ML20198T461
| ML20198T461 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/22/1997 |
| From: | Jaudon J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Richard Anderson FLORIDA POWER CORP. |
| References | |
| NUDOCS 9711170003 | |
| Download: ML20198T461 (156) | |
Text
~.
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-October 22, 1997--
Florida Power Corporation
-Crystal River Energy Complex-Mr. Roy A._ Anderson (SA2A)
Sr. VP. Nuclear, Operations-ATTN:
Mgr.. Nuclear Licensing 15760 West Power Line Street-Crystal River. FL 34428-6708 s
SUBJECT:
MEETING
SUMMARY
CRYf;TALRIVER3ENGINEERINGANDRESTARTMEETINGS C
Dear Mr. Andersoni a
This refers to the meetings on Sep e er 25, 1997, at your Training Center.
The purpose of the meetings were ty m scuss the status of your actions.to resolve technical issues and your progress-towara readiness for restart.
It is-our opinion, that these meetings were beneficial, Enclosed is a' List of Attendees-and Florida Power Corporation (FPC) Handouts.
-The discussions-included the following topics: Configuration Document Integration. Control Complex. Habitability Envelope Update. Failure Modes and Effects Analysis. Apaendix "R" U)date. Emergency Diesel _ Generator Update.
FWP-7 -(Aux-FW Pump) Jpdate, and lorida Power Corporation-(FPC) Restart Progress Status, The NRC staff informed FPC at the meeting that there were a number of-
- questions on two licensing submittals.
The first was the small' break loss of coolant accident submittal because of its size and complexity.
The second was the submittal on boron precipitation for which there were questions of adequacy raised in its initial review. These submittals will require additional correspondence to resolve the questions prior to restart, in accordance with Section 2.790 of NRC's " Rules of Practice "Part 2.
Title 10 Code.of Federal Regulations, a co)y of this letter and its enclosures will be placed in the NRC Public Document Room.
Should you have any questions concerning this letter, please contact us.
Sincerely.
Orig signed by Johns P. Jaudon John P. Jaudon. Director
/
Division of Rcactor Safety
Enclosures:
- 1. List of Attendees
- 2. FPC Handouts Docket No. 50-302 License No. DPR ggg gg 9711570dCbY71022 0FFICIAL CODY PDR ADOCK 05000302 P
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- FPCi 2
cc-w/encls:
John P. Cowan. Vice President-Nuclear Production-(NA2E)
FPC. Crystal River Energy Complex 15760 West Power. Line Street -
Crystal River. FL 34428-6708
. C. G. Pardee. Director Nuclear Plant Operations (NA2C)
FPC.' Crystal River Energy Complex 15760 West Power Line Street
. Crystal River. FL 34428-6708 Robert E. Grazio. Director Nuclear' Regulatory Affairs (SA2A)
= FPC. Crystal River Energy Corcplex 15760 West Power Line Street
- Crystal River. FL 34428-6708
. James S. Baumstark. Director Quality Programs-(SA2C)
FPC. Crystal River. Energy Complex
- 15760 West Power Line Street Crystal River. FL-34428-6708 R. Alexander Glenn. Corporate Counsel Florida Power Corporation MAC - ASA P. O. Box 14042 St, Petersburg FL 33733-4042 Attorney General Department of Legal Affairs The Capitol Tallahassee. FL.32304 Bill Passetti Office of Radiation Control De3artment of Health and Rehabilitative Services 1317 Winewood Boulevard Tallahassee FL 32399-0700 Joe Myers. Director Division of Emergency Preparedness Department of Community Affairs 2740 Centerview Drive
- Tallahassee. FL 32399-2100
- cc w/ercls:
Continued see page 3 4
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FPC 3
cc w/encls:
Continued Chairman Board of County Commissioners Citrus County 110 N. Apopka Avenue Inverness. FL 34450-4245 Robert B. Borsum Framatome Technologies 1700 Rockville Pike. Suite 525 Rockville, MD 20852-1631 Distribution w/encls:
'K. Landis. RII L. Raghavan NRR R. Schin, RII P. Steiner, RII PUBLIC NRC Resident Inspector U.S. Nuclear Regulatory Commission 6745 N. Tallahassee Road Crystal River FL 34428 l
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LIST OF ATTENDEES Florida Power Corocration R. Anderson, Senior Vice President, Nuclear Operations K. Baker. Administration Manager. CDIP J. Baumstark.-Director Quality Programs S. Browr., Administrator Power Marketing. OUC R. Colthorpe, General Manager, NUS Information Services J. Cowan, Vice President. Nuclear Production.
R. Finnin, Resident Engineer, Framatoma C. Garver. Senior Account. OUC R. Grazio. Director. Nuclear Regulatory Affairs G, Halnon, Assistant Plant Director, Nuclear Safety M. Hendrix, Senior Public Relations Coordinator J. Holden, Site Director. Nuclear Engineering and Projects B. Hickle. Director. Restart P, Holmes-Ray, Senior Licensing Engineer E. Kane. Vice President. Engineering, Framatome D. Kunsemiller, Manager, Nuclear Licensing M. Marano. Director. Nuclear Site and Business Support J. Maseda, Manager, Nuclear Engineering Programs C. Pardee. Director. Nuclear Plant Operations A. Petrowsky. Manager. Nuclear Operations Engineering S. Powell, Principal Nuclear Licensing Engineer J. Ranalli, Manager. Configuration Management M. Rencheck, Director. Engineering D. Shelton, Member. NERC T. Taylor. Director. Nuclear Operations Training J. Terry, Manager. Nuclear Plant Technical Support M. VanSicklen, Chief. Nuclear Operator E. Vilade Public Relations Consultant K. Ward Principle Licensing Engineer F. Wreath, Consultant. Smith, Wreath & Associates Nuclear Reaulatory Commission S. Cahill. Senior Resident Inspector, Crystal River H. Christensen, Chief. Engineering Branch. Region II (RII)
K. Clark, Senior Public Affairs Officer, RII S. Flander.i. Project Manager Project Directorate II-1, Office of Nuclear Reactor Regulation (NRR)
F. Hebdon. Director, Project Directorate 11-1 NRR J. Jaudon. Director Division of Reactor Safety (DRS)
C. Julian, Technical Assistant to Director, Division of Reactor Safety (DRS)
K. Landis, Chief, Branch 3. Division of Reactor Projects (DRP)
S. Ninh, Project Engineer. DRP L. Raahavan Project Manager. Project Directorate 11-1. NRR 1
l Public I. James Staff Writer, St. Petersburg Times l
l ENCLOSURE 1 l
l
Xuclear Regulatory Commission Florida Power Corporation Technical Meeting September 25,1997
AGENDA i
+ Configuration Document Integration l
+ Control Complex Habitability Envelope
[
(CCHE} Lpdate L
+ Failure Modes anc. Effects Analysis T< MEA >'!
+ Appendix "R" E pdate
+ Emergency Diesel Generator L~pdate
+ FW:?-7 (Aux. FW Pump) L~pdate' e
e
/
I I
i i
4 i
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Configuration Document i
Integration l
Future Configuration Contro L i
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CDIP Organization Jan Schroeder Greg Halnon j
ga gene Producia Ass;stard Plant Director y
Engmeering Vce speosast CDLP Ltanager
-i Presk! ant Fram atome
. s - ru.ud 2..
