changed to incorporate the resolution of the discrepancy.
- b. Determine the scope of any additional verification required.
- c. Document completion of the additional verification.
- d. Determine if additional training is required and, if so, notify the Training Department.
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.4 Validation Process 4.4.1 The validation method shall be selected using the following guidance:
- a. The simulator method is preferred and should be used, when practical, because this method:
More accurately demonstrates operator response to a specific scenario.
Effectively identifies discrepancies between instructions and Control Room hardware.
Effectively identifies discrepancies between instructions and the operators execution of them.
- b. The walkthrough method should be used when:
Use of the simulator method is impractical due to modeling constraints or other limitations.
In combination with the simulator method when the simulator method is partially impractical.
When the revision affects action taken outside the Control Room.
For changes which do not warrant simulator validation due to the nature or scope of the change.
NOTE:
The walkthrough method is more effective than a table-top discussion in ensuring that the instructions contain the necessary level of detail and are compatible with plant hardware and personnel.
- c. The table-top method should be used only when the simulator and walkthrough methods cannot be used effectively OR for minor editorial or technical revisions which do not involve plant hardware and do not warrant simulator or walkthrough validation.
INFORMATION USE Page 8 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.4.2 The Validation Team Leader shall be designated based upon the scope of the validation and the validation method(s) to be used. The Validation Team Leader should possess expertise in as many of the following areas as possible:
- a. Supervisory skills
- b. Plant Operations
- c. Operations Training
- d. Technical Bases
- e. Development of EOP/AOPs 4.4.3 The Validation Team members requirements should be based on the following:
- a. Technical changes to an EOP should be validated by a multi-disciplined team consisting of at least three members. Revisions to AOPs and minor changes to EOPs do not require a multi-disciplined team nor do they require a minimum of three team members.
- b. The Validation Team should collectively be knowledgeable in the following areas:
Plant Operations Training/Simulator Instruction Technical Bases Development of EOP/AOPs
- c. At least one member of the Validation Team shall be a licensed operator.
The operations personnel used as the operating crew for the validation scenarios may be included as part of the Validation Team.
- d. At least one member of the Validation Team shall be a simulator instructor (N/A for walkthrough or tabletop validation methods).
- e. The Validation Team members shall be listed on PBF-2103a, EOP Validation Form.
INFORMATION USE Page 9 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.4.4 The Validation Team Leader shall review the PBF-2102a, EOP Verification Team Meeting Form and any PBF-2102b, EOP Verification Discrepancy Form(s) to determine the validation methods to be used and identify significant changes incorporated into the new procedure revision.
4.4.5 The Validation Team Leader shall outline one or more scenarios encompassing the identified changes in the procedure. Select plant failures that will initiate the desire response, considenng the following:
- a. Use both single and multiple failures where practical.
- b. Use concurrent and sequential failures where practical.
- c. Use dual unit failures where practical.
- d. If the simulator is to be used, select simulator malfunctions that closely model the selected failures.
4.4.6 Each validation scenario shall be documented using on PBF-2103b, EOP Validation Scenario Form.
4.5 Simulator Validation Method 4.5.1 The Procedure Writer or Validation Team Leader should prepare for simulator validation as follows:
- a. Schedule licensed operators and a simulator instructor to participate in the simulator validation. Operators selected should be representative of the training level expected of all operators.
- b. Arrange for the needed resources to support the validation such as simulator time, copies of procedures and relate instructions, and copies of the scenarios to be covered.
- c. Review the purpose and objective of the validation with the operator(s) involved. Include a discussion of the procedure revision.
- d. Brief the operators on how the validation will be conducted.
- e. Evaluate any known simulator characteristics which are different from the actual plant responses for impact on the validation.
INFORMATION USE Pagre 10 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONJVALIDATION PROCESS TOTAL REWRITE
- f. Prior to beginning the scenario, the Validation Team will discuss any differences between units that may come into play during execution of the scenario. The Validation Team Leader should ensure that the operators are aware of these differences and what effect they have on execution of the steps to be validated.
