ML20045B196
| ML20045B196 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 05/31/1993 |
| From: | Garrett P, Schroeder C, Sykes K COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 93-0225, 93-225, NUDOCS 9306160401 | |
| Download: ML20045B196 (46) | |
Text
Commonwealth Edison i
Dresden Nuclear Power Stanon 6500 North Dresden Road Moms. lilinois 60450 Telephone 815/942-2920 June 1,1993 CWS LTR #93-0225 Director, Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 Attention:
Document Control Desk Genthsnen:
Subject:
Monthly Operating Data Report Dresden Nuclear Power Station Commonwealth Edison Company Docket Nos.50-010,50-237, and 50-249 Enclosed is the Dresden Nuclear Power Station Monthlv Operatins' Summarv Report for May,1993. This infonnation is supplied to your ofGee in accordance with the instructions set forth in Regulatory Guide 1.16.
Sincerely,
% hh// py Charles W. Schroeder Station Manager Dresden Station CWS/PG:sih Enclosure ec:
NRC Region III OfGce Illinois Dept. of Nuclear Safety, State of Illinois U.S. NRC, Document Management Branch Nuclear Licencing Administrator Site Vice Pres.
General Manager - Nuclear Services OPEX Engineer (2)
NRC Senior Resident Inspector Site Quality VeriGcation - Dresden Site Engineering and Construction Manager Nuck-ar Oversight Manager / R. Janecek Comptroller's OfGce INPO Records Center UDI,Inc.- Wash.,D.C.
File /NRC Op. Data File / Numerical
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9306160501 930531 PDR ADOCK 05000010 V
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L MONTIILY NRC
SUMMARY
OF OPERATING EXPERIENCE, CIIANGES, TFSTS, AND EXPERIMENTS ITR REGULATORY GUIDE 1.16 AND 10 CFR 50.59 FOR DRESDEN NUCLEAR POWER STATION COMMONWEALTII EDISON COMPANY FOR May,1993 UNIT DOCKET LLCENSE.
1 050-010 DPR-2 2
050-237 DPR-19 3
050-249 DPR-25 i
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m TABLE OF CONTENTS May,1993 NRC REPORT 1.0 Introduction 2.0 Summary of Operating Experience 2.1 Unit 2 Monthly Operating Experience Summary.
2.2 Unit 3 Monthly Operating Experience Summary.
3.0 Operating Data Statistics 3.1 Monthly Operating Data Report - Unit 2 3.2 Monthly Operating Data Report - Unit 3 3.3 Aserage Daily Power Level Data - Unit 2 3.4 Aserage Daily Power Lesel Data - Unit 3 3.5 Unit Shutdown and Power Reduction Data - Unit 2 3.6 Unit Shutdown and Power Reduction Data - Unit 3
'1 3.7 Station Maximum Daily Load Data 4.0 Unique Reporting Requimnents 4.1 Main Steam Relief and/or Safety Valte Operations - Unit 2 and Unit 3 4.2 Off-Site Dose Calculation Manual Changes 4.3 Major Changes to the Radioactive Waste Treatment 4.4 Failed Fuel Element Indications 4.4.1 Unit 2 4.4.2 Unit 3 5.0 Plant or Procedure Changes, Tests, Experiments, and Safety-Related Maintuunce 5.1 Amendments to Facility License or Technical Specifications 5.1.1 Unit 2 5.1.2 Unit 3
)
5.2 Changes to Procedures which are Described in the Final Safety Analysis Report (FSAR) (Units 2 and 3) 5.3 Significant Tests and Experiments Not Described in the FSAR (Units 2 and 3) 5.4 Safety Related Maintenance (Units 2 and 3) 5.5 Completed Safety-Related Modifications 5.6 Temporary System Alterations Installed 5.7 Other Required 10 CFR 50.59 Evaluations (Units 2 and 3)
S 1.0 Introduction Dresden Nuclear Power Station is a three reactor generating facility owned and operated by the Commonwealth l
Edison Company of Chicago, Illinois. Dresden Station is located at the confluence of the Kankakee and Des l
Plaines Rivers, in Grundy County, near Aforris, Illinois.
Drestien Unit 1 is a General Electric Boiling Water Reactor with a design net electrical output rating of 200 i
megawatts electrical (51We). The unit is retired in place with all nuclear fuel remosed from the reactor vessel.
Therefore, no Unit 1 operating data is prmided in this report.
Dresden Units 2 and 3 are General Electric Boiling Water Reactors with design net electrical output ratings of 794 AfWe each.
Waste heat is rejected to a man-made cooling lake using the Kankakee River for make-up and the Illinois River for blowdown.
The Architect-Engineer for Dresden Units 2 and 3 was Sargent and Lundy of Chicago, Illinois.
This report for Stay,1993, was compiled by Paul K. Garrett and Keiin W. Sykes of the Dresden Regulatory Assurance StalT, telephone number (815) 942-2920, estension 2713 or 2704.
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I 2.0
SUMMARY
OF OPERATING EXPERIENCE FOR May,1993 2.1 UNIT 2 MONTIILY OPERATING EXPERIENCE
SUMMARY
l 05/01/93 to 05/23/93 Unit 2 entered the month shutdown, and in week sixteen of outage D2R13. The Unit started the month shutdown in refuel mode of operation. The Unit was put into startup mode of operation 05/23/93.
05/23/93 Reactor became critical at 0525 hours0.00608 days <br />0.146 hours <br />8.680556e-4 weeks <br />1.997625e-4 months <br /> on 05/23/93.
05/25/93 to 05/26/93 The Unit 2 generator was synchronized to the power grid at 1137 hours0.0132 days <br />0.316 hours <br />0.00188 weeks <br />4.326285e-4 months <br /> on 05/25/93. The generator was removed from service at 0227 hours0.00263 days <br />0.0631 hours <br />3.753307e-4 weeks <br />8.63735e-5 months <br /> on 05/26/93 because of high drywell temperatures.
l 5/26/93 to 05/28/93 The reactor remained shutdown for this period while the high drywell temperatures were corrected.
05/28/93 to 05/31/93 Unit 2 became critical at 0520 hours0.00602 days <br />0.144 hours <br />8.597884e-4 weeks <br />1.9786e-4 months <br /> on 05/28/93. Unit 2 main generator was synchroniecd to the grid at 0151 hours0.00175 days <br />0.0419 hours <br />2.496693e-4 weeks <br />5.74555e-5 months <br /> on 05/29/93. Unit 2 remained on line for the remainder of the month.
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1 2.0
SUMMARY
OF OPERATING EXPERIENCE FOR May,1993 2.2 UNIT 3 MONTIILY OPERATING EXPERIENCE
SUMMARY
f 05/01/93 to 05/31/93 Unit 3 entered the month critical and on line and continued through the end of the month.
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3.0 OPERATING DATA REPORT 3.1 OPERATING DATA REPORT - DRESDEN UNIT TWO DOCKET No.
050-237 DATE June 1,1993 COMPLETED EY P. K. Garrett and K. W. Sykes TELEPHONE (815) 942-2920 t
OPER ATING STATUS
- 1. REPORTING PERIOD: May,1993
- 2. CURRENTLY AUTHORIZED POWER LEVEL (MWth): 2,527 MAXIMUM DEPENDABLE CAPACITY (MWe NET): 772 DESIGN ELECTRICAL RATING (MWe Net): 794
- 3. POWER LEVEL TO WHICH RESTRICTED (IF ANY)(MWe Net): N/A
- 4. REASONS FOR RESTRICTIONS (iF ANY): N/A REPORTING PERIOD DATA PARAMETER Tills MONTII YEAR TO DATE Cl31L'LATIVE 5.
HOURS IN PERIOD 744 3623 202.055 6.
TIME REACTOR CRITICAL Oiours) 1631 552.7 150.124.4 7.
TIME REACTOR RE5ERVE SHUTDOWN 01ours) 0 0
0 g
8.
TIME GENERATOR ON-LINE Olours) 90.7 4RO.3 143,713.0 9.
TIME GENERATOR RESERVE SHUTDOWN (Hours) 0 0
0 10.
THERM AL ENERGY GENERATED (MWHt Gross) 53.426 767.568 296,455,325 II.
ELECTRICAL ENERGY GENERATED (MWHe Gross) 15,988 242.535 94.627.175 12.
ELECTRJCAL ENERGY GENERATED (MWHe Net) 8.779 220,123 90.457,985 13.
REACTOR SERVICE FACTOR (5) 21.9 15.3 74.3 14.
