IR 05000413/1990010
| ML20042F136 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 04/30/1990 |
| From: | Merschoff E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Tucker H DUKE POWER CO. |
| References | |
| NUDOCS 9005070258 | |
| Download: ML20042F136 (81) | |
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. April 30, 1990 . Docket Nos. 50-413, 50-414 L License-Nos. NPF-35, NPF-52 l: L Duke Power Company ATTN: ~Mr. H. B. Tucker, Vice President Nuclear Production Department . " 422 South Church Street . Charlotte, NC 28242 Gentlemen: SUBJECT: ENFORCEMENT CONFERENCE SUMMARY, _ (NRC INSPECTION REPORT N0S, 50-413/90-10 AND 50-414/90-10)- , This' letter refers to the Enforcement. Conference-held at our _ request 'on - April 25, 1990.
This meeting concerned activities authorized for your _ Catawba- ' facility.
The issues discussed'at this conference related ' to-the Unit: 1~ pressurization event on March 20, 1990, the modification to.the control circuitry
logic for the residual heat ~ removal suction valve, ND-36 and the reportability of the pressurization event. A list of attendees, 'a summary and a copy of your . handouts are enclosed.
We are continuing our review of. these -issues to , ,' determine the appropriate enforcement action.
In accordance with Section 2.790 of the NRC's " Rules of Practice," _ L - Part 2, Title 10, Code of Federal-Regulations, a copy of -this' letter and its l enclosures will-be placed in the NRC Public Document Room.
Should'you have any questions concerning this matter, please contact us.
Sincerely, - ' (0RIGINAL SIGNED BY'C. A. JULIAN FOR) Ellis-W. Merschoff Acting Director Division of Reactor Safety Enclosures: 1.
List of Attendees 2.
Enforcement Conference Summary - 3.
Enforcement Conference Handout cc w/ encl: T. B. 0 wen, Station Manager i ' Catawba Nuclear Station P. O. Box 256 t, Clover, SC 29710 I (cc w/encls cont'd - See page 2) 9005070258 900430 l {DR ADOCK 050004.13 3- ! ' PDC ~ i
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.f-Duke Power Company
l April 30, 1990 , -(ccw/enclscont'd) A. V. Carr, Esq.
Duke Power Company
422 South Church Street i Charlotte, NC 28242 ! J. Michael McGarry, III, Esq.
I ' Bishop, Cook, Purcell end Reynolds 1400 L Street, NW Washington, D. C.
20005 North Carolina MPA-1-- . i 3100 Smoketree Ct., Suite 600 P. 0. Box 29513 , Raleigh, NC 27626-0513 Heyward G. Shealy, Chief s Bureau of Radiological Health
South Carolina Department of Health l and Environmental Control ! 2600 Bull Street l Columbia, SC 29201 j Richard P. Wilson, Esq.
I Assistant Attorney General ' S. C. Attorney General's Office P. O. Box 11549 Columbia, SC 29211 Michael Hirsch Federal Emergency Management Agency
500 C Street, SW, Room 840
' Washington, D. C.
20472 North Carolina Electric.
I I Membership Corporation ! 3400 Sumner Boulevard I l P. O. Box 27306 l Raleigh, NC 27611 ! l Karen E. Long l Assistant Attorney General l N. C. Department of Justice P. O. Box 629 Raleigh, NC 27602 l Saluda River Electric . u ' L Cooperative, Inc.
l P. 0. Box 929 Laurens, SC 29360 (cc w/encls cont'd - See page 3) i L-j
e- - p .- , Duke Power Company 3-April 30,1990 (cc w/encls. cont'd) S. S. Kilborn, Area Manager l Mid-South Area ESSD Projects j Westinghouse Electric Corporation ' MNC West Tower - Bay 239 P. O. Box 335 Pittsburg, PA 15230 County Manager of York County York County Courthouse York, SC 29745 Piedmont Municipal Power Agency 121 Village Drive Greer, SC 29651 State of South Carolina bec w/ encl: K.
- ?. Jabbour, NRR J. Lieberman, OE B. R. Bonser, RII M
. Shymlock, RII cument Control Desk NRC Resident Inspector U.S. Nuclear Regulatory Commission Route 2, Box'179-N York, SC 29745 RII:DRS RII:DRSr RII:DRS RII:DRP RII: P PKe gg TP e les / s hoff He dt
04/ /90 04/,)7/90 4d7/90 4/u /90 4/g 7 0 t RII:RA ~ an, 04/ /90 g l - ,
_ _, .- . ENCLOSURE 1 LIST OF ATTENDEES U. S. Nuclear Regulatory Commission.
l J. L. Milhoan, Deputy Regional Administrator-- C. W. Hehl, Deputy Director, Division of Reactor Projects E. W. Merschoff, Deputy Director, Division of Reactor Safety G. R. Jenkins, Director, Enforcement and. Investigation Coordination Staff A. R. Herdt, Branch Chief, Division of 'leactor Projects T. A. Peebles, Branch Chief, Division Of-Reactor Safety P. J. Kellogg, Section Chief, Division 0f Reactor Safety M. B. Shymlock, Section Chief, Division of Reactor Projects R. W. Borchardt, Office of Executive Director For Operations K. N. Jabbour, Project Manager, Office of Nuclear Reactor Regulation W. T. Orders, Senior Resident Inspector, Catawba W. M. Troskoski, Office of Enforcement P. T. Burnett, Reactor Inspector, RII C. W. Rapp, License Examiner, RII , i B. Uryc, Enforcement Coordinator, EICS l Duke Power Company H. B. Tucker, Vice President Nuclear. Production Department l T. B. Owen, Station Manager, Catawba L R. N. Casler,-- Operations Superintendent, Catawba T. E. Crawford, Superintendent of-Integrated Scheduling, Catawba ' l R. C. Bucy, Design Engineering, General Office A. S. Bhatnager, Performance Engineer, Catawba J. R. Ferguson, Operations Engineer, Catawba i R. M. Glover, Compliance Engineer, Catawba C.- L. Hartzell, Compliance Engineer, Catawba ! D. L. Rehn, Design Engineer, Duke Power Company i R. G. Morgan, Regulatory Compliance, General Office
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7, _ y ,.- . . ENCLOSURE 2 ' ENFORCEMENT CONFERENCE SUMMARY-
On April 25, 1990, representatives from Duke Power, Company (DPC) met with the.
NRC in 'the Region 11 office in Atlanta, Georgia to discuss -issues concerning the March 20, 1990, pressurization event, the modificatib1 to the control circuitry logic for. residual. heat removal' suction valve, ND-36 :and-the reportability-of the event.
The-first issue concerned the pressurization-of the Reactor Coolant System (RCS) on March 20, 1990, while conducting a~ fill-and vent evolution with the wide and narrow range RCS pressure instruments isolated.
The second issue concerned the modification-to the logic for-the operation of the residual heat removal suction valve, ND-36.
.This modification. led to the-operators not being able to electrically open the suction valve during the event.
The third issue concerned the reportability of the pressurization event to.the-NRC.
Following opening remarks by James-Milhoan, NRC.RII, Deputy Regional ~ Administrator,- DPC gave a presentation (Enclosure.3) on the issues.
T. B. Owen, the Catawba ! Station Manager, introduced DPC's presentations.
The first presentation covered the Unit 1 pressurization event and included a sequence.of events, system description, operator response, root cause: and contributing causes, corrective.
i actions, the safety significance 'and reportability of the event.
The second presentation covered the modification to ND-36 and included a-description of
ND-36 and FW-55 functions and interlocks, root'cause of the problem,: corrective actions and safety significance. A summary and concluding remarks were made by.
-l T. B. Owen and H. B. Tucker.
! The NRC -closed the meeting by stating that DPC's presentations had served to l enhance Regions II's understanding of the issues and DPC's corrective actions. > l ! ! i '! ! ! '
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. .. ... . -- - ENCLOSURE 3- -, , , ! AGENDA DUKE POWER COMPANY- '! - , CATAWBA = NUCLEAR STATION - ENFORCEMENT CONFERENCE j 04-25-90 -! f I ! L j . .f Introduction T. B. Owen , , o
! ND Pressurization Event ! E i Sequence of Events / System' Description / - ( Operator Response R. N. Casler .
Root Cause/ Inadequate Review of NMDB R. N.'Casler
-- Contributing Cause/ Scheduling Mechanism T. E.'Crawford - Corrective Actions T. E. Crawford/R. N. Casler ~ -
Safety Significance
- , ' Peak Pressure Calculations A. S. Bhatnager i . Design Evaluation of Component Effect R. C. Bucy
. Safety: Evaluation of LTOP Isolation R. C. Bucy
. Reportability R. M. Glover . , . N.D-36, Modification
Description of ND-36 and FW-55 Functions / ~ - l Interlocks - - D. L. Rehn
' Root Cause of the Problem 'R. N. Casler/D. L. Rehn - Corrective Actions R. N. Casler - . Safety Significance R. C. Bucy - . Summary T. B. Owen concluding Comments H. B. Tucker
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NC PRNARY COOLANT) SYSTEM
i NO SUCTM , ' t>i lND-37a cH te-me i ic s/9 to s/s ^ ^ v v E NCP gy nv v l . RX e uTunNLess "'""" " VESSEL '
, ' is wcP sa Nep . ' '- n n . ,, , .. . V U ' , 18 S/9 A / la S/S A V V ND-lodll ' ' , i m-nae t " NO SUCT m gH l l 4l ' C.PIR PONv's - - cH l l 9lf-estocamova , w NO SUCT. ION.M.LE.F =va ., n . PZR N) ' . [
PRT l' l: . . -. -. -. - - - -., _.. _. - _ - - _. _.... - -. - - - ... . _ _ ... - .. .. - _,. _ _ ,,..__..-..__.,-___) - -
- - . . _ --. .- -- - .. .. j .. . .- -SEOUENCE OF EVENTS /0PERATOR= ACTIONS-
0700 - Fill of Reactor Coolant System (NC) in progress.
-0708'- Pressurizer is full as indicated by Pressurizer Relief Tank (PRT) level increase. Closed pressurizer PORV's. Fill continues to raise.
pressure to 100 psig.
(This process normally takes 4 to 6 hours.)
i 0945-- Control Room Operator notices PRT. level increasing. Recognized this-as abnormal and reduced the charging rate toward the letdown rate.
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He notified supervision..
0950 - Operators began isolating PORV's one at a time to verify if they.
were leaking.
PRT level still increasing.
1000 - Operators notice RHR (ND) pump discharge pressure indicating 375 psig.. At this point, operators realize NC pressure is 175 psig even 1 though NC= pressure instruments-still-indicate no pressure.
Had-operator in containment look at-ND suction reliefs.
1008 - Operator in containment reported "B" ND suction relief was passing flow.
. 1030 - With ND pump discharge pressure reduced toward normal, "B" ND suction line was isolated from t?.a NC System to ressat relief valve.
(Per AP-19) ' Entered 8 hour Tech Spec action statement-for inoperable pressurizer - PORV's.
(Clock was started at 0708.). Notified IAE to investigate NC pressure indications.
- l 1100 - Obtained all points data base printout to analyze the. event.
Printout showed NC pressure reached 520 psig (uncorrected).