4.5.2 Conduct of the Simulator Method.
- a. The operators will use the procedures in response to the scenario enacted on the simulator. The procedure writer may be present but should not interfere or provide guidance during the scenario.
- b. The Validation Team will assess the procedures by noting any problems or deviations during the simulator run.
- c. At the conclusion of each simulator run, the Validation Team will conduct a debriefing as follows:
Evaluate the instruction using Attachment C, Validation Guidelines and document all discrepancies on PBF-2103c, EOP Validation Discrepancy Form.
Allow operators to present any problems or discrepancies that they identified during the simulator run. Document all discrepancies identified.
Discuss any deviations noted during the simulator run to identify discrepancies in the procedures.
- d. Any portions of the procedure or other procedures impacted by the revision which cannot be validated on the simulator should be validated separately using the walkthrough or tabletop methods.
INFORMATION USE Pagre 11 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.6 Walkthroueh Validation Method 4.6.1 The Procedure Writer or Validation Team Leader should prepare for walkthrough validation as follows:
- a. Schedule personnel to participate in the walkthrough. Individuals selected should be representative of the training level expected of all similarly qualified personnel.
- b. Arrange for the needed resources to support the validation such as copies of procedures and relate instructions, and copies of the scenarios to be covered, and related technical documentation.
- c. Review the purpose and objective of the validation with the personnel involved. Include a discussion of the procedure revision.
- d. Brief the personnel on how the validation will be conducted.
- e. Prior to beginning the walkthrough, the Validation Team will discuss any differences between units that may come into play during execution of the walkthrough. The Validation Team Leader should ensure that the personnel are aware of these differences and what effect they have on execution of the steps to be validated.
4.6.2 Conduct of the Walkthrough Validation
- a. Walkthrough validation should be performed at the in-plant location(s) where the procedure would be performed.
- b. If the procedure being validated is written for either unit, then a walkthrough should be performed on both units.
- c. The Validation Team Leader will use the scenario to direct the walkthrough by first pioviding the plant initial conditions and then providing appropriate cues while the personnel walk through each procedure step.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE
- d. The personnel will use the procedures in accordance with the scenario and walk through or talk through actions they would take in response to each instruction step. Personnel should:
Describe actions they are taking.
Identify information sources used to take actions.
Identify controls used to carry out actions expected system response(s), how response(s) are verified, and action(s) to be taken if response(s) did not occur.
- e. At any time during the walkthrough, personnel may stop to identify any problems or discrepancies in the procedures. Validation Team members may ask questions during the validation.
- f. The Validation Team will assess the procedures by noting any performance problems during the walkthrough.
- g. At the conclusion of each walkthrough, the Validation Team will conduct a debriefing as follows:
Evaluate the instruction using Attachment C, Validation Guidelines.
Review comments made during the walkthrough and document all discrepancies identified.
Discuss any performance deviations to identify discrepancies in the procedures which resulted in the deviation.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE 4.7 Table-Top Validation Method 4.7.1 The Procedure Writer or Validation Team Leader should prepare for table-top validation as follows:
- a. Schedule personnel to participate in the validation. Individuals selected should be representative of the training level expected of all similarly qualified personnel.
- b. Arrange for the needed resources to support the validation such as copies of procedures and relate instructions, and the scenarios to be covered.
- c. Review the purpose and objective of the validation with the personnel involved. Include a discussion of the procedure revision.
- d. Brief the personnel on how the validation will be conducted.
- e. Prior to beginning the scenario, the Validation Team will discuss any differences between units that may come into play during execution of the scenario. The Validation Team should ensure that the personnel are aware of these differences and what effect they have on execution of the steps to be validated.
4.7.2 Conduct of the Table-Top
- a. The Validation Team Leader will use the scenario to direct the table-top discussion by first providing the plant initial conditions and then providing appropriate cues while the performer discusses each procedure step.
- b. The personnel will use the procedures in accordance with the scenario, discussing the actions taken in response to each instruction step while identifying any problems or discrepancies in the procedure(s).
