ML030790044
| ML030790044 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 10/30/2002 |
| From: | Nuclear Management Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2003-0094 AOP 3, Rev 4 | |
| Download: ML030790044 (27) | |
Text
a Nuclear Power Business Unit -% r TEMPORARY CHANGE REVIEW AND APPROVAL rf+
Note: Refer to NP 1.2.3, Temporarv ProcedureChanges,for requirements.
Page ! of Doc Number AOP 3 I -INITIATIONZ Current Rev 4 Unit PBI Temp Change No. 2002-0840 Document Title Steam Generator Tube Leak Existing Effective Temporary Changes N/A Brief Description Add AFW Minimum Flow Requirements to Foldout Page.
(Tdentif specific changes on Form PBF-0026c, Document Revtew and Approval Contnuaton.
and include with the package) 0 Initiate PBF-0026h and include with the change.
Other documents required to be effective concurrently with the temporary change: n/a Changes pre-screened according to NP 5.1.8? 0 NO [I YES (Pr0de documentation a:ccording 4o,41?S IS)
Screening completed according to NP 5.1.8? El NA 0 YES (Aamch copy)
S a fe ty E v alu at io n R eq u ir ed ? 0E NO E Ye-rm EeSm y r wal, r-, ,. u a %h allbe o,,
, ed b efre, iY e,i nS)
Determine if the change constitutes a Change Of Intent to the procedure by evaluating (If any answers ameYES, a revision may be the following questions.
processed or final reviews and approvals shall be obtained before implementng)
Will the proposed change:
YES NO
- 1. Require a change to, affect or invalidate a requirement, commitment, evaluation or description in the Current or ISFSI Licensing Basis (as defined in NP 5.1.8 and NP 5.1.7)? 0
- 2. Cause an increase in magnitude, significance or impact such that it should be processed as a revision?
- 3. Delete or modify a prerequisite, initial condition, precaution, limitation or other steps that El could have safety significance or affect the procedure's margin of safety? '
- 4. Delete QC hold points, Independent Verification or Concurrent Check steps without the related step(s) that require the performance also being deleted?
- 5. Change Tech Spec or other regulatory acceptance criteria other than for re-baselining purposes?
- 6. Require a change to the procedure Purpose or change the procedure classification?
E 0 Initiated By (print/sign) Ross Groehler / j . "4 2t.,-Date 10/30/2002 1 - INITIAL APPROVAL This change is correct and complete, can be performed as written, and does no versely affect personnel or nuclear safety, or Plant.oerating conditions.
Group Supervisor (printisign) /, A' fnrA 1 #-I $ Date. o (Cannot be tht'Initiator This change does not adversely affect J*t-perating conditions. ( Relate
-dures only)
Senior Reactor Operator (print/sign) Ie. r e" ) I Date /} oA (Cannot be the Initiator or Group Supervisor)
Permanent III - PROCEDURE OWNER REVIEW C1 One-time Use 0l Expiration Date, Event or Condition:
El Hold change until procedure completed (final review and approval still required within 14 days of initial approval)
El QRIMSS Review NOT Required (Admin/NNSR only) E QR Review R red [] MSS Procedure Owner (print/sign) ewRequire-, R cel NP .s) 6 R-e
- _ 5) Z.
ThisChane and suvoortingfeauirements correctlv comnleted and nmc,.--
IV- NAL REVIEW AND APPROVAL e.-*ust be comoleted within 14 days ofipiJal aoLZ'o'ah) rThe Initiator. OR and A rov I Autr AMSS (print/sign) shall be indeoendent from ach oth)r)
( . C- _9?)-z 1 ) / 4'I L 7/ Date 61.3-7 Indicates 50-59/72.48 applicability assessed, any necessary screenings/evaluations perform cross-disciplinary review rmination made as to whethdraddiu6nal required, and if required, pe rformed.'O
'N tr7 Aooroval Authority (orint/sizn) C't/1~
/9 ~ h. / Da Post Typing Review (print/sign)
V- REVISION INFORMATIOr OR PERMA ANGES .4 D e indicates temporary change(s) incorporated exactly as approved and no other changes made to document.
Incorporated into Revision Number Effective Date _
PBF-0026e Revision 13 01/16102 References" NP 1.23
Point Beach Nuclear Plant DOCUMENT REVIEW AND APPROVAL CONTINUATION Page _ of Doc Number AOP 3 Revision 4 Unit PBI Title Steam Generator Tube Leak Temporary Change Number 2002-0840 Description of Changes:
Step
- Change/Reason Add AFW minimum flow requirement: Monitor and maintain minimum AFW discharge flow or stop the affected AFW pump as necessary to control S/G levels./See SCR 2002-0458, CAP029908, P-3 8A MD)
AFWP has inadequate Recite Flow during IT-10. See also CAP029952, Possible Common Mode Failure Foldout Page of Auxiliary Feed Recirculation Lines Other Comments o Note Recording of Step Number(s) is not requirtd for multiple occurrences of identical information or when not beneficial to reviewers.
PBF-0026c Revision 6 04/1&/1
References:
NP I.13.NP 1.2.3
Point Beach Nuclear Plant TEMPORARY CHANGE AFFECTED MANUAL LOCATION Page of Procedure Number AOP 3 Revision 4 Unit PB1 Title Steam Generator Tube Leak Temporary Change Number 2002-0840 I - IMMEDIATELY AFTER INITIAL APPROVAL ON PBF-0026e (Non-Intent changes)
(after Final Approval if change of intent involved)
This procedure change has been processed ft as follows: (Manual/ocation) Date Performed E Copy included in work package for field implementation. (WO No. )
I[I Copy filed in Control Room temp change binder (Operations only). ID -Jo-o z.
