ML20207B440
| ML20207B440 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 07/02/1986 |
| From: | Tucker H DUKE POWER CO. |
| To: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 8607180018 | |
| Download: ML20207B440 (19) | |
Text
_ _
040 DUKE POWER GOMPANY l'.O. nox 33180 CIIAHLOTTE, N.C. 28242 IIAL IL TUCKER t t j 1, J
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UM) 3DMM t-vers emessonsv NUCLEAS Peopt*CTION ibw h-July 2, 1986 Dr. J. Nelson Grace, Regional Administrator U.S. Nuclear Regulatory Connaission - Region II 101 Marietta St. NW, Suite 2900 Atlanta, GA 30323 Re: RII:PKV/PHS 50-413/86-15 50-414/86-16 Catawba Nuclear Station
Dear Dr. Grace:
Attached are our responses: to Violation 1413/86-15-02 ;and Violation 413/86-16-02, as identified in the subject Inspection Report.
Very truly yours, k
Hal B. Tucker LTP/05/jgm Attachments xc: NRC Resident Inspector Catawba Nuclear Station J
8607130018 860702 l
PDR ADOCK 05000413 hj G
DUKE POWEE COMPANY 4
RESPONSE TO VIOLATION 413/86-15-02
^
Technical Specification (TS) 3.2.1 requires that Axial Flux Difference (AFD) shall be maintained within a specified target band about a specific flux difference.
Action a.
4 requires that if indicated AFD is outside the target band and with thermal power greater than or equal to 90% of rated thermal power, within 15 minutes restore the indicated power to within the target band limits or reduce thermal power to i
less than 90% of rated thermal power.
Action b. requires that if indicated AFD is outside the acceptable operating limits of Figure 3.2.1 and with thermal power less than 90%
but equal to or greater than 50% of rated thermal power, reduce thermal power to less than 50% of rated thermal power within 30 minutes and adjust the power range neutron flux-high setpoints to less than or equal to 55% of rated i
thermal power within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
Contrary to the above, on March 17, 1986, AFD was not maintained in the specified target band and corrective t
action was not taken as required in that, rated thermal power was not reduced to less than 90% rated thermal power within 15 minutes as required by Action a, rated thermal i
power was not reduced to less than 50% rated thermal power l
within 30 minutes and the power range neutron flux-high setpoints were not reduced to less than or equal to 55% of i
rated thermal power within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as required by action b.
RESPONSE
1.
Admission or Denial of Violation Duke Power Company admits the violation 2.
Reasons for Violation if Admitted l
Mechanical Failure with contributory personnel error.
The failure of the Analog Scanner on the Operator Aid Computer (OAC) in conjunction with Operator at the Controls (OATC) relying solely on the OAC and the Excore Power Distribution Monitor (NUCL 06) Program for AFD j
indication resulted in the above stated violation of Technical Specification 3.2.1.
i 3.
Corrective Actions Taken and Results Achieved a.
Power was reduced to below 50%.
b.
Personnel began manual calculation of AFD %.
c.
The OAC was reinitialized, which returned the NUCL 06 program to service.
d.
Reactor power was held below 50% until there was no AFD penalty time accumulated in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
)
i e.
All the setpoints for Power Range NI were reset to 55%.
f.
AFD% was returned to within its required target band.
4 g.
A patch was placed on the NUCL 06 program that will cause automatic re-initialization when the analog scanner fails.
h.
The software of the NUCL 06 program was changed to give an alarm when the power range upper and lower e
channels' OAC points fail to change when the OAC updates their values.
i.
An Opera' tor Update was issued that stressed to Control Room Operators that the OAC should not be used as the primary indication for plant parameters where meters and recorders are available.
Also j
stressed was the importance of keeping shift supervision informed about concerns or questions
~on plant conditions, j.
An internal memorandum was routed to the responsible group that reminded personnel of expected responses to watch for during power changes.
t i
4.
Corrective Actions to be Taken to avoid further Violations Actions taken in Section 3 above ensure avoidance of j
further violations.
5.