- c. During the table-top, the Validation Team will discuss and evaluate the instructions against Attachment C, Validation Guidelines. All discrepancies from the checklist or from individual comments will be documented on an on PBF-2103c, EOP Validation Discrepancy Form.
- d. The Validation Team will assess the procedures by noting any performance problems during the walkthrough.
- e. At the conclusion of the table-top discussion, the Validation Team will discuss any deviations to identify discrepancies in the procedures which resulted in the deviation and document all discrepancies on PBF-2103c, EOP Validation Discrepancy Form.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE NOTE:
EOP/AOP changes to resolve verification/validation discrepancies may require repeating portions of the verification and/or validation process.
4.8 Resolution of Validation Discrepancies 4.8.1 Validation discrepancies are documented using form PBF-2103c, EOP Validation Discrepancy Form, so that future revisions will not undo corrections or improvements made as a result of the validation process.
4.8.2 The Verification Team Leader shall assign personnel (preferably those responsible for writing the procedures) to prepare a resolution for each discrepancy.
4.8.3 Discrepancies involving plant response from simulator validation shall be evaluated to determine if they were caused or aggravated by simulator modeling deficiencies.
4.8.4 The personnel assigned to resolve the discrepancy shall:
- a. Propose a resolution to correct the discrepancy on PBF-2103c, EOP Validation Discrepancy Form.
- b. Obtain concurrence from the Validation Team Leader, as applicable.
- c. If the Validation Team Leader does not concur with the resolution, coordinate efforts to assess and resolve the discrepancy.
- d. Document the final resolution on PBF-2103c, EOP Validation Discrepancy Form.
4.8.5 If the discrepancy cannot be resolved between the personnel assigned to resolve the discrepancy and the Validation Team Leader, then the Validation Team Leader shall recommend a corrective action and obtain approval from the Operations Manager or designee.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE 4.8.6 After resolution of the discrepancy has been determined, the Validation Team Leader shall:
- a. Ensure the procedure is changed to incorporate the resolution of the discrepancy.
- b. Determine the scope of any additional validation required.
- c. Document completion of the additional validation.
- d. Determine if additional training is required and, if so, notify the Training Department.
4.9 Final Approval of EOP/AOP Revisions NOTE:
Temporary changes to the EOPs and AOPs can be approved via NP 1.2.3, Temporary Procedure Changes.
4.9.1 Following completion of the verification and validation process, including resolution of all discrepancies, final approval is obtained.
4.9.2 MSS review and approval is required for technical revisions to the EOPs. If the basis and step deviation documents are affected by the change, the revised background document should be submitted with the EOP for MSS review.
4.9.3 All EOP/AOPs and background documents shall be approved by the Operations Manager or his designee.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE
5.0 REFERENCES
5.1 NUREG 0899, Guidelines for the Preparation of Emergency Operating Procedures 5.2 NRC Generic Letter 82-33, Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability 5.3 C. W. Fay letter to H. R. Denton, "Response to Generic Letter No. 82-33....," April 15, 1983.
5.4 OM 4.3.1, AOP and EOP Writers' Gui de 5.5 Westinghouse Owners' Group (WOG), Emergency Response Guidelines (ERGs) 5.6 Westinghouse Owners' Group (WOG), Abnormal Response Guidelines (ARGs) 5.7 PBF-2102a, EOP Verification Team Meeting Form 5.8 PBF-2102b, EOP Verification Discrepancy Form 5.9 Institute of Nuclear Power Operations (INPO) Guidelines, Emergency Operating Procedures Verification Guidelines,83-004, March 1983 5.10 PBF-2103a, EOP Validation Form 5.11 PBF-2103b, EOP Validation Scenario Form 5.12 PBF-2103c, EOP Validation Discrepancy Form 6.0 BASES NONE INFORMATION USE Page 17 of 28
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTACHMENT A TECHNICAL EVALUATION GUIDELINE Page 1 of 3 1.0 (EOP)
Are entry conditions consistent with those listed in the Owner's Group guidelines or are deviations justified in the basis and deviation documents (AOP/SEP)
Are entry conditions logical. (reflective of the expected conditions leading to performance of the instruction). Are the entry conditions observable.