[ Original change package provided to -16, to obtain Procedure Owner Review (eg., Owner review may be coordinated by In-Group OA IL Procedure Writer, Procedue Supervisor, etc.). / .,O -
ii
[1 Performed By (print and sign) Carol Schroeder / O.* .,J - Date 10/30/2002 II - PROCEDURE OWNER REVIEW ON PBF-0026e (may be performed by OA IL Procedure Writer, etc.)
Date This procedure change has been processed as follows: (Manual/Location) Performed
] Copy sent to Document Control Distribution Lead for Master File. 10 -3s -'d Z.
(Not required for one-time use change) f] Copy filed in Group satellite file. (Not required for one-time use changes)
El Copy filed in Group one-time use file.
[] Original Temp Change provided to d) Gb' to obtain Final Approvals / a -. 0 _
(e.g., final approval may be coordinated by In-Group OA Il, Procedure Writer, Procedure Supervisor, etc.)
SU1/U2 /o0.' oo?
[* PAB
[* OPS Shop
['Z OPS Office
, Simulator (Training OAII)
E]
Performed By (print and sign) Carol Schroeder / £.4(.1 . Date 10/30/2002 PBF-0026h Revision 5 06/13/01 Reference. NP 1.1-3
. I Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Venfy SCR number on all pages Page 1 title of Proposed Activity: AFW minimum flow requirement change to AOP, EOP, CSP, ECA, SEP, 01-62 A/B procedures Associated Reference(s) #: Removal of internals from AF- 117 and upgrade open function of AFW pumps minirecirc vlaves to safety -related (MR 02-029); SCR 2002-005-01 EOP/ARP actions for AFW mini-recirc requirement ; 2002-0055, P-38A/B mini recirc flow orifice replacment (MR 99-029 *A, *B);
Flowserve Corporation Pump Division letter dated March 2, 20012; CAP 29908; CAP 29952 Prepared by: Eric A. Schmidt / John P. Schroeder ,t ( A)4/, 420.
Name (Print) -- o ,
Reviewed by: Date: /0~.
N=a-( Print) Signature PART I (50.59/72.48) - DESCRIBE THE PROPOSED ACTIVITY AND SEARCH THE PLANT AND ISFSI LICENSING BASIS (Resource Manual 5.3.1)
NOTE: The "NMC 10 CFR 50.59 Resource Manual" (Resource Manual) and NEI 96-07. Appendix B. Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 10 CFR 50.59 and 10 CFR 72.48 screenings.
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 portions of the proposed activity may be documented via the applicability and pre-screening process requirements in NP 5.1.8.) Appropriate descriptive material may be attached.
This screening supports procedural uprgrades to address the Auxiliary Feedwater (AFW) System issue as identified in CAP 29908 and CAP 29952. Procedural guidance for operation of AFW System will be changed such that the operator must ensure that discharge flow for P-38 A/B must be greater than 50 gpm and 1/2 P-29 discharge flow must be greater than 75 gpm. If pump flow cannot be maintained within these requirements, the pump must be secured.
1.2 Search the PBNP Current Licensing Basis (CLB) as follows: Final Safety Analysis Report (ESAR), FSAR Change Requests (FCRs) with assigned numbers, the Fire Protection Evaluation Report (FPER), the CLB (Regulatory) Commitment Database, the Technical Specifications, the Technical Specifications Bases, and the Technical Requirements Manual. Search the ISFSI licensing basis as follows: VSC-24 Safety Analysis Report, the VSC-24 Certificate of Compliance, the CLB (Regulatory)
Commitment Database, and the VSC-24 10 CFR 72.212 Site Evaluation Report. Describe the pertinent design function(s),
performance requirements, and methods of evaluation for both the plant and for the cask/ISFSI as appropriate. Identify where the pertinent information is described in the above documents (by document section number and title). (Resource Manual 5.3.1 and NEI 96-07, App. B, B.2)
FSAR 10.2 Auxiliary Feedwater System (AF) - The AFW system shall automatically start and deliver adequate AFW flow to maintain adequate steam generator levels during accidents which may result in main steam safety valve opening, such as: Loss of normal feedwater (LONF) and Loss of all AC power to the station auxiliaries (LOAC). AFW system shall also deliver sufficient flow to the steam generators supporting rapid cooldown during such accidents as: steam generator tube rupture (SGTR) and main steam line break (MSLB).
Each pump has an AOV controlled recirculation line back to the condensate storage tanks to ensure minimum flow to prevent hydraulic instabilities and dissipate pump heat.
TS 3.7.5 Auxiliary Feedwater (AFW) System TS Bases B 3.7.5 Auxiliary Feedwater (AFW) System FSAR 7.3.3.4 Manual AFW Flow Control During Plant Shutdown Manual control of steam generator water level using the AF pumps to remove reactor decay and sensible heat.
FPER 6.6.4 Auxiliary Feedwater System The Auxiliary Feedwater Pumps are provided with a mini-recirc line to ensure a minimum amount of flow is established to keep the pumps from dead heading.