a Frank Wrean System informate l
l l
l Coordemor Ken Baker Garren Het,b usii Van Schien
- *9 CDtPAdm:1 Documen Techncal C apter 14 Pete Mc mes Ray oanager / Ch to Wagig RescMen Atem Informaron Coordnator Parnae Hamandez an ormaron Canto s
Ciers Bad Leon &d System Informatcc j
Coorocator l
l Tom Artau l
Mars Lisemenn Rchard Grumtr BE N4etsen l
Senor uechancas in*egrata toregratkm Pete Woannge, 48r Sam
- del Engerer Superwsor Supennsor i
S V $
U500Prepaar D r s i
i Fremator: e Wrder/ integrator Wrneracteystor Je seas u,1a,c, se_ %s P., Tem,e,m
- 8* *"
USOD Presser
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I I
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Joe Cudd Dean Andes Rabn Hayes T
Framew.e
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Wruerantegra:or Wrsernnse;rger J67 tiammond
- Dreams, US P
I Wes Johnson Don Lews Steve sommers USCO Prepaar uany Paece
~
Wrae Daily precursor screening > Minutes of all restart activities j i > Review of NRC correspondence l > Screening of modifications + CDIP Data aase o aerations j l l > Extensive and versatile database l + Auto-reporting feature l + Over 1866 Items collected + Conununication of issues between CDIP Gr6hps l !O w. i swm.cww-= ~ . g.g.. ..,.c. g.g 3..; g..w g l r W. ;,k <. AC AG.. ;. Ab.' 4 i = %...; u ,.l l =l Microsoft Access -[ Main Menu] I-0 l Mdit L/iew flecords Window Uelp 0 d$E E bbl? * .X. l%1 @ lMll$1 s! 9.Y Ul b l$'l O N'? [ we W2"'"'" f mi&SuaramMi[ ~
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_ I N l - --c g l l SEARCl1ES { REPORTS MISC fl]INiil fj 'l ilG Macros fl I TRG System 11-2 0 4 i l SISCO Rev 25 1120 Name 94 i j 2 4 4 p l Sys/Ilesolved 7j System l'rogres-input Data Form i ^ l t D) I Document # Franks form l SRR System I 4 ~~ G i Mall # [ i """""'"j*"'"'" i nevision 25 c i J 7, 1 t _ ~., - OUkt ACCl*SS [ l FormVew l l t
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f ,1466: 'l.. O l.,JJotinrire,q;.; Resdved Hold 1 Responsbie Manager Assigned to: D0YEL , S. ,, Closed fJo 2; . Assigned Due: 1 + Rev 25 ' TRG membec ' Jeff Hammond l2l e ' ' Cause 8HBI:D12 H 2; "O' i JJtEm _ Comments: . l' ; ! Resokton Need EDBD. FSAR ard ITS Bases chnges to resdve EDBD CHAf JGES RECIEVED WIT H COP / OF SA/USQD Fir 1AL ISSUE OF MAR ' Comments: DUE 10-7 97 >at: 8 Comments: 7 Integrator l S. K.8.yons @ Date l l - 1466 l 0 not inwher s rtie' Doc 6/6 FSAR Chapter: 9 Syn.m l ,DH SO g EDBD FSAR Sector - 94 Sub sys: l .l Doc Type: Revise..; l d ActionComplete l No W SISCO CompleteJNo .l Commentt I ..!p I4l>l y 14] 4 l Record l1 jof 1 l > ]>ll l -- 14j 4 l Record l6 jot 263 l>l>lj ~
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0 l System Information Management i + Coordinators are aoint of contact for aL information in and out of CDIP i > Keepers of the status of each issue > Oversight of resolutions + Hard copy files for c:.osure packages + Performance Indicators 9" i l l 1 (O32Ug5IcO3 QoOC3mU* 3nOcm$53 ',g'mO* l ) t, I 3 D i2C3 $,o+;@o gExm h c 4og~~ C3c* o% xm!.,,t r 8 88 \\ i mg* j 38: &8l n8l n8 l 88 : o8 m8 l m8 ; 1k#E7@ a8 : w8 : wE e$_8$.! o 7 7 0 7 s 7.d 7 7 7 7 7 7 7 7 7 7 7 7 7 7 7 9 9 3 9 9 9 9 9 9 9 9 9 9 9 9 9 9 0 9 / / / / / / / / / / / / / / / / / / / / 6 3 0 7 1 7 2 8 1 7 3 9 5 9 5 1 7 3 9 3 / / 1 1 2 2 0 / / 1 1 2 2 / 2 3 / 1 1 2 2 9 / / / / 6. / / 7 / / / / 8 8 / / / / 6 6 7 7 7 7 8 8 8 8 9 9 9 9 1 t 1 t i l i i l R-20 Determination (Rev 24) l + Rev 24 will contain the Significant l TechnicalIssues based on: l > Confirmatory Action Letter Items + Design Margin Improvements l + Extent of Condition Reviews + Emerging Items as screened by Restart Criteria + License Amendments j > Resolutions Received l > Review of Remaining Items i I. ll Technical Resolution Group + Receive Open Items from SISCOs and FSAR Writers + Work with Plant Departments to Facilitate Resolution of Open Items + Track anc. Prioritize All Open Items taat affect CDIP + Return Copies of Resolutions to the SISCOs and FSAR Writers > Receive inout from above when answer not sufficient - multi.ayers of checxs r 1 l m u o 8 8 8 8 8/18/97 /- 8/20/97 - 8/22/97 - o I 8/26/97 - o 8/28/97 - 9/1/97 - O 9/3/97 -- Q
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> Pilot for "Documentum"
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> Incorporate TSCRNs into ITS Bases
>Information source :for all departments PI
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+ Upc. ate the sa ~ety ana yses to clearly represent CR-3's accident mitigation s
strategy j
+ Framatome technical personnel directly l
involved with pro ect
> Senior Framatome Management 1
> Partnership in the resu ts i
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+ Framatome review of each safety anal sis
,.-,--,-----r-
Framatome Review Content
+ For Each Accident in the Safety Ana~ ysis:
> Key references
> Event Frequency Classification
> Analysis of Record
> Acceptance Criteria
> Analysis inputs to tae accident
> Operationalissues and operator actions
> Applicable computer codes
> Current status of FSAR vs Review L_
Chapter 14 Group
+ Com:parison to the EOPs
> Identification of operator actions which support the accic.ent mitigation strategy
> Ensure that appropriate actions are add.ressed in Chapter 14 and. the EOPs
+ Assist in resolution of identifiec. items which affect CDIP
+ Upc ate the Chauter 14 text to strp; port Restart Objective ND kW
CDIP (R-20) Chapter 14 Open items 35
-eOpen iteris 30 -
Oct. 6th Workoff 3-15-k 10-5 5-4 1
0 18-Aug 21-Aug 2-Sep S-Sep 1%Sep 22-Sep 29-Sep SOct 13-Oct O
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Revision 24 of the FSAR
+ Rev 24 will meet objectives of CAL
+ Will contain the significant issues existing and discoverec. this outage
> As screenec by Restart Criteria
+ Xew Issues
> Screen in as restart issue-incluc ec.
> Xot screened in-dependent on timing of the resolution tw
aassm9 After Restart
+ Revision 25 ~ay CDIP 6 months after S/L
+ Continued resolution of PCs ~ay line departments
> All Revision 25 :?Cs have been re-screened
+ Future revisions by new process
> Xew process will ae in : place by t:ae time Revision 25 is issuec.
> CDIP will be maintained until new. : process up and working i
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l Configuration Document Integration
+ CDIP Basic P:ailosop:ay
> Integration j
> Synchronization j
+ Restart Issue OP-8, NOD-56 Basic Philosop;ay
> Identification
> Control
> Verification and Validation
> Status Accounting
+ Converge the two into one program i
1 l
l Configuration Document Control FSAR I
i I
NOD-56 Configuration Documents Design Tech Basis Spec Document l
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Confguration Change CCN-1C I
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,i Summary After CDIP
+ NOD-56 will be revised to include I
integration and sync:aronization
+ Caange processes will:aave at a minimum a cross check of:
+ Corrective Action Tracking System will l
be used to initiate, track, and validate i
configuration changes
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ENGINEERING AGENDA l
l
+ Control Com: plex Habita aility Envelope (CCHE) Epdate
+ Failure Modes anc Effects Analysis (FMEA1
+ Ap;pendix "R" L~ pdate
+ Emergency Diesel Generator L'pdate
+ FWP-7 (Aux. FW Pump) L'adate i
M'
m i
Control Complex Habitability Envelope Demonstrate t:aat the actualin-leakage into the Contro:. Complex Ha aitability Envelope is 'aelow ::ae a:iowa ~ ale limit I
\\
- 1. - - ---
I
% AHD-1D Ewator Equp.
%g AHD-1 Room f71 f
s ska Abandoned
-l-TOXIC l l AHF-18A x
! gas I Abandoned HEPA
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l I
i
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%[/
CHARCOAL Abandoned AHD-2 %
AHD-24
_AHD-25
_g _ [g_.
t AHD-2D %
[,T I
NORMAL l
FILTER p
AHF-21 A/B AHF-17A AHD-26 AHD-27 AHF-19 A/B AHD-3 AHF-17B
(
l D
NORMAL i
FILTER
)
j g RM-A5 l j
blEPA S
j J
j CHARCOAL l
4==
HVAC Equipment Room AHF-183
==
164 Ft j
i l
C=s Control Room c-145 Ft j
4===
Cable Spreading Room
<==
134 Ft l
G--.
EFIC Rooms. Control Rod Dove Room. & ES 450 volt Sutch Gear Rooms 4
124 Ft AH 2D g
Battery Rooms, invener Rooms & ES 4160 volt Sutch Gear Rooms l
%g AHD-12 108 Ft 1
POST MOD RECIRCULATION MODE
'5
5 i
t
Control Complex
~
l Habitability Envelope
+
Assess and Seal Penetrations i
95% Complete, Completion on 10/1/97
+ Perform Information Test Test Complete Install Redundant Dampers in Each Isolation Path
+
Modification Package Issued, Completion on 11/15/97
+
Revise and Consolidate Calculations Preliminary Results Available, Final Due 9/26/97 Demonstrate Total Envelope In-Leakage is within Specifications
+
Test to Begin 10/8/97 i
- $NJ u
i Control Complex l
Habitability Envelope Assure Protection of the Control Room Personnel During Postulated Events, Within the Existing Design and l
Licensing Basis l
__________________________._____________-.___._-____.___,..-.____,-----_7
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Failure Modes and Effects l
Analysis (F MEA)
PURPOSE
+ Original 1E FMEA Did Not Fully Evaluate Component Failures on Rec undant Safety Systems
+ Previously Self Ic entified DC Concerns:
> HPI Flow Instrumentation
> ZFW Flow Control Valve and Instrumentation
> SW Flow Suaply to RB Fan Coolers
+ FPC Committed to Perform DC FMEA to the Extent That All " Safety Significant Problems" Id.entified i
?!
D Failure Modes and Effects Analysis (F MEA)
SCOPE
+ 21 DC Relatec Systems Were Identified
+ Scope Eraanded to Include 120V Vital AC Buses (Due to Loss of Inverters)
> Engineered Safeguards i
> Emergency Feedwater Control
> R.G.1.97 & Misc. Class 1E Control Comaonents l
i
?:
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Failure Modes and Effects Analysis (FMEA)
METHODOLOGY
+ Methodology / Basis Document Developed And Approved By FPC
> Parsons Power / Cy Crane & Assoc.
