ML20058G815
| ML20058G815 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 10/31/1993 |
| From: | Sykes K COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20058G813 | List: |
| References | |
| NUDOCS 9312100147 | |
| Download: ML20058G815 (39) | |
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MONTHLY NRC
SUMMARY
OF OPERATING EXPERIENCE, CIIANGES, TFSTS, AND EXPERIMENTS PER REGULATORY GUIDE 1.16 AND 10 CFR 50.59 4
FOR DRESDEN NUCLEAR POWER STATION COMMONWEALTII EDISON COMPANY FOR Odotxr,1993 USTT DOCKET LICENSE 1
050-010 DPR-2 2
050-237 DPR-19 3
050-249 DPR-25 i
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9312100147 931111 PDR ADOCK 050000'O n
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TABLE OF CONTENTS Octoba,1993 NRC REPORT 1.8 Intrna d*-_=
2.9 9-=ry of Operating Erpaience 2.1 Unit 2 Monthly Operating Experience Summary.
2.2 Unit 3 Monthly Operating Experience Summary.
3.8 Operating Data Statistics 3.1 Monthly Operating Data Report - Unit 2 3.2 Monthly Operating Data Report - Unit 3 3.3 Average Daily Powelevd Data - Unit 2 3.4 Average Daily Power level Data - Unit 3 3.5 Unit Shutdown and Power Reduction Data - Unit 2 3.6 Unit Shutdown and Power Reduction Data - Unit 3 3.7 Unit 2 Maximum Daily Electrical Load Data 3.8 Unit 3 Maximum Daily Electrical Load Data 4.8 Unique Reporting Requironmis 4.1 Main Steam Relief andler Safety Valve Operations - Unit 2 and Unit 3 4.2 OIT-Site Dose Calculation Manual Changes 4.3 M4or 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 Safdy-Related M= af=*nce 5.1 Amendments to Facility License or Technical Specifications 5.1.1 Unit 2 5.1.2 Unit 3 5.2 Changes to Procedurts which are Described in the Final Safety Analysis Report (FSAR) (Units 2 andIt 5.3 Significant Tests and Experiments Not Described in the ISAR (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) 1 1
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1.8 Introduction l
Dresden Nuclear Power Station is a three reactor generating facility owned and operated by the Commonweahi Edison Company of Chicago, Illinois. Dresden Station is located at the confluence of the Kankakee and Des Plaines Rivers, in Gnmdy County, near Morris, Illinois.
Dresden Unit 1 is a General El&tric Boiling Water Reactor with a design net electrical output rating of 200 megawatts electrical (MWe). The unit is retired in place with all nuclear fuel removed from the reactor vessel Therefore, no Unit 1 operating data is provided in this report.
Dresden Units 2 and 3 are General Electric Boiling Water Reactors with design net electrical output ratings of 794 MWe each.
Waste heat is rejected to a man-made cooling lake using the Kankakee River for make-up and the Illinois Rive-for blowdown.
The Architect-Engineer for Dresden Units 2 and 3 was Sargent and Lundy of Chicago, Illinois.
j This report for October,1993, was compiled by Kevin W. Sykes of the Dresden Regulatory Assumnce Staff, telephone number (815) 942-2920, extension 2704.
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i 2.9
SUMMARY
OF OPERATING EXPERIENCE IDR Ortahar,1993 2.1 UNIT 2 MONTIILY OPERATING EXPERIENCE
SUMMARY
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10/01/93 to 10/31/93 Unit 2 entered the month criScal and on line. The unit was shutdown at 2145 i
hours on 10/25/93 in order to repair the 2D Traversing Incore Probe (TIP) indexer and replace the 2B TIP det& tor. The unit ranained in shutdown thrsugh the end l
of the month.
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2.0
SUMMARY
OF OPERATING EXPERIENCE FOR Odober,1993 1
2.2 LWIT 3 MONTIILY OPERATING EXPERIENCE
SUMMARY
10/01/93 to 10/31/93 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 Nos ember 1,1993 COMPLETED BY K. W. Sy kes TELEPIIONE (815) 942 2920 OPERATING STATUS
- 1. REPORTING PERIOD: October,1993
- 2. CURRENTLY AUTiiORIZED POWER LEVEL (MWah): 2,527 MAXIMUM DEPENDABLE CAPACITY (MWe NET): 772 DESIGN ELECTRICAL RATING (MWe Net): 794
- 3. POWER LEVEL TO WIIICII RESTRICTED (IF ANY) (MWe Net): N/A
- 4. REASONS FOR RESTRICTIONS (IF ANY): N/A REPORTING PERIOD DATA FARAMETER TIIIS MONTII YEAR 1D DATE CUMULATIVE L
llOURS IN PERIOD 745 7296 204.984 i
G TIME REACTOR CRITICAL (Iloon) 613 4094 152,921 TIME REACTOR RESERVE SI!UTDOWN (lleen) 0 0
0 L
TIME CENERATOR ON-LINE (lloun) 611 4016 146,505 9.