REACTOR AV AILABILITY F ACTOR (%)
0 18.0 74.8 15.
GENERATOR SERVICE FACTOR (%)
12.2 13.3 71.1 16.
GENERATOR AVAILABILTTY FACTOR (%)
12.2 18.0 71.3 17.
CAPACITY F ACTOR (USING MDC Net) (%-)
1.5 7.8 57.9 18.
CAPACITY FACTOR (USING DER Net)(5) 1.4 7.6 56.4 19.
FORCED OUTAGE FACTOR (%)
35.9 86.7 28.9 20.
SHUTDOWNS SCHEDULED OVER THE NEXT 6 MONTHS (Type, Date and Duration of Each)
NONE.
21, IF SHUTDOWN AT END OF REPORT PERIOD, ESTIM ATED DATE OF STARTUP N/A
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3.0 OPERATING DATA REPORT 3.2 OPERATING DATA REPORT - DRESDEN UNIT THREE DOCKET No.
050-249 DATE June 1,1993
' COMPLETED BY P. K. Garrett and K. W. Sykes TELEPHONE (815) 942-2920 OPERATING STATUS i
- l. REPORTING PERIOD: May,1993
- 2. CURRENTLY AUTHORIZED POWER LEVEL (MWth): 2,527 MAXIMUM DEPENDABLE CAPACITY (MWe Nel): 773 DESIGN ELECTRICAL RATING (MWe Net): 794
- 3. POWER LEVEL TO WHICH RESTRICTED (IF ANY)(MWe Net): N/A 4, REASONS FOR RESTRICTIONS (IF ANY): N/A REPORTING PERIOD DATA 5.
HOUR $ IN PERIOD 744 3.6232,879 191,640 6.
TIME REACTOR CRTTICAL (Houn) 744 2,108.8 137,698.6 7.
TIME REACTOR RESERVE SHUTDOWN (Hours >
0 0
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8.
TIME GENERATOR ON-LINE (Hours) 744 2051.8 132,325.2
-l 9.
TIME GENERATOR RESERVE SHUTDOWN (Hours) 0 0
0 10.
THERMAL ENERGY GENERATED (MWH1 Gron) 1,796,263 4,686.937 272.216,39.
11.
ELECTRICAL ENERGY GENERATED (MWHe Gross) 574,812 1,498.678 87.475,885 12.
ELE /TRICAL ENERGY GENERATED (MWHe Net) 549,883 1.,447,987 83,058,167 13.
REACTOR SERVICE FAcrOR (5) 100
$8.2 71.9 14.
REACTOR AVAILABILTTY FACTOR (W) 100 58 6 72.0 15.
GENE.RATOR SERVICE FACTOR (%)
100 56.6 69.0 16.
GENERATOR AVAILABILTTY FACTOR (%)
100 56.6 69.0 17.
CAPACTTY FACTOR (USING MDC Net) (5) 95.6 51.7 56.1 la.
CAPACTTY FACTOR (USING DER Net)(%)
93.1 50.3 54.6 19.
FORCED OUTAGE FACTOR (5) 0 43.4 31.0 20.
SHUTDOWNS SCHEDULED OVER THE NEXT 6 MONTHS (Type, Date and Duration of Each)
NONE.
21.
IF SHUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP N/A
3.3 AVERAGE DAILY UNIT POWER LEVEL DOCKET No.
050-237 UNIT Dresden 2 DATE June 1,1993 COMPLETED BY P. K. Garrett and K. W. Sykes TELEPilOST (815) 942-2920 MONTII:
May,1993 DAY AVERAGE DAILY NET DAY AVERAGE DAILY NET POWER LEVEL (MWe)
POWER LEVEL (MWe) 1 0
18 0
2 0
19 0
3 0
M 0
4 0
21 0
5 0
n 0
6 0
23 0
7 0
24 0
8 0
25 76 9
0 26 11 10 0
27 0
11 0
28 0
12 0
29 129 13 0
30 151 14 0
31 274 15 0
16 0
17 0
3.4 AVERAGE DAILY UNIT POWER LEVEL j
l DOCKET No.
050-249 UNIT Dresden 3 DATE June 1,1993 COMPLETED BY P. K. Garrett and K. W. Sykes TELEPIIONE (815) 942-2920 MONTII: May,1993 DAY AVERAGE DAILY NET DAY AVERAGE DAILY NET POWER LEVEL (MWe)
POWER LEVEL (MWe) 1 534 18 768 2
665 19 773 3
730 20 775 4
589 21 771 5
773 22 730 6
768 23 734 7
771 24 728 8
764 25 743 9
729 26 751 10 794 27 747 11 771 28 753 12 770 29 743 13 770 30 757 14 769 31 756 15 713 16 723 t
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Docket No.
050-237 UNIT NAME:
Dresden Unit 2 DATE:
June 1,1993 Completed Ily:
P. K. Garrett 3.5 UNIT SilUTDOWNS AND POWER REDUCTIONS REPORT MONTil May,1)93 No.
DATE TYPE!Il M'R AtK)N REASON C)
MET 110D OF LETN5ti DDT SYSTEM (TMt449 n)MIT)NF.NT CTIM.($)
CAUSE A CURREITIVE ACif0N TO FFDDT RErt'RRENCE j
gla)t'RS) 5HtTTING IMIWN RFFORT#
- l' ACIORGl j
l 1,
049143 w 5
311 5 C
N.*A NM N 'A N>A N/A ci?)H 2
OW e v) to F
St F.
I, N'A VB N/A NOT YET ttTE RMINID. (HitR INVt3TIGATFW or::vs w.m. - w w= m m y.n
,v.
h. n s
=
(1)
E Operato-Training Licensee Enam
- 4. Other (Explain)
(5)
F Administrative 3.1.vad Reduction F: FURCED G Operational Error Edibit I Same %=rce as above.
S: SCliEDULED 11 Other (Explain)
(2)
(3)
Reason:
(4)
Method:
A Equirment Failure (Explain)
Edibit G Instructions for Preparation of Data B Mainienance or Test
- 1. blanual Entry Sheets for Licensee Event Reports (LER)
C Refueling
- 2. Manual Scram
' File (NUREG4161)
D Regulatory Restriction
- 3. Automatic Scram (GIIKLT92 / 2) /1
Docket No.
050-249 UNIT NAME: Dresden Unit 3 DATE:
June 1,1993 Completed By: P. K. Garrett Telephone (815) 942-2920 3.6 UNIT SilUTDOWNS AND POWER REDUCTIONS REPORT MONTII May,1993 No.
DATE TYPT1s t it1tATION REASON Q METitoD OF lEIN5f t EM.NT SYS'. l M CDDE f 41 WMPONENT CUtWd) cat %E & WERECTIVE ACTRW TO PREYDrT Rirl'DRENCE tHot'RSI
$HtTTPEG tw)WN 8 EFURT #
REACIORO)<
NONE n
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(1)
(2) 11 Other (Emplain)
F: FORCED Reason:
(3) 5: SCHEDULED (4)
A. Equipmera Failure (Explain)
Method:
B Maintenance or Test Eeihit G Instructions for Preparation of Data C Refueling
- 1. Manual Fsary Sheets fer Licensee Event Reports (LER)
D Regulatory Restriction
- 2. Manual Scram File (NUREG4)l61)
E Operator Training A Licenser Exam
- 3. Automatic Scram F l.dministrative
- 4. Other (Explain)
(3)
G Operathmal Error
- 5. Lead Reduction (GilKLT92 / 2) / 4
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3,7 COMMONWEALTH EDISON COMPANY f
DRESDEN NUCLEAR POWER STATION MAXIMUM DAILY ELECTRICAL LOAD FOR THE MONTH OF April,1993 f
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i
.i Day Hour Endine KWe I
n/a 0
1 2
n/a 0
i 3
n/a 0
4 n/a 0
5 n/a 0
6 n/a 0-I 7
n/a 0
8 n/a 0
9 n/a 0
10 n/a O
11 n/a 0
12 n/a 0
{
.,13 n/a 0
14 n/a 0
15 n/a 0
_t 16 n/a 0
i 17 n/a O
l 18 n/a 0
19 n/a 0
i 20 n/a 0
l 21 n/a 0
22 n/a 0
23 n/a 0
t 24 n/a 0
25 n/a 0
26 n/a 0
l 27 n/a 0
28 1300 186,000 29 n/a 0
30 2000 527,000 l
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3.7 COhiMONWEALTH EDISON COhiPANY DRESDEN NUCLEAR POWER STATION hi AXIMUM DAILY ELECTRICAL LOAD FOR THE MONTH OF May,1993 t
h pg Hour Endine KWe 1
1600 601,000 2
2400 807,000 l
.3 1100 810,000 4
2400 780,000 5
0400 810,000 6
0500 805,000 7
1700 810,000 8
1000 812,000
{
9 1700 807,000 10 1300 807,000 11 0700 807,000 12 2200 808,000 13 0800 809,000 14 0900 810,000 15 2400 754,000 16 1900 760,000 f
17 0700 812,000 18 2400 808,000 19 0200 807.000 l
20 1400 814,000 21 0100 11,000 22 1200 798,000 23 0900 795,000 24 2400 772,000 l
25 2200 895,000 26 0100 793,000 27 2400 784,000 l
28 2100 800,000 1'!