> Framatome j
+ Guidance From IEEE 352-1987 l
1
+ DC Failures Considered Up To And Including The Class 1E Batteries l
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i "A" BATTERY "B" BATTERY i
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l PANEL PANEL i
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@iFUSE RELAY
- $ FUSE RELAY I
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i Valve PEMP Valve PUMP 1
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i Failure Modes and Effects l
Analysis (FMEA)
RESULTS i
+ Approximately 10,000 Individual Checks and Reviews Were Completed (>8,000 Man Hours) l
+ 17 Precursor Cards Were Generated Which Were Evaluated for Operability / Reportability
+ 16 Were Determined to Be Within Design Basis 1
i
,)
Failure Modes and Effects Analysis (FMEA)
The Remaining PC Was Determinec. to be Reportable
+
+ LER 97-0021
> The Loss of the 250/125V "A" Class 1E Battery With LOCA/ LOOP Causes a Standing ES Train j
"B" Actuation That Cannot Be Bypassed i
i j
> Concern Now Identified and Tracked As Restart f
I Issue D-7-A j
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Failure Modes and Effects l
Analysis (F MEA)
Results Have Been Reviewed by:
> Parsons Power (Independent Verification)
> Failure Prevention International (Third l
Party)
> Operations
> System Engineering
> Design Engineering
> Licensing i
a,
Ap.pendix "R" Update Revalidation of FHA & Fire Study
> Completed 9/17/97 i
> Aparoximately 30 Items Have Been Identified As CAQ
> 8 LERs / Supplements Have Been j
Submitted l
> Items required For App. "R" Comphance Are Included In Restart Issues
- l!
i Appendix "R" Update i
l l
RCP Oil Collection System
> 8 Maintenance WRs Written. All Are m i
Progress j
> Modification Package Issued 9/8/97
> Field Construction Schec.ulec: to Be Completed 11/4/97 l
> Remote Oil Fill Line Exemption Submittec l
to XRC Staff On 9/5/97 l
Appendix "R" Update Modifications i
> List of Issues and Modifications Provic.ed to XRC (N o CSQs)
> Field Constructiort Underway l
> Design Work Continuing I
L l
,+ *,
-=,-w-w
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D raded Seal Materials 97-06-07-01 RCP Lube Oil Collection 97-06-16-01 Sho3?FatiS4 11/4/97 CT Secondary Protection 97-02-11-01 11/4/97 I
FCN-1 (Structural) 11/4/97
\\
FCN - 2 ( 50.59) 11/4/97 I
RSP Hot Shorts 97-03-02-01 11/15/97 FCN-1 (Separations)
/31/97 FCN-2 (New USQD)
FCN-3 ("A" Buss Items) 9/29197 FCN-4 ("B" Train /tems) 9/23/97 9/24/97 10/31/97 FCN-6 ("A" Train /tems) 10/20/97 10/25/97 10/26/97 11/15/97 t
IN 92-18 MOVs 97-06-13-01 9/24/97 9/28/97 9/29/97 11/22/97 I
FCN-1 10/3/97 10/31/97 11/1/97 12/2/97 j
Emergency Lighting 97-09-04-01 10/17/97 10/27/97 10/28/97 12/2/97 j
FCN-1 (SR) 10/17/97 10/27/97 10/28/97 12/2/97 i
}
FCN-2 (NSR) 0/17/97 11/8/97 11/9/97 12/2/97 AHF-22B, C, D Cables (EDG) 97-05-19-01 10/17/97 97-05-17-02 12/1/97 EDG Det e Riser 97-06-11-01 Thermo-La Circuit Rercutes 97-05-17-01
/23/97 9/24/97 11/12/97
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n _-,, - _ -n,- ---- - -, - - -, - -, - - - - _ _ _ _ _ _ _ _ _ _ _ -
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Appendix "R" Update l
Cold Shutdown Calculations / Capability.
> Ca:Lculation Completed (F-97-0010, Rev.1) i 1
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> Information Sent to XRC Staff 8/29/97 i
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> Field Valic ation of AP-990 (Shutdown l
Outside Control Room) Complete 9 23/97
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Appendix "R" Update l
IX 92-18
> Re-evaluation Completed.16 Hot Shutdown Valves
. being Moc.ified.
> Design / Field Work Package Issued 9/28/97 I
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> Completion Scheduled 11/22/97
> Review of New Plant Modifications continuing 1
EDG Loading Issue Part of a Taree-tiered Strategy
> EDG Power U:pgrad e
> EDG Load. 0;ptimization
> EDG Yargin Verification i
- i'l
Diesel Generator L pgrade 3500 kW 30 Min. Rating New 200 Hr.
3400 kW Rating 3250 kW 200 Hr. Rating New 2000 Hr.
3200 kW Rating 1
3000 kW 2000 Hr. Rating 2850 kW Contin. Rating Sept. '96 Sept. '97
EDG Load Optimization
+ kW Meters
+ Load Management
> P-T-L Control Hancies
> EFP-1 Trip Defeat Switch
+-Load Removal / Reduction
> ASV-204 C_25kW)
> DOP-2A/23 C.0kW)
Diesel Generator Upgrade l-l 30 Min. Rating i
3400 kTV i
3300 kW l
200 Hr. RatinE 3250 kW (ITS Upper SR Limit) i 3200 kW 1
ITS Lower SR Limit j
3100 kW I
3000 kW 2000 Hr. Rating i
)
20kW Contin. Rating l
Sept. '96 Sept. '97 f
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EDG Related Modifications Description NLmber Design lWorkPackage[5eid5Sork Tumover
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6.4 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDFRATIONS i
6.
4.1 INTRODUCTION
The use of normally operating equipment for Engineered Safeguards (ES) functions and the location of some of this equipment outside the reactor building requires that consideration be-given to direct radiation levels from the fluids circulating in these systems and the leakage from these systems after fission products have accumulated.
The shielding for components of the ES Systems are designed to meet the following objectives in the event of a Design Basis Loss-of-Coolant-Accident (LOCA),
a.
To provide protection for personnel to perform all operations necessary for mitigation of the consequences of the accident.
b.
To provide sufficient accessibility in all areas around the station to permit safe continued operation of the required equipment.
6.4.2
SUMMARY
OF POST-ACCIDENT RECIRCULATION Following a Reactor Coolant System (RCS) rupture, flow is initiated in the High Pressure Injection (HPI) and Low Pressure Injection (LPI) Systems from the Borated Water Storage Tank (BWST) to the reactor vessel. Flow is also initiated When by the Reactor Building Spray (BS) System to the building spray headers.
the BWST inventory reaches its mimimum operating level, recirculation from the Reactor Building emergency sump is initiated by the operators for both the LPI flow and the BS flow.
The post-accident recirculation flow paths include all piping and equipment external to the reactor building, as.shown in Figure 6-1, up to the valves leading to the BWST.
The net positive suction head available for the LPI and reactor building spray pumps during post-accident recirculation is summarized in Table 6-12.
6-29 (Rev. 23)
/
It was conservatively assumed that at the time of recirculation the available
,/
suction head was comprised only of static elevation head result %g from terminating BWST drawdown at the highest level (i.e., lowest Reactor Building Emergency sump level).
A design review and evaluation was performed to compara the calculated fluid frictional losses with the reactor building sump flow test data, relative to decay heat removal and reactor building spray pumps suction from the reactor building sump.
The results of this evaluation were submitted to the NRC (J.
Stolz) by Florida Power Corporation (J. T.
Rodgers) in a report entitled
" Evaluation of Calculated System Head Losses and Reactor Building Sump Test System Head Losses Relative to Crystal River Unit 3" by letter dated October 11, 1976.
6.4.3 BASES OF LEAKAGE ESTIMATE While the Reactor Auxiliary Systems involved in the recirculation complex are closed to the auxiliary building atmosphere, leakage is possible through component flanges, seals, instrumentation, and valves.
The leakage sources considered are:
a.
Valves 1.
Disc leakage when valve is on recirculation system boundary
~
2.
Stem leakage 3.
Flanges c.
Pump shaft seals While leakage rates have been assumed for these sources, maintenance and periodic testing of these systems will preclude all but a small percentage of the assumed amounts.
With the exception of the boundary valve discs, all of the potential leakage paths may be examined during periodic tests or normal operation.
The boundary valve disc leakage is retained in the cther closed systems and, therefore, will not be released to the auxiliary building.
6-30 (Rev. 23)
o
'i.
Detection of coolant leakage into the auxiliary building from the ES Systems during post-accident recirculation operation is expected to come from three main j
sources. These sources are:-
s.
Decay heat pumps b.
Reactor building spray pumps c.
Makeup and purification pumps Both the decay heat pumps and the spray pumps would leak into the decay heat sump while the makeup and purification pumps would leak into the auxiliary building sump via the compartment floor drains. -
The decay heat removal pumps and the reactor building spray pumps are located at a base elevation of 75 feet 0 inches in two individual cubicles.
Each cubicle contains one decay heat removal pump and one reactor building spray pump. These cubicles are constructed of steel reinforced concrete of which all joints hsve water stoss. The only antrance provided is through an equipment hatch at floor level 95 feet 0 inches. The equipme.it hatch slab is in place during normal plant.
l operation.
l Each cebicle contains a sump and sump pump. The sump s are capable of 30 gpa at 54 feet of head to accommodate normal equipment leak e.
The pump motors are NEMA Design B, totally enclosed water cooled and are set upon a concrete bed i
above base floor elevatica 75 feet 0 inches. Both decay heat pit sumps, 3A and 38, are 4 feet in depth with a 48 cubic foot capacity and have a pneumatic level located on the radioactive waste panel indicator (LI-32 and LI-34, respectively) level range of 0 to 4 feet.
(numbered AM and AN, respectively) with a Both sumps 3A and 38 have level switches (LS-133 and LS-134 respectively) controlling the i
{
corresponding sump pump. 8'oth level switches are such that they will stop the pumps as the water elevation reaches 72 feet 6 inches, start the pumps when the water elevation reaches 73 feet 0 inches, and sound a high level alarm when the r
~
water elevation reaches 74 feet 0 inches.
The auxiliary building sump is 8 feet in depth and contains two pumps, WDP-4A and 48, each with a head of 31 feet and a capacity of 150 gpe.
The sutsp has a pneumatic level indicator, LI-29, located on the radioactive waste panel, numbered AP, and has a range of 0 to 8 feet. The sump also has a level switch, LS-132, controlling the sump pump.
The level switch stops the pump when tha water reaches an elevation of 88 feet 6 inches, starts the pump at water elevation 89 feet 6 inches, and sounds a high level alars when the water in the sump reaches an elevation of 92 feet 6 inches.
j 1
4 6-31 i
I
l I
j Any leakage into these sumps would be observed on the radioactive waste panel by the level indicators as described above.
6.4.4 DESIGN BASIS LEAXAGE Values for the assumed design basis leakage and for the total leakage quantities for the various components which would be in contact with the recirculated fluid appear in Table 6-11.
Section 14.2.2.5.10 presents an analysis of the effects associated with the l release of the radioactive fluid.
G 6-32 (Rev. 13)
% 9 @dPd.