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Vcnfy SCR number on all pages Page 2 FSAR 10.2 Auxiliary Feedwater System (AF)
TS 3.7.5 Auxiliary Feedwater (AFW) System TS Bases B 3.7.5 Auxiliary Feedwater (AFW) System FSAR 7.3.3.4 Manual AFW Flow Control During Plant Shutdown FPER 6.6.4 Auxiliary Feedwater System 1.3 Does the proposed activity involve a change to any Technical Specification? Changes to Technical Specifications require a License Amendment Request (Resource Manual Section 5.3.1.2).
Technical Specification Change: I] Yes [0 No If a Technical Specification change is required, explain what the change should be and why it is required.
1 1.4 Does the proposed activity involve a change to the terms, conditions or specifications incorporated in any VSC-24 cask Certificate of Compliance (CoC)? Changes to a VSC-24 cask Certificate of Compliance require a CoC amendment request.
[]Yes EDNo If a storage cask Certificate of Compliance change is required, explain what the change should be and why it is required.
10 CFR 50.59 SCREENING PART II (50.59) - DETERMINE IF THE CHANGE INVOLVES A DESIGN FUNCTION (Resource Manual 5.3.2)
Compare the proposed activity to the relevant CLB descriptions, and answer the following questions:
YES NO QUESTION 0 El Does the proposed activity involve Safety Analyses or structures, systems and components (SSCs) credited in the Safety Analyses?
El 0] Does the proposed activity involve SSCs that support SSC(s) credited in the Safety Analyses?
0] El Does the proposed activity involve SSCs whose failure could initiate a transient (e.g., reactor trip, loss of feedwater, etc.) or accident, OR whose failure could impact SSC(s) credited in the Safety Analyses?
ED El Does the proposed activity involve CLB-described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, orders or technical specifications?
El 0 Does the activity involve a method of evaluation described in the FSAR?
El 0D Is the activity a test or experiment? (i.e., a non-passive activity which gathers data)
El 0D Does the activity exceed or potentially affect a design basis limitfor a fission productbarrier(DBLFPB)?
(NOTE: IfTHI.S 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 conclusion section (Part IV), then proceed directly to Part V - 10 CFR 72.48 Pre-screening Questions.
-any of the above questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Venfy SCR number on all pages Page 3 "AR-02-029 upgraded the open function of the AFW pumps mini-recirc AOV to safety-related. The safety-related boundary includes Ane recirc orifice and all associated upstream components and piping. It is postulated that a failure of the piping downstream of the recirc orifice will not have any adverse affects on the AFW system. The availability of the recirculation flowpath provides an additional flowpath to support minimum flow requirements. This procedure change will improve the reliability of the AFW pumps by not relying upon the recirc flow path for operability as it has been concluded that the restrictions in the recirc orifice may not be adequate for use. Whereas current guidance mandates that the operator verify the position of the recirc AOV and the status of the Instrument Air system, these procedural changes will only require the operator to monitor pump discharge flow.
PART M (50.59) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (Resource Manual 5.3.3)
If ALL the questions in Part II are answered NO then Part II is [] NOT APPLICABLE.
Answer the following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a 10 CFR 50.59 Evaluation is required; EXCEPT where noted in Part II.3.
I1.1 CHANGES TO THE FACILITY OR PROCEDURES YES NO QUESTION E] E] Does the activity adversely affect the designfunction of an SSC credited in safety analyses?
El [] Does the activity adversely affect the method of performing or controlling the designfunction of an SSC credited in the safety analyses?
If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers are NO, describe the basis for the conclusion (attach additional discussion as necessary):
Minimum flow requirements will be maintained within recommendations from the vendor by monitoring pump discharge flow and securing the pump as required. Starting and stopping of the AFW pumps has been previously evaluated in 50.59 Evaluation 2002-005, which addressed procedural changes to reduce the potential of pump damage as a result of the loss of the recirculation flow path.
111.2 CHANGES TO A METHOD OF EVALUATION (If the activity does not involve a method of evaluation, these questions are 0 NOT APPLICABLE.)
YES NO QUESTION E- [I Does the activity use a revised or different method of evaluation for performing safety analyses than that described in the CLB?
El El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in the CLB?
If any answer is YES, a 10 CFR 50.59 Evaluation is required. If both answers are NO. describe the basis for the conclusion (attach additional discussion, as necessary).
111.3 TESTS OR EXPERIMENTS If the activity is not a test or experiment, the questions in IH.3.a and lII.3.b are [D NOT APPLICABLE.
- a. Answer these two questions first:
YES NO QUESTION 0l El Is the proposed test or experiment bounded by other tests or experiments that are described in the CLB?
0l 0l Are the SSCs affected by the proposed test or experiment isolated from the facility?
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59172.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 4 If the answer to BOTH questions in V.3.a is NO, continue to III.3.b. If the answer to EITHER question is YES, then describe the basis.
- b. Answer these additional questions ONLY for tests or experiments which do NOT meet the criteria given in III.3.a above.
If the answer to either question in IH.3.a is YES, then these three questions are E] NOT APPLICABLE.
YES NO QUESTION El El Does the activity utilize or control an SSC in a manner that is outside the reference bounds 'ofthe 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?
0l El Does the activity place the facility in a condition not previously evaluated or that could affect the capability of an SSC to perform its intended finctions?