29 1100 803,000 30 0200 967,000 31 2200 1,312,000 i
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4.0 UNIQUE REPORTING REQUIREMENTS i
4.1 MAIN STEA51 RELIEF VALVE OPERATIONS Relief vahe operations during the May,1993 reporting perioJ are summarized in the following table. The table includes information as to whi;h relief vahe was actuated, how it was actuated, and the circumstances resulting in its actuation.
Unit Date Valves Actuated No. and Type of Plant Conditions Description of Actuations Events l
l 2
05-04-93 203-3 A 1 (automatic)
Refuel See Note 1 Below 1
l 203-3B 1 (automatic)
Re fuel See Note 1 Below 203-3C 1 (automatic)
Refuel See Note 1 Below 203-3D 1 (automatic)
Refuel See Note 1 Below 203-3E 1 (automatic)
Refuel See Note 1 Below 2
05/24/93 203-3A 2 (manual)
Startup See Note 2 Below 203-3B 2 (nunual)
Startup See Note 2 Below 203-3C 2 (manual)
Startup See Note 2 Below 203-3D 2 (manual)
Stanup See Note 2 l
Below l
203-3E 2 (manual)
Startup See Note 2 Below I
3 N/A N/A N/A N/A N/A I
Note 1: While perfonrung Dresden instrument Surseillance (Di$) 0250-04, Auto-Blowdown lepic Test, and Dresden Operating Surveillance (DOS) 1500-06, LPCI System Pump Operability Test With Torus Available, concurrently, all five Automatic Depressurization Ss tem (ADS) valves received an open signal. This event was reponed under Licenwe Event Report (LER) 0500023?g3007.
Note 2: DOS 0250-04, Relief Yalve Cy,rability Test At Imw and At Rated Pressure, was performed to test the operability of the relief vahes at low eactor t ressure and at high reactor pressure.
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4.2 -
OFF-SITE DOSE CALCULATION MANUAL (ODCM) CHANGES The following changes were made to the ODCM for this reporting period.
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l'i 4.3 MAJOR CllANGES TO TIIE RADIOACTIVE WASTE TREATMENT SYSTEMS DURING May,1993 Current Status of Radioactive Waste Treatment System Upgrade Project:
No significant change to report for the month of May,1993.
I Floor Drain Collection Pump replacement is 95 % complete. The replacement is being performed under modification number M12-2/3-82-002N.
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4.4 FAILED FUEL ELEMENT INDICATIONS i
4.4.1 Unit 2 Unit 2 fuel performance during May,1993, continued to show no indications of leaking fuel. This is based on the
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sum of the activities of the six (6) Noble Gases as measured at the Recombiner. Therefore, Unit 2 had excellent fuel performance.
j 4.4.2 Unit 3 Unit 3 fuel performance during May,1993, continued to show no indications of leaking fuel. This is based on the sum of the activities of the six (6) Noble Gases as measured at the Recombiner. Therefore, Unit 3 had excellent fuel-j performance.
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5.0 PLANT OR PROCEDURE Cil ANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED M AINTENANCE 5.1 Amendments to Facility License or Technical Specifications during 51ay,1993.
Amendment 122 to Unit 2 and 117 to Unit 3. These amendments relocate Primary Containment Isolation Valse Table 3.7.1 to DATR 3/4.18 " Component Lists." The DATR Table includes all talses credited as Primary Contairunent Isolation salves according to General Design Criteria (GDCs) 54, 55, 56, and 57, as resiewed by the NRC StafT under SEP Topic VI.4. Installed 05/19/93.
Amendment 123 to Unit 2. This amendment revises Figure 3.6.1, Vessel Pressurization Temperature (P/T)
Curves. Implementation of the resised cunes was accomplished prior to Unit 2 startup on 08/06/92 when the NRC allowed use of the Unit 3 P/T cunes for Unit 2 startup from D2F18. Installed 05/19/93.
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5.3 Significant tests and experiments not described in the FSAR (Units 2 & 3)
Significant special pmcedure involving tests rxit decritxx! in the FSAR which were approved during the month of April,1993, are listtxt twlow:
Special Procedure No.
Description SP 93-4-44 Unit 2 Reactor Water Cleanup System Surge Tank Ikron Hush Administrative Procedure (approved for use April 28,1993)
This prxicedure pmvidts the necessary steps to apply contmis for efTective managtment of the RWCU Surge Tank Decon Hush.
Sper2al Proctxture Safety Evaluation Summary SP 93-444 1.
The probability of an occurrence, the crnisapxnce of an accident, or malfunction of equipment important to safety as previously ev;duated in the Mnal Safety Analysis Report has not increased. Tinre are no ESAR amried accidents.
2.
The possibility for an accident or malfunction of a difiennt type than any preiiously evaluated in the ISAR has not txsn created. The SP ckxs not change die normal operation of the systtm and is performed during a plant optssting mode wlxn containment integrity is not rtxguirtx1, The procedure establi.sixs controls to restore the affected systtms and in no way degrades those systtms. No new harards or failure mod (s are intmduced.
3.
The margin of safety, as defined in the basis, for any Tahnical Specification, has not hmn reduced. (Itasis-affecial Tahnical Sperification: 3.6.c/ 3.7.Al 4.7.A/ 3.7.D/ 4.7.D). There are no Technical Spaifications wtxre the rrquirunents, associated action ihms, awriated surveillancts, or basis are afTecial.
Significant spaial procedures imo! ing tests not described in the FSAR which were apprmed during the month of May,1993, are listed below:
special Procedure No.
Description
- P 93-4-42 llandling And Loading of the Tramnuclear TN-RAM Shipping Cask (apprmed for use May 11, 1993)
The purpose of this procedure is to allow for the handling and loading of the 9
4 Transnuclear TN-RA si cn in a safe and efficient manner.
SP 93-4-43 Processing of Irradiated liardware Using the Underwater Shear / Compactor (approved for use May 10, 1993)
This procedure describes and controls the method by which control blades currently stored in the Dresden 2(3) spent fuel pools will be processed.
SP 93-4-45 Reactor Vessel Water Level Instrument System (ROVLIS) indication compaison Modification Test Modifications MI12-2-89-004A and M12-2-89-004B (approtea for use May 20, 1993) f The purpose of this modification test is to verify the proper installation of the Reactor Vessel Level Instrumentation System (RVWLIS) reference legs and sariable legs being installed under modifications M12-2-89-004A and M12-2 004B by comparing the indication readings between the dilTerent instrument channels. The special procedure will also document that the required Dresden Instrument Suneillance (DIS), and Dresden Operating Surveillances (DOS), and Dresden Technical Suncillances (DTS) have been performed and that the pressure i
compensation has been removed from narrow range level prior to startup of 32R13.
SP 93-4-46 Reactor Water Lesel Instrumentation System (RVWLIS) Operability Test, Modifications M12-2-89-004A and M12-2-89-004B (approved for use May 20,1993)
The purpose of this operability test is to verify the proper installation of the Reactor Vessel Water Level Instrumentation System (RVWLIS) reference legs and sariable legs being installed under modifications M12-2-89-004A and M12-2 004B by verifying level indications, monitoring the thermal growth of the e
condensing chamber temperatures. This Special Procedure will also serify that no sisible leakage can be deserted from the excess flow check valve and from unions downstream of the excess flow check vahes.
SP 93-4-47 Modification Test M12-2-90-057A Isolation Condenser Upgrade incomplete Modification (approved for use May 18, 1993) i The purpose of this procedure is to insure that diesel oil can be transferred to the Unit 2 Diesel Generator Day Tank via the U-2 Diesel Oil Transfer Pump. It also verifles that MO-2 4399-74 strokes properly.