I 1*
CR3 PROPOSED FSAR. REV. 24 CHAPTER 6.4 7n,c ENGINEERED SAFEGUARDS LEAKAG -
~~
2 AND RADIATION CONSIDERATIONS,ex THIS IS A PROPOSED FSAR CIIANGE
};" K Tills FILE REFLECTS THOSE CHANGES CURRENTLY BEING MA,DE THIE tNFORMATION MAY CHANGE ON A DAILY BASIS IT IS FOR USE AS AN INFORMATION SOURCE THE INFORMATION CONTAINED HEREIN MAY NOT BE APPROVED ANY CHANGES THAT ARE APPROVED WILL BE ANNOTATED AS SUCH IN THAT SECTION.
PLEASE OPEN AND READ THE README. DOC FILE LOCATED IN THIS DIRECTORY. THIS FILE IS PROVIDED TO ASSURE UNDERSTANDING OF THE CHANGE PROCESS USED.
For questions or comments about this FSAR chapter, please contact Sharon Lyons at extension 6038 EOF Trailer 29 NU47 l
7:
,q i
\\1 6.4 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS
. ~. ' 'q
- L J j! (
6.
4.1 INTRODUCTION
[
The use of normally operating equipment for Engineered Safeguards (ES) functions and the location of some of this equipment outside the rEeictor bDuilding $Hlisxul_ l requires thr.1 consideration be given to direct radiation levels from the fluids circulating in these syrste;ns and the leakage from these systems after fission }m> ducts have accumulated.
m.
The shielding for components of the ES Systemstsxul are designed to meet the following objectives in the event of a Design Basir. Loss of _ Coolant _ Accident (LOCA)lsxul.
- a. To provide protection for personnel to perform all operations necessary for mitigation of the conseqttences of the aEciderit,
- b. To provide sufficient accessibility in cll areas around the statiAn to permit safe continued operation of the required equipment.
6.4.2 SUMM ARY OF POST-ACCIDENT RECIRCULATION Following a Reactor Coolant $Q System-(RG)tskul rupture, flow is initiated in the High Pressure Injection (HPI) and Low Pressure Injection (LPI) Systemstsxts) from the Borated Water Storage Tank (BWST) to the reactor vessel. Flow is also initiated by the Reactor-Building Spray (BS) Systemisxul to the building-sprayHStsxn1 headers. When the BWST inventory reaches its miamimumisrul operating level, recirculation from the Reactor-BuildingIstt91 eEmergency sSumpistuoiis initiated by the operators for both the LPI flow and the BS flow. The post-accident recirculation flow paths include all piping and equipment external to the reactor-buildingED,tsxtui as shown in Figure 6-1, up to j the valves leading to the BWST.
The net positive suction head available for the LPI and reactor 4Huilding sSpray pEumps (BSPshsru21 during post-accident recirculation is summarized in Table 6-12.
It was conservatively assumed that at t$1e time of recirculation the available suction head was comprised only of static elevation head resulting from terminating BWST drawdown at the highest level (i.e., lowest Reactor-Buildingisxual Emergency l
Ssumptsru41 level),
l A design review and evaluation was performed to compare the calculated fluid l
frictional losses with the EDreactor4uildingtsxust sSump flow test data, relative to l SU
3,.
i j
i decay heat removal and reactor-buikling-spray-pumpsDSEisxusi suction from the reactor-buildingED sSumptsru71. The results of this evaluation were submitted to the Nuclear Reculatorv Commission (NRQlsKUBs G. Stolz) by Florida Power Corporation IEEenu*10. T. Rodgers) in a report entitled " Evaluation of Calculated System Head Losse, and Reactor Building Sump Test System Head Losses Relative to Crystal River Unit 3" by lette: dated October 11,1976.
ya 6.4.3 BASES OF LEAKAGE ESTIMATE While the RIeactor Aguxiliary Ssystemstsxuol involved in the recirculation codiplex are closed to the aAuxiliary bHuildingisguil atmosphere, leakage is possible through component flanges, seals, instrumentation, and valves.
The leakage sources considered are:
- a. Valves
- 1. Dise leakage when valve is on recirculation system boundar 3
- 2. Stem leakage
- 3. Bonnet flangeleakage
- b. Flanges
- c. Pump shaft seals While leakage rates have been assumed for these sources, maintenance and periodic testing of these systems will preclude all but a small percentage of the assumed amounts. With the exception of the boundary valve discs, all of the potential leakage paths may be examined during periodic tests or normal operation. The boun.dary valve disc leakage is retained in the other closed systems and, therefore, will not be released to the aAuxiliary bDuildingistu21 Detection of coolant leakage into the aAuxiliary bHuildingisKu31 from the ES Ssystemstsxu41 during post-accident recirculation operation is expected to come from three main sources. These sources are:
- a. Decay hHeat Removal pEumps (DHPGsxust
- b. Reactor-bDuilding sSpray pEumps (BSPshsxu61
- c. Makeup and pEurification pEumps (MUPs]Isxu71 Both the decay 4wat-pumpsDHPsisxusi and the spray-pumpsDSEsisxu91 would leak into the dRecay hHeat sSump,istuoi while the makeup +nd-purification-pumpsMUEsisxuil would leak into the aAuxiliary bDuilding sSumpistu21 via the compartment floor drains.
The decay---hea t-rem oval-pu m ps Dilhisnimi and the reactor-buil spray pumpslLTitsKr.ul are located at a base elevation of 75 feet 0 inches in two
'idual cubicles.
Each cubicle contoins one dany-imat-removal-pumpIlllfisxi and one r eado r-bu ild i ng-s p ra y-pu m pHSI'isxtuj.
These cubicles are constructed of steel-reinforced concretersxt271 of which all joints have water stops. The-only-eEntrance h11.gprovidal each cubicle is through individual equipment hatches at floor 5 feet 0 inches. ILithThe equipment hatch slab has five removable hatch covers wit ivin place during normal plant operation.IsKimi Each cubicle contains a sump and sump pump. The sump pumps are capable of 30 gpm at 54 feet of head to accommodate normal equipment leakage. The pump motors are NEMA Design B, totally enclosed water cook'd and are set upon a c bed above base floor elevation 75 feet 0 inches. Both decay heat pit sumps,3A a 3B, re 4 feet in depth with a 48.-cubic-foot capacityisxta91 and have a-pneu evel indicator 3[sniaol(LI-32 and LI-34, respectively) located on the radioactive waste panel (numbered AM and AN, respectively) with a level range of 0 to 4 feet. Both sumps 3A and 3B have level switches (1S-133 and LS-134, respectivelyistiaij) cor gthel corresponding sump pump. Both level switches are such that they will stop umps as the water elevation reaches 72 feet 6 iriches, start the pumps when e water elevation reaches 73 feet 0 inches, and sound a highrlevel alarmisxia21 wh te water l elevation reaches 74 feet 0 inches.
The at\\.uxiliary bDuilding sSumptsxtail is 8 feet in depth and contains t
- ips, WDP-4A and 4B, each with a head of 5031 feet and a capacity of125150 gp:
xia41 The sump has a pneumatic level indicator, LI-29, located on the radioactive s
- 132, controlling the sump pump. The level switch stops the pump when the water reaches an elevation of 88 feet 6 inches, starts the pump at water elevation 89 feet 6 '
s,and sounds a high-level alarmisxtast when the water in the sump reaches an ele
'o of 92 l feet 6 inches.
Any leakage into these sumps would be observed on the radioactive waste panel by the level indicators as described above.
6.4.4 DESIGN BASIS LEAKAGE h3hmdhd.Vyalues for the assumed design basis leakage and for the total leakage quantities for the various components which would be in contact with the recirculated fluid appear in Table 6-11. These values rentesent the minimum values to
')
analyses related to releaws. The estimated total is not automaticallv adiustet witl ch channe_in1he_munb.uflemnene n t sisxt asi l
l Section 14.2.2.5.10 presents an analysis of the effects associated with the release of the radioactive fluid.
l
Annotations Chapter 6.4 Engineered Safeguards Leakage and Radiation Considerations Page: 2
[SKL1).
n/a Editorial Changed " reactor building" to a proper noun and included the acronym "RB."
Page: 2
[SKL2) n/a Editorial Changed " System" to a common noun.
Page: 2
[SKL3) n/a Editorial Deleted hyphens for style consistency in accordance with (iaw) Administrative Instructions 400G, Florida Power Corporation, Crystal River Unit 3, Procedure Writing Reference Manual (Al-400G), Rev. O.
Page: 2
[SKL4]
n/a Editorial Replaced the acronym "RCS" with "RC" and changed " System" to a common noun.
Page: 2
[SKL5]
n/a Editorial Changed " Systems" to a common noun, Page: 2
[SKL6) n/a Editorial Changed " Reactor Building Spray" to " Building Spray" for consistercy of information and changed " System" to a common noun.
Page: 2
{SKL7]
n/a Editorial Replaced the words " building spray" with the previous!) defmed acronym "IlS."
ii
s Page: 2 -
[SKL8) n/a Editorial Corrected misspelled word (rdinimum).
Page: 2-
[SKL9]
n/a Editorial Replaced " Reactor Building" with the previously defined acronym "RB."
Page: 2
[SKL10] n/a Editorial Changed " emergency sump" to a proper noun.
Page: 2
[SKLil] n/a Editorial Replaced " Reactor Building" with the previously defined acronym "RB."
Page: 2
[SKL12] n/a Editorial Changed " Reactor Building Spray" to " Building Spray" for consistency of information, changed " pumps" to a proper noun, and included the acronym "BSPs."
Page: 2
[SKL13] n/a Editorial Replaced " Reactor Building" with the previously defined acronym "RB."
Page: 2
[SKLl4] n/a Editorial Changed " sump" to a proper noun.
Page: 2
[SKL15] n/a Editorial Replaced " reactor building" with the previously defined acronym "RB" and changed
" sump" to a proper noun.
l l
1 Page: 3
[SKLl6] n/a Editorial-Replaced " reactor building spray pumps" with the previously defined acronym "bSP."
- Page: 3
[SKLl7) n/a Editorial Replaced " reactor building" with the previously defined acronym "RB" and changed
" sump" to a proper noun.