1112 - Attempted to realign 'B' ND Suction to the NC System. ND-36B would , not.open. Dispatched operator per procedure to manually'open ' ND-36B.
{ 1205 - ND-36B is opened.. 'B' ND realigned to NC System.
Relief reseated.
' f l 1345 - PORV's are-reopened.
' 1420 - NC pressure instruments were unisolated. Declared PORV's operable.
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____ __ _ __ _. _. - _. __________ _ ___ _ ' .. , . . e,. . REACTOR C00LANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the 'following.0verpressure Protection Systems shall be OPERA 8LE: a.
Two power operated relief valves (PORVs) with a lift setting of less than or equal to 450 psig, or j b.
The Reactor Coolant System depressurized with a Reactcr Coq 1 ant System vent of greater than or equal to 4.5 square incoes.- APplSCA81LITY': MODE 4 when the temperature of any Reactor Coolant System cold- . leg s less than or equal to 285'F, MODE 5 and M00E 6 with the reactor vessel head on.
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' .- q ' ACTION: . - . a.
With one PORV inoperable, restore the inoperable PORY to OPERABLE. status within 7 days or depressurize-and vent the Reactor Coolant , l System through at least a 4.5 square inch vent within the next-8 hours, , b.' With both PORVs inoperable, depressurize and vent'the Reactor' Coolant System through at least a 4.5 square inch vent within 8 hours.
. c.
In the event either the PORVs or the Reactor Coolant System vent (s) - are used to mitigate a Reactor Coolant System pressure transient, a Special, Report shall be prepared and submitted to the Commission pursuant to' Specification 6.9.2 within 30 days.. The report shall describe the circumstances initiating the transient, the effect of the PORVs or Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence. - d.
The provisions of Specification 3.0.4 are not applicable.
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CATAWBA - UNITS 1 & 2 3/4 4-37 ,--- - .- - . - . - - - - -
(9 . , . . . -. , . ? IMMEDIATE CORRECTIVE ACTIONS o Stopped the NC System fill and vent t , b o Conducted a review of all outstanding pape'rwork '
o' Verified station surveillances'we're up-to-date .i o Walked down the RHR System piping and components o Verified instruments needed for existing plant condition and for the:- remainder of the fill and vent evolution were available'and valved in .
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. ' - l . .- OPERATOR ACTIONS TAKEN o Thoroughly discussed event with all shifts,_and followed that with an-operator update o Procedure changes completed or being made-o Added information for alternate indications , o Added information about static pressure with NC filled o Added caution to remind operators pressurization may occur at any time- '
. , , o Event will be covered in requal t o Presently determining if simulator can be utilized to allow operator to train on the fill and vent evolution
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UIIIT 1 OPERATIONS islEGER UNIT 2 OP R ATIOMB MalBG E , D Tower C. Muse i ununumummummm usminunummuum munummmmmmunimummmmmmum-m CLERE - B. Currence > ! ? " - IfDC PROD NUC PROD - EXIGINEER ElIGINEER I I i J.14 ethers N.~Nicholace J. Reese R. Green l I J. Gregog L. Benjamin ! - SHIrf - SHIFF - SUPERVISOR SUPERVISOR . R. W. Salth .J. Hill-ASST / ASSOC - - ASST / ASSOC - ENGINER ENGINER i i j ' L. Blaakaa= hip D. Coforth J. Suptela - D. Miller T. Simril T. Earl i
, ' D&fSTAFF -
M. Finamhat==r W. Seegle NUC PROD - i l SPEC m l Stan Marice
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i ' METHOD USED TO ENSURE PROPER.
EQUIPMENT IS OPERABLE DURING OUTAGE ! o Project 2 Schedules and Tracks Work i o Tech Spec Requirements t o Containment Closure [ o Mid Loop Operations o Prior to work starting, Operations must approve all work requests o Operations protects operability requirements for the equipment needed for " that mode or condition change , o Operations has the'following tools to aid in.their review prior to changing modes or condition: o Nuclear Maintenance Data Base (NHDB) o Project 2 o Removal and Restoration (R&R's) f J o Periodic Tests o Proper System Alignments o Detailed knowledge of work status by Unit Manager's staff engineers-o Technical Specification Action Item Log l .
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ROOT CAUSE . i . Management deficiency due to an incomplete. review. of equipment stated . Indicators prior to a condition change ' ,
CONTRIBUTING CAUSE Scheduling Mechanism , - ! l ! i i I
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Outage Management Philosophy Outage Schedules Process of Scheduling a Worz Request 3 vents That Led to Pressure Transient ' Root Cause I Program Enhancements ' , - - - - - _. .. . ... .. .. .. . -. ... _... _. _.,
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. . ' Basis of Outage Philosophy l Operations - /s th control of the Plant < Will sip on work to be done in , - a mamer that the plant wi// remath th a safe condition , l Interated SchedJ/thg . l - Responsible for Scheou/e D.velopment - Inte7ation of a// Grom Wc k Activities thto an effective Outage Plan - Keeps all voms thformed cf where outage is headed and what work needs to take place , . - . -. -..
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! , ! l Generic Outare Schedule i , i l s > l Rovides Work Window for - . ! Major System / Equi lcment ( i ! l l Work Windows are arranged to: - .
i ! - Conply with Tech @ecs
- Rectice ALARA Ribchles . \\ - Perform Work Safely and . , Efficlently - Maintain Emergency Power Sqqoly (D/G) for a// equipment required for Decay Heat Removal - Mihih7tze Tih7e in Recticed Inventory Goeration . . - - - -. - -. -. -. -
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Mode / Condition Change '
i ! i j MODE - Identified in Tech Spec, Table 1.2.
l CONDITION CHANGE - A change in the Tech Spec requirements due to a change in some plant parameter or configuration (temperatura, level, reactor vessel head j installed, etc.) without changing from one ! mode to another.
There is not a l consolidated list of condition changes in l Tech Specs.
They are identified in the various specifications under applicability statement.
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! i , i REACT!v!TY s AAtt0 AvtAAst co0LANT
E CONDITION. K;ff THERMAL POWER" TEMPERATURE ' ' 1.
POWR OPEAAHON 1 0.99 > ts > 360*F i 2.
sTARTup 3 0.99 < as > ss0*P ! i 3.
NOT STANOSY < 0.99
> 390*F 4.
NOT $NUTDihm < 0.99
350*P > T8 > 200*F
5.
COLO $NUTDOW < 0.99
< 200*F ' 6.
REFUELINB** < 0.95
< 140*F ', , . . . < . "pcTustng escay heat.
- Fuel in the reacter vessel with the vessel head closure bolts less than ful tensioned or with the head removed.
.. t l l l . l-I t CATAWB4 - UNITS 1 4 2 1-9 ._ .. _. _. ~. _ _.... _... _. - _ _.... _ _. _.... _..., _ _. _ _.. _. _ ... .. ... .. . . -. -
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i LWleT M , ! LIMITING CONDITION FOR OptRATION l I
i 3.9.8.2 ! at least one residual heat removal loop shall be in operation.*Two
i i I APPLICABILITY: MODE 6, when the water level above the top of the reactor t vessel flange is less than 23 feet.
, , ACTION: ! , L ! a.
With less than the required residual heat removal loops 0PERA8LE, ! imediately initiate corrective action to return the required resi-l dual heat removal loops to OPERA 3LE status, or establish greater ' than or equal to 23 feet of water above the reetter vessel flange, I as soon as possible.
.
b.
! With no residual heat removal loop in operation suspend all I operations involving a reduction in boron concen,tration of the i Reactor Coolant System and imediately initiate corrective action ' to return the required residual heat removal loop to operation.
' Close all containment penetrations providing direct :::ess from the containment atmosphere to the outside atmosphere within 4 hours.
l SURVE!LLANCE REQUIREMENTS t
, be verified in operation and circulatin9 reactor coolant at ! ! 4) greater than or equel to 1000 gps, and i } b) sufficient to maintain the RCS temperature at less than or equal to 140'F.
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- Prior to initial criticality, the residual heat removal loop s
C0RE A TERATIONS in the vicinity of the reactor vessel hot legs.
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- i h ! !, CATAWBA - UNITS 1 & 2 3/4 9-11 Amendeont'No. 69 Amenenent No. 63 [ Unit 11 LUnit20
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Process of Scheduling a Typical l Work Activity i 1.
A// outage related cr potentially . outage related Wcrk Requests are routed to the Lhit QJtage Manager , for anproval to be wcyked otnng , I the outage.
If he has questions, he ncyme//y contacts the Cperations lhit Manager': Groip - fcr gJ/ dance.
. 2.
After the Qitage Manager qqcroves the wcyk request is routed to air . Plamhg Section.
- . . F i ... _,. _, _. _., _ --- - - - - - - - - - - - - - - - - - - - - - - - - - " - - - - - - - - - - " ' ' " " ~ * ~ ~ ~ " " ' " ~ ~ ' " ' ' ~ " ~ ~ ~ ~ ~ ~ '
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\\ Process of Scheduling a Typical .
. Work Activity ' i (Continued) '
1 3.
After the work request is plameo,' i a copy of the plamed work
request is routed back to the ^ , , . Outage Manager along with the , following: . . . ' - 9W Go40s ad syncWements/ wcvk requests requhed to sunocrt Ws Wak Request - 9% nequhunents sta:h as scaffoldha Inntiation remonn4 pola crams etc - odw Infannetten that Pfamirty Secticn cletamshee is neceaney to hop scheathe We Wcyk Requset- .,
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, p Process of Schedullar a Typical Work Activity (canusne$ ! . I
4.
A copy of the planned Work Request is sent to the Operations Unit Manager's Group and the Performance Group.
, Operations Prorides: , ad Rwstar Berninmuto meet en cantatsmut auere er interrity
Meet en Paneiretion !\\notlag ' Xede maage Antairemente candittens the plast has to be la to rark this aettrity . , I Per/ormance Prorides: notenting nerairemuts t t 5.
With this information a scheduler will enter the work activity into our outage schedule.
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i Typical Work Activity l Investigate and Repair Packing Leak on 1NV-181 t ' . The following information is available to the l Job Supervisor: -York Request for the Repair Work (275750PS) -IAE Support is required per Work Request 275750P-2.
-The tagout this work is to be performed under.
-ALARA pre-job meeting is required.
-Scaffolding is required.
-Hanger Removal / Replacement is required.
, l The Job Supervisor is responsible for the support activities.
. r . -. .. _.,.. _..... _ _ _. _. _.,.. -.., _ _ _,.. _ . . . _,,... _ _ _ _ _ _ _ _ _ _ _ _ _.. _ _ _ _ _ _ _. - _ _ _ _ _. _. _.. _. _ _ _, _,.. _ _ _. _ _ _ _ _. _ _
_ - - _ - - - _ - _ _ _ - _ - _ _ - _. - _ - _ -. _. _ _ - - - - ,
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! Events That Led to Pressure Transient
we had previoumey rienced primary evetem ientese on Nah - , j prenew. oormression ta*w fitting aoosted inside contahmenti Work Regmoto were generated to repleos oormroesten fitting with - - socket weidad fitting.