TIME CENERATOR RESERVE SIILTDOWN (Iloun) 0 0
0 10.
TIIERMAL ENERCY GENERATED (MWilt Cross) 1,443,525 8,5.*.4,529 302,593.349 11.
ELELTRICAL ENERGY CENERATED (MWHe Cross) 454.820 2,691,633 96.561,096 12.
ZLECFRICAL ENERCY CENERATED pfWHe Net) 431,503 2,531,408 92.298,708 13.
REACTOR SERVICE FACTOR (%)
82J 5&I 74.6 14.
REACTOR AVAILABILITY FACTOR (%)
82.3 56.1 74.6 IL CENERATOR SERVICE FACTOR (%)
82.0 55.0 71.5 1&
CENERATOR AVAILABILITY FACTOR (%)
82.0 510 71.5 17.
CAPACITY FACTOR (USINC MDC Net)(%)
750 44.9 58.3 IL CAPACITY FACTOR (USINC DER Net)(%)
72.9 43.7 56.7 19.
FORCT.D OUTAGE FACIOR (%)
18_0 4.8 11.9 30.
SIIUTDOWNS SCllEDULED OVER TIIE NEXT 6 MONTIIS (Type, Date and Duration of Each)
NONE.
21.
IF SIIUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP The unit n as shutdown at 2145 hours0.0248 days <br />0.596 hours <br />0.00355 weeks <br />8.161725e-4 months <br /> on 10/25/93 in order to repair the 2D Tras ersing Incore Probe (TIP)inderer and replace the 2B TIP detector. The unit remained in shutdown through the end of the anonth. The projected startup date is 11/29/93 1
l 3.0 OPERATING DATA REPOR"I 3.2 OPERATING DATA REPORT - DRESDEN UNIT TliREE DOCKET No.
050 249 DATE Neveruber 1,1993 COMPLETED BY K. W. Sykes TELEPliONE (815) 942 2920 OPERATING STATUS
- 1. REPORTING PEh* D: October,1993
- 2. CURRENTLY AUTIIORIZED POWER LEVEL (MWah): 2,527 MAXIMUM DEPENDABLE CAPACITY (MWe Net): 773 DESIGN ELECTRICAL RATING (MWe Net): 794
- 3. POWER LEVEL TO Wi!ICII RESTRICTED (IF ANY)(MWe Net): N/A
- 4. REASONS FOR RESTRICTIONS (IF ANY): N/A REPORTING PERIOD DATA 5.
IlOURS IN PERIOD 745 7296 195,313 6.
TIME RIMCTOR CRITICAL Olaun) 745 5653 141,242 TIME REACTOR RF. SERVE SirLTDOWN Olaun) 0 0
0 8.
TIME CENERATO t ON-LINE Ofoun) 745 5587 135,851 0.
TIME CENERATOR RESERVE SIIUTDOWN Oloun) 0 0
0 10.
TilERMAL ENERGY CENERATED @! Wilt Cron) 1,727,953 12.957,082 280,486.537 II.
ELECTRICAL ENERGY CENERA1ED(MWile Grou) 552,284 4.159,073 90,124,022 12.
ELECTRICAL ENERGY CENERATED OfWile Net) 525.848 3.962,409 85,583,621 13.
REACTOR SERVICE FACTOR (%)
100 77.5 72.3 16 REACTOR AVAILABILITY FACTOR (%)
100 77.5 72.3 Q.
CENERATOR SERVICE FACTOR (%)
200 76.7 69.6 IS CENERATOR AVAILABILITY FALTOR (%)
100 76.7 69.6 17.
CAPACITY FACTOR (U5ING MDC Net)(%)
91.3 70.3 56.7 IL CAPACITY FALTOR (USING DER Net)1.%)
88.9 68.4 512 10.
l'ORCED OUTAGE FACTOR ('4) 0 23A 11,8 20.
SHUTDOWNS SCHEDULED OVER THE NEXT 6 MONTHS (Type, Date and Duration of Each)
Refuel Outage 13, D3R13, is scheduled for March 1994. The scheduled duration for this rehe; outage is 13 weeks.
IF SHUTDOWN AT END OF REPORT PERIOD, ESTDChTED DATE OF STARTUP N/A
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3.3 AVERAGE DAILY UNIT 2 POWER LEVEL l
DOCKET No.