Page: 3 s
[SKL18] n/a Editorial Added the words " Nuclear Regulatory Commission" to define the acronyra "NRC."
Page: 3
[SKL19] n/a Editorial Added the acronym "FPC."
Page: 3
[SKL20] n/a Editorial l
l Changed ' Reactor Auxiliary Systems" to a common noun.
Page: 3
[SKL21] n/a Editorial Changed " auxiliary building" to a proper noun.
Page: 3
[SKL22] n/a Editorial Changed " auxiliary building" to a proper noun.
i Page: 3
[SKL23] n/a Editorial Changed " auxiliary building" to a proper noun.
Page: 3
[SKL24] n/a Editorial Changed " System" to a common noun.
Q (J
Page: 3
[SKL25] n/a Editorial Replaced " Decay heat pumps" with " Decay Heat Removal Pumps" and added the acronym "DHPs."
Page: 3
[SKL26] n/a Editorial Changed " Reactor building spray pumps" to " Building Spray Pumps" and added the acronyut "BSPs."
Page: 3
[SKL27] n/a Editorial Changed " Makeup and purification pumps" to a proper noun and added the acronym "MUPs."
Page: 3
[SKL28] n/a Editorial Replaced " decay heat pumps" with the previously defined acronym "DHPs."
Page: 3
[SKL29] n/a Editorial Replaced " spray pumps" with the previously defined acronym "BSPs."
Page: 3
[SKL30] n/a Editorial Changed " decay heat sump" to a proper noun.
Page: 3
[SKL31] n/a Editorial Replaced " makeup and purification pumps" with the previously defined acronym "MUPs."
Page: 3
[SKL32] n/a Editorial Changed " auxiliary building sump" to a proper noun, Page: 4
[SKL33] n/a Editorial Replaced " decay heat removal pumps" with the previously defined acronym "DHPs."
.)
t
)
Page: 4 --
[SKL34] ? n/a Editorial
_c
- Replaced " reactor building spray pumps" with the previously defined acronym "BSPs."
Page: 4
[SKL35] n/a
. Editorial -
ILplaced " decay heat removal pump" with the previously defined acronym "DHP."
Page: 4
_- [SKL36] n/a Editorial Replaced " reactor building spray pumps" with the previously defined acronym "BSP."
Page: 4
[SKL37] n/a Editorial Included hyphen between " steel" and " reinforced" because they modify concrete.
Page: 4
[SKL38] PC97-2225 (MAR 79-05-77)
FSAR6.4-R24-xxx "S AR Section 6.4.3 states that the only entrance to the decay heat pits is through an equipment hatch at floor 95'-0" and the equipment hatch is in place during normal operation. The entrance to the decay heat pits is open all the time. No documentation can be found as to what prompted the decision to leave a decay heat pit hatch (one of five for each pit) removed during normal operation contrary to the information currently in the FSAR. Discussions with Operations indicated that there are normal operator rounds which require entrance into the pit. An evaluation was performed to determine if l
operation of the plant as described above is acceptable, and it was determined that it is acceptable to operate the plant with one of the five hatches removed in each pit.
Also reference IOC NOE97-1026, PC97-2373, and PC97-3128.
Page: 4
[SKL39] n/a Editorial Included hyphen between "48," " cubic," and " foot" because they modify capacity.
l Page: 4 l
[SKL40) n/a Editorial Corrected subject verb disagreement.
i i
,,L Page: 4
- [SKL41) n/a Editorial Included comma to offset "respectively."
Page: 4
[SKL42) n/a Editorial Included a hyphen between "high" and " level" because they modify alarm.
Page: 4
[SKL43] n/a Editorial Changed " auxiliary building sump" to a proper noun.
Page: 4
[SKL44) PC97-2373 (MAR G-90-08-02-01, MAR G-90-08-02.02)
FSAR6.4-R24-xxx Sump pumps for the Reactor Building Sump, WDP-2A & -2B, and the auxiliary Building Sump, WDP-4A & -4B were replaced by MARS G-90-08-02-01 and G-90 02-02, respectively. Calculation M91-1012 is the calculation of record that specified the design flow and 1.ead requirements for these replacement pumps. The identical model pump was ordered for both applications and exceeds / bounds the flowrate and head requirements of the above calculation. These changes will correct the above specified documents so they will be consistent with the design values determined in calculation M91 1012. In addition, FSAR Section 6.4.3 is being revised because it includes a duplicate error noted in PC97-3128. Also reference PC97-3128 and PC97-2225.
Page: 4
[SKL45] n/a Editorial Included a hyphen between "l'igh" and " level" because they modify alarm.
Page: 4
, [SKL46] MAR 97-03-03-01, Rev. O FSAR6.4-R24-xxx These changes are to clarify the meaning and use of Table 6-11. The conclusions of the USQD do not depend on this change. The change is being made to clarify the use of the leakage estimates in the ITS 5.6.2.4 leakage comp.!iance program (CP-149) and in the radiological dose assessment. That is, the leakage compliance program and dose assessment use the FSAR Table 6-11 result as an initial estimate of total leakage from which to cerive the FSAR 14.2.2.5.10.2 analytic limit, which is then used in the radiological dose assessment (Calculation 1-86-0003). The FSAR change describes this purpose and notes that the estimate of Table 6-11, therefore, needs not be revised due to m
m m-
changes in the number of components that may exist on the system boundary. Also Reference NOE97-0842.
O
6.4 ENGINEERED SAFEGUARDS LEAKAGE AND RADIATION CONSIDERATIONS 6.
4.1 INTRODUCTION
The use of normally operating equipment for Engineered Safeguards (ES) functions and the location of some of this equipment outside the Reactor Building (RB) requires l that consideration be given to direct radiation levels from the fluids circulating in these systems and the leakage from these systems after fission products have accumulated.
The shielding for components of the ES systems are designed to meet the following objectives in the event of a Design Basis Loss of Coolant Accident (LOCA).
- a. To provide protection for personnel to perform all operations necessary for mitigation of the consequences of the accident.
- b. To prc, vide sufficient accessibility in all areas around the station to permit safe continued operation of the required equipment.
6.4.2 SUMM ARY OF POST-AC.CIDENT RECIRCULATION Following a Reactor Coolant (RC) system rupture, flow is initiated in the High Pressure Injection (HPI) and Low Pressure Injection (LPI) systems from the Borated Water Storage Tank (BWST) to the reactor vessel. Flow is also initiated by the Building Spray (BS) system to the BS headers. When the BWST inventory reaches its minimum operating level, recirculation from the RB Emergency Sump is initiated by the operators for both the LPI flow and the BS flow. The post-accident recirculation flow paths include all piping and equipment external to the RB, as shown in Figure 6-1, up to the l valves leading to the BWST.
The net positive suction head available for the LPI and Building Spray Pumps (BSPs) during post accident recirculation is summarized in Table 6-12.
It was conservatively assumed that at the time of recirculation the available suction head was r.omprised only of static elevation head resulting from terminating BWST drawdotvn at the highest level (i.e., lowest RB Emergency Sump level).
A design review and evaluation was performed to compare the calculated fluid frictional losses with the RB Sump flow test data, relative to decay heat removal and BSP suction from the RB Sump. The results of this evaluation were rubmitted to the Nuclear Regulatory Commission (NRC) (J. Stolz) by Florida Power Cortcration (FPC)
(J. T. Rodgers) in a report entitled " Evaluation of Calculated System Head Losses and
Reactor Building Sump Test System Head Losses Relative to Crystal River Unit 3" by letter dated October 11,1976.
(
6.4.3 BASES OF LEAKAGE ESTIMATE 4
While the reactor auxiliary systems involved in the recirculation complex are closed to the Auxiliar/ Building atmosphere, leakage is possible through component flanges, seals, instrumentation, and valves.
The leakage sources :onsidered are:
- a. Valves
- 1. Disc '4cakage when valve is on recirculation system boundary
- 2. Stem leakage
- b. Flariges
- c. Pump shaft seals While leakage rates have been assumed for these sources, maintenance and periodic testing of these systems will preclude all but a small percentage of the assumed amounts. With the exception of the boundary valve discs, all of the potential leakage paths may be examined during periodic tests or normal operation. The boundary valve disc leakage is retained in the other closed systems and, therefore, will not be released to the Auxiliary Building.
Detection of coolant leakage into the Auxiliary Building from the ES systems during post accident recirculation operation is expected to come from three main sources.
l These sources are:
- a. Decay Heat Removal Pumps (DHPs)
- b. Building Spray Pumps (BSPs)
- c. Makeup and Purification Pumps (MUPs)
Both the DHPs and the BSPs would leak into the Decay Heat Sump, while the hRJPs would leak into the Auxiliary Building Sump via the compartment floor drains.
The DHPs and the BSPs are located at a base elevation of 75 feet 0 inches in two l
individual cubicles. Each cubicle contains one DHP and one BSP. These cubicles are l
constructed of steel reirdorced concrete of which all joints have water stops. Entrance I
into each cubicle is through individual equipment hatches at floor level 95 feet 0 inches.
Each equipment hatch slab has five removable hatch covers with one open during j
normal plant operation.
1 (j
l
Each cubicle contains a sump and sump pump. The sump pumps are capable of 30 gpm at 54 feet of head to accommodate normal equipment leakage. The pump motors are NEMA Design B, totally enclosed water cooled and are set upon a concrete bed above base floor elevation 75 feet 0 inches. Both decay heat pit sumps,3A and 3B, are 4 feet in depth with a 48-cubic-foot capacity and have pneumatic level indicators (LI-32 l and LI-34, respectively) located on the radioactive waste panel (numbered AM and AN, respectively) with a level range of 0 to 4 feet. Both sumps 3A and 3B have level switches (LS-133 and 15134, respectively) controlling the corresponding sump pump. l Both level switches are such that they will stop the pumps as the water elevation reaches 72 fat 6 inches, start the pumps when the water elevation reaches 73 feet 0 inches, and sc,und a high-level alarm when the water elevation reaches 74 feet 0 inches.