Date of Full Compliance 4
Duke Power Company is now in full compliance i
l i
I I
DUKE POWER COMPANY RESPONSE TO VIOLATION 414/86-16-02 10 CFR 50.59(b) requires that the licensee shall maintain records of changes in the facility as described in the safety analysis report and that these records shall include a written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question.
Contrary to the above, a record of a change to the facility as described in the safety analysis report did not include a written safety evaluation that included the bases for the determination that the change did not involve an unreviewed safety question, in that, Nuclear Station Modification CN-2-0084 was processed to delete the feedwater pump turbine trip on feedwater isolation and the safety evaluation for this modification did not include the bases for the determination that this change did not involve an unreviewed safety question.
RESPONSE
Duke Power Company denies the violation. The bases for the determination of no unreviewed safety question are a part of the records maintained in the Design Engineering Department.
Form 160.2 is only intended to provide documentation that a 50.59 assessment has been made and that for this particular modification no unreviewed safety question exists that would prevent implementation.
Attached is the complete 50.59 package from the Design Engineering Department records which indicates that for the purpose of meeting the requirements of Section 50.59, Chapters 7, 10, and 15 of the Catawba FSAR were reviewed in the determination of no unreviewed safety question. The reference to Chapter 7 made on Form 160.2 refers only to the section of the FSAR reveiwed by the Electrical Division in their determination of what Electrical changes were required to the FSAR for update to assure that the FSAR reflected the as-built condition of the plant.
I
.j C. N C O 8 4 p.../
, j 02637 (R11-85)
FORM 160.2 REVISION 3.
SHEET I
OF 2 4
j ISEARCil P. iWCTS DUKE POWER COMPANY i ec. J A N 3 ' ~ ^
NUCLEAR STATION MODIFICATION t
SAFETY EVALUATION STATION:
Ca QW bO-C N '2 - 0 0 84 NSMIDNO.
2 REVISION:
O I}
PART:
C/E M/N h
R H. A ller Q-820 QA CONDITION:
I N0ne-CONTACTS:
PART A-SAFETY EVALUATION (RPSS)
DESCRIPTION OF MODIFICATION 5ee c,Ho.c.ke.c) shed 2 of 2 TECHNICAL SPECIFICATION CHANGES REQUIRED. (Yes@ Attach applicable pages if yes.
NO 10 CFR 50.59 EVALUATION - UNREVIEWED SAFETY QUESTIONS Section 50.59 states that an Unreviewed Safety Question is involved (i)if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be in-creased; or (ii)if a possibility for an accident or malfunction of a different type than evaluated previously in the safety analysis report may be created; or (iii)if the margin of safety as defined in the basis for any technical specification is reduced.
Address each part separately in the Safety Evaluation with appropnatejustification. 8)ed f;cd ion M preWnt OVer p reSSu ri Zo. tion of h;cjh pressur, fee ho.Yer hecaers c<Ftec reacJor-e c
trip. No unres;ewec) so.Msf guest.co,
s b.ted above., is gel ej -h k as g
involve cl Oc c.rmfeJ by Mis meMiccdio n. No s deh sys b c W 11 Le.
Jeya}ec) ctnc) fund:oaa.l ckonges a ll b e made to any sysb a.s ct no resd o9
%.s rnoch0caba.
If the ar.swer to any of the above is yes an Unreviewed Safety Question is involved, which requires a License Amendment pursuant to 10 CFR 50.90.
SIGNATURE:(RPSS)
WA DATE:
/~/Y $
REVIEWED:(RPSS)
[(d4/r.
i_/
DATE:
/-/bb PART B -SAFETY ANALYSIS REPORT (RE)
SAFETY ANALYSIS REPORT SECTION(S) REVIEWED:3OMf CHANGED://CWG' pjt //t1/24 SAPPENDIX R OF 10 CFR 50 REVIEW CONDUCTEDLum 12 yV4 5010 CFR 50.49(I) REVIEW CONDUCTED SIGNATURE: N[
B1.EAD ENGINEER DATE: /2 8
8,.