050-237 UNIT Dresden 2 DATE Nomnber 1,1993 COhlPLETED BY K. W. Sykes TELEPIIONE (815) 942-2920 hlONTil: Odolxr,1993 e
DAY AVERAGE DAILY NET DAY AVERAGE DAILY NET POWER LEVEL (hnVe)
POWER LEVEL (hnVe) 1 733 18 705 2
489 19 733 3
657 20 741 4
748 21 746 5
699 22 752 6
710 23 751 7
738 24 730 8
743 25 722 9
744 26 212 10 713 27 0
11 736 28 0
12 731 29 0
13 737 30 0
14 747 31 0
15 739 16 695 17 682 s
4 3.4 AVERAGE DAILY UNIT 3 POWER LEVEL DOCKET No.
050-249 UNIT Dresden 3 DATE Nomnber 1,1993 COMPLETED BY K. W. Sykes TELEPIIONE (SIS) 942-2920 MONTII: October,1993 DAY AVERAGE DAILY NET DAY AVERAGE DAILY hTT POWER LEVEL (MWe)
POWER LEVEL (MWe) 1 757 18 673 2
742 19 683 3
759 20 712 4
754 21 724 5
734 22 734 6
735 23 740 7
747 24 714 8
732 25 734 9
374 26 749 10 699 27 743 11 732 28 574 12 742 29 638 13 741 30 742 14 694 31
'13 8 15 690 16 669 17 671 i
3.5 UNTT 2 SHlITDOWNS AND POWER REDUCI1ONS REPORT MONTH OF October,1993 NO.
DATE TYPE (1)
DURATION REASON (2)
METHOD LICENSEE SYSTEM COMPO-CORREC-(HOURS)
OF EVENT CODE (4)
NENT TIVE SHLTTTING REPORT #
CODE (5)
ACTD G DOWN COM-REACTOR MENTS 0) 9 931002 S
0 H
5 NA NA NA SEE NOTE 1 BIIDW 10 931025 F
134 B
1 NA NA NA SEE NOTE 2 BELOW NCrTE 1: Power was reduced to perform Dresden Technical Procedure (iYTP) 8270. Deep / Shallow Control Rod Swap.
NOTE 2: The unit was shutdown to repair te 2D Treversing lacore Probe (TIP) indexer (NWR 121960) and replace the 2B TIP detector (NWR (D21961L TABLE KEY:
(1)
(3)
F: Forced Method:
S: Scheduled
- 1. Manual
- 2. Manual Scrim (2)
- 3. Automade Scram Reason:
- 4. Other (Explain)
A Equipment Failure (Explain)
- 5. Imad Reduction B Maintenance ce Test C Refueling (4)
D Regulsiory Restriction Exhibit G Instmetions for Preparation of Data Entry Sheets for Ucensee Event Repar.: M2.
File (NUREO-016!)
E Operator Training & Ucensing Exam F Administrative (5)
G Operational Error Exhibit I Same Source as above.
H Other (Explain)
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3.6 UNIT 3 SHUTDOWNS AND POWER REDUCTIONS r
REPORT MONVH October,1993 i
NO.
DATE TYPE (1)
DURATION REASON (2)
METHOD LICENSEE SYSTIN COMPO.
mm w.
(HOURS)
OF EVENT CODE (4)
NENT TIVE
'l SHUTTINO REPORTf CODE (5)
Acno!G DOWN COM-REACTOR MENTS.
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(3)
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6 931009 S
0 B
5 NA NA NA SE M i
1 REIDv i
7 931028 3
0 B
5 NA NA NA SE M i
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NOTE 1:
Power reduced to perfonn Main Steam laoladen Vatve timing. Dresden Operadng Surveillance (DOS) 0250 02. Full CZusure hing am!
Exerciairig of Main Steam laolation Valves, requires s'reduedon in power to less than 50% of rated (or na specified by the Operadans SLt Supervisor) prior to tirning the valves.
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NOTE 2:
Power reduced to repair D.C. power supply to turbine supervisory equipment (NWR (D20927). Cause of problem beicg imestigased mer Problem Investigadon Report (PIR) 12-3-93428, land Drop to Repair Main Turbine EHC 24V Power Supply (NTS #20-23936Xh.
i TABLE KEY:
(1)
(3)
F: Forced Method:
3: Scbeduled
- 1. Manual
- 2. Manual Scram (2)
- 3. Automatic Scram Reason:
- 4. Other (Explain)
A Equipmers Failure (Explain)
- 5. land Redusdon B Mairnenance or Test C Refueling (4)
D Regulatory Restricdon Exhibit O Instrucdons for Preparation of Data Entry Sheets for licensee Escra Repons (IEU L j
(NUREG-Ol61)
E Operator Training & IJcensing Exam F Administradvs (5)
Exhibit I Same Source as above.