This auteos, we were scheeled to oormiete the remainho fitting - on the Remotor Coolant System, The work motMty to weld the fitting on the Remoter Coolant ' - . Wide Range Prenews Loop canaleted of a primary work rommet , to weld tho' fitting and a emplemental work recpest for IAE to ' isolate and unlooiste the inotnsnentation loop, i
- O
'to setthe the Remotor Vessel Head.
The empien;-atal work recuest wee a list item mder the primary work activity.
, - S " I the job complete.
The Planning section determhed that the work to be performed - by the emplemental work recpost wouci be covered by another work reemet, that was achecMed to be completed prior to entry hto Mode 4, Not knowing that thle work needed to be oormleted now, they planned to work both work recpeste teosther trier to Mode 4L l , t ! !
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. __ __._. _. _. _ ___ _ _ _ _ _ _ _ _ _ ___ _. _ _ _ __. _ _ ____ _ _.... _.
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! k i Root Cause j , ! i taitiany, we theaght the teet eense-to be taadequasies la the I estage sehedsung presses.
i
! Reverer, we learned that if we had rettered the Naatser [ ] Mat =&====== Data Base (NMDB) prior to doetartag the . i PORT's operable the entstanding Work Requests would have ! - l bees identified (per previess itseassieas).
This weald have j taaleded both the primary and supplemental Werk Bequests.
i ! ! , ) Additional Facts After Further Review i
- The immediate eerrective acties was to held the sait and perform a rettow of the NMDB for any work requests that m4y affect egalpment i ! l aseded for fut and vent.
i - I retter meet be performed of NMDB besasse work eeste have started ! ! that was met la ear estage network.
' . , Conclusion After Further Review Our sekodale is adequate and met the rest easse, as we determined earner.
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- -. .. - - - - . - -. 3.m w.-. -. - - - - - -
. -. - - ... _ _. _ - - _ - - - _. - . - _ _ - _ _ _ _ - _- _ _ _ _
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Operations
i Provide Integrated Scheduling with a list of all Tech Spoo Condition Changes that occur in i Modes 1 through 6.
Develop a list of equipment required operable to support these condition changes.
- l Revise operating procedures to require signoffs by the appropriate station sections that the equipment operability requirements for condition change are i met prior to the change occurring.
Intearated Schedulina Establish codes in our outage network for all condition changes identified by Operations.
All activities identified by Operations as being required for a Tech Spec Condition Change will be flagged in our ' outage schedule.
This will allow Integrated Scheduling to generate liste of York Requests that need to be , completed prior to changing conditions.
This will - L be independent of what has been previously ! reported as completed in our Outage Schedule.
l , i
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__-- .._-- - _... __ -___. -_ ______ _-
. - - . i .
t Program Enhancements _ ELkana ' Revise the existing administrative controls for instrument valve position verifications.
t
- Currently IAE completes an instrumentation valve checklist for all critical safety related
instruments prior to entering Mode 4 (these t checklists include root valves, manifold valves, drain valves, etc.)
, l ' l - These checklists will be subdivided into individual - . l checklists to align root valves needed for a condition change.
These checklists will be ' completed prior to changing ' conditions.
Develop a program (in conjunction with Operations).
that will identify Control Room instrumentation that is either out of service (e.g. root valves isolated, etc.) or
known to be out of calibration.
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___. _ _.. _.___.. _ _. _ _. _.._ _ _. _.____ _.__ _ _...-.. _ _._____ __ _...__ _ _ j
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SUMMARY i
! Our Outage Scheduling Process has worked well.
' - ' ! It has been effective in completing six refueling - outages.
Our Scheduling Program aids Operations in - - maintaining Plant Configuration Control by . logically scheduling work to comply with l . Technical Specifications and other operating ' requirements.
i - Containment Closure - Mid Loop Operations
' . Our Scheduling Process did contribute to our - - Configuration Control problems in that it did not provide a second level verification.
. . . The enhancements. planned for the scheduling - process will provide the needed verification of work -
completed.
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. . f' PLANNED CORRECTIVE ACTIONS ' FOR OPERATIONS MANAGER GROUPS { . o The existing program for the review process for condition changes and mode changes is currently not documented in our Operations Management Procedures.- This program will be documented in our Operations Management procedures to ensure uniform understanding and implementation by all Operations' groups , o A review session will be conducted with the Operations Manager groups on this procedure prior to Unit #2 EOC3 outage (currently scheduled for June 1990) ,' , o Discussed event with Operations Manager groups emphasizing what a proper review consists of prior to changing modes or condition changes , l , T
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l l DMAE1 , . The process in place for modo changes or condition changes is a very good one.
We have successfully completed many mode and condition changes without any probloos. We recognise a problem occurred for this enndition change where a work request was not identified in the review procesr. We have taken ! corrective actions and have other planned corrective actions for Operations that will prevent a similar occurrence in the future. Also coupled with the enhancements to our scheduling process as Tom stated earlier, our very good Operation's review process will be even better.
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/nEPuEt G% ~, 8e027 . 1 MATER OnAGE ! EG - I Ccr > I ses. ass GAL.
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uTeon . , I )$ I seDee mm _ senny PnT g y ~ HDfL IA ', ; sed 26 NS43A O A
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CONTAsseneEM' ,, g e s f Suaer 1r SAAAPLE W136 I -- -- ) -. NV 18 D
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. O O . .... .. _ _ _,.._ __._ - - -. CNS OAC DATA FOR UNIT 1 FOR R1484 ND PMP R DISCHRRGE PRESS 800 SIG -
P 760 - I I I
- .r -- -- - (gweyym esu,m-prse - 720 z--- L.
-savrwpswrt+1 knA.. . ..~ 660.i 'M Pvy.., .- ......j,.. .... . 9.,..y.......,.s... t......... , ,... ... G40 ' - - - -- -; - - - - , - - ..- - - - 600 l - . - - - - -- - - --- - - - - - 560,: - - -- - - ' - - - -- - - - 520 :.- - +- - - A80 ~ -- - - - < . - .- ---
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10.5
ogpa: TIME (HOURS) FROM STRRT OF MRR 20 ,
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-= . CNS ORC DATR FOR UN1T 1 FOR R0820
' CHRRCING LINE FLOU CONTROL
GPM 110 m ...,, s--u-a4 . -.. - - + -- s . 100 _ l . i -
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9. 5
10.5
TIME (HOURS) FROM START OF MRR 20
CNS DAC DATR FOR UNIT 1 FOR A0452 NV LETDOWN FLOW G 120 _ PM - -
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. egas se.
- 100.* e
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9. 5
10.5
TIME CHOURS) FROM START OF' MAR 20 ! t I i .
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. . !' L CATAWBA NUCLEAR STATION , ND PRESSURE EVENT: 3/20/90 i t i Peak Pressure Calculation ! ' t
. ! ! !
i ! . All ND Pump A discharge pressure data lowered by 65 l , ! psig for "as found" instrument error
l
. Peak recorded pressure 588.4 psig @ 0940 hrs
. . Extrapolated pressure between 0935 & 0940 hrs
calculated to be 625 psig i i , ! ' l . ND Pump A suction pressure (NC system pressure)
calculated to be 460.8 psig from pump head curve
. l - at 3,100 gpm flow - developed head of 390 ft l
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_ __
- __ , , . . r ! CATAWBA NUCLEAR STATION ND PRESSURE EVENT i '
Peak Pressure Calculation , 1000 f-800 - - - , ( j 600 y -- N \\\\ m gl \\ .
J > _ /N sH7 8 k / i %s: :: _.....
- :: :: ::_g
u "- -: p - <
/ k 200 / %
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900 920 938 955 1015 1035 1055 TIME Uncorrected Discharge + Corrected Discharge
+ Calculated Suction Extrapolated r .. .. . -. .
.. O e . .. . . , , CNS OAC DATR FOR UNIT 1~FOR A0879 ~ ' PRT LEVEL
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10.5
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TIME CHOURS) FROM STRRT OF MAR 20 Twf: oft M P4nf/?Md > psAN.".o?$$Y,fe*h ^ s. n; o ;. * l.o ? ? /* f,.? e, was msa - / n, S . O m ce p,e t sv unt C;;= fr6 ava %*s' in'u 144NtA9ff Wi%f 1>hs tynerfs sofrrD, run's R*1fi 9t"'O 80 n4%Q 'n D% dufL-12n6 cf fM 7>**K AfrG4 oNd, f /.07if 7.I/,u e', + o795%/',+i,[ = /,/5F[j *<$ d k t4~<4 Ta cAwuS pCuom$/ht e$ fosa^.o in14i atm Os* 6P/i/phycs/o i CVRHC '?.27 e PF6s(4Pitr.wt ReutPrast (per>%Mw) pet 7>6 (paM op
ItJT6 @ $f: ' Cr? */o cotw1Sfuet Tb 9500 MLk"'l '2900 %$ -c2- - - 0 h CoRA65"ffn95 76 l 2,'l O O M itlM $ '3h ~ /31. 9 "*fy; . o ~ m m,sn, w t.wr e us. ne - m ins 7>a u n,is : ' ( l.15? f &*" h ' ' " / 5 '2. S' g, s= Wf d mm a/seks cacmcen ep hA// a/zJg6_ l i?cacam rey: w p,.m - C/ / /-
, ,
m, . . - 4 i CATAWBA NUCLEAR STATION ! ND PRESSURE EVENT: 3/20/90 - Confirming Evidence - PRT/ND Relief Valve ~
Input to Pressurizer Relief Tank calculated at 153 gpm
, ! . At full lift ND relief valve will pass 1040 gpm . , - Cold Set Pressure is 463 psig (< 250 F) ' - Full Lift at 45 psig above set pressure - Designed capacity greater than one NV pump , l ' Upon rescating of relief valve, weepage detected past.its . seat. Experience shows this as a sign of chattering of the relief valve near its lift setpoint of 463 psig
Calculated ND suction pressure was 460.8 psig .
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..._ , t - . . . O - i CATAWBA NUCLEAR STATION i ND PRESSURE EVENT: 3/20/90 Confirming Evidence - OAC i , 800
psig
11 volt (770 psig) ,. ' hl ki[ " 700 4000 counts i psig . (700 psig) . - .
.
l i
0
0 psig psig volts counts , . Transmitter Calibrated Process Operator Capability Range Control Aid Cabinet Computer ,. Output i . Absence of Scanner Overflow Alarm on OAC . , t P ., . ,,. -.,.- -..,. .$ , ..- - - - - - m -- - --- - -e-- -- +m.
- - - - - - - - - - - - - - - - - - -
... _ _ _ _. _. _ _ _. _. _. _.. _. _ _. _ _ _. _ _ _ _ _ _ _ _ . - _ _ _ _ _ _ _.. _ _ _ _ _ _ _ _ _ _ _ i . . i.
DESIGN ENGINEERING EVALUATION METHODOLOGY NC SYSTEM PRESSURE I 1.
CALCULATED PEAK NC PRESSURE BASED ON ND PUMP 1 A i DISCHARGE PRESSURE GAGE.
I 2.
COMPARED TO LIMITS FOR RCS IN T.S. 3.4.9.1 , . l 3.
EVALUATED SIGNIFICANCE OF 2 TRAINS OF LTOP BEING j INOPERABLE FOR NC CONDITIONS PRIOR TO MODE 4.
l
l l - l .
+
1 !