If any answer in HI.3.b is YES, a 10 CFR 50.59 Evaluation is required. If the answers in 1IT.3.b are ALL NO. describe the basis for the conclusion (attach additional discussion as necessary):
Part IV - 10 CFR 50.59 SCREENING CONCLUSION (Resource Manual 5.3.4).
Check all that apply:
A 10 CFR 50.59 Evaluation is [I required or 0 NOT required.
A Point Beach FSAR change is El required or 0D NOT required. If an FSAR change is required, then initiate an FSAR Change Request (FCR) per NP 5.2.6.
A Regulatory Commitment (CLB Commitment Database) change is [E required or [D NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.
A Technical Specification Bases change is [E required or Z NOT required. Ifa 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 0 NOT required. If a change to the Technical Requirements Manual is required, then initiate a Technical Requirements Manual change per NP 5.2.15.
-1 10 CFR 72.48 SCREENING NOTE: NEI 96-07, Appendix B, Guidelines for 10 CFR 72.48 Implementation should be used for guidance to determine the proper responses for 72.48 screenings.
PART V (72.48) - 10 CFR 72.48 INITIAL SCREENING QUESTIONS Part V determines if a full 10 CFR 72.48 screening is required to be completed (Parts VI and VII) for the proposed activity.
YES NO QUESTION
'I 0] Does the proposed activity involve IN ANY MANNER the dry fuel storage cask(s), the cask transfer/transport equipment, any ISFSI facility SSC(s), or any ISFSI facility monitoring as follows: Multi-Assembly Sealed Basket (MSB), MSB Transfer Cask (MTC), MTC Lifting Yoke, Ventilated Concrete Cask (VCC), Ventilated Storage Cask (VSC), VSC Transporter (VCST), ISFSI Storage Pad Facility, ISFSI Storage Pad Data/Communication Links, or PPCSIISFSI Continuous Temperature Monitoring System?
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Venfy SCR number on all pages Page 5 L Does the proposed activity involve IN ANY MANNER SSC(s) installed in the plant specifically added to support cask loading/unloading activities, as follows: Cask Dewatering System (CDW), Cask Reflood System (CR.F), or Hydrogen Monitoring System?
El 0] Does the proposed activity involve IN ANY MANNER SSC(s) needed for plant operation which are also used to support cask loading/unloading activities, as follows: Spent Fuel Pool (SFP), SFP Cooling and Filtration (SF),
Primary Auxiliary Building Ventilation System (VNPAB), Drumming Area Ventilation System (VNDRM),
RE-105 (SFP Low Range Monitor), RE-135 (SFP High Range Monitor), RE-221 (Drumming Area Vent Gas Monitor), RE-325 (Drumming Area Exhaust Low-Range Gas Monitor), PAB Crane, SFP Platform Bridge, Truck Access Area, or Decon Area?
El 0] Does the proposed activity involve a change to Point Beach CLB design criteria for external events such as earthquakes, tornadoes, high winds, flooding, etc.?
El 0] Does the activity involve plant heavy load requirements or procedures for areas of the plant used to support cask loading/unloading activities?
E] z Does the activity involve any potential for fire or explosion where casks are loaded, unloaded, transported or stored?
If ANY of the Part V questions are answered YES, then a full 10 CFR 72.48 screening is required and answers to the questions in Part VI and Part VII are to be provided. If ALL the questions in Part V are answered NO, then check Parts VI and VII as not applicable. Complete Part VIII to document the conclusion that no 10 CFR 72.48 evaluation is required.
PART VI (72.48) - DETERMINE IF THE CHANGE INVOLVES A ISFSI LICENSING BASIS DESIGN FUNCTION (If ALL the questions in Part V are NO then Part VI is El NOT APPLICABLE.)
"ompare the proposed activity to the relevant portions of the ISFSI licensing basis and answer the following questions:
YES NO QUESTION El 0] Does the proposed activity involve cask/ISFSI Safety Analyses or plant/cask/ISFSI structures, systems and components (SSCs) credited in the Safety Analyses?
El 0] Does the proposed activity involve plant, cask or ISFSI SSCs that support SSC(s) credited in the Safety Analyses?
E] 0 Does the proposed activity involve plant, cask or ISFSI SSCs whose function is relied upon for prevention of a radioactive release, OR whose failure could impact SSC(s) credited in the Safety Analyses?
El 0 Does the proposed activity involve caskfISFSI described SSCs or procedural controls that perform functions that are required by, or otherwise necessary to comply with, regulations, license conditions, CoC conditions, or orders?
El 0D Does the activity involve a method of evaluation described in the ISFSI licensing basis?
El 0] Is the activity a test or experiment?. (i.e., a non-passive activity which gathers data)
El 0] Does the activity exceed or potentially affect a cask design basis limitfor a fission product barrier(DBLFPB)?
(NOTE: If THIS questions is answered YES, a 10 CTR 72A8 Evaluation is required.)
If the answers to ALL of these questions are NO, mark Parts VII as not applicable, and document the 10 CFR 72.48 screening in the conclusion section (Part VIH).
If any of the above questions are marked YES, identify below the specific design function(s), method of evaluation(s) or DBLFPB(s) involved.
ART VII (72.48) - DETERMINE WHETHER THE ACTIVITY INVOLVES ADVERSE EFFECTS (NEI 96-07, Appendix B, Section B.4.2. 1)
(If ALL the questions in Part V or Part VI are answered NO, then Part VII is 0 NOT APPLICABLE.)