DATE:#, /4, /$
REVIEWED:
/-27 74
CN 0084 02637 (R11-85)
FORM 160.2 REVISION 3' SHEET I
OF 2 DUKE POWER COMPANY NUCLEAR STATION MODIFICATION SAFETY EVALUATION STATION:
ba QW bO-NSM_ID NO.
CN 0084 REVISION:
O 2
(I PART:
C/E M/N h
CONTACTS: M N MiI Er (3-592 OA CONDITION:
h None PART A-SAFETY EVALUATION (RPSS)
DESCRIPTION OF MODIFICATION TECHNICAL SPECIFICATION CHANGES REQUIRED. (Yesh Attach applicable pages if yes.
bO 10 CFR 50.59 EVALUATION - UNREVIEWED SAFETY OUESTIONS Section 50.59 states that an Unreviewed Safety Question is involved (i) if the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the safety analysis report may be in-creased; or(ii)if a possibility for an accident or malfunction of a different type than evaluated previously in the safety analysis report may be created; or (iii)if the margin of safety as defined in the basis for any technical specification is reduced.
Address each part separately in the Safety Evaluation with appropriate justification. //)c),f; cod ion dN prWent oVerpreSSuri Ecdion o f h;g h pressure, feecjwa}er hea+ers a.ftec reacJor-frip. No unreviewec) safety guestico,
sfo.ted above, is judged f6 b e.,
as involve ci Oc c.recJed by Mis mediRecdioa. Mo sdeb systen,c a3ill b e.
deyded and
&nSoad ckonges sli be. <nade to any sys b a.s a.
no reswl4 09
%,s mo c)i0c.a.4 ion.
If the answer to aay of the above is yes an Unreviewed Safety Question is involved, which requires a License Amendment pursuant to 10 CFR 50.90.
SIGNATURE:(RPSS)
WA DATE:
/N'$
REVIEWED:(RPSS)
,4/r.
DATE:
hM'
\\f PART B - SAFETY ANALYSIS REPORT (RE)
SAFETY ANALYSIS REPORT SECTION(S) REVIEWED:
CHANGED:
O APPENDIX R OF 10 CFR 50 REVIEW CONDUCTED O 10 CFR 50.49(1) REVIEW CONDUCTED SIGNATURE:
O LEAD ENGINEER DATE:
REVIEWED:
DATE:
'~
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[ Al CG84
'; P Shesh 2 of Q 1
NSM CN-20084 Page 2 of 2 i
OBJECTIVE:
Prevent overpressurization of high pressure feedwater heaters after reactor trip. Presently the heater relief valves have been regularly challenged and actually blown off the heaters due to high feedwater pump discharge pressure.
PROPOSED MODIFICATIONS:
Delete the present feedwater pump turbine trip on feedwater isolation (safety inj or Hi-Hi S/G water level or Rx trip and Lo Tav.) and replace with the 3
original Westinghouse provided trip signal '(safety injection or_ Hi-Hi S/G
~i water level).
Add a signal to the main feedwater pump turbine (CFPT) speed control circuit to cause the turbine to run its speed back to the low speed stop (LSS) setting when the reactor trips. This signal (Rx trip) should be independent of low Tav.
Also, the signal should be independent of any Rx trip breaker testing.
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December 18, 1985 W
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A.
P.
Cobb
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Attention:
J.
D.
Hager 7'g, ;
C.
C.
Rolfe Attention:
M.
S.
Tully*
/.,/ 7
((
SUBJECT:
Catawba Nuclear Station i
NSM CN-10729/00: Taske 2135B: MA# 531.17 NSM
CN-1459.02 Attached for your information and use are the lists of the involved groups and their scheduled completion dates for the subject NSMs.
The RPSS Section should proceed with the completion of the safety evaluations for the Electrical portion of the subject modifications.
For additional scope information, please contact M.
H.
Miller.
If there are any problems in supporting the schedules, please contact me immediately.
M iCL v 592b I
45 5 T.
C.
McMeekin, Chief Engineer L
(2 IS) Og Electrical Division l
OWVom BY:
D.