.. . -............,., -. .. ~ -. -. -. -.. - -.. .. -.. .. - - .. - .. - - - - - - - . . . .
, i - y . . CATAWRA NUCl FAR STATION J - t , ! NC (PRIMARY COOLANT) S$1 ITEM j ' No suction . . ~ H [NO-37A , cH(N0-300 ic s/s 10 s/e i A A v v
' IC NCP 10 NCP ! A n l v v ! . . RX
- " " "
MD RETURN LINES i VESSEL
' l
l IS NCP IA NCP , n A . v v . , IA S/g i 1884 n / n v v . ' '
NO-It ' " NO-RA. l ' NO SUCTION % PER PORV'S ' SLDCK MhLVES i l- % NO SUCTION RELIEF VALVES NO-3, N-SS : i - PZR
l- , lXM l PRT '
! i ,,,,,, _,.. -, _,,..,. -... -..,. -,.. -,,.....,..... -. _.,. - _.. _.. _ _. -. _ _,.. -. _,. .. .. -.,,,.., _ _......... _..,,,,., -
_ __ ._ - - - - - - - - . . . . . SAFETY RELIEF VALVE DATA SHEET PLANT: Catauba Nuclear Staties ~ . . FLOW DIAGRAM SERIES: G-1561-CN-2561 ^#
SYSTEM MAME: Red " - - 1 M==t W1 - W sy: B.E. Cauthem
J SYSTEM FILE: CN-1223 -11 b Non-Active . DOCUMEllT NUMBER: ramac_i m _ii an aair ~ -t a IM REVISION: 0* DATE:. .- Q4 CONDITION:
Valve 13 3 IBM O Number (s) 23 3 29 3 Temperature F Desle -
!- Max Relieving Valve i Function ' andy SS ' Nozzle SS Inlet Outlet "'7" Trim SS Size 4" 0, "# E * " * I Size Other Duke Piping Inlet Design y , , Class g Pressure (PSIG) '
ASME Code
IW Inservice inlet Facing ' Class' Inspect Category Requirements (if any) ! p Process Fluid Radioactive, Berated Mater Outlet facing Requirements (if any) Capaci ty p g gpg (State Units) Accumulation le Set Pressure PSIG 450 (% Set Pressure) i Tolerance - % or FSI Blowdown
R:nge % or PSI ' (t-Set Pressure)
t Back Pressure SUPerimPOSCd 9~i Special (PS IG) Built Up Accessories kIIW5 ! , !
HOTES: OR$4NAL Tills DATA HAS BEEN REVIEWED FOR (I) DW - Dut t Weld ACCURACY Also IS SATISFACTORY FOR ' F - Flanged DO _NOT REMOVE FROM Fitt E@lNNT E 3TERIR ECIFICATIONS.
PREPARED BY P f DATE 1/st/s4 , SW - Socket Wcld CHECRED BY 47/ h_ DATE g-/f-f[ i SCRD - Screwed Rev.'7/30/81 APPROWED BY4 ff J-cfm DATE1_+ n . - (
,4 ,- , ,,. . -. , .. -... -. . .. . .
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l DESIGN ENGINEERING EVALUATION SIGNIFICANCE l NC / ND OVERPRESSURE EVENTS VS. ND RELIEF j
CAPACITY , l i ' l EVENT FLOW (GPM) l 3/20/90 PRESSURE EXCURSION 153
! i . FSAR LOSS OF AIR EVENT 400
. HYPOTHETICAL 1 CHG PUMP RUNOUT 550 ! i ! E ANALYZED 2 CHG PUMP CASE 600 < ' f (T.S. REQUIRES 1 CHG PUMP RACKED OUT ' l LESS THAN 285* F) l-i HYPOTHETICAL 1 S1 PUMP RUNOUT 650 i f (T.S. REQUIRES BOTH S1 PUMPS RACKED OUT) , SPECIFIED ND3, ND38 CAPACITY (EACH) 900 . ACTUAL RATED CAPACITY ND3, ND38 (EACH) 1040 - NOTE: ND SAFETY RELIEFS EXCEED PORV CAPACITY.
CONCLUSION: THE ND SUCTION RELIEF VALVES HAVE THE CAPACITY TO RELIEVE ANY CREDIBLE EVENT WHICH COULD OCCUR DURING MODE 5 DURING NC FILL AND VENT.
A - _.,.,.._.m .-_.._.,..,_...._-,,...,,-..............._.m..... ....,_,..-,,,,,,,.%m__,._,.m_ ,,,,
.. .. .-.. - -. - - - - - - -. - - - - - - - -- -.-.- -. -. _. - - -. -. - - ". .? . .. .' DESIGN ENGINEERING EVALUATION - RESULTS.
NC SYSTEM PRESSURE CALCULATED PEAK NC PRESSURE = 460 PSIA AT 114 F . . T.S. 3.4.9.1 LIMITS APPROX. 530 PSIG AT 114 F - ' , k CONCLUSION: NC SYSTEM ALLOWABLE PRESSURE WAS NOT - EXCEEDED, NOR COULD IT HAVE BEEN EXCEEDED =f SINCE AT LEAST ONE TRAIN OF ND SUCTION RELIEF VALVE WAS AVAILABLE AT ALL TIMES.
.
i' . ! [.- - , .
1
4 . -e *s -- - ..,......--,----...e--
.-.--e --- ---, -
- e _
_ _.-. - - -.. -. - _ _ - _ ___ _ - _ -__ _ _ -- _. ___ _ __ .. , t . ..: . ! DESIGN ENGINEERING EVALUATION SAFETY SIGNIFICANCE ! ~ NC SYSTEM PRESSUR5 i . LTOP MODE OF PORV'S WAS INOPERABLE (2 TRAINS) PER T.S. 3.4.9.3 . DUE TO INSTRUMENT ROOT VALVE ISOLATION.
' ( MET T.S. ACTION STATEMENT BY DEPRESSURlZING & PROVIDING 4.5 - -. ' SQUARE INCH VENT WITHIN 8' HOURS.
[ ONE TRAIN ND IN OPERATION PER T.S. 3.4.1.4.2 THROUGHOUT THE . EVENT, OTHER TRAIN OF ND WAS OPERABLE.
, PROCEDURES REQUIRE THAT EACH OF 4 RCP'S HAVE BEEN RUN PRIOR TO ' .- , DECLARING " LOOPS FILLED".
ONE TRAIN ND WOULD STILL BE IN OPERATION IN MODE 5 WITH LOOPS . FlLLED PER T.S. 3.4.1.4.2.
. . NC PRESSURE INSTRUMENT ROOT VALVE POSITION WOULD HAVE BEEN . DETECTED BY PERFORMING ROOT VALVE CHECKLIST PRIOR TO MODE 4.
LTOP MODE OF PORV'S WOULD HAVE BEEN ASSURED PRIOR TO . ENTRY TO MODE 4 AND NC TEMP GREATER THAN 200* F BY PERFORMING THE MODE 4 CHECKLIST.
, CONCLUSION : 1. LTOP OPERABILITY OF PORV'S WOULD HAVE BEEN ASSURED PRIOR TO 1 SOLATION OF ND SUCTION RELIEF VALVES.
$ j 2. WHiLE NC IS OPEN TO THE ND SYSTEM, SUCTION RELIEF I VALVES ARE LARGER CAPACITY THAN PORV'S AND THUS SATISFY ANY LTOP CONDITION.
, i i l
4 - -,,. . - ,, . ..... -... . . ... . ...... -__________,___________,____m_____
_ _ _. _. _... _ _ _... _ _ _. _ _ _ _. _ _. _. _ _... _ _ _ _ _ _. _ _ _. _ _ _.... _ _. _ _ _ ' .
.. . -. , , DESIGN ENGINEERING' EVALUATION - METHODOLOGY , , ND SYSTEM PRESSURE I . l-1.
GIVEN CORRECTED ND PUMP 1 A DISCHARGE PRESSURE = 625 L PSIG.
I ' i-2.
DETERMINED THE BOUNDS OF PRESSURIZATION EVENT l (ND PUMP 1 A UP TO 1ND26,1ND27,1ND28A,1NV135 AND - f - ALL NORMALLY. CLOSED VALVES).- . . i 3.
RESEARCHED THE MAXIMUM DESIGN CONDITIONS OF THE' j-FOLLOWING PIPING AND COMPONENTS INSTALLED PER FLOW DIAGRAM DESIGN CONDITIONS:-
, ! i PIPING MATERIALS,-INCL. FITTINGS AND~ FLANGES- - l VALVES - , ! ' INSTRUMENTS, INCL. THERMOWELLS, - TR ANSMITTERS, GAGES-MECH. EQUIPMENT, INCL. PUMPS AND HX - . i '
r ! . .., - - +..... -. -. - - - - _ -. _. -, . - -. ,,,. _ _ - _ -., _ _. - - - - -,.. _ - - -,.. -
._ _ ____ ______-_____- _-__ - - _ - - _ ___ _ _ _ _
- -
. CATAWBA-NUCLEAR STATION - . BOUNDARIES OF ND PRESSURE: EVENT - - ' . -. . . I b ND: ' y To cces . IND275 [O ' r , 'INO28A ~ ' '450 PSIG SET PRESS.
d O 600 PSIS ET PRESS.
, PRT C ND 3 - RHTg IND34
ND SPRAY PT ND [[ = IFR0tt NOT PUMP INS 43A l. LES S 'l ' .- M=
IA > h ' TO C. L* l.
R = d' =
= = l NDC ND2A IND26'
i i: INO24A X ..X
NO i LETOOWN ' O ~ ' . 38vl35 ) [IN0588 ~X
, i-1f.TO TRAIII18 1 I ' . _ - . . -. - . - . . . _ - -. - .-. .
.... -. -. -. - -. -.. .. - _ -. _ - - _ -.. ..- -. - _ - . - - -. -... ____ _ _ - _ _ -- _ _ _. _ . . DESIGN ENGINEERING EVALUATION SUMMARY RESULTS - ND PRESSURE EVENT-
, CATAWBA UNIT 1 . ' . SYSTEM / COMPONENT FLOW DIAGRAN NAX DESIGN COND* HYDROTEST PEAK CALC.
. REMARKS ' i, PORTION OF SYSTEN DESIGN PRESSURE PER PIPE / VALVE SPEC.
PRESSURE ~ PRESSURE-ND PUNPS I A,15 525 PSIG N/A 460 PSIG NOT OVERPRESSURIZED l SUCTION PIPING ND PUNP IB NOT OVERPRESSURIZED DISCHARGE PIPING 600 PSIG 792 PSIG 900 PSIG 460 PSIG SINCE ND PUNP IB - , (AT 120 F) WAS NOT OPERATING . ' \\ ND PUNP I A
PEAX PRESSURE WAS
' DISCHARGE PIPING 600 PSIG 792 PSIG 900 PSIG 625 PSIG WITHIN BOTH HYDRO & !
( AT 120 F) PIPE SPEC DESIGN COND.
' ND PUNP 1 A PEAK PRESSURE WAS'
CASING 600 PSIG N/A ' 936 PSIG 625 PSIG WITHIN HYDROTEST !
NECHANICAL SEAL 600 PSIG N/A 1200 PSIG , PRESSURE' .; a ND HX 1 A PEAK PRESSURE WAS . ! SHELL SIDE (KC) 150 PSIG N/A 225 PSIG.