The Auxiliary Building Sump is 8 feet in depth and contains two pumps, WDP-4A and 4B, each with a head of 50 feet and a capacity of 125 gym. The sump has a pneumatic level indicator, LI 29, located on the radioactive waste panel, numbered AP, and has a range of 0 to 8 feet. The sump also has a level switch, LS-132, controlling the sump pump. The level switch stops the pump when the water reaches an elevation of 88 feet 6 inches, starts the pump at water elevation 89 feet 6 inches, and sounds a high-level l alarm when the water in the sump reaches an elevation of 92 feet 6 inches.
Any leakage into these sumps would be observed on the radioactive waste panel by the level indicators as described above.
6.4.4 DESIGN BASIS LEAKAGE Estimated values for the assumed design basis leakage and for the total leakage quantities for the various components which would be in contact with the recirculated fluid appear in Table 6-11. These values represent the minimum values to be used in l
analyses related to releases. The estimated total is not automstically adjusted with each l
change in the number of components.
Section 14.2.2.5.10 presents an analysis of the effects associated with the release of the radioactive fluid.
l l
l e-
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FLORIDA POWER CORPORATION CRYSTAL RIVER - UNIT 3 b?S FPC RESTART PROGRESS l
September 25,1997
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5 AGENDA FPC CR-3 l
Opening Remarks J.P.Cowan Restart Status / CAL Status J.P.Cowan Operations Status / Update C. G. Pardee Technical Issues M. W. Hencheck CDIP / Corrective Action J. J. Holden Regulatory Submittal Status R. E. Grazio Restart Readiness / CH-3 Metrics B. J. Hickle j
l Concluding Statements J.P.Cowan 9/25/97 2
Current Status
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Secondary Side Vacuum Evolution Diesel Generator Work l
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9/25/97 3
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Objectives FPC CR-3
- 1. Frequent Open Communications
- 2. Resolution of Confirmatory Action Letter Issues
- 3. Close Book on First 20 Years of Operation
- 4. Invest for the Next 20 years of Operation
- 5. Integrate Startup Activities Into Long Term Improvement Plan i
Do What Is Right For The Long Haul, Not Just What is Required For Startup 9/25/97 l
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~
Operations Performance i
- Operations Organization Responding WeII to New Challenges 9/25/97 24 E
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Technical Issues FPC CR-3 Control Complex Habitability Envelope Failure Modes and Effects Analysis
=
Appendix 'R' Diesel Generator FWP-7 9/25/97 25 m m,,s
N":
Control Complex P
Habitability Envelope FPC Assess and Seal Penetrations CR-3
- 95% Complete, Completion on 10/1/97 Perform Information Test
- Test Complete Install Redundant Dampers in Each Isolation Path
- Modification Package issued, Completion on 11/15/97 Revise and Consolidate Calculations
- Preliminary Results Available, Final Due 9/26/97 Demonstrate Total Envelope In-Leakage is within Specifications 8/ [,/87
- Test to Begin 10/8/97
E-Failure Modes and Effects Analysis (FMEA)
FPC CR-3 j
~
SCOPE t
21 DC Related Systems Were Identified c
l l
Scope Expanded to include 120V Vital AC Buses (Due to Loss of Inverters) l J
r 1
~.
9/25/97 27 j
i
.i
Appendix 'R' Revalidation of FHA and l
Fire Study Cold Shutdown Calculations / Capability Information Notice 92-18
~
Modifications
~
~
9/25/97
I Em Appendix "R" Modifications l
~
l 2
W ork FPC oescription Number Design l Package ! Field Work ! Turnover CR-3 Degradeo seal Materials 97-06-07-01 ISSUED ISSUED b WORKING r?)
10/30/97 RCP Lube Oil Collection 97-06-16-01 ISSUED Shop Fab Shop Fab 11/4/97 11/4/97 CT Secondary Protection 97-02-11-01 ISSUED ISSUED (WORKING 3 7
FCN-1 (Structural) i ISSUED ISSUED iWORKING ?
11/4/97 FCN - 2 ( 50.59) f ISSUED ISSUED dh0NK'INGd 11/4/97 RSP Hot Shorts
- 97-03-02-01 ISSUED ISSUED WORKING;l 11/15/97
~
ISSUED [EWORKINGN Complete
$ORKING 10/31/97 FCN-1 (Separations)!
ISSUED ISSUED I
~
FCN-2'(New USQD) E ISSUED
~
ISSUED hkORKING][~~i0/3 9/29/97 FCN-3 ("A ~ Bhss Items) i ISSUED
+.
~ ~~ FCN-4 ("B" Train llems) [ ~
ISSUED 9/23/97 l 9/24/97
~
~ FCN-6 ("A" Train itein s) I
' -' ~ ~ ~
10/20/97 [~10/25/97 i 10/26/97 T ~1T/15/97 DHV-3, 4 Ca ble Se pa ra tion 97-02-18-01 ISSUED ISSUED (WORK 1NGg 12/2/97 i 97-02-18-02 ISSUED ISSUED #WO.RK!NGin 12/2/97 IN 92-18 MOVs j 97-06-13-01 9/24/97 9/28/97 i 9/29/97 l
11/22/97 FCN-1
'10/3/97 10/31/97 I
~11/1/97' f
12/2/97 i
12/2/97 10/28/97 Emergency Lighting 97-09-04-01 10/17/97 l 10/27/97
~10/28/97' ) ~12/2/97 FCN-1 (SR)
~f 10/17/97 I10/27/97
12I2/97
~
~
FCN-2 (NSR)
~!
~ ~~
10/17/97 ! ~~11/8/97~ ~~11/9/97 ' ~ ~ ~
97-05-19-01 ISS UED ISSUED pWORKINGJ 10/17/97 AHF-22B, C, D Cables (EDG) ;
971 5-17-02' ISSUED ISSUED kWORKINGS ~~12/1/97 -
0 ~
~~
~
~
~
EDG Deluge Riser 97-06-11-01 ISSUED ISSUED Complete Com plete 29
l l
l Diesel Generator Margin FPC 3500 kW CR-3 30 Min. Rating 3400 kW 3300 kW 200 Hr. Rating
.e 3250 kW, (ITS Upper SR Limit) n l
3200'kW 3159 kW Auto Connect Load
' " ^ *sw"a
' MARGIN l
ITS Lower SR Limit
. 3060 kW, 3100 kW 3000 kW 2000 Hr. Rating 2850 kW l
Contin. Rating 9/25/97 Sept. '96 I
Sept. '97 30 l
1 E
EDG Re ated Modifications i
FPC CR-3 Description Number ! ~ Design jWork' Package] Field Workl Turnover 1 lEDG Power Upgrade
~
~
j 96-10-05-01 issued Issued Complete
'11/10/97{
l lDOP-2A/2B Mod ~
' ~ 196~1237-01 ~
lasued issued Complete 9/30/97; l
ASV-204 Mod' ~
~~196-11-01-01 ~
issued
. Issued Compiete ~10/1/97j
! Kilowatt Meter Repiscement ' ~ ~ ~I96-03-12-01 ~
lasued issued Complete
~10/28/37l l
lEDG~ Standby Keep warmSpsts196-07-15-01~
Issued Issued Complete ~~ 9/30/97l ISW / RW PTL Contro! Handles ~ l97-04-02-01~
Issued issued Complete 9/30/97) l lEFP-1 Trip D fsat SwitrX ~ ~~ ~ 197-04-0141-Issued issued EWorkingW10/22/97!
{
(3Workihg]s] ~
lEDG Room Ventilation Mods ~ ~ ~ ~~ l97-0443-02 ~
sWorking T '10/10/97l l
~
issued issued
}
lEDG AHF-22 Cables '
97-05'19-01 ~
lasued Issued '
lEDG AHF-22 Structural
~
~ l97-05-17-02 lasued issued
$Wo' king
~~12/1/97l r
lEDG~ Radiator Mods (Mech)~~]97-0515-01^lasued l
p5WorkingM[W6tkingE;
~10/25/97j
[fiWortiingD Working >
10/25/97j
}
lEDG Radiator Mods ~(Ele ~ct)~ ~ ~~ 197-05'15 Issued
~
[EDG'Protsctive'Relaps 197-08-0141TWorking 1'O i'il97 ]
YO3/97 11/10/97j
^
lEDG Radiator Exhaust '
l97-08-04-01 pdrking}; ---
- -..L Schbduling ; Scheduling'
~
10/13/97 ~~
i i
l t
l 9/25/97 l
31 t
1
i FWP-7 Diesel Power Modification
~
i FPC CR-3
- Modification Field Complete
-;:l@
- Increase Defense-in-Depth ze Maximize FWP-7 Independence i
Minimize Operator Burden
~
l 9/25/97 j
l 32 l
t
i 2/
Extent of Condition 5b FPC CR-3 l
Reasonable Assurance That CR-3 Will Operate Within Its Design and Licensing Bases CH-3 Can Mitigate All Design Basis Accidents Modifications System Configuration and Focused Readiness Document Review Review Integration j
of Programs Program Project j
9/25/97 l
l 33 i
n.
Attribute Evaluation i
3:.w)LP i
l i
FPC J
l CR-3 Attribute-Checks 16044-
~
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l Work Controls Restart Process issue l
j
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9/25/97 1149 375 34 t
i
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Corrective Action Program
$,rNM e
FPC CR-3 i
- CARB Effectiveness
-Ensuring Quality of Root Causes
- Reduction of CAP open items
- Decreasing trend
- New CAP Information System i
- Available 10/1/97 9/25/97
{
35 l
i
Quality of Root Causes FPC CR-3 20,
A l
18 -
EXCELLENT Goal = 16 or Gmater y
16
-x 4
$14 7
MEETS
~
7 STANDARD
$ 12 -
g s.