SP 93-5-56 Turbine Deck Auxiliary Air Compressor Operation (approved for use May 20, 1993)
This procedure provides guidance for Auxiliary Compressor Pre Start-Up Checks, Compressor Start-Up, Compressor Operation and Shutdown when connected to the Service Air System. A portable air dryer has been connected to the Auxiliary Compressor discharge to provide dry filtered air.
Special Procedure Safety Evaluation Summary SP 93-4-42 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. There are no FSAR afTected accidents.
2.
The possibility for an accident or malfunction of a difTerent type than any previously evaluated in the FSAR has not been created. This special procedure will not n'fTect or degrade any safety-related systan, structure or component.
i All cask handling is subject to restricted mode operation of the overhead crane, limiting the cask movement to the predetermined restricted path.
The TN-RAM is an approved shipping cask for irradiated hardware. No new malfunction or accident type is created.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. (Basis - alTected Technical Specification: 3.10.F/ 4.10.F) The action items, surveillances, and hasis hase not been afTected.
SP 93 4-43 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. There are no ISAR afTected accidents.
l 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created.
No new failure modes are introduced. The Underwater Shear / Compactor (USC) will not he mosed above the fuel and the USC drop is bounded by a
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cask drop accident. Also, no new failure modes are introduced and no Technical Specifications are affected.
3.
The margin of safety, as defined in the basis, for any Technical-Specification, has not been reduced. (Basis-affected Technical Specification: None). There are no Technical Specifications where the requiranents, associated action items, associated surveillances, or basis are afTected.
j SP 93-44-45 1.
The probability of an occurrence, the consequence of an accident, or j
malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. There are no FSAR i
affected accidents.
l 2.
The possibility for an accident or malfunction of a different type than any
)
1
previously evaluated in the FSAR has not been created. No new hazards or failure modes are introduced. This SP verifies proper installation of RVWLIS modification and will not impact the function of any systems.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. (Basis-affected Technical Specification: 3.1/ 3.2/ 3.5/ 3.7). There are no Technical Specifications where the requirements, associated action items, associated surveillances, or basis are affected.
SP 93-4-46 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. There are no FSAR affected accidents.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. This SP verifies proper installation of RVWLIS modification and will not impact the function of any systems. No new hazards or failure modes are introduced.
3.
The margin of safety, as defined in the basis, for any Tahnical Specification, has not been reduced. (Basis-afTected Technical Specification: 3.1/ 3.2/ 3.5/ 3.7). There are no Technical Specifications where the requirements, associated action items, associated surveillances, j
or basis are affected.
1 SP 93-4-47 1.
The probability of an occurrence, the consequence of an accident, or I
malfunction of equipment important to safety as previously evaluated in
'l the Final Safety Analysis Report has not increased. There are no FSAR alTuted accidents.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. Stroking MO2-4399-74 is already done in the automatic mode as part of normal operation.
The design of the system allows for manual fill as necessary. MO2-4399-74 is in the clean deminerallier admission valve for the ISCO. This test will be done when the ISCO is not required and therefore, no new hazards or failure modes are introduced.
i 3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. (Basis-affected Technical Specification: 3.5.El 3.9). There are no Technical Specifications where the requirements, associated action items, associated surveillances, or basis are afTected.
SP 93-5-56 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. There are no FSAR l
affected accidents.
s t
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2.
The possibility for an accident or malfunction of a different type than any j
previously evaluated in the FSAR has not been created. Loss of service air i
system was evaluated in the FSAR. Loss of instrument air is an anticipated event throughout the life of the plant. No new hazards or failure modes are introduced.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. (Basis-afTected Technical Specification: None). There are no Technical Specifications where the requinments, associated action items, associated surveillances, or basis are alTected.
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5.2 Changes to Procedu es Which are Described in the FSAR (Units 2 and 3) during May,1993. Only those procedures that required a new or an additional 10CFR50.59 review of changes are included.
I Procedure Description DTS 1600-07 Dresden Technical Surveillance 1600-07, Unit 2(3) Integrated Primary Containment Leak Rate. The purpose of this procedure is to detail the steps necessary to perfonn the Primary Containment Type A Integrated Leak Rate Test (ILRT) required by Technical Specifications: Section 4.7. A.2.a, Primary Containment Primary changes: various valve line-up changes made. (Revision l
14, approval date May 11, 1993)
DOS 0500-11 Dresden Operating Surveillance (DOS) 0500-11, Verification of Neutron Flus Less Than Safety Limit. The purpose of this procedure is to document the Tech Spec Neutron Flux Safety Limit has not been siolated following an anticipated reactor scnun.
Primary changes: added clarity concerning acceptance criteria.
(Revision 03, approval date May 5,1993)
DOS 0500-15 DOS 0500-15, Operators Surveillance of Thermal-flydraulic Limitations on Power Distribution. This procedure outlines the method to be used in perfonning the daily surveillance of the Local Steady State Linear IIcat Generation Rate (LIIGR), the Local Transient LIIGR, the Aserage Planar Linear IIcat Generation Rate (APLIIGR), and the Minimum Critical Power Ratio (MCPR).
Primary changes: added clarity concerning acceptance criteria.
(Revision 02, approval date May 14, 1993)
DOS 0500-18 DOS 0500-18, Verification of Flow Control Line and Average Core Thennal Power. This procedure verifies the average Core Thermal Power (CTP) remains less than or equal to 2527 MWth during the course of a nonnal shift and the Flow Control Line (FCL) runains below the calculated FCL limit.
Primary changes: added clarity concerning acceptance criteria; revised corrective actions for esceeding the FCL limit. (Revision 13, approval date May 14, 1993)
DSSP 0100-F Dresden Safe Shutdown Procedure (DSSP) 0100-F, Ilot Shutdown Procedure - Path F. This procedure provides guidelines to achieve Dresden Unit 3 shutdown, directed from the Control Room, using Ilot Shutdown Path F following a fire in which critical plant components were rendered inoperable.
Primary changes: added attachments for actions of individual plant personnel. (Revision 05, approval date May 19, 1993)
DSSP 0200-T6 DSSP 0200-T6, Temporary 4 kV Feed Connections - SDC, LPCI, RBCCW, CCSW. This procedure provides instructions for connecting temporary 4 kV power to Shutdown Cooling (SDC),
Low Pressure Coolant Irdection (LPCI), Reactor Building Closed Cooling Water (RBCCW), and/or Containment Cooling Service Water (CCSW) pumps, as applicable, after a fire which may have damaged normal supply cables or switchgear, i
Primary changes: addressed electrical safety concerns; provided greater detail in various steps. (Revision 02, approval date May 6, 1993)
DSSP 0200-T9 DSSP 0200-T9, Cable Connections For Monitoring RPV Water, Shell, and Flange Temperature Locally. This procedure pro / ides instructions for connecting instruments to monitor parameters locally at drywell penetrations, using temporary instruments, after a fire which may have damaged normal instrument cables or rendered normal monitoring locations uninhabitable.
Primary changts: in place of RTDs, connections to be made to thermocouples because the RTDs do not have sullicient range.
(Resision 02, approval date May 6,1993)
DSSP 0200-T13 DSSP 0200-T13, Supplying Unit 2(3) Reactor Building 125 VDC Distribution Panel #2(3) From Unit 3(2) Reactor Building 125 VDC Distribution Panel #3(2). This procedure provides instructions for supplying power to Unit 2(3) Reactor Building 125 VDC Distribution Panel #2(3) from Unit 3(2) Reactor Building 125 VDC Distribution Panel #3(2) after a fire which may have damaged the normal source of power or supply cables.
Primary changes: provided more detailed instructions on removing bus links; added requirement to obtain a DC voltmeter
.)
or equivalent. (Revision 02, approval date May 6,1993) 4 l
1
Procedure Safety Evaluation Summary DTS 1600-07 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. There are no ISAR alTected accidents.
2.
The possibility for an accident or malfunction of a difTerent type th.m : ny preiiously evaluated in the FSAR has not been created. The test method remains unchanged. Only the details necessary to perform the test have changed. The integrity of primary contairunent will be maintained and tested per the FSAR and current regulations and testing practices.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. There no Technical Specifications where the requirements, associated action items, associated surveillances, or basis are affected.
DOS 0500-11 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as presiously evaluated in the Final Safety Analysis Report has not increased. This procedure provides serification that the 1.5 second trip of RPS from the APRMs occurs as required. If it is detennined that the 1.5 second limit is exceeded then the necessary actions are provided as delineated in section 6.4 of the Technical Specifications.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. This procedure provides the verification of the 1.5 second trip of RPS from the APRMs occurs as required. The necessary actions are provided in the unlikely event the limit is exceeded.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. This procedure verifies the safety limit after the fact ot ensure plant startup will not occur if it has been violated.