N/A WITHIN HYDROTEST !. TUBE SIDE (ND) 600 PSIG N/A 503 PSIG - 625 PSIG PRESSURE , j INSTRUNENTATION 600 PSIG VARIOUS VALUES FOR. M/A 625 PSIG PEAK PRESSURE WAS
" SAFE WORKING PRESSURE" WITHIN SAFE WORKING . (ALL ABOVE 625 PSIG) PRESSURE ] ! VALVES 600 PSIG VARIOUS. VALUES _ PER N/A 625 PSIG PEAK PRESSURE WAS i VALVE SPEC.
WITHIN MAXINUM DESIGN i , (ALL ABOVE 625 PSIG) CONDITION PER VALVE SPEC.
~ ! ! .' . l a NO CREDIT TAKEN FOR ADDITIONAL MARGIN AVAILABLE UNDER CODE ALLOWABLES f
! - -,
e ' ) I . - -. . -. . - - -- - -- ~~ , +, s a ms
. . t ,.- ., d~ DESIGN ENGINEERING EVALUATION - CONCLUSIONS . NC / ND SYSTEM PRESSURE ND SUCTION RELIEF VALVES PROVIDED COLD OVERPRESSURE
PROTECTION FOR NC SYSTEM AS DESIGNED
- -
\\ ND SUCTION RELIEF VALVES PROVIDED OVERPRESSURE.
' .
PROTECTION FOR ND PlPING AND COMPONENTS WITHIN , NORMAL SAFETY RELIEF VALVE SET PRESSURE +10% , ACCUMULATION ALLOWED PER-ASME lli.
540 PSIG + 54 PSIG = 594'PSIG FOR ND SUCTION' - - LINE 600 PSIG + 60 PSIG = 660 PSIG FOR ND i . - , DISCHARGE LINE SIGNIFICANCE: EVEN THOUGH PRESSURIZER PORV'S WERE UNAVAILABLE DUE TO INSTRUMENT ROOT VALVE l-CLOSURE, ND SUCTION RELIEF VALVES RESPONDED PROPERLY TO PRESSURE EXCURSION CAUSED BY , - CHARGING PUMP OPERATION.
i OPERATOR RESPONSE WAS AS DESCRIBED IN THE FSAR. SYSTEM RESPONSE WAS AS DESCRIBED IN . THE FSAR.
FROM A DESIGN ASPECT,-THIS PRESSURE ' EXCURSION WAS OF NO SAFETY SIGNIFICANCE, i
k' l l . - - . - -. - .... -.. -
_ _ _ _ _ __ _ - o ..
REPORTABILITY ISSUE QUESTION: DOES THIS EVENT, ISOLATION OF LTOP, FALL IN THE REPORTABILITY CATEGORY, "AN EVENT OR CONDITION WHICH ALONE=COULD PREVENT FULFILLMENT OF A SAFETY
FUNCTION REQUIRED FOR MITIGATION OF AN ACCIDENT"7 s DUKE'S POSITION: IT DOES NOT MEET THIS CATEGORY'S INTENT , BASIS: ON THE DAY OF THE EVENT, OUR REASONS FOR NOT REPORTING THIS UNDER-THE ABOVE CATEGORY ARE AS'FOLLOWS: 1.
WE DETERMINED LOSS OF LTOP LEADING TO- ! . OVERPRESSURIZATION WAS NOT A' CHAPTER 6 OR 15 ACCIDENT.
THIS CAUSED US TO QUESTION THIS EVENT , FITTING THIS CATEGORY.
l 2.
NUREG-1022 STATES, "IF THE CONDITION IS NOT , PROHIBITED BY TECH SPECS IT'IS NOT REPORTABLE".. AS: TECH SPECS ALLOWS THIS CONDITION FOR 8 HOURS AND OUR EVENT MET THE ASSOCIATED ACTION STATEMENT ! IT WOULD NOT BE REPORTABLE.
3.
ASSUMING EQUIPMENT INOPERABILITY THAT COULD LEAD TO EVENTS OTHER'THAN CHAPTER 6 AND 15 " ACCIDENTS" ARE REPORTABLE, THIS PRESSURIZATION TRANSIENT WOULD STILL'NOT BE REPORTABLE.
AS RUSS AND OTHERS HAVE INDICATED - AN " ACCIDENT" CONDITION WAS NOT CAPABLE OF- .. BEING PRODUCED EVEN WITH LTOP ISOLATED.
BOTH ND SUCTION RELIEFS WERE~AVAILABLE - + 1800 GPM CAPACITY FSAR DESCRIBES OVERPRESSURE PROTECTION + ' ' TO INCLUDE ND SUCTION RELIEFS IN THIS CONDITION.
l NRC'S SER INDICATES ND RELIEFS AS A
l BACKUP TO LTOP ADMINISTRATIVE CONTROLS WERE IN PLACE TO + LIMIT POSSIBLE INPUT TO THE SYSTEM.
(REACTOR COOLANT PUMPS, SAFETY-INJECTION PUMPS, AND ONE CHARGING PUMP TAGGED OUT.)
EVEN AFTER ONE TRAIN OF ND WAS ISOLATED
THE ONE AVAILABLE RELIEF COULD HANDLE THE FULL INPUT OF ONE CHARGING' PUMP WITHOUT INCREASING ND/NC PRESSURE.
i THUS, THIS EVENT OR CONDITION "ALONE"
! COULD NOT HAVE PREVENTED FULFILLMENT OF ' THE SAFETY FUNCTION (OVERPRESSURE PROTECTION)-NEEDED TO MITIGATE THE CONSEQUENCES OF AN ACCIDENT.
- . _ - . __ _ . _ ~ __ _
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y N, .. . ' -) , . l ,
$ L ' L i
. + !
- . T Supp'ose during shutdown we' are doing maintenance on both SI pum - - - - , ~ . '" 7.10 Since this system is not required to be' operational,.-I assume-this1 situation is not reportable? I also understand that if something , happens that-would cause-both SI pumps not to be operational at power, " , - that would be reportable.
- Is that correct? Answer: Removing both.SI pumps from service' to do maintenance.is not
l reportabledf the resultino s;/ stem'tonfigurationL is?n6t ~ prohibited '- l by the clant's Technical-Spect rications.
I However,:1f a sttuation ! is discovered during the maintenance that-could have caused both pumps to fail (e.g. they are both improperly lubricated). then that l condition is reportable even-though the pumps'were not required.to-be operational at the time that the condition was discovered.- As another example, suppose the scram breakers were tested-during shut-down conditions, opening times wer.and it was found that for more than one breaker,.
e in excess of those specified, or that UV trip attachments were: inoperative.: reportable in an LER.
Such potential generic problems-are '
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.-. ,-. . - --_ -. - _ ~ _ - - - -. _ _ _ - - A' =. . .- . . REACTOR COOLANT SYSTEM l 0VERPRESSURE PROTECTION SYSTEMS ,; . LIMITING CONDITION FOR OPERATION- !
3.4.9.3 At least one of the following overpressure Protection Systems shall! be OPERABLE: ' , Two power operated relief valves (PORVs) with a lift setting of a.
i less than or equal to 450 psig, or.
.- ~ ' i b.
The Reactor Coolarit System depressurized with e Reactor Coolant:
System vent of greater than or equal to 4.5' square inches.
. APPLICABILITY: MODE'4 when the temperature of any Reactor Coolant System cold - L . leg is less than or equal to 285'F, MODE 5 and MODE 6 with the reactor vessel ' head on.
ACTION:- O With one PORV inoperable, restore the inoperable:PORVfto OPERABLE a.
Q status within.7_ days or depressurize and vent the Reactor Coolant: - System through at least a 4.5 square inch vent within'the.next'
,
8 hours.
- , \\ . b.
With both PORVs inoperable, depressurize and vent the Reactor Coolant' l System through at least a 4.5 square inch. vent within 8 hours.
- ' ' \\ . I In the event either the PORVs or the Reactor Coolant Sy c.
are used to mitigate a Reactor Coolant System pressure, stem vent (s)- transient, a Special Report shall be prepared and submitted to the Commission-- pursuant to Specification 6.9.2:within 30 days.
The report shall l describe the circumstances initiating the transient, the effect of.
' the PORVs'or Reactar Coolant System vent (s) on the transient, and . any corrective action necessary to prevent recurrence.
l d.
The provisions of Specification 3.0.4 are not applicable.
!
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i - )
L JATAWBA-UNITS 1&2.
_ . 1fa t-Jtt . . - - - ~.
-_.
- - Z1 1_ _ _ L.- , . .. ! .. CNS Residual Heat Exchanger Two residual heat exchangers are installed in the system.
The heat exchanger l design is based on heat load and temperature differences between reactor coolant and component cooling water existing twenty hours after. reactor shutdown when , the temperature difference between the two systems is small.
The installation of two heat exchangers in separate and independent residual heat removal' trains assures that the heat removal capacity of the system is-only partially lost if one train becomes inoperative.
The residual heat exchangers are of the shell and U-tube type.- Reactor coolant circulates through the tubes, while component coo!ing water circulates through.
the shell. 'The tubes are welded to the tube sheet to prevent leakage of reactor . coolant.
The residual heat exchangers also function as part of the ECCS (See Section 6.3).
Residual Heat Removal System Valves Valves that perform a modulating function are equipped with two sets of packings l and an intermediate leakoff connection that discharges to the drain header.
The design bases for the RHRS isolation values are Branch Technical Position i RSBS-1 and ICSB-3.
Manual and motor operated valves have backseats to facilitate repacki where required by valve size and fluid conditions. ge connections aren limit stem leakage when the valves are open.
Leaka provided- , 5.4.7.2.3 Control Each inlet line to the RHRS is equipped with a pressure relief valve sized to . relieve the combined flow.of all the charging pumps at the relief-valve set pressure.
These relief valves also protect the system from inadvertent over-pressurization during plant cooldown or startup.
Each valve has a reli,ef flow capacity of 900 gpm at a set pressure of 450 psig.-
The RHR suction relief valve design assumes that one RHRS train is isolated ' from the RCS thereby requiring the injection flow from two charging pumps be accomodated by one relief valve.
The combined flow delivered by two charging pumps is not, however, twice that of a single pump since the two pumps deliver to a common charging header.
The combined flow has been calculated and found to be less than 600 gpm at the valve set pressure of 450 psig.
' _ Two limiting situations were analyzed to confirm the capability of the RHRS relief valve to prevent overpressurization in the RHRS.
The first consists of the Reactor Coolant System (RCS) in the initial phase of the RHRS cooldown.
RCS temperature and pressure are 350 F and 450 F psig , respectively and one charging pump is in operation.
The operator initiates - . 5.4-30 l _ _.._, ' _ . .. ~
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CNS
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U RHRS operation by opening one suction line and starts the pump.
At this point a complete loss of plant air occurs, the charging line flow control valve fails open and the low pressure letdown flow control valve fails closed.
The maximum charging pump injection rate is 400 gpm for Catawba at 450 psig RCS pressure.
To avoid overpressur,izing the RHRS, the suction relief valves must pass these flows at set pressure plus accumulation.