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6-NEEDS O
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=
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Apr May Jun Jul Aug Sep Oct Nov Dec 1997 9/25/97 36 l
Open Precursor Cards i
"l$ =e% r 1
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^
CR-3 iOld System
- Total
-+-Initiated
[23 Grade B i
iGrade C m Grade D i
Open j
i 2500 -
i 2000 -
ae N
l o 1500 -
l D
4 O 1000 -
E Z
y 500 -
i 3
O I I I. E E, li E
,Em,n e, re m_,
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i 8/8 8/15 8/22 8/29 915 9/12 9/19 9/26 10/3 9/25/97 1997 37 1
i
Regulatory Submittal
~-
i
$r34#
Status w
E FPC f
CR-3
{
Total License Amendments 47
~
i i
j FPC Actions Completed 37 FPC Remaining Actions 10
(
+
j I
9/25/97 l
38 i
Regulatory Readiness License Amendments and 1
&I[$
Tech Speo Submittals j
FPC CR-3
~
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j
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20 -A
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Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 9/25/97 39
"C.
Regulatory Submittal
$@g Status 3
Total Potential USQs 19 Resolved by FPC 12
~
Submitted by FPC 3
f Remaining Submittals 4
9/25/97 40
i
=* H Regulatory Submittal i
6%%F Status
,ms FPC CR-3 l
Other Resta:rt Submittals 40 l
l Completed by FPC 28 i
Remaining 12
~
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9/25/97 j
41 i
Regulatory Readiness Other NRC Restart Submittals FPC m Additions" CR-3 25 -
--*-Goal = 0
~ *'* #
20 -
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0-3 Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 1997 9/25/97 42
l l
"""* V '
Resolution of Submittal
&%mgP Review Issues l
FPC FPC Actions Support Prompt Identification and Resolution
- FPC Licensing Staff Presence in Washington 4
l
- Status Checks With NRC Project Manager
- Quick FPC Turnaround on NRC Information Requests
- Prompt FPC Response Meetings With Staff 9/25/97 43
l FPC Ensures NRC Remains m7 Cognizant of Plans
$;y,j$_.
l s
l FPC j
CR-3 Technical Briefings i
i I
- Frequent Status Updates Communicate Emerging Issues
(,
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9/25/97 l
44
i 71 Regulatory Readiness 1
Y-Total 1997 Commitments FPC CR-3 200 -
^
j 180 -
s_
'x m 160 -
N y
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"N 140 -
g y
120 -
N
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j 100 -
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Action Plan m
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Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 1997 9/25/97 45
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- Systeen Discovery Complete I
O tc.opeae -
Review Of Operating Esperience O (Ce=We*o l
Power Ascension Program Developed ~
l
- 6 tComp <es Perseemet Fqr Augmented Startup Organiastien Selected i
O (Camp.64 Special Testtag - Secondary Systems Commenced
~6 gc.mp, ) -
?[fr' Operating Procedures Resised O
! Functional Tuting Scope & Schedule Freren I
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I Startup Se vence Commenced I
CAL Items Complete I
h I
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l Plant Ready For Ifeatup h
i l
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l l
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0 g
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l i
Results from Secondary System Startup FPC I
Operator Performance Was Very ca-3
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Very Few Equipment Problems j
Acceptable System Chemistry Established t
PMT Program Weaknesses identified j
Productive Dry Run Using AI-256 9/25/97 48
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{
Restart issue Closure 0
MCAP 11 Restart item Closure 0
r Self-identified Precursors 60 % or Greater Self-Identified Violations 33 % or Greater i
System Readiness Corrective Maintenance 200 Open Work Requests or Less Temporary Modifications 5 or Less l
Requests for Engineering Assistance 200 Open or Less Restart Modifications 0
j Systems Remaining to be 105 Systems Accepted l
l Accepted by Operations l
'~
Operations Readiness l
Control Board Deficiencies 10 or Less j
Operator Work Arounds 7 or Less Operating Procedures 25 Comments or Less
[
l Emergency Operating Procedures AII E0Ps Issued l
l 9/25/97 Abnormal Procedures 15 Comments or Less j
So
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CR-3 Performance Indicators Goal at Startup.
Regulatory Readiness l
I License Ame.ndments and Tech Spec l
Submittals AII Submittals Made Other NRC Restart Submittals All Submittals Made Total 1997 Commitments Complete c
Submittal Quality 5.5 or Greater l
{
Communications Readiness i
Employee Surveys Improving Results l
Community Leaders Meet Prior to Startup State & Federal Officials / Staff Meet Prior to Startup Opportunity to Tour Plant All Elected Officials and Staff i
Invited to Tour i
Open Communication with Medic Editorial Boards, Intesviews, and Plant Tours Conducted for Those who Routinely Cover CR-3 i
a 9/25/97 o1 l
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I Organizational Readiness E_.
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200 -
j I
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Organizational Readiness
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Organizational Readiness Self-Identified Violations l
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FPC CR-3 100 -
C"3 National (Average) 90 -
Goal = 33% or Greater u g 80 -
~ Actual (Average) 3 3m 70 -
Q E 60 -
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Total Number of Open Precursors FPC j
l CR-3 i
i 2000 -
E
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x l
E 1500 -
s E
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1000 -
i W.PCs Opened l
E 500 -
C"IPCs Closed j
- PCs Total I
~PCs Restart g,
Action Plan i
Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 9/25/97 56
System Readiness Eg--
Corrective Maintenance FPC CR-3 1000 -
900 -
e l
800 -
E 700 -
z N
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600 -
f h
500 -
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g 300 -
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a x
j j
100 -
Actual 0
Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 1
1997 9/25/97 57 I
i l
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FPC CR-3 Goal = 5 or Less Prior to Heat Up 10 -
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Requests for Engineering Assistance (REAs)
FPC CR-3 800 -
700 -
y 600 -
g
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- Actual i
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0 Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 9/25/97 l
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l System Readiness i
m gy Restart Modifications g-v-gj l
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+ Actual t5m 50 -
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.8 i i
l2 20 -
z 10 -
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Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 1997 9/25/97 60
l i
l l
Operations Readiness Control Board Deficiencies i
FPC CR-3 i
i 25 -
- Goal = 10 or Less at Start Up i
- Actual l
l
'5 g 20 -
2: ;3 c
c l
.g 15 -
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61 i
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Operations Readiness c-Operator Work Arounds FPC CR-3 Goal = 7 or Less Prior to Heat Up o
40 -
--o Workoff 35 -
Actuai S
30 -
@E E g 25 -
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Ns
$x 15 -
g o
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5-0 Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec f
9/25/97 1997 62 i
Regulatory Readiness
=
1 i
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~
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~ouality index
-*- Goal j
g 7-x i
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j Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec 1997 i
9/25/97 63 i
l Communications Readiness FPC FPC Community Meetings:
- Crystal River June 19,1997 Held
- Inverness August 12,1997 Held
- Beverly Hills October 9,1997 Scheduled
- Inglis/Yankeetown October 1997 Planned
- Crystal River November 1997 Planned Speakers Bureau Established s-Local Officials Continue to be Briefed l
and Tour Plant
~
l 9/25/97 64
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-w SEPTEMBER SUBMITTALS - Completed l
STAWS Description Draft Review Resolved /
Projected Submitted Date ADV - No USO Information Letter
/
/
9/02/97 Proprietary and Non-proprietary Version of LBLOCA FT1 Analysis
/
/
9/02/97 Appendix R - Exemption for Lube Oil Fill Remote Fill Line
/
/
9/05/97 Generic Letter 96 Summary Report
/
/
9/05/97 LTOPs - Response to Request for Additional Information - 12 questions
/
/
9/05/97 Letdown Line Rupture USO
/
9/09/97 Boron Precipitation
/
/
9/12/97 EDG Protective Trips USO
/
/
9/12/97 LTOPs Supplemental Information
/
/
9/12/97 EOPs 3,4,8 and AP770 Request for Additional Information
/
/
9/17/97 HPl Special Test - No USO Information Letter
/
/
9/17/97 Appendix R - Hot Shorts
/
/
/
/
9/24/97 H OAVTISE PT Ste 9l74
p
=~ 4 USQs Status / Resolution
.N s
1 TOTAL POTENTIAL USQs 19 4
RESOLVED 12 i
SUBMITTED 3
1 REMAINING 4
.l 2
n oneuso stat es I
4 Y-
c-USQs YN2W Status / Resolution N
brue Resch e.
Status EDG Fan Upgrade - fans under capacity USO
'o be submitted Submitted 8/26 Letdown Line Rupture - mar sal operator action in place USO - preparation of submittal ongoing Submitted 9/9 of automatic EDG Protective Trips -installation of overvoltage trips USQ - design in progress USOD to follow, USQ remains Submitted 9/12 could result in trip of EDG valid and submittal required EOP operator actions Review ongoing Open - due 9/30 AHF-1C - this modification makes control logic changes Not a USQ for installation, however, use of it wi!! be due Open - due 9/30 to allow use of AHF-1C by providing interlocim with its to reliability of circuitry being mod:fied and use of MCC-3AB transfer switch operator actions. Modification will be made but not tumed over until LAR approved.
Boron Precipitation Mitigation USO -implementation of Hot Leg Injection Open - due 9/30 RB Cooler /SW System Temperature Review ongoing - potential USO Open H 1DAVt1USQ PEV We f.
5 USQs Y b--
wd- '*
-~
Status / Resolution
-d v.
Issue Resolution Status Control Complex (Issues)
Current!y no USO identifed - test planned Resolved HPl Cavrtating Venturi Flow Test FPC determined that attemate approach would be developed Resolved (fo!!ow-up for insta!!ation in Refueling Outage 11 R letter)
Main Steam Line Break Not a USO Resolved RWP Cyclone Separator Fouling - strainers installed upstream Not a USO - new strainer tested and no clogging resulted Resolved might have clogged resutting in a loss of RWPs HPl Pump Recirc. Line - MAR 96-11-02-01, proposes to insta!!