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on all pages Page 6
.nswer the following questions to determine if the activity has an adverse effect on a design function. Any YES answer means that a 10 CFR 72.48 Evaluation is required; EXCEPT where noted in Part VII.3.
VII.I Changes to the Facility or Procedures YES NO QUESTION El E] Does the activity adversely affect the designfunction of a plant, cask, or ISFSI SSC credited in safety analyses?
El 0l Does the activity adversely affect the method ofperforming or controlling the designfunction of a plant, cask, or ISFSI SSC credited in the safety analyses?
If any answer is YES a 10 CFR 72.48 Evaluation is required. If both answers are NO. describe the basis for the conclusion (attach additional discussion, as necessary):
VII.2 Changes to a Method of Evaluation (If the activity does not involve a method of evaluation, these questions are E] NOT APPLICABLE.)
YES NO QUESTION Ml El Does the activity use a revised or different method of evaluation for performing safety analyses than that described in a cask SAR?
El El Does the activity use a revised or different method of evaluation for evaluating SSCs credited in safety analyses than that described in a cask SAR?
If any answer is YES, a 10 CFR 72.48 Evaluation is required. If both answers are NO. describe the basis for the conclusion (attach additional discussion, as necessary):
VII.3 Tests or Experiments (If the activity is not a test or experiment, the questions in VII.3.a and VII.3.b are [] NOT APPLICABLE.)
- a. Answer these two questions first:
YES NO QUESTION El E] Is the proposed test or experiment bounded by other tests or experiments that are described in the cask ISFSI licensing basis?
El El Are the SSCs affected by the proposed test or experiment isolated from the cask(s) or ISFSI facility?
If the answer to both questions is NO, continue to VII.3.b. If the answer to EITHER question is YES, then briefly describe the basis.
- b. Answer these additional questions ONLY for tests or experiments which do not meet the criteria given in VII.3.a above.
If the answer to either question in VII.3.a is YES, then these three questions are [] NOT APPLICABLE:
Point Beach Nuclear Plant SCR 2002-0458 10 CFR 50.59/72.48 SCREENING (NEW RULE) Verify SCR number on 211pageS Page 7 YES NO QUESTION 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?
El [: Does the activity utilize or control a plant, cask or ISFSI facility SSC in a manner that is inconsistent with the analyses or descriptions in the ISFSI licensing basis?
I] [] Does the activity place the cask or ISFSI facility in a condition not previously evaluated or that could affect the capability of a plant, cask, or ISFSI SSC to perform its intended functions?
If any answer in VII.3.b is YES, a 10 CFR 72.48 Evaluation is required. If the answers are all NO, describe the basis for the conclusion (attach additional discussion as necessary):
PART VIII - DOCUMENT THE CONCLUSION OF THE 10 CFR 72.48 SCREENING Check all that apply:
A 10 CFR 72.48 Evaluation is [] required or Z NOT required. Obtain a screening number and provide the original to Records Management regardless of the conclusion of the 50.59 or 72.48 screening.
A VSC-24 cask Safety Analysis Report change is F1 required or Z NOT required. If a VSC-24 cask SAR change is required, then contact the Point Beach Dry Fuel Storage group supervisor.
A Regulatory Commitment (CLB Commitment Database) change is [] required or Z NOT required. If a Regulatory Commitment Change is required, initiate a commitment change per NP 5.1.7.
A change to the VSC-24 10 CFR 72.212 Site Evaluation Report is [I required or 0 NOT required. If a VSC-24 10 CFR 72.212 Site Evaluation Report change is required, then contact the Point Beach Dry Fuel Storage group supervisor.
-. 11 -
-POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 1 of 17 A. PURPOSE
- 1. This procedure provides guidance to identify the leaking steam generator, control the spread of contamination, perform a plant shutdown, isolate the leaking steam generator, and stop the leak by equalizing RCS and steam generator pressure.
- 2. This procedure is applicable when RCS hot leg temperature is greater than or equal to 3500F with SI accumulators in service.
B. SYMPTOMS OR ENTRY CONDITIONS
- 1. The following are symptoms of a steam generator tube leak:
a; Any of the following area-and prodess radiation monitors alert or high alarm:
o 1RE-215. Air ejector radiation o 1RE-219. Steam generator blowdown radiation o IRE-22-2. Steam generator blowdown tank radiation o IRE-231. Steam line "A" radiation o iRE-232. Steam line "B" radiation o RE-225. Combined Air ejector radiation
'b. Ris[ in' chai-ini pumpspe&d.
- c. Feedwater flow lowering with steam-generator level stable or rising.
d.-Unexpected steam generator level deviation alarms.
- e. Steam generator level rising with no change in auxiliary feedwater flow.
- f. Steam generator chemistry samples of abnormal activity or the abnormal presence of boron.
- 2. This procedure may be entered from the following:
"o AOP-lA Unit 1. REACTOR COOLANT LEAK "o PBF-2034. Control Room Shift Log Unit 1. when confirmed S/G tube leakage greater than 75 gpd or rate of change greater than 30 gpd within a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.
"o NP 3.2.4. Primary To Secondary Leak Rate Monitoring Program
POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL'OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 2 of 17 C. REFERENCES
- 1. NP 3.2.4. Primary To Secon'dary Leak Rate Monit6ring Program
- 3. Commitment Change Evaluation CCE 2001-009 to VPNPD-87-510 Letter -to NRC dated November 20. 1987 as revised by CCE #1999-008.
POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page-3 of 17 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTANED NOTES
" Foldout page shall be monitored throughout this procedure.
"*Due to Improved Technical Specifications (ITS) implementation, in order to allow use of this procedure prior to hnd after implementation. both the Custom and Improved Technical Specification information is shown, with the ITS information inbraces.
Example CTS info {ITS info)
I Check Safety Injection Not Required: Perform the following:
- PZR level - WITHIN 10% OF PROGRAM a. Manually trip reactor.
LEVEL
- b. Manually initiate.safety AND injection.
- RCS subcgoling - GREATER THAN730OF c. Manually initiate containment isolation.
1TI-970 I
1TI-971 1 d. Go to EOP-0 Unit 1. REACTOR TRIP OR SAFETY INJECTION.
- 2 Check Reactor Trip Not Required: .
- a. Check reactor - CRITICAL a. Go to Step 3.
- b. Check charging pump suction - b. Perform the following:
- ALIGNED TO VCT
- 1) Manually. trip reactor. *
- 2) Stabilize plant conditions *
- using EOPs while continuing *
- with this procedure. *
- 3) Go to EOP-0 Unit 1, REACTOR *
- TRIP OR SAFETY INJECTION. *
- 3 Check PZR Level - STABLE AT OR IF PZR level trending lower. THEN *
- TRENDING TO PROGRAM LEVEL . perform the following: *
- a. Control charging flow as *
- necessary to maintain PZR level. *
- b. IF.PZR level continues to lower. *
- THEN isolate letdown.
- POINT BEACH NUCLEAR PLANT AOP-3 Unit I:
ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 4 of 17 SiI STEP I ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED IF PZR pressure trending lower. THEN i
I 4 Check PZR Pressure - STABLE AT OR TRENDING TO DESIRED PRESSURE perform the following:
- a. Ensure spray valves shut.
"*Normal spray valves 1PCV-431A 1
"*Auxiliary spray valve 1CV-296 1
- b. Operate PZR~heaters as necessary to establish PZR pressure at program.
5 .Check Reactor Makeup Control: Initiate manual makeup as necessary to maintain VCT level between 17%
. Makeup sqt at proper boric acid and 56%. -
concentration
- Makeup armed and in auto 6 Notify DCS, Chemistry. And Implement Emergency Plan
- j
POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL-OPERATING PROCEDURE SAFETY RELATED Revision 4" 11/8/2001 STEAM GENERATOR TUBE LEAK Page 5 of 17 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED
- 7 Identify Leaking S/G:
- o Abnormal S/G sample results
- o Boric acid,
- 'o activity
- PH
- o High main-steam line radiation
- e 1RE-231 for S/GA *
- lRE-232 for S/G B 1
- o Unexpected S/G level deviation
- alarms
- o Unexpected rise in any S/G level
- o Isolate '/G sample valves one at a
- time and check blowdown activity *
- for trends
- - lMS-2083 for S/G A
- - IMS-2084 for S/G B 8 Determine Leak Rate:
o Direct Chemistry to perform CAMP 014. STEAM GENERATOR TUBE LEAK RATE CALCULATION o PBF-2034. Control Room Shift Log Unit 1 o 01-55. PRIMARY LEAK RATE CALCULATION
POINT BEACH NUCLEAR PLANT AOP-3 Unit I ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 6 of 17 STEP ACTION/EXPECTE RESPONSE L RESPONSE NOT OBTAINED NOTE Primary to secondary steam generator leakage in excess of 500 gpd in either steam generator will place Unit I in CTS 15.3.1.D.4 {ITS 3.4.13. RCS Operational Leakage).
9 Check Reactor Shutdown Required: Go to NP 3.2.4. PRIMARY TO SECONDARY LEAK RATE MONITORING PROGRAM.
Confirmed leakage - GREATER THAN OR EQUAL TO 75 GPD I!
CAUTION Action response time clock starts when the first RMS monitor indicates a tube leak has developed.
- 10 Determine.Action Response Based On *
- S/G Leakage:
SS/G LEAKAGE ACTION RESPONSE Greater than or Greater than Reduce power to less
- equal to 75 gpd AND or equal to than or equal .to 50%
- 30 gpd/hr in one hour AND*
Be in RSD (Mode 3)in
- next two hours
- Greater than or
- equal to 75 gpd Less than Be in HSD {Mode 31 in
- and sustained for AND 30 gpd/hr less than or equal
- greater than or to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
- equal to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
- Greater than or AND Less than Be in HSD [Mode 31 in equal to 150 gpd 30 gpd/hr less than or equal to
- six hours
- ***************************1**********************************
POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK "Page 7 of 17 I I p
- I STEP II ACTION/EXPECTED RESPONSE I RESPONSE NOT OBTAINED I 11 Place Unit In HSD (ITS MODE 31 o OP-3A. POWER OPERATION TO HOT SHUTDOWN OR o AOP-17A. RAPID POWER REDUCTION 12 Notify Chemistry Of Leak Rate And Rate Of Change 13 Monitor Leakage Every 15 Minutes 14 Direct Radiation Protection To Perform Exposure'And Contamination Evaluations 15 Check Leaking SIG - IDENTIEIED Perform the following:
- a. Continue with plant shutdown.