The second consists of < the RCS in the last part of cooldown.
RCS temperature and pressure are less than 200*F and 450 psig, respectively.
The additional conservatism of a second charging pump in operation was added since the RHRS is used for extended periods below 200*F (cooldown of a second charging pump in operation was added since the RHRS is used for extended periods below 200'F) (cooldown from 350'F to 200*F occurs in approximatel The combined flow of the charging pumps is less than 600 gpm.= y 5 hours).
Each relief valve has a relief flow capacity of 900 gpm at a set pressure of 450 psig..This capacity provides adequate protection for the RHRS overpressurization.
'll credible events were examined for their potential to overpressurize the A RHRS.
These events included normal o transients, and abnormal occurrences.perating conditions, infrequent , The analysis confirmed that one relief valve has the capability to maintain the RHRS maximum pressure within code , limits.
. . - -Each discharge line from the RHRS to the RCS is equipped with a pressure relief valve to relieve the maximum'possible back-leakage through the valves separating Os the RHRS from the RCS.
Each valve has a relief flow capacity of 20 gpm.at a set pressure of 600 psig.
These relief valves are located in the ECCS (See Figures 5.4.7-1 and 5.4.7-2).
The fluid discharged by the suction side relief valves is collected in the pres-surizer relief tank.
The fluid discharged by the d.ischarge side relief valves is collected in the recycle holdup tank of the boron recycle system.
The operator is alerted to the' lifting of the RHR relief valves by increasing . i pressurizer relief tank level, pressure and temperature indications and alarms ! or by increasing recycle holdup tank level indication and alarm.
The design of the RHRS. includes' two. motor-operated gate isolation valves in series on each inlet line between the high pressures RCS and the lower pressure . RHRS.
They are closed during normal operation and are only opened for residual heat removal during a unit cooldown after the RCS pressure is reduced to , ' .approximately 385 psig and RCS temperature is reduced to approximately 350 F.
During a unit startup the inlet isolation valves are' shut after drawing a bubble in the pressurizer and prior to increasing RCS pressure above approxi- ~ ' mately 385 psig.
These' isolation valves are provided with both " prevent open" and " auto-closure" interlocks which are designed to prevent possible exposure of the RHRS to normal RCS operating pressure.
The two inlet. isola. tion valves in-each subsystem are separately and independently interlocked wi.th pressure signals to prevent their being opened whenever the RCS pressure is greater than approximately 385 psig.
The two inlet isolation valves in each subsystem , are also separately and independently interlocked with pressure signals to Q automatically shut if RCS pressure increases to 600 psig during a plant V star. tup.
A reverse check valve-(spring loaded lift check) in parallel with . 5.4-31 -- . . - - .- .. - -
._.
_ ___ _ _ _. _ _ _ _ _ _ . _ _ - _. _ _ _ \\ ' .. . . . a
CNS l tail the types and number of pressure relief devices employed, relief device description, locations in the systems, reliability history, and the details of i ! the methods used for relief device sizing based on typical worst case transient , conditions and analysis data for each transient condition.
As stated in i WCAP-7769, Topical Report Overpressure Protection for Westinghouse Pressurized ' Water Reactors ( Revision 1), the pressurizer safety valve is sized based upon , j the peak surge rate into the pressurizer following a complete loss of load without reator trip and with energy relief only through the steam generator i
safety valves and pressurizer safety valves.
The actual safety valve capacity must be greater than or equal to this required capacity.
The ratio between the actual s'afety valve capacity and the peak surge rate into the pressurizer , . is an entry in Table 2-2.
If this ratio is greater than the ratio for that ' type of plant in Table 2-2, then the assumptions in WCAP-7769 envelope the: I plant under consideration.
This is the case for Catawba and the pertinent . values are given in FSAR Table 5.2.2-1.
The description of the analytical model used in the analysis of the overpressure protection system and the basis _' for its validity is discussed in Reference 3.
The protection against low temperature overpressure transient fanditinnE is ~ nenvidQb a combination of interlocks, design features and(administrative]
(procedures The low temperature overpressure protection is enablea only on
coincidence of: 1.
RCS temperature decreasing to a predetermined set point; , and 2.
The operator placing the key-lock switch to the LOW PRESSURE position.
l Once the low temperature overpressure is enabled the PORV requirements are provided via the-plant instrument air or the Nitrogen supply from the cold leg accumulators whichever has the higher pressure (normally instrument air has h'igher pressure).
The PORVs are thus provided a seismically qualified source of Nitrogen from the cold leg accumulators when the cold leg accumulators are isolated from the RCS at low temperature and pressure and not required for ECCS function.
The Nitrogen supply valves NI438A and NI439B are shown in , figure 6.3.2-2.
Thus a combination of interlocks and design features assure a supply of Nitrogen to the pressurizer PORVs under the conditions requiring low temperature overpressure protection without compromising the effectiveness of the. cold leg accumulators of the ECCS.
, Administrative procedures associated with reducing the potential for , overpressure events utilize a sequence of operations which ensure that a '
pressure relieving path is always available.
A steam bubble is formed in the pressurized early in the startup sequence.
This provides a cushion against pressure surges and overpressurization when the Reactor Coolant System is isolated from the Residual Heat Removal System.
' , . The Low Temperature Overpressure Protection (LTOP) would remain functional in . the event of a postulated single failure.
There are two independent trains and associate.d PORV to relieve pressure at low temperature.
In the event of a DC power bus failure supplying Train A PORV, the Train A PORV would fail close and the normal letdown flow path would be automatically isolated.
A postulated single failure of train B PORV (closed position) would not fail all mitigating systems for assuring LTOP.
. 5.2-4 _ __m---
- . - -. . . -. -. . . . - - . .. - . . . f CNS Wheii~the~ Re s i dus1 'Heit" Re'moilil (RHR) l sis tem i s 'i n 5% ados /Et@cE ori ~ relief valve's provide "oveirpres'sure protectio'n...These relief.l valves 7are " sized J to relieve.the_ combined flow of all the' charging pumps at~.their' set' pressure' ~ ~~ of 450 psig (see S.ection 5.4.7.1).
When the RHR system isTisolate"d from the 7 Reactor Coolant System (RCS), a pressurizer steam bubble is maintained? If the postulated scenario were to occur under those conditions, adequate time is ~ available for the operator to mitigate the event.
The Technical Specifications for Catawba impose limiting conditions of operation as well as surveillance requirements to assure the validity of the assumption' used in the low temperature overpressure design analyses.
Catawba s Nuclear Station is in conformance with the applicable items of Branch Technical Postion RSB 5-2.
A description of the pressurizer safety valves performanc'e characteristics along with the design description of the incidents, assumptions made, method of analy-sis and conclusions are discussed in Chapter 15.
. 5.2.2.3 Piping and Instrumentation Diagrams Overpressure protection for the Reactor Coolant System is provided by pressurizer safety valves shown in Figure 5.1-2.
These discharge to the pressurizer relief tank by a common header.
/ The main steam safety valves are discussed in Section 10.3 and are shown on Figure 10.3.2-1.
5.2.2.4 Equipment and Component Description The operation, significant design parameters,' number and types of operating l, cycles and environmental qualification of the pressurizer safety valves are discussed in Section 5.4.13.
A discussion of the equipment and components of the steam system overpre'ssure system is discussed in Section 10.3.
' 5.2.2.5 Mounting Westinghouse provides Duke with installation guidelines and suggested physical layout.
This information is transmitted to Duke as part of a systems standard ~ design criteria document.
Duke is required by Westinghouse to limit the piping reaction loads on the safety valves to acceptable values.
Westinghouse provides mounting brackets on the pressurizer which can be used to support the pressurizer safety valves.
Duke is responsible for the design and mounting of the supports for these valves.
They are also responsible for deter-mining reactions on the pressurizer mounting brackets.
Design and installation details for the pressure relief devices are provided in Section 3.9.3.3.
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5.2-5 - __ _ _ _ _ _ .. _ _.
. _ _ . _ _.._ _ _, ' , . .. a , i j' to enable the PORV low pressure set point.1 Should a pressure excursion occur
with the low pressure mode enabled when the plant temperature is below the h temperature set point, system pressure would be limited to acceptable values, j i and excess mass would be relieved to the pressurizer relief tank.
An annunci- , ator in the control room would alert the operator to system overpressure.
g i The PORVs and associated block valves are required to have safety grade emer- , , . gency power supplies in accordance with Item II.G.1 of NUREG-0737.
Section' ' y 8.4.12 of this SER provides a discussion of Catawba's compliance with this requireuent.
As' a backup to the low-temperature overpressure protection system, the~ residual
heat removal system (RHRS) has two suction relief valves with a capacity of 900 gpm each at a set point pressure of 450 psig. :The' relieving' capacity of ) each valve is adequate to relieve the' combined flow of the two centrifugal' charging pumps.
The RHRS suction. relief valves-provide overpressure protection after the RHRS is put into operation and the RHRS suction isolation valves are open at-an RCS pressure of less than 425 psig. - Also, operating procedures' require that the operator lock out the cold-leg accumulator isolation valves in the closed position during shutdown.
The applicant has discussed a' postulated failure of a'dc power bus that would - initiate.a potential. low-temperature overpressure condition by both isolating ,_. letdown and disabling one train of the low-temperature overpressure protection , ' - system, coupled with the single failure (closed) of the PORV in the unaffected train. The applicant has stated that the RCS would be protected by RHR suction side relief valves when the RHRS is in operation and by alarm-initiated operator action when the RHRS is isolated.
To ensure adequate time for operator action.
l the applicant's operating' procedures call for a pressurizer bubble to be main- , l .tained when the RHRS is isolated.
This steam bubble will be'of sufficient size l to allow at least 10 min after the operator is alerted to the transient to ' terminate the worst possible transient under these conditions-without violating , < Appendix G limits.
This steam bubble size will be specified in the Catawba Technical Specifications.
/ i Since the PORVs are equipped with a nonsafety grade air supply, a backup supply i of nitrogen is provided to two of the PORVs through seismic Category I piping i and seismic Category 1 motor-operated valves-(MOVs)'and check valves connected l to the nitrogen space of two of the four cold-leg accumulators. The applicant ! stated that the two MOVs, 438A and 439B, are supplied by separate 1E power sources and stated that they are qualified to operate following the design-basis i i LOCA.
These MOVs are kept closed until the key-lock switch is placed on the l
low pressure protection mode of operation, at which time they open.
The nitro- [ gen is passed through a pressure reduction valve that reduces it to a pressure lower than that normally provided by the air system. Therefore, the nitrogen , l
supply is only called upon if the air supply pressure substantially drops below
its normal value.
In conformance with BTP RSB 5-2 and SRP Section 6.3,- the applicant is requested to commit to the following: ' E (1) test the low temperature overpressure protection system to ensure its f operability before each shutdown s Catawba SER 5-5 ,>ll - - - ._
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RESULTS: 1.
EVENT DID,NOT SEEM TO FIT THE CATEGORY.
2.
AS THE EVENT BECAME KNOWN AND WAS-DISCUSSED WITH THE' RESIDENTS THEIR INTEREST AND THE REGION'S BECAME EVIDENT.
3.