DOS 0500-15 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety
P Analysis Report has not increased. This procedure monitors the nuclear (thermal) limits of the core, thereby ensuring the analyzed portions are within the constraints of the FSAR.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. This procedure monitors the nuclear (thermal) limits of the core, thereby ensuring the analysis of the FSAR remains in effect.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. By verifying the core nuclear (thermal) limits the assurance of Tech Spec adherence is defined.
DOS 0500-18 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. This procedure verifies the CTP and FCL are within the prescribed limits and prmides corrective actions should they be found out of the limits, thus providing assurance that the analyzed limits are adhered to.
2.
The possibility for an accident or malfunction of a ditTerent type than any preiiously evaluated in the FSAR has not been created. This procedure verifies the CTS and FCL are within the prescribed limits and provides corrective actions should they be found out of the limts, thus providing assurnace that the plant will stay within the FSAR limits.
3.
The margin of safety, as def~med in the basis, for any Technical SpeciGcation, has not been reduced. This procedure verines the CTP and FCL are within the prescribed limits and provides corrective actions should they be found out of the limits, thus pr9 tiding assurance that the plant will stay within licensed cmstraints.
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DSSP 0100-F 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously av;i:md h t Final Safety Analysis Report has not increased. The accident has already occurred when this procedure is implemented.
The procedure would mitigate the consequences of equipment failure by using alternate paths to perform the shutdown.
l 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR i
has not been created. No new failure modes are anticipated. Equipment has already failed due to the fire. This procedure shuts down the plant using j
alternate methods described in the Safe Shutdown Report. FSAR 10.7.4.2 references Safe Shutdown Report which is used as the basis for this procedure.
1 1
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. Use of this procedure is outside the requirements of the Technical Specifications per 10 CFR 50.54X, Appendix R, and the S: fe Shutdown Report.
DSSP 0200-T6 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. The accident has already occurred when this procedure is implemented.
This procedure repairs equipment damaged by the fire to allow safe shutdown per 10 CFR 50, Appendix R.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. Equipment addressed in this procedure has already failed due to the fire. This procedure enhances capability to protect the public by placing systems that were damaged back in service to allow safe shutdown.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. Use of this procedure is outside the requirements of the -
Technical Specifications per 10 CFR 50.54X and the Safe Shutdown Report.
DSSP 0200 T9 1.
The probability of an occurrence, the consequence of an
accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. The accident has already occurred when this procedure is implemented.
Probability will not increase. This procedure would mitigate the consequences by placing failed equipment back in service.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. Since the systan is already inoperable, the possibility of a malfunction does not exist. When returned to service after emergency repairs, the probability would be no greater (no increase) than before.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. Use of this procedure is outside the requirements of the Technical Specifications per 10 CFR 50.54X, Appendix R and the Safe Shutdown Report.
DSSP 0200-T13 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report has not increased. The accident has already occurred prior to implementing this procedure.
The procedure would mitigate consequences by placing safety equipment back in senice after failure due to fire.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. This procedure mitigates an accident by placing sital equipment back in serSice.
3.
The margin of safety, as defined in the basis, for any Technical Specification, has not been reduced. The use of this procedure is outside the requirements of the Technical Specifications per 10 CFR 50.54X and the Safe Shutdown Report.
)
)
4 TABLE 5.2.1 CilANGES TO PROCEDURES WillCII ARE DESCRIBED IN Tile FSAR ( UNITS 2 AND 3 ) FOR May,1993 PROCEDURETYPE PROCEDURE NO.
PROCEDUR E TITLE /DESCRIITION SUMM ARY OF CilANGES DTS 1600 07 SEE SECTION 5.2 FOR DESCRIITION 2
SAFETY RELATED PROCEDURE CllANGES FOR Tile MONTil OF May ARE DESCRIBED IN DOS 0500-11 4
SECTION 5.2 OF Til!$ REPORT.
DOS 0500-15 4
DOS 05001 A 2
DSSP 0100 F SEE SECTION 5.2 FOR DESCRIPTION I
DSSP 0200-T6 2
DSSP 0200-T9 4
DSSP 0200-T 13 SEE SECTION 5.2 FOR DESCRII410N 2
1 intent of rocedure unchanged.
540TES:
- 1. Administrative change:
P
- 2. Changed for clarification. intent of procedure unchanged.
- 3. Changed to incorporate requirements for new equipment; intent of procedure unchanged.
-1. Changed to irnplement improved testing I colit' ration methodology; intent of procedure unchanged.
1 (GilKLT92 / 2) / 6
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5.4 Safdy Related Maintenance (Unit 2 and 3)
Safety related maintenance activities for May,1993, are summarized in the attached tables.
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bA6LiY RLLAlLU N;i1HILdFNLt i
tJATURE tW LER UR OtJTAGE mat FilNCTIntt EQUIPMENT MdIN TE NANCE NilHE4 R C A U".0 FESUIf f:0RRECT)VE ACTION 1404-31 CORRECIIVE N/A
~ - - - - -
NORK COMPLI:T EI) [N At:I:0RDollEl: Hf Til FCII otul$ER CORE SPRAY LINE 1116707-tluRK PACK (die 01473/-01 Allu INR, 1404-10' wo t A.?.1 I'l<EVLHT IVE tt/A HORK COHrlt:TEt) (N AttoRnANCl~ HITil l~CII f!Ul*ER RECIRC RING llEADER Ill6700 MORK PAi:K AGE lt167:'M-01 Atih flWR.
3204-24 PREVENTTVE N/A HORK PERFORMED ANIJ COMPLI:# E D IN
.t!Ul*FR IEEDWATER LINE fil6/93 ACCORI'AtlCE WilH FCil WORK PACKAGE l
-C040-12' h167V t-01 ANlt HMR, i
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301YA9 7 CURRECTIVE n/A HORK CUNFLl:TED [N ACCORDONCI: M ITil T P '
"HilDDER l'ARGEl ROCK 3A DUTLET D16/94 HHR AND FCII WORK 1:EOUl:St D16794-01, t.lHE 2-3019A-n*
2-3019A-60 CORRECTIVE N/A HORK COMPLETED PER lilIS H.W.R. AND FCII SHUDBER TARGET ROCK 3A QUTLET D16790 HORK PACKAGE 014716-01.
IINE 2-3019A-O' l
l J-3019C-U5 CORRECTIVE N/A WORK COMPLETED PEG TllIS NMR AND FCII
!;Hul$ER lARGEI ROCK 3A OUTLET 016/94 HORK l'ACKAGE D147V4-01.
1INi? 2-3019C-M' 2-3019f3-53 CORRECTIVE N/A HORK COMPLCTED PER TilIS NMR AND FCII SNiilSER TARGEl~ ROCK 3A DUTLET D16790 HORK PACKAGE D167V3-01.
!INE 2-3019D-O' 2-301VE.
CORRECTIVE N/A WORK COMPLETED PER TilIS NMR AND FCII SHUDBER TARGET ROCK 3A OUTLET D16799 WORK PACKAGE D16799-01.
LINE 2-3019E-n' 2-24200 CORRECTIVE N/A DETORQUEING SWITCll NOT ENGAGING. BENCH' RECORDER DRYWELL HIGH RANGE D16312 CALED PER DATA SHEET 5 0F DIS 1600-16 B.
MONITOR RADIATION REINSTALLED IN V02-56 PANEL.
3 02018-36 CORRECTIVE N/A CORK COMPLETED IN ACCORDANCE WITil FCII l
SHUI'$ER RECIRC PUMP OUTLET LINE 1116053 HURK PACKAGE D16850-01 AND HHR.
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e bAlLiY htLr "NLL NATURE OF LER OR OUTAGE t u, ION EQUIPMENT MAINTENANCE NUMBER RESULT CORRECTIVE ACTION 2-1301-10 FFEVENTIVE N/A VALVE MD ISOLATION CONDENSER D06401 HORK COMPLf:TED FER TilIS NNR AND FCII J-------
/ FILL INLET #10 WP D06431-01 M-11780-10 PREVENTIVE N/A
%UPPORT llEAD SPRAY 0214-2" D066SH READJilSTED SPRING llANGER FRDM i U/G" ElAHS 1 M-11700-10 COLD SETIING TO 3/: ".