W.
Vass Design Engineer 11 TCM/DWV/cp Attachment cc:
All w/ attachment Central Records S.
G. Whisenant M.
E.
Efaro M.
H.
Miller B.
W.
Brewer J..M. Butler D.
E.
Thomas
- C.
E.
Robinson.
Jr.*
T, A.
Sanders F.
N.
Mack J.
F.
Canborn Ray Catoe a wfngus
tl NSM CN-10729 REV 00 TASK #
21358 DATE 12-18-85 ELECTRICAL DESIGN ACTIVITIES RESP START SCHEDULED GROUP ACTIVITY NEEDED BY DATE COMPLETION l
ECPS/MHM Provide T-O Package ECPD 12-06-85c ECPD/JMB Design Release PCNS 12-06-85 02-28-86 ECPD/JMB Cables to Routing EPES 12-06-85 01-10-86 EPES/DET Route Cables EPCT 01-10-86 02-21-86 EPCT/RAR issue Cables PCNS 02-21-86 02-28-86 i
RPSS Provide SE (Elec)
ESPD/DWV 0 3,1.4.- 8 6 S c h,
c-Complete
- 0 2 - 1 1 - 8 6 E x py/-
y NSM CN-20084 REV 00 TASK #
21368 DATE 12-18-85' ELECTRICAL DESIGN ACTIVITIES RESP START SCHEDULED GROUP ACTIVITY NEEDED BY DATE COMPLETION ECPS/MHM Provide T-O Package ECPD-12-06-85c ECPD/JMB Design Release PCNS 12-06-85 01-15-86 ECPD/JMB Cables to 1cuting EPES 12-06-85 12-18-85 EPES/DET Route Cables EPCT 12-18-85 01-10-86 EPCT/RAR issue Cables PCNS 01-10-86 01-15-86 i
RPSS Provide SE CElec)
ESPD/DWV 01-29-86S3g 01-07-86Expx c:
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,e October 11, 1985 PCNS-85-1066 i
T C McMeekin Attn: D W Vass Re: Catawba Nuclear Station, Unit - 2 NSM CN-20084/00; Task 2136B Modify CF System Feedwater Pump Trip Control Scheme W0 #10573, CA #531.17 File: CN-1459.02 The subject work item has been received in Design Engineering and should be completed as indicated below:
WORK CATEGORY EngineeringDesign(ListB)
ACTION REQUIRED
- INDICATES PUNCHLIST COMMITMENT
~
Evaluate and provide scope by 11/11/85.*
Submit engineering schedule to the undersigned by 11/11/85.
l ADDITIONAL INFORMATION Originating Department / Contact / L J Benjamin - NPRJ Lead Responsibility Designation / Responsible Engineer - ESPD/DWV Support Responsibility Designations - ECPS, ECPD, MSYS DE Received Date - 10/10/85, Requested DE Completion - 11/22/85 Initial Category - 11 Applicable Procedures: DED II.4.2 and DE QA PR-160 Coments: NSM is to be installed prior to mode 3.
{
Please call if you have any questions.
A P Cobb, Jr., Manager
~ ( i. - w r ' m, Project Management Division
!._13 By: L E Suther, Jr.
Project Engineer LES/dgl/S10-15 i
Attachment cc: S L Misenheimer (w/ attach)
R R Weidler (w/ attach)
M D Hopkins (w/ attach) - Catawba Constr/MMT l
J F Sanborn (w/ attach) j E M Couch (w/ attach) - Catawba Constr S D Alexander (w/ attach)
.. Tenn 34950 (R9-84)
TRANSMITTAL TO DESIGN ENGINEERING TO:
F. G. TEMPLETON, JR.
DATE:
9 - // - g5 i
Cesign Engineering - PMD
SUBJECT:
CATAWBA Nuclear Station Project No. C A/ 2 o o P 4
, Rey, o
Part
, Unit 2
Project type:
00M SOM Design Study Exemet Change SPR This work item is being sent to you for the following:
PROJECT !.1AflAGE11ENT NvisIO.'i Budget Cost Estimate Infomation Only R EC E.I'V,E D Detailed Cost Estimate Cancellation SEP 1 61985 OlVISION! RECORDS CO?v Detailed Design Other flLE:0.
s
.i Schedule infomation:
U# 0
.; w;: - - ;.