.t H Other (Explain) i i
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3.7 UNIT 2 MAXIMUM DAILY ELECTRICAL LOAD FOR THE MONTH OF Octoter,1993.
Day Hour Ending UNIT 2 MAXIMUM DAILY ELECTRICAL LOAD (KWe) 1 0900 801,000 i
t 2
0100 788,000 3
2400 795,000 4
1000 796,000 f
5 0900 749,000 6
1400 791,000 7
0700 791,000 8
1100 788,000
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2400 791,000 10 0100 791,000 f
II 0700 794,000 12 1300 793.000 I3 1100 791,000 14 0500 791,000 i
15 1000 791,000 16 1200 750,J00 17 1100 759,000 18 0700 762,000 19 1100 789,000 20 0600 785,000 2I i100 790,000 22 1200 795,000 23 0100 795,000 l
24 0700 782,000 I
25 0800 781,000 l
26 0100 472,000 27 0
i 28 0
29 0
30 0
31 0'
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P 3,8 UNIT 3 MAXIMUM DAILY ELECTRICAL LOAD FOR October,1993 Day llour Ending UNIT 3 MAXIMUM DAILY ELECTRICAL LOAD (KWe) 1 0800 803,000 t
2 120]
802,000 3
0600 803.000 4
1100 800,000 0100 775,000 6
0900 781,000 7
0700 792,000 8
0100 781,000 9
2400 655,000 i
10 2100 783,000 i
i ii 1100 796.000 12 0900 795,000 13 0800 794,000 14 0100 756,000 15 1000 727,000 16 1800 719,000 17 1300 715,000 18 1800 717,000 19 1100 727,000 20 1200 767,000 i
21 1800 7 3,000 22 1203 792.000 23 0800 796,000 24 1900 752,0]O 25 1300 793,000 1
26 0600 792,000 l
27 1600 783,(0 0 28 1100 787,000 29 1300 806,000 30 1100 807,000 31 0100 785,030
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4.8 UNIQUE RF10RTING REQUIREMENTS 4.1 MAIN STEAM RELIEF VALVE OPERATIONS None
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4.2 OFF-SITE DOSE CALCULATION MANUAL (ODCM) CIIANCES None 9
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l 4.3 MAJOR CHANCES TO THE RADIOACTIVE WASTE TREATMENT SYSTEMS DURING Octeher,1993 None.
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4.4 FAILED FUEL ELEMENT INDICATIONS 4.4.1 Unit 1 I
Unit 2 fuel performance during October,1993, continued to show no Indications ofleaking fuel This is based on the sum of the activities of the sis (6) Noble Cases as measured at the Recombiner. Therefore, Unit 2 had excellent fuel performance.
4.4.2 Unit 3 Unit 3 fuel performance during October,1993, continued to show no indications of leaking fuel. This is based on the sum of the methities of the sis (6) Noble Cases as measured at the Recombiner. Therefore, Unit 3 had excellent fuel performance.
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5.8 PLANT OR PROCEDURE CIIANCES, TESTS, EXPERDIENTS, AND SAI'ETY RELATED MAINTENANCE i
5.1 Amendments to Facility License or Technical Specifications during October,1993.
None.
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I 5.2 Additional Changes to Procedures Which an Described in the FSAR (Units 2 and 3) during the months prior :s July 1993. Only thwe pmcedures that required a new or an additional 10 CFR 50.59 review of changes an included. ' Changes implemented during July,1993 and thereafter are expected to be reported along with the FSAR updates in accordance with 10 CFR 50.71(c).
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PROCEDURE TYPE PROCEDURE NUMBER PROCEDURE TITLEl SUM 3fARY OF DESCRIFFION CHANGES j
None I
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5.3 Significant tests and crperiments not described in the FSAR (Units 2 & 3)
No Special prardures invoiring tests rxd decribed in the ISAR, which were apprmed during the months pHiar to July,1993, wtre reported to Regulatory Assurance.
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5.4 Safety Retafevi Mat =f== (Unit 2 and 3) i Safety related maintenance activities for October,1993, are summarized in the attached tables.
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5.5 C-H Safefy Rdatal Modification (Units 2 and 3)
Additional modifications authorized for operation during the months prior to July,1993 which were reported to Regulat 7 Assurance are listed below. As stated in last month's report, changes implemented dudng July,1993 and the..ufter will be reported with the FSAR updates in accordance with 10 CFR 50.71(e).