M-1176D-Z f'REVENTIVE ft/A SilPPORT f4SDN 7-30078-1.G' CLASS D066S" UtlPPORT WAS Afh.7tJ3 rf'D TO NOTED DIMENSIOff.
1 M-11760-2 AftJUSIED SPRING CAN IO DDfAIN PROPER TENSION ANH SET TING.
2-0203-20 PREVENTIVE N/A VALVE A0 OUTBOARD MAIN STEAM D07217 BONNET LEAK OFF LINE WAS NELDED Cl 10.
ISOLATIDH 2D REPACED VLV. SATIGFIED ALL DMPS.
. 2-0203-2D PREVENTIVE N/A VALVE A0 OUTCOARD STEAM ISOL 2D 00721~7 DONNET L EAK OFF LINE WAS HELDED CLOSED.
RLPACKED VIV. SATIbFIEI) AlL OMPS.
20?293-1D PREVENTIVE N/A UALVE A0 INPDARD MAIN STEAM D07223 INSTALLED NEW PACKING, 1 i:RAIDEDs 3 ISOLATION 10 GRAPHITE, tAST ONE CRAIDEU. BRASE DUSHING AND SNUG UP 10LLOWER AND NUTS.
=
FOLLOWER IS' LEVEL.
2-2301-2Y CORRECTIVE N/A e _ req 1 VALVE GLOBE A0 HPCI STEAM LINE 007704 INSTALLEDNEWINTERNALS.REASBtMBLEDM 10ROUED. SET LEFT PRESSURE.& REPACKEO.y.
o DRAIN ISOL.
EMD REINSTALLED LIMITS.
<?1 j.-
- s.. i n w a me. q %
2-4399-72 PREVENTIVE N/A 4
VALVE RELIEF A LPCI HT EX SHELL 007753 LAPPED VALVE. RETESTED. VALVE PASSED.ATIS s
SIDE RNED DVER TO M.M.S.G. FAILED TEST.
e HEN NWR D17859 HRITTEN.
' Q
,p s 2-1601-56 PREVENTIVE N/A REBUIL' TAD.HONEDCYLINDERWALLS'T0b OPERATOR A0 VALVE DRYMELLL VENT 007801 456 0.00V, COULDN'T DET LINE GUT. TESTEQi>
.35 LB IN 5 MIN, RAMSEY SAID TO INSTALL AG, REPLACED CHECK VLV.
2-2301-51 PREVENTIVE N/A
~
VALVE CHECK HPCI AUX COOLING D08884 WORK COMPLETED IN ACCORDANCE WITH THIS WATER PUMP DISCil NWR AND FCII WORK PACKAGE 008884-01.
4 w
!l NATURE Or'
- LER OR UUTAGE MALFUNCTION EQUIPMENT
' MAINTENANCE NUMBER CAUSE
-RESULT CORRECT]VE ACTION y
E-1402-?A PREVENTIVE N/A
'" VALVE CHECK CORE SPRAY INLET TO 009482
=--
DISA35EMBLED, INSPECTED AND FIEA35EMii!8P Xf-0 2-0263-17A PREVENTIVE N/A RLMOVFD FLOH CHI CK,1 APED GPCit1NUS, VALVE FLOW LIMITING CHECK RPV 009498 CLEANED VALVE AND INIERNALS, ICUfED Ott, INSTRUMENTS.
Oi 2-0263-2-15A PREVENTIVE N/A REMOVED FLOWCHECK, TAPED OPENINOU, VALVE FLOW LIMITING CHECK VAR D09499 CLEANED VALVE INTI:RNALS AND PERFORNEO O_. LEG FOR 2202,
FLOW CHECK.
INSTALLED FLOW CllECK HIfil 3.. Q.
APPROVED SEALANT.
4.2-0263-2-11 PREVENTIVE
- /A
~n ' VALVE FLDW LIMITING. CHECK LOW
'D09501 REMOVED FLOW CHECK AND TE31ED. INST All Elt-LT-263-61 FLOW CHECK WITH APPROVED SEALANT.
yshs(.Mir t
]t1001-0A 5
^
PREVENTIVE N/A PERFORMED MORK A5 PER WORK INSTRUC1IDMG 0PERATOR MOV 2-1501-5A 009698 DEP 0040-09. COMPLETED DEP 0040-27 N<
D ',
%-g y;&.
L SIGNATURE.
a.
001A-38 PREVENTIVE N/A
',# -1001A-16"-B D{0020 HORK COMPLETED IN ACCORDAHCr W11H lH10 BBER SHUTDOWN COOLING LINE NWR AND FCII WORK PACKAGE D10020 -01.
yr 2-1501-20 PREVENT 1VE.
N/A COMPLLTED HORK IN ACCORDANCE HIlli THm
---~~ - - ~ ~ - - - - - -
SNUBBER TORUS RING LEADER LINE D10021 t
, 2-1501-20 NWR AND FCII HORK PACKAGE D10021-01.
2-1522-23 PREVENTIVE N/A NORK COMPLLIED IN ACCORDANCE HIllt lHD SNUGBER LPCI TEST LINE D10022
' 2-1522-23-14" (B-LOOP).
NHR AND FCII D100?2-01.
2-2305-01 I
PREVENTIVE N/A COMPLETED MORK IN ACCORDol>Cf M11H 110 '
SNUBBER HPCI STEAM SUPPLY LINE D10023 2-2305-10" NHR AND ICII WORK PACKAGE D100?3 01.
I 2-0202-1 PREVENTIVE N/A HORK' COMPLETED IN ACCORDANCI.' H[161 fHt4 SNUBBER D RECIRC HOTOR 1 D10025 4
NWR AND FCII HDRK PACK AGr. D1002R01.
'pR >
8; I
t
's i
I rw-s e-nt te e emp y y ?
___.____________,______.._______.__m
- .___.__m-.
.....s
_. _,... - _ = _., -., _ _ -.. _ -..
bat t l Y ht.LH I LU PlH L N I Af4eNLL NATURE OF 1.ER GR OUTAGE M ALFilNC T ION EQUIPMENT MAINTENANCE NtlM6f R CAUSE RESUIT CORRECTIVE ACTION 0!O3-33 PREVENTIVE ff/A WORK COMPLCTED IN ACCORDAffCC HITil THIS s u llSFR TARGET ROCK VALVE D10036 0?O3 3A tlWR AND FCII WORK PACK AGl: D 100 %-01.
m2R-7C1 PREVENTIVE N/A link F%t' RX RfCIRC PP SUCT VLV D10%6 PilLLED BRK. Cl.EANFO AND RtPiACEn THE AUX 20"-4A l.ONTACTS, COILS, INANSIGRMER, OVERHEAf1 RFLAYS, I E STF n 'l Hl: PRK ant i OVERHEAD Rt L AYS. lESTEU OK.
- iT47 PREVENTIVE N/A allOR HPCI ROOM COOLER Ot2731 Ct.EANED FAN 9 REPLnt E D f-:El 13, RFPLACCD DOLIS AND GASKEl 10 PECKE R llEAli, PERFORMED DEP 5700-04 klV.
1.
1402-30A PREVENTIVE N/A niERATOR MOV CORE SPRAY PP MIN D13392 RFMOVED LIMIT & THRQ SWTCtt. RFMOVED 110W VLV 11 EXES /TIEli 1 HEM Ul F.CONIh0L WIRES R J 10 CE REtVGGEO. Mul0R L EAliS WL RE RAYCHEM WL CUT THEN OF F IN CI N1ER OF tlT T SPLICE.
I fiO 1 -228 PREVENTIVE N/A OVERHAULED AND REAUSEMCLI0, SIGNATURE R di'ERATOR MOV D1375M DID MOD.
DOOR WATERTIGHT EAST LPCI PREVENTIVE M/A REMOVED & REPLACEO PACKING IN LATCH &
.HRHER ROOM D13773 HANDLE. LUCEO HANDLES ANtt CLEANED NEW.
TURNED DOOR SEAL AROUND, ADDRESSED ALL PMV AND OC HOLD PulNTS.