Not a USQ - actons identifed for operator acton were Resolved an attemate manual recirculaton !ine from HPI discharge conc!uded to be equrvalent to existing actons header and the RB (operator action)
MUV-31 (Pressunzer Level Control Valve) control logc Not a USQ - design change is like for like replacement Resolved response during transient conditons - replacement of some parts HPI Nozzle Cycles - planned change in methodology reflected Not a USQ - change not implemented; submitted with FSAR Resolved in FSAR increissed allowable thermal cycles from 11 to 80 R23 letter, will correct FSAR back to 11 from 80 and maintain approved methodology AHFL - 1/2 testing not in compliance with FSAR Not a USO - FPC will comply with testing requirements Resolved MUV-27, - MAR 97-02-17-01, HPl autoclosure, add automate Not a USO - acton is an aid to the operator, no credrt for Resolved closure upon receipt of 1500# RCS low pressure signal actons in DBA ana!ysts; this is reflected in SBLOCA submittal Remote shutdown panel fuses as onginally intended to be Not a USO - revised design Resolved installed would not extinguish MCR frghts on fa!!ure CT protectors additon could reduce availability of bus Not a USQ - overall bus avatlabtitty determined no discemible Resolved impact ADV capacity does not meet FSAR values (301.000 lb/hr vs.
No USQ - ADVs had overcapacity and can perform functon Resolved 419.700 lb/hr) with existing intemals; acton is to clanfy FSAR informaton and resolve ambiguity regarding ADV actual capacity
^ !
NRC RESTART LICENSING SUBMITTALS 9/24 S7 i
r I
luut hubmittal Date
{
i I
Restart License Amendments l
1.
Small13reak LOCA Analysis TSCRN 210 Complete 6/14/97 i
TSCRN 210 (SBLOCA) Supplement I (Fuel Storage, Appendix D) 9/25/97 (N)(R) a.
2.
Decay llent Flow instrument Removal IJR 220 Complete 6/26/97 Supplemental Response (re: 13oron Precipitation 9/12/97 Letter) 10/3/97 (N) a.
3.
Appendix R Thermo Lag Resolution Schedule (Letter to Address Complete 7/26/97 (N)
Exemption Rea vst) 4.
Appendix R (111.0 - Reactor Coolant Pump Oil Cooling) Potential Exemption 10/8/97 (N) a.
Withdrawal / Revision - When "A" Pump is in Compliance i
b.
Exemption Request for Lube Oil Remote Fill Line Complete 9/5/97 (N) 5.
Low Temp Over Pressurization (LTOPs) Code Case N-514 Exemption Complete 4/7/97 6.
L' lops TSCRN 213 a.
15 EfTective Full Power Years (EFPY) License Amendment Complete 7/18/97 b.
30 EFPY License Amendment (not needed until cycle 12)
Resolved (N) c.
Supplemental Information Complete 9/12/97 (N) d.
RAI Complete 9/5/97 (N) t e.
Revision 1 10/3/97 (N) 7.
Letdown Rupture Line (USQ)- /JR 2/8 Complete 9/9/97 (R) 4 8.
Integrated Leak Rate Test Appendix J. Option B - 75CRN 212 Complete 2/1/97 i
9.
Integrated Leak Rate Test, Appendix J, Option B, Rev.1 Complete 5/1/97
- 10. Appendix J Option B - AdditionalInformation Complete 8/l1/97 (N) 1
- 11. System Readiness Review (Potential License Amendment)
Resolved
- 12. License Condition Amendment (if needed)
Resolved items i3 31. Potential USO Submittals
- 13. Emergency Diesel Generator Protective Trips During Engineered Safety Complete 9/12/97 (R)
Actuation and Loss of Offsite Power - IJR 219
- 14. IIPl Anal > sis (Without Cavitating Venturis)
Complete 6/14/97 l
- 15. IIPI Pump Recire Resolved
.._~._.,-_.--.e_.,~_-.-~n w.
.-.,-.._..,....-,-x_.
._-y.
r.
y.
o e..
e.,v,-,
litus Submittaljhtig
- 16. liigh Pressure injection Nozz.le Cycles (Contained in the 1 SAR R.23, Resolved 7/3/97 Letter, and not a USQ)
- 17. Air llandling 1ilter l/2 Testing Requirem-nts Resolved
- 18. hiakeup Supply Valve (h1UV 27)- Add Automatic Closure upon Receipt Complete 6/14/97 of #1500 Signal
- 19. liDU Fan Upgrade - L4R 216 Completc 8/26/97 (N)
- 29. Atmospheric Dump Valve (ADV) Design Capacity Resolved' (N)
- 21. Operator Actions for liOPs and APs 9/30/97 (R)
- 22. Remote Shutdown Panel Added Companents Pesolved' (N)
~
- 23. RW Cyclone Separator Fouling Resolved
- 24. h1UV 31 Control Logic Resolved
- 25. CT Protectors Resolved'(N)
- 26. htSLil Resolved (N)
- 27. IlPI Cavitating Venturis Special Test Resolved'(N)
- 28. AliF lC Fan Auto Transfer - LAR 2/7 9/30/97 (N)
- 3. Iloron Precipitation hiitigation - llot Leg injection 9/30/97 (N)
- 30. CCilli Leakage / Dose hiethodology Resolved (N) l
- 31. SW Cooling T13D (N)
- 32. Once Through Steam Generator (OTSG) License Amendment 211, R1 Complete 3/27/97 a.
OTSG, License Amendment 211. Resubmit Sholly Notice Complete 5/1/97 (N) b.
OTSG Response to AdditionalInfonnation Request from NRC Complete 8/19/97 (N) c.
Revision I to License Amendment 211 Complete 8/20/97 (N)
- 33. OTSO License Amendment IGA Growth Tube Flugging Information 9/30/97 (N)
- 34. PAhl Table Revision, TSCRN 209, RG 1.99EDG kW hieters)
Complete 7/29/97 l
l i
I
' infenuation letter to NRC 90 (C)
- Information letter to NRC 8'IC (C)
' Information leder to NRC 9/l7 (C) l
laAuf hubmhtml Date
- 36. CCllE Surveillance TS (Submit Prior to Restart) MR 222 11/24/97 i
- 37. EDO Surveillance Testing Exigent License Amendment TSCRN 215 Completc 8/4/97 (N) i a.
TSCRN 215. Revision 1 Complete 8/16/97 (N) 38.
100 kV llackfeed Tech Spec Change (not needed before restart)
Resolved (N) t t
I Licensing Restart Submittals (Non-License Amendments) 1.
13oron Precipitation Letter Complete 9/12/97 2.
Appendix R issues n.
Independent Assessment Summary Complete 7/3/97 b.
Appendix R - Revise Response to Address 72 N. Cold Shutdown Complete 8/29/97 (N)
(NRC RAI 7/7/97) c.
Appendix R - Thermo Lag Ampacity (NRC RAl)
Complete 7/3/97 (N) d.
Appendix R - NRC Infonnation Notice 92 18 (llot Shorts)
Complete 9/18/97 (N)(R) c.
Appendix R - Submit Description ofissues Related to Restart issue 9/25/97 (N)
D 11 to Support NRC Plans for Close-out (NRC RAI 7/7/97) l 3.
System Readiness Review-Summary Report i1/5/97 4.
Generic Letter 96 01 - Summary Report Complete 9/3/97 5.
FSAR Revision 23 - USQ Review Complete 7/3/97 l
6.
Generic Letter 96-06 Summary Report 10/3/97 (R) 7.
Tendon Surveillance Report (if needed) 11/1/97 (R) l 8.
hiaintenance Rule Structural liaseline - Walkdown Summary Report (if 10/6/97 (R)
L needed) l 9
Integrated Safety Assessment of Current Outage hiodifications -
10/27/97 Summary Report
- 10. FSAR Rev,24 Upgrade 12/4/97
- 11.. Restart Request Letter 12/1/97
- 12. RCS Attached Piping - Respond to NRC 4/7/97 Letter Complete a.
Program P.lan for Fatigue Analysis 10/30/97 (N)
- 13. Loss of DC Power. Failure hiodes and EITects Analysis - Summary 9/24/97 (R)
Report
- 14. FSAR Rev. 24/25 Upgrade Program Description Complete 7/17/97 (N) l m --
m3 ye m
y
hinc submittui nate
- 15. 50.54(f) Letter to Address Discrepancies (if needed)
Resolved (N)
- 16. Review of implementation of License Conditions Complete 5/20/97 Supplement to Address Additional Condition Closed (IR 97-08)
Resolved (N) a.
- 17. Submittals to Support SBLOCA 'ISCRN 210:
a.
Letter to Confirm Calculations (EDG Loading, Block Valve 9/25/97 (N)(R)"
Cycling) b.
Letter to Confirm EOPs do not Result in a USQ 9/25/97 (N)(R)"
c.
Submit Draft EOPs 3 and 8, AP 770 Complete 8/4/97 (N)
RAI for EOPs 3,8 and AP 770 Complete 9/17/97 (N) d.
Submit Proprietary and Non proprietary Version of LBLOCA Complete 9/2/97 (N)
Analysis of Record c.
EDO Testing Plan 9/25/97 (N) (R)"
- 18. Licensing liasis for Dynamic Loads a.
Initial Response (Pipe Whip) to NRC 4/10/97 Letter Complete 6/5/97 (N) b.
Supplemental Response (re: Jet Impingement) 9/29/97 (N)
- 19. Low Pressure injection - hiission Time Complete 5/15/97
- 20. Appenux R - Atmospheric Dump Valve 30 Day Response Complete 4/30/97
- 21. System Readiness Review Plan Complete 3/27/97
- 22. EOP/AP Operator Action - Commitment Date Change Complete 3/25/97
- 23. GL 96-06, Letter regarding Service Water Fittings Complete 4/30/97
- 24. ash 1E Section XI - Code Case Relief Request Complete 3/24/97
- 25. h1 CAP II, Rev. 2 Complete 4/25/97
- 26. ITS Ilases Submittal Complete 4/18/97
- 27. Letter on Pressure Locking re: EFV 12, 32,-33 (Supports SER for OL 95-Complete (N) 8/l1/97 07)
Complete (N) 8/1/97
- 29. Temporary Waiver of Compliance for CCHE hiodifications (FSAR Resolved (N)
Requirement)
.._ __..a
- 31. IWE/lWL Supplemental Response TBD (N)
(N) New (R) Revised
" Consolidated with TSCRN 210 Supplement Due 9/25/97 n omsw.v m l
l