Modification Description M12 2-92-001B Blowdown Valve Upgrade - This modification replaced the MO212012 actuator gearing set. The modification was performed in response to NRC Generic Letter 89-10 Supplement 3, to ensure that the valve actuator can generate sufficient thrust to dose against a high energy line break.
(Authorized for operation April 25,1993)
Safety Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as l
previously evaluated in the FSAR did not incresce. (Affected FSAR accidents: Loss of Coolant Accident, FSAR Sections 6 and 14; Control Rod System Failure, FSAR Section 6.7)
Loss of Coolant Accident (LOCA):
The probability of the accident did not increase: The gear set replacement provides greater assurance of valve dosure capability J
in - LOCA condition.
l The consequences of the acddent did not increase (offsite dose):
The gear set replacement provides greater assurance of valve dosure capability in a LOCA condition.
i The probability of a malfunction of equipment important to safety.
did not increase: The gear set replacement provides greater assurance of valve dosure capability in a LOCA condition.
i The consequences of a malfunction of equipment important to safety did not increase: The gear set replacanent provides greater assurance of valve dosure capability in a LOCA condition.
Control Rod System Failure:
The probability of the accident did not increase: the Control Rod System win not be afected; the gear set replacement provides
l greater assurance of valve closure capability in a LOCA condition.
The consequences of the accident did not increase (offsite dose):
the FSAR/UFSAR assumptions concerning isolation of the Re= Mar Water Cleanup Systan (RWCU) upon actuation of tie Standby -
Liquid Contml System (SBLC) (via the interlock with the RWCU containment isolation valves) am not affected by the modification.
The probability of a malfunction of equipment important to safety i
did not increase: the gear set replacanent provides greater assurance of valve closure capability in a LOCA condition, thereby improving the reliability of existing equipment.
The consequences of a malfunction of equipment important to safety did not increase: the system functions rtmain the same.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created.
The gear set replacement provides greater assurance of proper i
valve operation in blowdown conditions, and does not adversely impact any systan or function.
3.
The mari;in of safety, as defimed in the basis, for any Technical Specification, was not reduced. None of the parameters used to establish the Technical Specification limits were changed.
hil2-2-92-001C Blowdown Valve Upgrade - This partial modification replaced valve 2-1301-1 and 2-1301-4 actuators, valve yokes and associated cables, and included changes to the thennal overload heaters. The modification was performed in response to NRC Generic Letter 8910 Supploment 3, to ensure that the valve can generate sufficient thrust to close against a high energy line break (Authorized for operation hiay 20,1993).
Safety Evaluation I
1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the ISAR did not increase. (Affected FSAR accidents: Large break LOCA of Isolation Condenser line (HELB), Sections 6 and 14; Loss of External Power, Sections 6, 14; Fire, Section 10.7)
Larce break LOCA:
The probability of the accident did not increase: Increased s:roke time does not affect accident probability.
The consequences of the accident did not increase (offsite dose):
~
The effect of increased stroke time on offsite dose has been evaluated and found to be acceptable.
The probability of a malfunction of equipment important to safety did not increase: The effects of additional steam release on safay equipment has been assessed and found to be acceptable.
The consequences of a malfunction of equipment important to safety did not increase: The effects of additional steam release en safety equipment has been assessed and found to be acceptable.
Loss of External Power The probability of the accident did not increase: Increased s:rt.Le time will not affect the probability of the loss of external power.
The consequences of the accident did not increase (offsite dose):
Increased stroke time will not increase the consequences of a lass of external power.
The probability of a malfunction of equipment important to safey did not increase: Increased stroke time will not increase the probability that safety equipment will fail.
The consequences of a malfunction of equipment important to safety did not increase: Increased stroke time will not have any effect on safety equipment failure cor> sequences.
UEU i
The probability of the accident did not increase: The probability of a fire is not increased since there is no increase in ignition sources, fuel and oxygen level.
i The consequences of the accident did not increase (offsite dose):
The consequences of a fire are not increased since there are no changes to actions required to mitigate an accident, because the amount of cable insulation has been evaluated as not significanz.
)
The probability of a malfunction of equipment important to safety I
did not increase: The change to the actuator, and the changes to the normal and alternate feeds have been evaluated and it has bee =
determined that there is no increast in the probability of equipment malfunction.
' 1 i
a
~
The consequences of a malfunction of equipment important to safety did not increase: The consequences of a malfunction of equipment important to safety did not increase since there wtre m changes to the actions required to mitigate an acdde it.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report luu not been created.