_-700-APRM2 CORRECTIVE N/A HuMITOR AVERAGE POWER RANGE D15197 REPLACED POWER SUPPLY, SI 506C14. SET ALSO CHECKED M CitANNEL 2 OllTPUT PER DIS 0700-09.
klPPLE OF NEW POWER SUPPLY PER
.4 DIS 0700-09. RED TAGGED OLD POWER SUPP : i
. r.&
+
2-1641-2008 PREVENTIVE N/A liFCORDER SUPPRESSION POOL TEMP D15530 TURNED RECORDER ON, COULD NOT FIND'* N N PROBLEM., MONITORED FOR TWO SHIFTSt % S SIILL NO PROBLEM. INKED RECORDER.~,
^
b n
' 1501-26 A CORRECTIVE N/A HAND OPERATED LIMIT TO CLOSED POSITIDH SWIICH LIMIT CLSD LPCI MANUAL D16401 UI V 26A VERIFIED LITE INDICATION CAME UP.-...i..t 3
REPLACED OLD SNAPLOCK. RELUGGED TERMI -t ATIONS. REPLACED ARM ON OPEN LMT SWITCH 4 '
_-305-18-23 CORRECTIVE H/A HYDRAULIC CONTROL UNIT #13-23 D16409 CtlT 2 LOWER' STAND OFF BLOCKS OM GAUGE TO INSTALL GUAGE.
REMOVED SCREWS W/ NUTS OH I:ACK. REMOVED PER ATTACHMENT J AND REPLACED.
e
?
.mrt..
.u. t i. e u.
..e m, u,..n u NATURE OF LER OR OUTAGE MALFUNCTION EQUIPMENT NAINTFNANCE NUMBER CAUSE RESUI.T,
CORRECTLVE ACIION 2-7n28-ZD3 CURRECTIVE N/A
~
REPLACED Dil HEATI-R5 ON U2 A RPS MG set SWGR BRK RX PROTECTION SYS MG D16713 j SET 2A PERFORMED AMP CilEtt UN A RPS St T AFTER 00S WAS CLI ARED. OA 13.3 AMPS, OB 38.6 AMPS, OC 40.2 AMPS.
2-263-73A-LD CURRECTIVE N/A RI PLACED PROPER F1 TTING ll E'S ft TUDING VALVE U2 CORE COVERAGE LT D16721 PER INSTRUCTIONS. DID 1.EAK IEST OF 2-263-73A LO DRN XMIITER 2-23-151A 0F 1500 PSI. TIGHIENs LOOSE FITTINGS R PFES5tlRI LOOKED FINE.
M-1151D-2 CORRECTIVE N/A WORK COMPLETED IN ACCORDANCE HITil THIS -
SUPPORT llPCIS 2305-10' CLASS 1 D1676" NWR AND FCII WDRK l ACKAGE D16763-01.
2-1102-B CORRECTIVE N/A AhJUSTED PACKING PER DNP 1100-03 *'TESl PllMP STANDBY LIQUID CONTROL fB D16775 REV.
2-1303-01 CORRECTIVE N/A WORK COMPLETED IN ACCORDANCE WITil THIS SNUl*ER ISO COND RETURN LINE D16'/80 NWR AND FCt1 WORK l'ACKAGE D16700-01.
2 1303-12' 2-1303-02 CORRECTIVE tt/A HORK COMPLETED PER TilIS NWR ANO FCII SNUBBER ISO COND RETURN LINE D16701 WORK PACKAGE D16701-01.
2-1301-12' 2-1404-01 CORRECTIVE N/A WORK' COMPLETED PER THIS NWR AMD FCII;/'
SNUBBER CORE SPRAY LINE D16733
,HORK PACKAGE D16783-01.
2-1404-12' (B-LOOP)
' 7 l$
2-1404-02 CORRECTIVE N/A COMPLETED WORK PER THIS NWR AND'FCII?.
SNUBBER CORE SPRAY LINE D16784 HORK PACKAGE D16784-01.
2-1404-12' (B-LOOP) 5,.
d.{ n y
19 y, FERTHISNNMkND'FCIkk' 2-1404-03 PORRCCTIVE N/A COMPLETED MORK SNUBBER CORE SPRAY LINE D16705 HORK PACKAbE D16705-01..
2-1404-12' (B-LOOP)
~
,,a 2-1404-04 CORRECTIVE N/A COMPLETED WORK PER'THIS NWR AND FCI SNUDGER CORE SPRAY LINE D16786 WORK PACKAGE D16706-01.
2-1404-12' (B-LOOP) e 8
m
~*1 e
($
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I SAFE 1Y RLLAILU NAINILNANGL NATURE OF LER OR OUTAGE M ALFilNC TION EQUIPMENT MAINTENANCE NUMBER CAUSE RESUIT CORRECTIVE ACTION 0-305-130-26-35 CORRECTIVE N/A b------
3HT TCH At.CUMUL ATOR FRESSURE CR0 D 16'/ ]6 CHANGED PRISSURE SWITCH AND CALIDRATED
, ?6-35 PLR D15 300-2 TO DATISFY PMV.
2-702-IRM-13 CORRECTIVE N/A INSTAL LED NEW DRIVL IUDE, INS FALLED NEC]
INTERMEDIATE RANGE MONITOR MISC D17129 lhM DElECitiR.
.2-750-78 CURRECTIVE N/A RI CONNECTEli CAE:t E AT FREAMP.
I h0NITOR INTERMEDIATE RANGE D17151 l
CH 15 DISCONNECTID PER HR D063.?O lER D1 S 700- 31.
J'-1140-3D CURRECTIVE N/A METER SOLC SQUIB MONITOR RELAY D17.?66 1.IFTED LEADS FROM OLD CONNECTOR AT
.HINCT ION OtiX, RI. MOVED CUNNECTOR.
INST ALLED HEW CONNEC10R,1 ANDF11 LEADS.
2-1501-5B PREVENTIVE N/A MOTOR MOV LPCI PP B SUCTION VLV D17499 OVERHAULED OPERATOR. EMD PERFORMED MOV SIGNA 1DRE 1)K.
2-750 CORRECTIVE N/A INSTRUMENTATION STARTUP RANGE D l?!i? 3 RFPLACED MFAN SQUARE ANALOG HODULE ON 1RM 14, REPLACEll l'REAMP ON IRM 10.
PiRFORMED INPUT VOLTAGE 1151S AS PER WORK INSTRUCTIONS.
2-750-7A CURRECTIVE N/A INSTRUMENTATION STARTUP RANGE D17593
' REPLACED NEAN SOUARE ANALOG MODULE ON
WORK INSTRUCTIONS.
[
2-750-78 CORRECTIVE N/A INSTRUMENTATION STARTUP RANGE D17593 RFPLACED'MEAN SQUARE'ANAE00 NODULY DN.
CH 15
'IRM 14, REPLACED PREAMP DN IRM 18.
a PERFORMED INPUT VOLlAGE TESTS AS PF" g.
WORK, INSTRUCTIONS.
., q.
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2-750-7C CORRECTIVE N/A IREPlikCdiFMeAR 900ARECANAtt10 OIODULY INSTRUMENTATION STARTUP RANGE 017593
'IRM 14,- REPLACED PREAMP DN IRM 18,dR CH 12 M
PERFORMED INPUT VOLTAGE TESTS AS P j.
WORK INSTRUCTIONS.
- q
- ,z,E3 2-750-7D CORRECTIVE N/A
' REPLACED MEAN SQUARE ANALOG 'N000LE' bH INSTRUMENTATION STARTUP RANGE D17593 Cil 16 IRM 14, REPLACED PREAMP DN IRM 18
.g PLRFORMED INPUT VOLTAGE 1ESTS AS PER WORK INSTRilCTIONS.
38 e'-
NATURE OF LER OR OUTAGE MAI FuttCTION EQUIPMENT MAINTENANCE NUME'ER CAUSE RESULT CORRECTIVE ACTION
- - 750-7E CORRECTIVE Pf/A
--W----
REPLACED NEAN SQUARF ANAL.OG MODULE ON INSTRUMENTATION STARTUP RANGE 1117573 cli 13 IRM 14, REPLACEO PREAMP DN IRM 13, F ERFORMED INPUT VOI.1 AGE Tl: SIS AS PER HORK INS 1RIJCTIONS.
.: ~ 750 -7F CURRECTIVE N/A INSIFUMENTATION STARTUP RANGE D17593 REPLACED MFAN SQUARE ANALOG MODULE OH tH 17 IhM 14, REl'LACEO lREAMP ON IRM 18.
IT RFORMElf ]NPUT VOLTAGE ll-515 AS PER FORK INSTRUCTIONS.
.'-750-7G CORRECTIVE H/A INSIRUMENTATION STARTUP RANGE D17593 REPLACED MEAN SOUARE ANAt.00 MODULE ON I'll 14 IRM 14, REPLACED PREAMP DN IRM 18, PERFORMED INPUT VOLTAGE TESTS AS PER HORK INSlRUCTIONS.