- b. WHEN leaking S/G identified.:THEN continue with Step 16.
S..... . ... . ... NOTE .. .....
" "St 16 throu-gh Ste-p" 26 align sys-is t o control the "spread of contkmination.
These steps may be performed in any order.
16 Adjust Affected S/G Atmospheric Steam Dump Controller To 1050 psig o 1HC-468 for S/G A OR o IHC-478 for S/G B
POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page.8 of 17 LIS 17 II ACTION/EXPECTED RESPONSE Isolate Blowdown On Affected Steam
"-I
! I . RESPONSE NOT OBTAINED I-I--F J
Generator o SIC A
"*IMS-5958
"*lMS-2042 OR o S/G B lMS-5959 I
e IMS-2045 18 Shut Affected Steam Generator Sample Isolation Valve o IMS-2083 for S/G A
- OR "o lIS-2084 for S/G B 19 Ensure Condensate Storage Tank Isolated From Condenser Hotwell:
- a. Ensure condenser reject level a. IF valve can NOT be manually shut. THN_*r-dly shum upsfteam isolation valve.
1CS-2130 I
ICS-112
- b. Ensure condenser reject control bypass valve shut
- ,ICS-113 20 Locally Align Low Pressure Trap Header To Condenser:
- a. Check condenser available a. Go to Step 21.
- c. Shut low pressure trap header inlet to atmospheric blowoff tank a lMS-165
POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 9 of 17 1
I -. III .i i ISTP ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I
21 Locally Isolate Building Heating Steam Supply:
- a. Check Unit 2 available to supply a. Perform the following:
heating steam
- 1) Secure boric acid evaporators.
- 2) Startup house heating boilers per 01-63. HEATING BOILER OPERATION.
- b. Place heating steam control valve selector switch to 9 o'clock (unlabeled) .position 22 Locally Shift Air Ejector After Condenser Drains To Hotwell:
- a. Open air ejector after condenser drain trap inlet
- IFD-125
- b. Open air ejector after condenser drain trap outlet IFD-126 1
c Shut a' i -cj-o F afteiiondensi drain bypass IlFD-124
- d. Shut air ejector after condenser drain to funnel IFD-124A 1
23 Isolate Turbine Building Sumps:
- IB52-116K for IP-64B
POINT BEACH NUCLEAR PLANT AOP-3 Unit I ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 10 of 17 i Ip n n STEI ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 24 Shut Known Steam Release Paths To Clean Side:
e Ensure affected MSIV Bypass Valve shut o IMS-234 for S/G A OR o IMS-236 for S/G B
- Locally ensure S/G Header Bypass Drains are shut and capped
"*IMS-123A
"*IMS-120A 9 Locally ensure upstream drains to blowdown tank are shut e IMS-229
- IMS-:239 S.Locally ensu re the following elbow vents are shilit:
- HP Turbine Steam Supply Inlet "PX-2625 I
- IMS-1OA
- lMS-10B
- HP Turbine Steam Supply PX-2024
"*lMS-15A
"*lMS-15B 25 Check Radwaste Steam Aligned To Perform the following:
Unit 2
- a. Locally align radwaste steam to Unit 2 per OI-18A, SHIFTING RADWASTE STEAM SUPPLIES.
- b. IF Unit 2 is shut down. THEN shut down strippers and blowdown evaporator and secure radwaste steam.
POINT-BEACH NUCLEAR PLANT AOP-3-Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK .Page 11 of 17 I IIN/RESPNSE I I
I RESPONSE NOT OBTAnED I
I NOTE Shutting the steam supply valve to the turbine driven AFW pump will place Unit 1 in CTS 15.3.4.A {ITS 3.7.5,Auxiliary Feedwater}.
26 Check If 1P-29 AFP/Radwaste Steam Isolation Valve From Affected Steam Generator Can Be Shut:
- a. Check either-of the following a. WHEN Step 25 is comiplete. THE.N do.-,
completed: Step 26.b. Continue with o Radwaste steam aligned to Step 27.
Unit 2
.ORR o Radwa"e steam is secured
- b. Locally shut 1P-29 AFP/radwaste steam isolation valve from affected S/G o IMS-235 for S/G A
... . .""- _'R . .
o IMS-237 for S/G B 27 Check Change Iii Reactor Power - LESS Direct Chemistry to sample RCS for THAN 15% IN ANY 1 HOUR Iodine within 2 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following the power change.
28 Check Reactor - SHUTDOWN .WHEN reactor shutdown complete. THEN continue with Step 29.
29 Shut Main Steam Isolation Valve For Affected S/G o lMS-2018 for S/G A OR o lMSI2017 for SIG B
POINT BEACH NUCLEAR PLANT. AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED
- Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 12 of 17 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED 30 Locally Shut Affected Main Steam Trap Isolation Valve o lMS-228 for SIG A OR o IHS-238.for S/G B -.
31 Borate The RCS: "
- a. Borate RCS to establish either of the following:
o Establish a 1%'shutdown margin.
(70F .to 3500F..-ARI. Xenon free) per Rod 9 Column 3 OR o Value$lpecified by Reactor Engineering
- b. Maintain PZR boron concentration within 50 ppm of RCS:
- 1) Energize all available PZR heaters
- c. Check required RCS and PZR boron c. DO NOT continue until boron concentration established concentratlon established.
POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED S Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 13 of 17 STP[ ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I 32 Initiate Plant Cooldown To Between 490'F And 5000F:
- a. Manually plot cooldown per 01-105, RCS HEATUP/COOLDOWN
, PLOTTING
- b. Establish a cooldown rate of less b. Establish a cooldown rate of leas than'1000F/hr by dumping steam to than 1000F/hr using atmospheric condenser steam dump on ilntact SIG.
- a. Maintain S/G levels between 60%"
and 75%
- d. Maintain PZR level at 30%:
- 1) Check RCS temperature less 1) WHEN RCS temperature less than than 540*F - 5400 F. THEN do Step 32.d.2.
Continue with Step 33.
- 2) In instrument rack 1C-107.
place Pressdfizer. Level Programmer in manual
- TC-401C
- 3) Adjust controller output to 30%
. 46 POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED 1- Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 14 of 17 iSTEPI ACTION/EXPECTED RESPONSE I I "RESPONSE NOT OBTAINED i 33 Depressurize RCS And Block SI:
- a. Check SI BLOCKED status light a. Go to Step 34.
-NOT LIT
- b. Check RCS temperature - BETWEEN b. OBSERVE CAUTION PRIOR TO STEP 32 490OF AND 500OF and return to St*e 32.
- c. Place PZR pressure controller in manual
- d. Place one spray valve in manual -d. Perform the following:
and idjust as necessary to establish depressurization rate .1) F letdown in servi eTHN of less than 100 psig per minute. establish a depressurization rate.of less-than 100 psig per minute using auxiliary spray.
- 2) IF letdown NOT in service.
THEN consult with plant-staff to determine alternate means of RCS depressurization.
o PORV o'PZR vents o Ambient losses
.9.. Energize all banks of PZR heaters
- f. Check RCS pressure - BETWEEN f. Return to Step 33.d.
- g. Check Lb PZR Press Block SI permissive light - LIT
- h. Block SI
- Train A - BLOCKED e Train B - BLOCKED
- 41 POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY. RELATED Revision 4 11/8/2001 STEAM GENERATOR TUBE LEAK Page 15 of 17 STPII ACTION/EXPECTED RESPONSE I I RESPONSE NOT OBTAINED I l
34 Continue To Depressurize RCS To
- Between 900 PSIG And 1100 PSIG:
a.. Establish a depressurization rate a. Perform the following:
of less than .100 psig per minute
.using normal spray and backup 1) LF letdown in service. THEN heaters establish a depressurization "rate of less than 100 psig per minute using auxiliary spray
.and backup-heaters.
- 2) JF letdown NOT in service.
THEN consult with plant staff to deteimine alternate means
-of RCS depressiirization.
o PORV o PZR vents o Ambient losses
- b. Check lvaking SIG pressure b. Go to Step 35.
BETWEEN 900 PSIG AND 1100 PSIG
- c. Reduce RCS pressure to equal leaking SIG pressure
- d. Go to Step 37 35- Continue Plint Cooldown To Between 370*F And 380 0 F:
- a. Manually plot cooldown per 01-105. RCS HEATUP/COOLDOWN PLOTTING
- b. Establish a cooldown rate of less b. Establish a cooldown rate of less than 1000F/hr by dumping steam to than 1000F/hr using atmospheric condenser steam dump on intact SIG.'
- c. Maintain SIG levels between 60%
and 75%
- d. Maintain PZR level at 30%
POINT BEACH NUCLEAR PLANT AOP-3 Unit 1 ABNORMAL OPERATING PROCEDURE SAFETY RELATED "Revision 4 11/8/2001
-STEAM GENERATOR TUBE LEAK Page 16 of 17 STEP ACTION/EXPECTED 'RESPONSE: I RESPONSE NOT OBTAINED i 36 Depressurize RCS To Match Leaking Steam Generator Pressure:
- a. Establish a-depressurizationrate" a. Perform the following:
of less than 100 psig per minute using normal spray and backup 1) IF.letdown in service.' THEN heaters -establish a depressurization rate of less than 100.psig per minute using auxiliary spray and backup heaters.*
- 2) IF letdown NOT.in service, THEN-consult with plant staff to determine alternate means of RCS depressurization.
o PORV o PZR vents o Ambient losses
- b. Reduce WS pressure to equal leakingS/G pressure 37 Check If SI Accumulators Should Be Isolated:
-a.Check RCS pressure - LESS THAN a. Go to Step 38.
1000 PSIG
- b. Rest6re power to both accumulator discharge HOVs a IB52-324F for lSI-841A
- IB52-424F for 1SI-841B
- c. Shut both accumulator discharge MOVs
"*ISI-841A. accumulator A
"*1SI-841B. accumulator 1B
- d. Remove power from both accumulator discharge MOVs
"*IB52-324F for lSI-841A
- POINT BEACH NUCLEAR PLANT- AOP-3 Unit I ABNORMAL OPERATING PROCEDURE SAFETY RELATED Revision 4 11/8/2001
. STEAM GENERATOR TUBE LEAK Page 17 of 17 STEP ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED.
- 38 Maintain Stable Plant Conditions ,
- RCS'pressure - EQUAL TO LEAKING.
- S/G PRESSURE *
- * . RCS temperature - STABLE *
, 39 Consult With Plant Management .To .
SDetermine Long Term Corrective Action
"-END
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