2.0 (EOP)
Is the sequence of steps consistent with that in the Owner's Group Guidelines or are deviations adequately justified in the basis and deviation documents.
(AOP/SEP)
Are the steps sequenced logically. Does the sequence follow good operations principles.
3.0 (EOP)
Are all steps consistent with the intent of those in the Owner's Group Guidelines or are deviations adequately justified in the basis and deviation documents.
(AOP/SEP)
Is the intent of each step understandable. Does the step provide adequate detail.
4.0 (EOP)
Have all applicable Owner's Group Guideline steps been incorporated into the procedure or are deviations adequately justified in the basis and deviation documents.
(AOP/SEP)
Are the steps necessary instructions provided to the user.
5.0 (EOP)
Are differences from the Owner's Group Guidelines consistent with the intent of the Owner's Group Guidelines.
6.0 (EOP)
Is documentation adequate to explain the intent of complex steps.
(AOP/SEP)
Is documentation adequate to explain the intent of complex steps.
7.0 (EOP)
Is all Owner's Group Guidelines "bracketed" information, pertinent to the plant design, incorporated.
(AOP/SEP)
Is applicable plant design and components clearly addressed by the instruction.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTACHMENT A TECHNICAL EVALUATION GUIDELINE Page 2 of 3 8.0 (EOP)
Have all references to systems or components in the Owner's Group Guidelines that are applicable to the plant design been included.
(AOP/SEP)
Are all references to system, component and plant design clear and correct.
9.0 (EOP)
Are required computations, specified in the procedure. consistent with Owner's Group Guidelines or deviations adequately justified within source documents.
(AOP/SEP)
Are all required computations specified in the procedure. Has adequate guidance been given and is space available for working and recording computations.
10.0 (EOP)
Are the cautions and notes, as specified in the procedure, consistent with the Owner's Group Guidelines or are deviations adequately justified in the basis and deviation documents.
(AOP/SEP)
Are cautions and notes specified in the instruction clear and concise. Do they provide adequate information to convey the message.
11.0 (EOP)
Are the contingency actions in the procedure consistent with those specified in the Owner's Group Guidelines or are deviations adequately justified in the basis and deviation documents.
(AOP/SEP)
If specified/used, are contingency actions clear and easily understood. Do they provide adequate detail for implementation.
12.0 (EOP)
Is there a conflict between the foldout page requirements and the action steps of the procedure.
(AOP/SEP)
Is there any conflict between steps and required actions.
13.0 (EOP)
Are the required steps to be performed consistent with the plant design.
(AOP/SEP)
Are the steps consistent with plant design.
14.0 (EOP)
Are the quantitative ranges as specified in the procedure consistent with the plant design.
(AOP/SEP)
Are the quantitative ranges as specified in the procedure consistent with the plant design.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONIVALIDATION PROCESS TOTAL REWRITE ATTACHMENT A TECHNICAL EVALUATION GUIDELINE Page 3 of 3 15.0 (EOP)
Are the limits, as specified in the procedure consistent with those specified in the Owner's Group Guidelines or are deviations adequately justified in the basis and deviation documents.
(AOP/SEP)
Are limits clearly specified.
16.0 (EOP)
Are the charts, tables, and curves presented in the procedure consistent with the Owner's Group Guidelines or are deviations adequately justified in the basis and deviation documents.
(AOP/SEP)
Are the charts, tables, and curves consistent with the information provided in source documents.
17.0 (EOP)
Do parameter values, numerical values, and setpoints in the procedure correspond with the parameter values, numerical values, and setpoints specified in Setpoints Document.
(AOP/SEP)
Do parameter values, numerical values, and setpoints in the procedure correspond with the parameter values, numerical values, and setpoints specified in supporting technical documentation.
18.0 (EOP)
If the revision involves a change to a setpoint, have all the procedures affected been revised.
Verify against the list of affected procedures contained in the setpoints document.