2-750-7H CORRECTIVE N / r, INSTRUMENTATION STARTUP RANGE D1759]
REFLACED MEAN SQUARE ANAIDO MODULE 0" ril 10 1RM 14, REPLACED lREAMP ON IRM 10.
N:RFORMED INPUT VOLTAGE !! SIS AS PEh WORK INSTRilCTIONS.
2-1554A CORRECTIVE N/A SWI1CH LPCI PUMP DISCHARGE D17640 REPLACED VALVE DTIrli NEW ONE FROM STORES.
PRESSURE S1 4808B24.
2-1501-4D PREVENTIVE N/A VALVE MANUAL GATE
'B' CONT D17670 TIGHTENED PACKING.
~~-
El.GilX SERV WTH OUT
'C-
- L~.
- '-2320-GSEF PREVENTIVE N/A NOTOR CONTROL CENTER HPCI GLND D33699 NORK COMPLETED PER THIS NWR ' ANDTCIINf '
y*
b STM EXHST FAN.
WP D83699-01.
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f 5.5 Completed Safety Related Modifications (Units 2 and 3)
Modifications which have been authorized for operation during April,1993 are listed. For case of reference, the changes have been identified by their design change control modification number. Previously, only modifications that had bten completely closed out were reported.
Modification No.
Description M12 2-89-011C To install a new fire protection control panel to replace the existing Panel 2252-44. To replace the local alann hell, manual pull stations, and deluge soler.oid valve. Also, to remove the tamper switch on the deluge gate vahe. This modification will upgrade the existing 112 seal oil unit fire protection deluge system and will provide supervision for the ekttrical components associated with this deluge system ensuring compliance with NFPA 72D,1975 Edition.
Modification No.
Safety Evaluation Summary 2
M12-2-89-011C 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not been increased. There are no FSAR affected accidents. This modification does not change the results and conclusions of the Fire llazards Analysis and Safe Shutdown Analysis and, i
therefore, it does not create the possibility of an accident or malfunction of a type different from those evaluated in the SAR.
2.
The possibility for an accident or malfunction of a different type than any presiously evaluated in the FSAR has not been created as a result of the j
modification. The modification has no effect on 4
operating modes or equipment functions.
3.
The margin of safety, as defined in the basis, for j
any Technical Specifications, has not been reduced.
The change does not affect any parameters upon which the Technical Specifications are based.
i I
Modifications which have been authorized for operation during May,1993 are listed. For ease of reference, the changes have been identiGed by their design change control modification number. Previously, only modiGeations that had been completely closed out were reported.
Modification No.
Description M12-2 89-017B This change prosides an online vibration monitoring system to view reactor recirculation pump sibrations to provide for early indication / detection of recirculation pump shaft integrity and imminent shaft failure.
M 12-2-91-002 The proposed change consists of remosing a section of isolation (ISO)
Condenser piping and repairing the inaccessible pipe to Dued head welds for two containment penetrations. The replacement piping and weld repairs will be accomplished using IGSCC resistant materials.
ModiGcation No.
Safety Evaluation Summary M12-249402 1.
The pruhability of an occurrusice or die consequence of an accident, or malfunction of equipmesit important to safety as previously evaluated in the ISAR has not been increased. During the postulated aaident, the main steam line isolation valves will close, and the ISO condenser will be utilized. The piping replacanent and weld repair das not charige the liklihood of a main steam line break. The new pipe and repaired penetration maintain the systan pressure boundary, just as the existing piping systan did before implementation of the modification. Thus, the consequences of the acddent are not incrt-ad.
2.
The possibility for an acrident or malfunction of a differait type than any previously evaluated in the ISAR has not been created as a rrsult of the modification. The modification has no effat on operating modes or equipment functions.Tids modiGcation dms not change the configuration of the existing systan, and eliminates the possibility of any failure modes associata! with IGSCC of piping systans. Therefore, there will be no acri& sits or malfunctions of a type diffensit from those evaluatal in the SAR 3.
The margin of safety, as defined in the basis, for 4
1
Modification No.
Safety Evaluation Summary
- 3. (con't) any Technical Specifications, has not been reduced.
The change does not affect any parameters upon which the Technical Specifications are based.
M12-2-89-017B 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not been increased. There are no ISAR affected accidents. This modification does not change 2.
The possibility for an accident or malfunction of a difTerent type than any previously evaluated in the FSAR has not been created as a result of the modification. This modification installs a passive j
on-line vibration monitoring system on the reactor l
recirculation pumps in order to detemnine pump j
shaft integrity and provide indication / warning of imminent shaft failure. There is no effect on any plant system, structures, component or interface with any system under any plant operating mode.
The modification has no elTect on operating modes or equipment functions.
3.
The margin of safety, as defined in the basis, for any Technical Specifications, has not been reduced.
The change does not affect any parameters upon which the Technical Specifications are based.
5.6 Tunporary Syst4m Alterations Installed (Unit 2 and Unit 3)
A " Temporary System Alteration" refers to eintrical jumpers, lifted leads, remosed fuses, fuses turned to non-conducting position, fuses moved from normal to reserse holder, temporary power supplies, test switches in alternate positions, temporary blank flanges, and spool pieces. Alterations controlled and documented as part of a routine out-of-service or other procedure, alterations which are a normal feature of systan design, and hoses installed as part of a senting or draining process are not included.
TIIE FOI.LO. WI.:N_G_I_NFY_)RM.- A.. TION REHECTS T..EM.M)R ARY A. I..TERATIONS WillCII WERE w.-
n-
-- - - - - ~
~m INSTAL. LED DURIIiG TIIE MONTII OF MAY 1993 LOG #
WR$
DE$C RFPTIO N R[ ASON LihGTH ifv5T RCMVD El i
5-32 93 D19568 f te massory er cornpe,emoe was used to The A.inil.sev er compeeesoe won 5 Months 19 May 43 e
proviele taedup serwece of du'eng corp cled to the Uret 2 Seevice Apr henaev-trout =estwsoong ered reper of tre Utet 3 beewsce An compeessor f ree compeceepr tu mi 3 E Al Pee estarteced degemoed pettormeesce oevd miei tse pornowed teure serve 6e i
5 m 33 43 D18963 Tees temp at mvolved deiltrug a hule in The wave poco.etig for 2 220102 ve ve hetuel Cyde 27 Mov92 W
s tr.e yuee secnon of veve 2 2201C2 tot was semeaeig A huse was d*Jeed and the tevecton of Furenervie oemerit erwected with f uemorate to stop Une 4eenage I
l
+
mcs 93 D18573 True temporary esteesoon peuv*ood powee t o provute eeecie.cei powee to He 3 Montes 20 Me, 8 3 e
to a toeripo*ecy mi compieseot fearn t'ee tempureev ese cump*essor support top ttie j
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I 5.7 Other Units 2 and 3 Required 10 CFR 50.59 Evaluations Other required 10 CFR 50.59 evaluations include Set Point Changes (SPC), Rigging Evaluations and changes to equipment not reported in Sections 5.2 through Section 5,6.
The following SPC was implemented in May,1993:
i Installation No.
Description SPC 2-93-037 To provide proper relay settings to support the replacement of the existing GE type EC trip desice with the R51S-9 model for the S. Turbine Room Vent Fan 2 B breaker -
on 480V Switchgear Bus 29 as stated on the relay setting order dated 11/16/92.
Installation Safety Evaluation Summary SPC 2-93-037 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previousiv evaluated in the Final Safety Analysis report has not increased. There are no FSAR affected
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accidents.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. This setting change for the safety related circuit breaker to non-safety related South Turbine i
Room Vent Fan 2B is to facilitate the replacement of the EC type trip device with an R3tS-9 model. The settings for the Rh1S-9 device are more stringent than the settings for the present device (EC.2A) with respect to the long time delay and instantaneous current settings. The more stringent overcurrent trip requirements are to present an equipment fault from propagating, damaging the bus and other equipment. The replacenwnt device incorporates an improved design which decreases the failure probability associated with the present device. The design replaces the dashpot device trip device with solid state circuity because dashpot trip devices are prone to leaking. These improsements increase the reliability of the associated systems. IIence, the probability of an accident due to setpoint, device or associated systems failing, or the consequences of an accident are not increased. No unevaluated failure modes are thus generated.
i 3.
The margin of safety, as defined in the basis, for any Technical Specificahon, i
has not been reduced. (Affected Technical Specification: 3.9/4.9) The change does not affect any parameters upon which Technical Specifications are based.
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