Procesed Design Engineering Completien Date:
W 2-is On u'
l,J Proposed Implementation Start Date:
/ 2 -/ 6 -R5 Accounting information:
RESP. N0.
LOCATION WORK ORDER CLOSING DESIGN LABOR fXXx 7332
/o573 o s 31.17 MATERIALS CONSTR. LABOR REMARKS:
T4S K 2.l36 XRE F. C N 10 72 9 R usH )
u,,
- f., c For t chnical questions, contact the Accountable Engineer:
c J.a m u s,n For a :ounting questions, contact:
E. E. HARWELL Any o 1er questions should be directed to:
T. A. SANDERS Xc:
.. E. SCHMID - G.O.
J. F. SANBORN (Constr.)
J. R. THOMAS - CNS TOM LOVE (Constr.)
w sacw, NUCLEAR STATION MODIFICATION REQUEST Task 2136 (1) NSM # C_2L
.2 0_. _0_ _a__4_ Rev. # 0 f
Sys CF; CNPR 00982 (2) Description of Problem Overoressurization of Main Feedwater Pumo (MFWP) discharge piping has occured (on Unit 1) following a reactor trip.
This has caused failure of the High Pressure Feedwater Heater Relief Valves and assomiated piping.
(3) Requested Modification Delete the MFWP trip that occurs at low Tava following a reactor trip.
Provide a control scheme to run feedwater cump speed back to minimum setting on reactor trip.
(4) Design Responsibility O SDM %DDM (Red-marked drawings requested from DE? O Yes [3 No)
(4a) Verbal Approval (SDM's only) Provided By:
N/A Date N/A (4b) Verbal Approval (SDM's only) Requested By:
Date 1
(5) Implementation Responsibility: O NPD orTSSD OA Condition (6) Accountable Engineer:
L.
.T. Benjamin 8 30/85 (7) Category Assigned (1 thru 16):
11 Urgency (chec,, one): C5 High O Medium O Low (8) Commitment Dates:
N/A i
N/A Commitment To:
4 OGSi&h nn ni + n 1ne- /2 85 QlH)
Dutage Related? CD Yes O No (9) Date, Desired by Station: nnnign Reason: To supoort initial unit testing.
/
(f0)NSMmRequest
.a pproveyay_,3,4,_ g y-s -gs n-s (11) Estimated Cost of dification $
~-
g Estimate By:
y Estimate Attached: 0 Yes O No 8
.f (12) Proposed Schedule h
Proposed Impiementation Complete Date:
cc g
Scheduled By:
.9 3
Schedule Attached: O Yes O No j
-t-s (13) Approved 89:
~ ~ ~ ~ ~
e Name Date
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Attn:
f.k).A/w,/e,-
.Re: Catawba Nuclear Station, Unit 2.
NSM CN-200ti//00 ; Task 2zost8
- /cd.'4 Cf Ae4 m Se e dw+/c < h,o
&W Ce ~ be f.5~e1,a m e,
~
- WO 9 /os*73; MA 9 CTI. /7 File:
Ct4-1459.02 4
-The ' subject work item has been received in Design Engineering and should be -
completed as indicated _below.
.._m WORK CATEGORY Engineering (List B)
~
ACTION RE0VIRED
- INDICATES PUNCHLIST COMMITMENT
~
soNccs c'verad
-~~ ~~~EvalHate'~and pr6 vide scope document by ro/icles *.m4*Mc-j Submit en pugh,g,ineering schedule to S L Misenheimer (Project Scheduling Group) by_
i ADDITI0f:AL INFORMATION o
- Originating Department / Contact - /yar / LJ Peui4~W Lead Responsibility Dpsignation/ Responsible Engineer - mrefs /_ A. D. Le e_
DE Received Date - %/rr, Requested DE' Completion - _ /z-z+c Category -
//
NSM
~
Applicable-Procedures - DED VI.E.2 and DE' QA PR-160 Comments:
Please call the undersigned at ext. 2891 if you have any questions.