EVEN WITH THE EVALUATION COMPLETED, WE DECIDED TO MAKE AJ COURTESY REPORT TO ENSURE FULL UNDERSTANDING OF THE EVENTS-OCCURRENCE.
i4.
SPECIAL REPORT PREPARED AND WILL BE SUBMITTED TOMORROW.- . 5.
RECOGNIZED THE'NEED FOR A " TEAM" APPROACH WITHIN COMPLIANCE
AND OPERATIONS AT EACH STATION IN THE DUKE SYSTEM-(SIMILAR . < .TO OUR RECENT' APPROACH ON EAL's) TO-EVALUATE CONSISTENT AND
RESPONSIBLE REPORTING.
>
6.
NRC WORKSHOP IS PLANNED FOR LATER THIS YEAR TO UPDATE THE UTILITIES =ON NRC'S NEEDS AND-LESSONS LEARNED SINCE-1984 ON REPORTING.
WE'LL BE THE FIRST IN LINE TO' SUPPORT THIS: EFFORT.
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- CATAWBA NUCLEAR--
STATION , SUMMARY 'OlAGRAM-VALVE INTERLOCKS - u , i ND-PUMP SUCTION -
,. FROM NC-SYSTEM.
- ! LOOP C HOT LES - i 1 f
( FWST ) XIND368*
. l TO NS PUMP' .t i [lFW553 [lND37A g i
- -
. . i IND38- . o - ' [lNS18
v ! TO PRT- ! I
'E 'E - ' , . - ND PUMP
- $1NI1848 a
FROM CONTAIMMENT.- RECIRCULATION sump I- , - - ~ - . _ _ _.... -. _., _. ,. _ . . . .... - .. . -. . _ _ -.. _ _.. _......,. - _..,. _ _.., _ _.,. - - -
__ _ _. _, _.. _ _... ? ] ~ . n.
,
- -
L SUMMARY DESIGN BASES 3OR INTERLOCKS VALVE OPENE VALVE CLOSED ~
INTERLOCK INTERLOCK PROVIDES INTERLOCK PREVENTS l' .FW553 FLOWPATH FROM FWST - INADVERTENT .q !' FOR NORMAL REFUELING FLOODING OF- > i OR INJECTION MODE-CONTAINMENT SUMP l FROM FWST DURING' , , ' - NORMAL REFUELING - OR INJECTION MODE - L l . ND365 FLOWPATH FROM NC PRESSORIZED NCSYS'- ' ! ND37A SYS TO ND PUMPS FOR - FROM FLOODING NORMAL COOLDOWN CONTAINMENT SUMP.. , l REFUELINGi AND OR BEING RELEASED ! ! STARTUP VIA FWST- ! - l . Nt1848 RECIRCULATION MODE INADVERTENT j-FROM CONTAINMENT FLOODING OF.
. i SUMP TO ND AND NS - CONTAINMENT SUMP-I PUMPS FROM FWST DURING NORMAL COOLDOWN.
OR INJECTION MODE l . NSIB FLOW FROM FWST TO NS PUMP OVER , I NS PUMPS PRESSURIZATION. - . FROM PRESSURIZED , , e NC SYS. i FLOW FROM CONTAINMENT SUMP < TO NS PUMPS .
. . .
1 . . ... . . . .. . .. _ _.
_ _ .. . . ..
. < l VALVEINTERLOCKS ND PUMP SUCTION VALVES USED IN ALIGNING THE ECCS ARE INTERLOCKED TO PREVENT INADVERTENT MISALIGNMENT.
l.
FW55B CLOSED ND368 OPEN' , ' N D37A - OPEN ' N11848 OPEN.
- i FW558 QEEtt ONLY IF... N1184 B CLOSED NSIB CLOSED.
FW558 CLOSES AUTOMATICALLY WHEN N1184B IS QP.fJf.
' i 11.
ND368, ND37A CLOSED N11848 OPEN ' ND368,ND37A QEgli ONLY IF... FW558 CLOSED N11848 CLOSED ! . l ND368, ND37A INTERLOCKED TO CLOSE AUTOMATICALLY FOR LTOP.
' lil.
N11848 CLOSED ' FW55B - OPEN [ ND368 OPEN , ND37A ' OPEN . $ j N1184B QEEli ONLY IF... FW55B CLO3ED !- ND36B OR CLOSED
ND37A' , b< a $
_ _. . _ ._- _. __ _._.. . .. _ _ - _ _ _ _ _ _ _ _. _ _ ___ _ _. _ _ o , - , .. 1' ., ! NSM: CN-10942 & CN-20330 , ' PURPOSE:. IMPLEMENTED AS A RESULT OF FINDINGS IN.
NRC IEB 85-03 "MOV COMMON MODE. FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS" . WHAT IT DID: TORQUE SWITCHES ON CERTAIN VALVE OPERATORS $ WERE BYPASSED DURING-VALVE-OPENING'TO. ENSURE SUFFICIENT TORQUE WAS AVAILABLE TO OPEN VALVE EFFECTED VALVES: 23 VALVES PER UNIT- , . SYSTEMS INVOLVED: AUXILIARY FEEDWATER OhhhhhhlNkl.
D Y HEAT-REMOV ' ' NU EAR SPRAY MAINSTEkhh hM0 SPHERE ' M{E{EAM I \\ - H .
...... - .. .. ....... -. _. _ - .,.s.
., _ _, , -..
PERMISSIVES REQU:: RED TO' OP$N NJ036B
l I I ' 1.
FW 055B o:.0 sed.
t ! 2.
.N::184 3 c:.osed
3.
NI13.6B closed 4.. NS0383 closed 5.. NC press < 375 psig - .. e e -..- ..., .. .. . .
. _ _. . . __ ____.
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. 'L J . PO R BEFORE NSM CN-10942_ SU LY )CONTROLROOM'PUSHBUTTON " ' ' cc BREAKER 4 ' . " PY/403AX - ELECTRICALLY-OPERATED RELAY CONTA di FWO558 - MECHANICALLY OPERATED LIMIT SWITCH - ' P NI1848 - MECHANICALLY OPERATED LIMIT SWITCH ) BC - ELECTRICALLY.0PERATED RELAY CONTACT -
_ . _. LCh4TA'C i ' NS0388 - MECHANICALLY OPERATED LIMIT SWITCH OTO - fMOTORSTARTERRELAY VALVE i
ND036B , . l ' . R AFTER NSM CN-10942 ICONTROLROOMPUSHBUTTON BREAKERl ' ,
- PY/403AX - ELECTRICALLY OPERATED RELAY CONTA
!
- DA - ELECTRICALLY OPERATED RELAY CONTACT
, p NI1848 - MECHANICALLY OPERATED LIMIT SWITCH T MTDR "I BC - ELECTRICALLY OPERATED RELAY CONTACT ~~" ~
" CONT CTS FE - ELECTRICALLY OPERATED RELAY CONTACT
- . VALVE fMOTORSTARTERRELAY OTOR ND0368
_ _ _ _ _ _ _ _ _ - - _ - - - - - - - - - - - - - - - - - - - - - . . .
s-10 CFR 50.59 EVALUATIONS 'l , o PURPOSE: L 10 CFR 50.59 EVALUATIONS PROVIDE A HIGH DEGREE OF CONFIDENCE: l THAT THE HEALTH AND-SAFETY OF THE PUBLIC.IS NOT COMPROMISED- , l BY DESIGN CHANGES. :THIS IS DONE BY PRESERVING THE BASIS OF THE SAFETY ANALYSIS REPORT
l I METHOD:' ' "10 CFR 50.59 IS BASED UPON WHAT IS IN THE SAFETY ANALYSIS-l REPORT."
IN PERFORMING A 10 CFR 50.59 EVALUATION ON A DESIGN MODIFICATION A DETAILED COMPARIS0N IS MADE BETWEEN l THE MODIFICATION AND THE RELEVANT REQUIREMENTS IN THE FSAR, TECHNICAL SPECIFICATIONS, SER & SUPPLEMENTS AND OTHER RELATED DOCUMENTS (SUCH AS STANDARD REVIEW-PLAN AND DUKE - " ,
! NUCLEAR GUIDES).
THIS COMPARISON DETERMINES THE IMPACT ~
0F THE MODIFICATION ON THE RELEVANT'NRC REQUIREMENTS AND , DUKE COMMITMENTS REGARDING THE AFFECTED STRUCTURES, SYSTEMS, l AND COMPONENTS. THE PROCESS BY WHICH THIS COMPARIS0N IS
i
PERFORMED IS CONSISTENT WITH THE INDUSTRY'S POSITION ON " p
PERFORMING 50.59 EVALUATIONS.AS DESCRIBED IN.NSAC-125,. > GUIDELINES FOR 10 CFR 50.59 SAFETY EVALUATIONS.
BOTH ' NSAC-125 AND DUKE'S IMPLEMENTATION OF ITS' PRINCIPLES HAVE BEEN SUBJECTED TO NRC REVIEW.
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L i FAILURE MODE DETERMINATION FOR 10 CFR 50.59 EVALUATIONS i FAILURE MODES ARE: DEFINED AS THE-FAILURE" INTERACTIONS-BETWEEN o STRUCTURES.-SYSTEMS,~AND COMPONENTS-(SSC)~. THESE_ INTERACTIONS DO L NOT INCLUDE-THE LIMITLESS' NUMBER OF WAYS-A DEVICE MAY FAIL, AS . L LONG'AS.THE DEVICE FAILURE DOES NOT CREATE A SSC FAILURE THATEIS: ' NOT DESCRIBED-IN.THE FSAR.
EXAMPLE - A PUMP, WHICH IS COMPOSED OF MANY INDIVIDUAL DEVICES, ' FAILS TO START.
THE' FAILURE MODE IS'A-FAILURE T0' START AND IS NOT DEPENDANT'UPON THE~ CONDITION OF THE PART THAT FAILED, BUT ONLY THE END RESULT OF THE FAILURE (i.e. NO FLOW).
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. .. r i ( PROBABILITY DETERMINATION FOR 10'CFR 50.59LEVALUATIO " UTILIZE REASONABLE ENGINEERING PRACTICES, ENGINEERING t o ..g JUDGEMENT, AND PRA TECHNIQUES, AS APPROPRIATE,1 " TO DETERMINE CHANGES IN PROBABILITIES I } o. " A CHANGE IN PROBABILITY S0'SMALL OR THE UNCER , f . DETERMINING-WHETHER A CHANGE IN PROBABILITY i
ARE SUCH THAT IT CANNOT BE' REASONABLY' CONCLUDED , , ! THAT THE PROBABILITY HAS ACTUALLY~ CHANGED.(i.e.,~THERE IS NO CLEA {- TREND TOWARDS INCREASING THE PROBABILITY), THE CHANGE - NOT BE CONSIDERED AN INCREASE IN' PROBABILITY."
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.. , l ' . l P0kNhMgEDMAY 10, 1989 FROM CHARLES.ROSSI,-NRC, TO MS L ! - ' - i- $
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{ , CN-10942 50.59 DISCUSSION " FUNCTIONALLY,-THESE VALVES WILL OPERATE IDENTICALLY T0 THE - l-WAY THEY PRESENTLY OPERATE."
, ! o THE FW VALVES MODIFIED-BY THIS'NSM D0 FUNCTION EXACTLY-THE SAME AS BEFORE.