No new failure modes have been created as a result of tie changes to the actuator and alternate power supply; therefore, there are m changes to the operability or reliability of safe shutdown components.
3.
The margin of safety as defined by Technical Specification Table 3.7.1 (Amendment 114) was reduced. Table 3.7.1 listed the maximum operating time of the subject valves as 30 seconds. The new maximum time is 40 seconds. The new time has been evaluated and found acceptable.
Amendment 122 removed Table 3.7.1 from the Technical Specifications; valve operating time is now controlled by the Dresden Administrative Technical Requirements (DATR). ne new 40 second operating time was appruved under DATR 3.18.1, Primary Containment Isolation Valves.
5112-2-92-001D Blowdown Valve Upgrade - This partial modification replaced valve 2-1301-2 motor and associated discharge resistor, gearing set, circuit breaker, and thermal overload heater. The modification was performed is reponse to NRC Generic Letter 8910 Supplement 3, to ensure that the valve can generate sufficient thrust to close against a high energy line batak (Authorized for operation hiay 20,1993).
Safety Evaluation 1.
The probab 'Ay of in t, currence or the consequence of an accident, or nWJurdim of equipment important to safety as previously ev.a.uted la he FSAR did not increase. (Affected FSAR accidente I.e. break LOCA of Lwhtion Condenser line I
(HELB), Sections 6 and 14; Loss on' Exter.u Power, Sections 6, 14; Fire, Section 10.7) l l
Larre Break LOCA of Isolation Condenset; j
The probability of the accident did not increase: increased stroke l
F time has no eHect on accident probability.
The consequences of the accident did not increase (offsite dose):
the elTect of increased stroke time on offsite dose was evaluated and found acceptable.
The probability of a malfunction of equipment important to safety did not increase: the effects of additional steam release on safety equip; ant was assessed and found acceptable.
The consequences of a malfunction of equipment important to safety did not increase: the eHects of additional steam release on safety equipment was assessed and found acceptable.
Ims of External Power:
The probability of abe auident did not increase: increased stroke time has no effect on event probability.
The consequences of the accident did not increase (offsite dose):
increased stroke time does not increase event consequences.
The probability of a malfunction of equipment important to safety did not increase: increased simke time does not increase the probability of safety equipment failure.
The consequences of a malfunction of equipment important to safety did not increase: increased stroke time will not have any effect on the consequences of a safety equipment failure.
EIt1 The probability of the accident did not increase: there was no increase in ignition sources, fuel or oxygen level.
The consequences of the accident did not increase (offsite dose):
there were no changes to actions required to mitigate an accident.
The probability of a malfunction of equipment important to saftsy-did not increase: the changes to the actuator were evaluated and were found not to increase the probability of equipment malfunction.
1 The consequences of a malfunction of equipment important to safety did not increase: there were no changes to the actions required to mitigate an accident.
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The possibility for an accident or malfunction of a different type l
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than any previously evaluated in the Final Safety Analysis Report has not been created.
This conclusion is based on an evaluation of the effects of additional steam release due the increased valve stroke time. Basic functions remain unchanged, no new failure modes are created.
l 3.
The margin of safety as defined by Technical Specification Table 3.7.1 (Amendment 114) was reduced. Table 3.7.1 listed the maximum operating time of the subject valves as 30 seconds. The new maximum time is 40 seconds. The new time has ban evaluated and found acceptable.
Amendment 122 removed Table 3.7.1 from the Tecimical Specifications; valve operating time is now controlled by the Dresden Administrative Technical Requiranents (DATR). The new 40 second operating time was approved under DATR 3.18.1, Primary Containment Isolation Valves.
M12-2-92-001E Blowdown Valve Upgrade - This partial modification replaced valve 2-1301-3 actuator and yoke. The modification was performed in respome to NRC Generic Letter 89-10 Supplement 3, to ensure that the valve will function properly during the blowdown condition (Authorized for operation May 20,1993).
Safety Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the BAR did not increase. (Affected FSAR accidents: Large break LOCA of Isolation Condenser line (IIELB), Sections 6 and 14; Loss of External Power, Sections 6, 14; Fire, Section 10.7)
Larre Break LOCA ofIsolation Condenser:
The probability of the accident did not increase: increased stroke time has no effect on accident probability.
The consequences of the accident did not increase (offsite dose):
the enat of L. creased stroke time on offsite dose was evaluated and found acceptable.
The probability of a malfunction of equipment important to safety did not increase: the effects of additional steam release on safety equipment was assessed and found acceptable.
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1 The consequences of a malfunction of equipment important to safety did not increase: the effects of additional steam release om safety equipment was assessed and found acceptable.