19.0 (EOP)
If the revision affects a "standard" step, have all of the procedures affected been revised. Verify against the list of affected procedures contained in the standard step document.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONNVALIDATION PROCESS TOTAL REWRITE ATTACHMENT B STATUS TREE EVALUATION GUIDELINE Page 1 of 3 1.0 WRITERS'GUIDE CONVENTIONS 1.1 Procedure Title 1.1.1 Is the title 10 words or less.
1.1.2 Are the important words placed at or near the beginning of the title.
1.2 Identification Information 1.2.1 Does the procedure number include the required information:
- a. Instruction type
- b. Instruction number 2.0 STATUS TREE FORMAT 2.1 Page Format 2.1.1 Does the Status Tree clearly show the transitions.
2.2 Symbol Coding 2.2.1 Are the symbols used correctly.
2.2.2 Are arrows positioned correctly.
2.3 Function Flow and Branching 2.3.1 Does the flow path move from left-to-right.
2.3.2 Is sufficient spacing allowed between flow paths.
2.3.3 Are the number of arrowheads sufficient to indicate flow.
2.3.4 Does the flow path go down for each favorable response.
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- 4)
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTACHMENT B STATUS TREE EVALUATION GUIDELINE Page 2 of 3 3.0 READABILITY 3.1 Text 3.1.1 Is the text in black type against a light background.
3.1.2 Is the text readable at arms length under degraded lighting conditions.
3.1.3 Is the typeface legible and consistent.
3.1.4 Is spacing between letters and words adequate.
3.1.5 Is the correct line spacing used.
4.0 WRITING STYLE 4.1 Step Construction 4.1:1 Does each step contain only one statement.
4.1.2 Are the statements simple and precise.
4.1.3 Are double negatives avoided.
4.1.4 Are terms used consistently within and among status trees.
4.1.5 Does each decision step clearly indicate a yes or no answer.
5.0 MECHANICS OF STYLE 5.1 Spelling 5.1.1 Is the spelling correct.
5.2 Abbreviations and Acronyms 5.2.1 Are abbreviations and acronyms used consistently.
5.2.2 Are abbreviations used in accordance with the Writers' Guide.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATIONNALIDATION PROCESS TOTAL REWRITE ATTACHMENT B STATUS TREE EVALUATION GUIDELINE Page 3 of 3 5.3 Curves and Tables 5.3.1 Are the curves and tables legible, consistent with the instructions, and usable.
5.3.2 Are the safe and unsafe regions of curves labeled.
5.4 Hyphenation 5.4.1 Are hyphens used correctly.
5.4.2 Is hyphening at the end of a line avoided.
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x1 r POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTACHMENT C VALIDATION GUIDELINES Page 1 of 4 1.0 USABILITY 1.1 Level of Detail 1.1.1 Are the introductory sections of the instruction sufficient.
1.1.2 Is there sufficient information to perform the specified actions at each step.
1.1.3 Are the alternatives adequately described at each decision step.
1.1.4 Are labeling, abbreviations, and nomenclature as provided in the instruction sufficient to enable the operator to find the needed equipment.
1.1.5 Does the instruction have all information or instructions needed to manage the emergency condition.
1.1.6 Are the actions sufficient to correct the condition.
1.1.7 Are the titles and numbers sufficiently descriptive to enable the operator to find appropriate instructions.
1.2 Understandability 1.2.1 Is the instruction's typeface easy to read.
1.2.2 Are the figures and tables easy to read with accuracy.
1.2.3 Can the values on figures and charts be easily determined.
1.2.4 Are the cautions and note statements readily understandable.
1.2.5 Are the individual instruction steps readily understandable.
1.2.6 Were the step sequences understood.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTACHMENT C VALIDATION GUIDELINES Page 2 of 4 2.0 OPERATIONAL CORRECTNESS 2.1 Plant Compatibility 2.1.1 Can the actions specified in the procedure be performed in the designate sequence.
2.1.2 If alternate success paths exist, does the procedure use. the best method to accomplish the task.
2.1.3 Can the information from the plant instrumentation be obtained, as specified, by the instructions.
2.1.4 Are the available Control Room instrumentation and annunciators adequate for the Operator to recognize the entry or prerequisite conditions.