A P Cobb, Jr., Manager Project Management Division AM By: Coite S Moss, Technical Specialist CSM/M10 cc: w/ attachments S L Misenheimer E M Couch M D Hopkins R G Eble J F Sanborn J K Ray i
J R Thomas
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F. p les O URGENT REPLY NECESSARY R
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SIGr/D R
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Y SIGNED DATE
- 7..
C n a-o o g nt-
-t November 11, 1985 7')jk A.P.
Cobb A
Attention:
J.
D.
Hager
SUBJECT:
Catawba Nuclear Station NSM CN in790/00: Task # 2135B: MA# 531.17
_N h N-20084/0 D ak# 21368: WO/MA# 10573/531.17 Mod: Iy GP system Feedwater Pump Trip Control Scheme File:
CN-1459.02 Attached for your information and use are the scope documents for the subject modifications. By copy of this letter, the involved groups should begin their review of the requested changes and provide me by 11-21-85 with their schedules for completing their portion of the NSMs.
Please note that the information requested from NOCS (E.
W.
Zimmerman) and NCNS (J.
R.
Wallace) should be expedited to, enable the Electrical Division to schedule work on the subject NSMs.
Please contact me if there are any questions.
T.
C.
McMeekin, Chief Engineer Electrical Division W,o Qgf f'$
9 O u> h BY:
D.
W.
Vass Design Engineer 11 TCM/DWV/cp Attachment cc:
All w/ attachment Central Records M.
E.
Efird M.
H.
Miller 8.
W.
Brewer J.
M.
Butler M.
S.
Tully S.
L.
Misenheimer T.
A.
Sanders F.
N.
Mack J.
F.
Sanborn Ray Catoe E.
W.
Zimmerman(CNS)
J. R. Wallace(CNS) i
Form 02726 (R8-84)
DUKE POWER COMPANY 1
2' DESIGN ENGINEERING DEPARTMENT PAGE OF NSM DESIGN SCOPE DOCUMENT REVISION Catawba 2
STATION UNIT (S) 20084 00 N
NSM NUMBER NSM REVISION PART TASK El.ect/ECPS LEAD DIVISION / GROUP REQGREU/ DESIRED DE COMPLETION DATE 11/22/85*
M. H. Miller PREPARED BY DATE 11/11/85
(*Per: PCNS-85-1066)
NSM DESCRIPTION:(SEE ATTACHED NSM PACKAGE)
EXCEPTIONS: None at this time.
NSM SCOPE:(ATTACH ADDITIONAL PAGES AS REQUIRED)
See Attached Sheet AFFECTED GROUPS:
NEED BY RESP. GROUP WORK ITEM DESCRIPTION GROUP N0CS Provide add / delete sheets and ECPS marked drawings NCNS Provide present CFPT low speed ECPS stop setting ECPS Provide Engineering T/0 ECPD ECPD Design Documentation KELC/PCNS RPSS Safety Evaluation Chk List ECPS/ECPD/KELC COMMENTS:
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NSM CN-20084 Page 2 of 2 OBJECTIVE:
Prevent overpressuri, ration of high pressure feedwater heaters af ter reactor trip. Presently the heater relief valves have been regularly challenged and actually blown off the heaters due to high feedwater pump discharge pressure.
PROPOSEDj0DIFICATIONS:
Delete the present feedwater pump turbine trip on feedwater isolation (safety inj or Hi-Hi S/G water level or Rx trip and Lo Tav.) and replace with the orig'inal Westinghouse provided trip signal (safety injection or_ Hi-Hi S/G water level).
Add a signal to the main feedwater pump turbine (CFPT) speed control circuit to cause the turbine to run its speed back to the low speed stop (LSS) setting when the reactor trips. This signal (Rx trip) should be independent of low Tav.
Also, the signal should be independent of any Rx trip breaker testing.
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