Loss of External Power:
The probability of the accident did not increase: increased stroLe time has no effect on event probability.
The consequences of the accident did not increase (offsite dose):
increased stroke time does not increase event consom The probability of a malfunction of equipment important to safery did not increase: increased stroke time does not increase the probability of safety equipment failure.
The consequences of a malfunction of equipment irnportant to safety did not increase: increased stroke time will not have any efftet on the consequences of a safety equipment failure.
EIti The probability of the acddent did not increase: there was no increase in ignition sources, fuel or oxygen level.
The consequences of the accident did not increase (offsite dose):
there were no changes to actions required to mitigate an accident.
The probability of a malfunction of equipment important to safey did not increase: the changes to the actuator were evaluated and were found not to increase the probability of equipment malfunction.
The consequences of a malfunction of equipment important to i
safety did not increase: there were no changes to the actions
{
required to mitigate an accident.
I 2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created.
This conclusion is based on an evaluation of the effects of additional steam release due the increased valve stroke time. Basic functions remain unchanged, no new failure modes are created.
3.
The margin of safety as defir.ed by Technical Specification Table 3.7.1 (Amendment 114) was reduced. Table 3.7.1 listed the maximum operating time of the subject valves as 30 seconds. The
t new maximum time is 40 seconds. The new time has been evaluated and found acceptable.
M12-2-92-001F Blowdown Valve Uprmde - This partial modification replaced valve 2-2331-l 4 actuator, motor, cables and thennal overload heater. The modificatian was perfonned in response to NRC Generic Letter 89-10 Supplement 3, ta ensure that the valve will function properly during the blowdown con:!Iti:n i
(Authorized for operation hiay 20,1993).
Safety Evaluation t
e 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the ISAR did not increase. (AfTected FSAR accidents: Small break LOCA, Sections 6 and 14; Maia Steam Line break, Sections 6 and 14; Loss of External Powe.
Section 6)
Small Break LOCA:
The probability of the accident did not increase: no pressure related components were affected by the modification, and the increased stroke time does not affect accident probability.
l The consequences of the accident did not increase (offsite dose):
the modification does not affect the IIPCI System response to the small break LOCA.
The probability of a malfunction of equipment important to safe:y did not increase: the modification does not affect the IIPCI System response to the small break LOCA.
The consequences of a malfunction of equipment important to safety did not increase: the modification does not affect the HPCI System response to the small break LOCA.
Main Steam Line Break:
The probability of the accident did not increase: no pressure related components were affected by the modification, and t!w increased stroke time does not affect accident probability, i
The consequences of the accident did not increase (offsite dose):
i the off-site dose would be greater than previously ana.lyzed duri::3 a IIPCI line break but would be well within the 10 CFR 100 ll=Its.
The probability of a malfunction of equipment important to safay m.
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did not increase: the effects of additional steam could impact safety equipment, but era!uation has found this to be acceptable.
The consequences of a malfunction of equipment important to safety did not increase: the effects of additional release could increase the consequences of malfunction of Safety equipment, but evaluation has found this to be acceptable.
Loss of External Power:
The pmbability of the accident did not increase: the modification has improved valve reliability.
The consequences of the accident did not increase (ofTsite dose):
the modification has improved in!re reliability.
The probability of a malfunction of equipment important to safety did not increase: Increased stroke time will not increase the probability that safety equipment will fail.
The consequences of a malfunction of equipment important to safety did not increase: Incrrased struke time will not have any efTect on safety equipment failure consequences.
Etti The prubability of the accident did not increase: there is no significant increase in ignition sources, fuel and oxygen.
The consequences of the accident did not increase (offsite dose):
the actions required to mitigate an accident have not changed.
The probability of a malfunction of equipment important to safety did not increase: era!uation of the changes did not identify an increase in the probability of equipment malfunction.
The consequences of a malfunction of equipment important to safety did not increase: the actions regired to mitigate an accident have not changed.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report has not been created.
No new failure modes have been created as a result of the modification.
3.
The margin of safety as defined by Technical Specification Table
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6 3.7.1 (Amendment 114) was reduced. Table 3.7.1 listed the maximum operating time of the subject valves as 25 seconds. The new maximum time is 50 seconds. The new time has been craluated and found acceptable.
Amendment 122 runored Table 3.7.1 from the Technical Specifications; valte operating time is now controlled by the Dresden Administrative Technical Requirements (DATR). The new 50 second operating time was approved under DATR 3.13.1 Primary Containment Isolation Valves.