2.1.5 Are the instructions entry or prerequisite conditions appropriate for the plant symptoms displayed to the operator.
2.1.6 Is all the equipment required to accomplish the task specified in the instruction.
2.1.7 Do the plant resources agree with the instruction.
2.1.8 Are the instrument readings and tolerances stated in the instruction consistent with the instrument values displayed on the instruments.
2.1.9 Is the instruction physically compatible with the work situation (e.g., too bulky to hold, binding would not allow them to lie flat in the work space, no place to lay the instruction down to use).
2.1.10 Are the instrument readings and tolerances specified by the instruction for remotely located instruments accurate.
2.1.11 Can plant parameters be maintained within limits or bands specified in the procedure.
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C.
POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13. 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTACHMENT C VALIDATION GUIDELINES Page 3 of 4 2.2 Operator Compatibility 2.2.1 If time intervals are specified, can the instruction action steps be performed on the plant within or at the designated time intervals.
2.2.2 Will environmental conditions permit completing the required actions.
2.2.3 If concurrent or sequential steps are required by more than one individual, can the required actions be coordinated adequately.
2.2.4 Can personnel follow the designated action step sequences.
2.2.5 Can a particular step, set of steps, or other information be readily located when required.
2.2.6 Can instruction branches be entered at the correct point.
2.2.7 Are place keeping aids utilized as required by the user's guide.
2.2.8 Are instruction exit points adequately specified.
2.2.9 Are the procedures compatible with the operating shift manning.
2.2.10 If steps and instructions are verified with signoffs, are provisions adequate.
2.2.11 Do Operators interfere with each other physically.
2.2.12 Is there adequate Radiation Protection support and/or provisions to make the required entries into contaminated areas.
2.2.13 Does plant staffing support procedure requirements.
2.2.14 Is the procedure adequate to allow properly trained personnel to complete the task without errors.
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POINT BEACH NUCLEAR PLANT OM 4.3.2 OPERATIONS MANUAL Revision 2 May 13, 2002 EOP/AOP VERIFICATION/VALIDATION PROCESS TOTAL REWRITE ATTACHMENT C VALIDATION GUIDELINES Page 4 of 4 2.3 Additional Guidelines for Validation of Local Operator Actions 2.3.1 Can the Operator easily locate the component from a combination of the information in the procedure and operator training/knowledge.
2.3.2 Is the component clearly identified by name and/or number.
2.3.3 Is the component easily accessible.
2.3.4 Are special tools needed to operate the component.
2.3.5 Is the environment at the component location suitable to allow the operator to perform desired actions.
2.3.6 Do the local actions require more than one operator.
2.3.7 Are communications available from the remote location.
2.3.8 Is the Operator performing the local actions familiar with the procedure and does he/she understand the objective and/or consequences of his/her actions.
2.3.9 Are the local actions required to be performed in a specific time period. If so, can the actions be completed within this time period.
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POINT BEACH NUCLEAR PLANT OPERATIONS MANUAL EOP/AOP VERIFICATION/VALIDATION PROCESS OM 4.3.2 Revision 2 May 13. 2002 TOTAL REWRITE ATTACHMENT D PRA CORE DAMAGE RISK MATRIX Page 1 of I Procedure EVENT SGTR Turbine Trip LOOP Loss of CCW Steam Line without the Break Condenser EOP 0 X
X X
X X
X X
EOP 0.2 X
X EOP 0.3 X
X EOP 0.4 X
X EOP 1 X
X X
E O P 1.1 EOP 1.2 X
X EOP 1.3 X
EOP 2 X
EOP 3 X
X EOP 3.1 X
EOP 3.2 X
EOP 3.3 X
ECA 0.0 X
ECA 0.1 X
ECA 0.2 EC A 1.1 ECA 1.2 ECA 2.1 X
X ECA 3.1 X
ECA 3.2 X
ECA 3.3 X
CSP C.1 X
X X
X CSP H.1 X
X X
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