THE ONLY DIFFERENCE IS.THAT THE TORQUE SWITCH WILL NOT q l ENGAGE UNTIL LATER IN THE VALVE'S.OPEN STROKE.- -
"WITH THE NEW CONTROL CIRCUIT WIRING AND " TORQUE BYPASS. SWITCH"' I SETTING, THE VALVES SHOULD BE MORE RELIABLELIN ATTAINING THE '
DESIRED POSITIONS, WITHOUT' LETTING THE TORQUE SWITCH PROTECTION
DEVICE INTERFERE.
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} o THIS IS TRUE.
IMPROVING-VALVE RELIABILITY WAS THE PURPOSE OF'THE IE BULLETIN (IEB 85-03) AND THE SUBJECT NSM.
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"ALSO, OTHER INDICATIONS AND INTER' LOCKS ASSOCIATED WITH VALVE OPERATION, POSITIONS ETC., WILL NOT BE AFFECTED BY THESE CHANGES."
, o THE FW VALVES MODIFIED BY THIS NSM WILL CONTINUE TO HAVE THE SAME INTERLOCKS AND INDICATIONS.AS THEY DID PRIOR TO THE MODIFICATION.
l FOR EXAMPLE, THE PREVENT OPEN PRESSURE INTERLOCKS BETWEEN FW AND ND l CONTINUE T0-PREVENT ND VALVES 1B, 2A, 36B, 37A FROM BEING OPENED' - REMOTELY.IF NC PRESSURE EXCEEDS 385 PSIG OR IF THE ASSOCIATED FW l L VALVES (27A AND SSB) ARE'OPEN.
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. _ _ _ - - - -. .-- - _ . . _. _ _ _ _ _ _ - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - . .. y, "NO.NEW FAILURE MODES WILL BE CREATED AS A RESULT OF THIS ~ MODIFICATION."
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N0 NEW FAILURE MODES ARE CREATED.
ONLY THE FAILURE-MODES THAT [ ' EXISTED PRIOR TO THE MODIFICATION (i.e., -FOR FW-558, FW-27A, ND-2A ' ' ND-18) EXIST AFTER THE MODIFICATION.
FOR EXAMPLE, THE'FSAR FMEA ! 'FOR THE-ND SYSTEM ADDRESSES FAILURE 0F THE AFFECTED ND VALVES T0.
OPEN 0N DEMAND.
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" APPENDIX R AND ENVIRONMENTAL QUALIFICATION HAVE BEEN APPROPRIATELY
, CONSIDERED AND THERE ARE NO CONCERNS IN THESE AREAS."
i o ALL ELECTRICAL CABLE ROUTING BEING DONE IS CONSISTENT WITH THE , t l REQUIREMENT 3 0F. APPENDIX R.
ALSO, ALL EQUIPMENT T0.BE ADDED MEETS THE ENVIRONMENTAL QUALIFICATIONS FOR THE' AREA IT WILL BE ' LOCATED IN.
c i "N0 POWER SUPPLIES OR BREAKERS ARE AFFECTED EITHER."
o ALL FW VALVES IMPACTED BY THE NSM ARE POWERED BY.THE SAME ' MOTOR CONTROL-CENTER AND BREAKERS AS THEY WERE PRIOR TO THE.
) i MODIFICATION.
THEREFORE NEITHER WAS AFFECTED.. ! ! i i L l i- . ., ,, , . -... ~. ....,, , . - -.. - - - -.. . . -- -
.. -. . . .. - - - - - _. _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ _. > .. .. i INSPECTION REPORT ISSUES o 'AN ADDITIONAL FAILURE MODE,. LOSS 0F POWER-TO 1FW-55B', WAS INTRODUCED INTO THE ND SYSTEM.
l BY NATURE OF ITS REDUNDANT TWO TRAIN DESIGN, THE ND SYSTEM IS DESIGNED TO ACCEPT ALL MAJOR COMPONENT SINGLE-FAILURES WITH THE ONLY EFFECT BEING AN EXTENSION IN'THE REQUIRED C00LDOWN TIME.
1ND-36B WILL' FAIL TO OPEN ON DEMAND GIVEN THE LOSS OF. POWER TO.1FW-55B.
HOWEVER, FAILURE OF IND-36B TO OPEN-ON DEMAND.IS'A PREVIOUSLY ANALIZED FAILURE MODE (FSAR TABLE 5.4.7-3).
' E o THE RESULT OF THE NSM WAS THE INCREASE IN THE-PROBABILITY OF-A MALFUNCTION.
THE' SUBJECT NSM SIGNIFICANTLY REDUCED-THE PROBABILITY OE A-MALFUNCTION OF THE AFFECTED FW VALVES.
THE INTRODUCTION OF A i RELAY IN THE INTERLOCK CIRCUITRY DID INTRODUCE ANOTHER DEVICE WITH AN ASSOCIATED FAILURE PROBABILITY.
HOWEVER, FAILURE' PROBABILITY OF THE. RELAY IS BOTH SMALL AND WELL WITHIN THE UNCERTAINTY.
ASSOCIATED WITH THE FAILURE PROBABILITY OF IND-36B
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THE NSMs ON RELEVANT NRC REQUIRMENTS AND DUKE COMMITMENTS, l ,1 NOT INTENDED TO: .. ci o BE A DETAILED DESCRIPTION OF THE NSM
o BE AN' EVALUATION OF THE INSTALLATION,0F THE NSM o COVER-THE PROCEDURAL OPERATION, TESTING OR MAINTENANCE OF: ! THE STRUCTURE, SYSTEM, OR COMPONENT.
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o CREATE NEW INFORMATION, ONLY REVIEW FROM THE ABOVE I , . PERSPECTIVE' WORK ALREADY COMPLETED ON-THE NSM l - , l
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i - . . - - IIPO ' Design IIPO ' e Operations Engineering
Projects C e plaintenance Department Section e Chemistry e Performance e etc e Review problem e Assist in defining problem o Identify Problem e Provide solutions e Assist in defining solution e Evaluate Solution e Assist in implementation e Assist in implementing
- Assess impact of e Assist in station review modification modification of modification impact e Facilitate the review of the modification for impact on various station operations
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NUCLEAR STATHON MOBHMCATHON ' DOCUMENTA7 HON O INITI AL SCOPE DOCUMENT o FINAL SCOPE DOCUMENT Project Description - Functional Description ' Civil Description of Changes to Equipment or Components Civil Description of Changes to Pipe Supports Electrical Description of Changes to Equipment or Components . Mechanical Description of Changes to Equipment or Components PMT Objectives / Acceptance Criterta Summary Design Summary Dose assessment Civil - List of Calculations Electrical - List of Calculations , Mechanical - List of Calculations ' O FSAR CHANGES ' . O 10CFR50.59 EVALUATION O DESIGN PACKAGE EXPIRATION NOTICE o DESIGN COMPLETION NOTICE l i ! i
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NUCLEAR STATHON MOs#HCATHON l HNTERFACES ' J ! O STATION PROBLEM REPORT {
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i ) O SCOPE MEETING -
- O INITI AL SCOPE DOCUMENT (ISD) l
\\ O NPD R$ VIEW & ACCEPTANCE OF 150 ' , l t
i ) O PRE-DESIGN SURVEY i ' ' - ! i
CONCEPTUAL NPD/CMD MEETING i ,
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O DETAILED NPD/CMD MEETING .
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O INTEGRATED DESIGN REVIEW t
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. O FINAL SCOPE DOCUMENT s i ! O DESIGN COMPLETION NOTICE '
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.- o PRE-EVENT OTERATIONS NSM IMPLEMENTATION REVIEW Review of Paperwork
$0.59 Evaluation (Operation Group Tech Spec Perspective)
Functional Description
Review of flow diagrams /EE drawings
Get help from Projects and/or Design if necessary
Prepare procedure changes as needed
Prepara Operator Training packages as needed
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. a , ROOT CAUSE Inadequate review of the "EE" drawings by the Operations NSM implementation group for this modification due to personnel error.
(The NSM Implementation Process Procedure was not fully followed) REASONS FOR PROGRAM DEVIATION o Modification was very similar to others i o 70 modifications completed last year o 30 modifications completed this outage o Wording in the modification package led us to the conclusion that the interlocks were not affected
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CORRECTIVE ACTIONS Short Term I Changed Operations procedures to reflect the modification done on FW-SSB
that affected ND-36B.
, ' i Reviewed other NSM's of this type for similar problems e No problems found
Reviewed other electrical type NSM's implemented since January 1989
No problems found
, IIeld a meeting between station groups and Design to fully discuss and
understand the modification process Re-emphasized the requirement that the Operations NSM review will always
include a review of 'EE" drawings with the Operation NSM group Lonn Term
Projects accountable engineer will drive the interface of the NSM review
by the station groups and Design All affected station groups participate in the 80% design complete NSM > e meeting that will be enhanced tot Discuss the operational aspects of the modification.
- Ask standard questions for each modification to facilitate thought
Operations provide input at the modification initial scope meeting
i Establish an NSM audit committee to follow an NSM through the process to
determine other improvements . Revise the Operation Procedure for the NSM Implementation Process to
clearly state all components of the NSM package and the review requirementa for each
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I SUBSEARY Station / Design NSBI Process is an excellent program.
+ Operations NSBB Implesmatation Process has a proven
effective track record.
459 CEVNs/NSB8s were evaluated and no additional . problems were found.
corrective actions will further strengthen our NSIE . process.
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_- ~ .. u DESIGN ENGINEERING EVALUATION SAFETY SIGNIFICANCE 1ND369 FAILURE TO REOPEN FOLLOWING l SOLATION AUTO CLOSURE INTERLOCK OF VALVES ND18, ND2A, AND . ND369, ND37A WITH NC PRESSURE IS A SAFETY FUNCTION TO PRECLUDE GROSS OVERPRESSURIZATION OF THE ND SYSTEM PER FSAR SECTION 5.4.7.
, THE NSM DID NOT AFFECT THE CLOSURE CIRCULI OF ND- . ISOLATION VALVES. THUS THE ACI CLOSURE AS WELL AS MANUAL CLOSURE OF THESE VALVES (INTERLOCK SATISFIED) WAS UNAFFECTED.
FSAR TABLE 5.4.7 3 SHOWS THAT PAILURE OF ANY TRAIN B . VALVE OR INTERLOCK IS ACCEPTABLE SINCE 100% ' l . REDUNDANCY IS PROVIDED BY TRAIN A, INCLUDING THE ' FAILURE OF 1ND36B TO OPEN ON DEMAND. MANUAL OPENING OF THIS VALVE IS ACCEPTABLE AND 18 OF NO SAFETY SIGNIFICANCE. ADEQUATE TIME EXISTS PRIOR TO - i SIGNIFICANT NC HEATUP.
I I PROCEDURES DIRECT THE OPERATOR TO MANUAc.Y OPEN . 1ND388 AFTER VERIFYING THAT INTERLOCKS ARE SATISFIED.i
CONCLUSIONS: 1. OPERATIONS AND THE PLANT RESPONDED PROPERLY TO THE FAILURE OF 1ND368 TO OPEN.
2. FROM A DESIGN ASPECT, THIS EVENT WAS OF NO SAFETY SIGNIFICANCE.
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