M12-2-92-00lG Elowdown Valve Upgrade - this modification replaced the valte hf02-2331-5 actuator gearing set. The modification was performed in response to NRC Generic Letter 89-10 Supplanent 3, to ensure that the valve art rstv-can generate sufficient thrust to close against a high energy IIne break.
(Authorized for operation May 20,1993)
Safety Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the ISAR did not increase. (Affected ISAR accidents: Small break LOCA, Sections 6 and 14; Main Steam Line break, Sections 6 and 14; Loss of Extemal Poww, Section 6; Fire, Section 10)
Small Break LOCA:
The probability of the accident did not increase: no pressure related components were affected by the modification, and the increased stroke time does not affect accident probability.
The consequences of the accident did not increase (offsite dose):
the modification does not affect the IIPCI Systan risponse to tie small break LOCA.
The probability of a malfunction of equiptnent important to safer.=
did not incruse: the modification does not affect the HPCI Systan response to the small break LOCA.
The consequences of a malfunction of equipment important to safety did not increase: the modification does not affect the HPCI Systan response to the small break LOCA.
Main Steam Line Break:
P
,a v.
1 The probability of the accident did not increase: no pressure related components were affected by the modificatior and the 1
s increased stroke time does not affect accident probability.
The consequences of the accident did not increase (offsite dose):
the off-site dose would be greater than previously analyzed during a HPCI line brrak but would be well within the 10 CFR 100 tim
- The probability of a malfunction of equipment important to safety did not increase: the effects of additional steam could impact safety equipment, but evaluation has found this to be acceptable.
The consequences of a malfunction of equipment important to safety did not increase: the effects of additional release could increase the consequences of malfunction of safety equipment, but evaluation his found this to be acceptable.
Loss of External Power:
The probability of the accident did not increase: the modification has improved valve reliability.
The consequences of the accident did not increase (offsite dose):
the modification has improved valve reiIability.
The probability of a malfunction of equipment important to safety did not increase: Increased stroke time will not increase the probability that safety equipment will fail.
The consequences of a malfunction of equipment important to safety did not increase: Increased stroke time will not have any effect on safety equipment failure cortquences.
EIG The pmhability of the accident did not increase: there is no significant increase in ignition sources, fuel and oxygen.
The consequences of the accident did not increase (offsite dose):
the actions required to mitigate an accident have not charged.
The probability of a malfunction of equipment important to safety did not increase: evaluation of the changes did not identify an increase in the probability of equipment malfunction.
The consequences of a malfunction of equipment important to safety did not increase: the actions required to mitigate an accident have not changed.
d ',
e 2.
The posibility for an acddent or malfunction of a different 1,pe than any previously evaluated in the Hnal Safety Analysis Report has not been created.
No new failure modes have been created as a result of the modification.
3 3.
The mare n of safety as defined by Tecimical Specification Table i
3.7.1 (Amendment 114) was reduced. Table 3.7.1 listed the maximum operating time of the subject valves as 25 seconds. The.
I new maximum time is 50 seconds. The new time has been evaluated and found acceptable.
Amendment I?? removed Table 3.7.1 from the Tecimical Specifications; ' valve operating time is now controlled by the Dresden Administrative Technical Reg'drements (DNIR). The '.
new 50 second operating time was approved under DNIR 3.18.1, Primary Containment Isolation Valves.
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- l 5.6 Tanporary Systan Alterations Imta!!ed (Unit 2 and Unit 3)
A "Tunporary Systan Alteration" rrfers to electrical jumpers, lifted leads, renoved fuses, fuses turned to nons:enducting position, fuses moved from normal to reserve holder, tanporary power supplies, test switches in alternate positions, tanporary blank flanges, and spool pieces. Alterations controlled and documented as part d a routine out-of-service or other procedure, alterations which are a nonnal feature of systan design, and bases installed as part of a venting or draining process are not included.
No additional tanporary alterations approved for use during the months prior to July,1993 were reported to Regulatory Assurance. As stated in last month's report,10 CFR 50.59 evaluations perfonned during July,1993 and thereater will be reported with the FSAR updates in accordance with 10 CFR 50.71(e).
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6 Y
f 5.7 Othe Units 2 and 3 Required 10 CFR 50.59 Evaluations Other requirrd 10 CFR 50.59 evaluations include Set Point Changes (SPC), Rigging Evaluations and changes to equipraent not reported in Sections 5.2 through Swtion 5.6.
No additional evaluations perfonned during the months prior to July,1993 were reported to Regulatory Assurance. As stated in last month's report,10 CFR 50.59 evaluations performed during July,1993 and thereafter will be reported with the FSAR updates in accordance with 10 CFR 50.71(e).
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