ML20045H164
| ML20045H164 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 06/30/1993 |
| From: | Garrett P, Spedl G, Sykes K COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GFSLTR-93-0015, GFSLTR-93-15, NUDOCS 9307190147 | |
| Download: ML20045H164 (41) | |
Text
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@ Commonwealth Edison Dresden Nuclear Power Station 6500 North Dresden Road Morns, Illinois 60450 Telephone 815/942-2920 July 12,1993 GFSLTR #93-0015 Director, Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, DC 20555 Attention: Document Control Desk Gentlemen:
Subject:
hionthly 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 hionthly Onerating Summary Renort for June,1993. This information is supplied to your office in accordance with the instructions set forth in Regulatory Guide 1.16.
Si cerely, j
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Gary F s pe Station hianager Enclosure cc:
NRC Region III Office Illinois Dept. of Nuclear Safety, State of Illinois U.S. NRC, Document hianagement Branch Nuclear Licensing Administrator Site Vice Pres.
General hianager - Nuclear Services OPEX Engineer (2)
NRC Senior R(sident inspector Site Quality Verification - Dnsden Site Engineering and Construction 51anager Nuclear Oversight hianager/ R. Janecek Comptroller's Onice INPO Records Center UDI,Inc. Wash.,D.C.
File /NRC Op. Data File / Numerical r
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PDR ADOCK 05000010 L R
PDR t,
'4 MONTilLY NRC
SUMMARY
OF OPERATING EXPERIENCE, CIIANGES, TESTS, AND EXPERIMENTS PER REGULATORY GUIDE 1.16 AND 10 CHC 50.59 1
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DRESDEN NUCLEAR POWER STATION COMMONWEALTli EDISON COMPANY FOR June,1993
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UNIT DOCKET LICENSE 1
050-010 DPR-2 2
050-237 DPR-19 3
050-249 DPR-25 l
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TABLE OF CONTENTS June,1993 NRC REPORT 1.0 Introduction 2.0 Sununary of Operating Experience 2.1 Unit 2 Afonthly Operating Experience Summary, 2.2 Unit 3 Afonthly Operating Experience Summary.
. 3.0 Opwating Data Statistics 3.1 hionthly Operating Data Report - Unit 2 3.2 Afonthly Operating Data Report - Unit 3 3.3 Average Daily Power Level 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 Station hiaximum Daily Load Data 4.0 Unique Reporting Requirements 4.1 Main Steam Relief and/or Safety Valve Operations - Unit 2 and Unit 3 4.2 Off-Site Dose Calculation Afanual Changes 4.3 Afajor 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 Procedum Changes, Tests, Experiments, and Safety-Related Alaintenance 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 hiaintenance (Units 2 and 3) 5.5 Completed Safety-Related Afodifications 5.6 Temporary System Alterations Installed 5.7 Other Required 10 CFR 50.59 Evaluations (Units 2 and 3)
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I 1.0 Introduction Dresden Nuclear Power Station is a three reactor generating facility owned and operated by the Commonwealth Edis(m Company of Chicago, I!!inois. Dresden Station is located at the confluence of the Kankakee and Des Plaines Rivers, in Grundy County, near Aforris, Illinois.
Dresden Unit 1 is a General Electric Boiling Wate. Reactor with a design net cicetrical output rating of 200 megawatts electrical (h!We). The unit is retired in place with all nuclear fuel removed from the reactor sessel.
Therefore, no Unit 1 operating data is provided in this report.
Dresden Units 2 and 3 are General Electric Boiling Water Reactors with dtsign 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 River for blowdown.
The Architect-Engineer for Dresden Units 2 and 3 was Sargent and Lundy of Chicago, Illinois.
This report for June,1993, was compiled by Paul K. Gairett and Kevin W. Sykes of the Dresden Regulatory Assurance Staff, telephone number (815) 942-2920, extension 2713 or 2704.
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2.0
SUMMARY
OF OPERATING EXPERIENCE FOR June,1993 2.1 UNIT 2 MONTIILY OPERATING EXPERIENCE
SUMMARY
06/01/93 to 06/30/93 Unit 2 entered the month critical and on line and continued through the end of the month, l
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l 2.0
SUMMARY
OF OPERATING EXPERIENCE IX)R June,1993 2.2 UNIT 3 MONTilLY OPERATING EXPERIENCE
SUMMARY
06/01/93 to 06/30/93 Unit 3 entered the month critical and on line and continued through the end of the month.
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3.0 OPERATING 'OATA REPORT 3.1 OPERATING DATA REPORT - DRESDEN UNIT TWO DO"KET No.
050-237 DATE July 1,1993 COMPLETED BY P. K. Garrett and K. W. S kes 3
TELEPlIONE (815) 942-2920 OPERATING STATUS
- 1. REPORTING PERIOD: June,1993
- 2. CURRENTLY AUTIIORIZED POWER LEVEL (MWth): 2,527 MAXIMUM DEPENDABLE CAPACITY (MWe NET): 772 DESIGN ELECTRICAL RATING (MWe Net): 794
- 3. POWER LEVEL TO WIIICil RESTRICTED (IF ANY) (MWe Net): N/A
- 4. REASONS FOR RESTRICTIONS (IF ANY): N/A i
REPORTING PERIOD DATA i
PARAMETER tills MONTII YEAR TO DATE CUMUIATIVE 5.
IlOURS IN PERIOD 720 4343 202.775 6.
TIME REACTOR CRITICAL (Iloun) 720 1272.7 150.844.4 7.
TIME REACTOR RESERVE SilUTDOWN (lloun) 0 0
0 N.
TIME GENERATOR ON-LINE (Iloun) 720 1200J 144.433.0 9.
TIME GENERATOR RESERVE SilUTlX)WN (lloun) 0 0
0 10 TilERMAL ENERGY GENERATED (Mullt Gnm) 1.395.744 2,163.342 297.851.099 11.
ELECTRICAL ENERGY GENERATED (Mulle Gnm) 435,732 678,267 95.062.907 13.
EllCTRICAL ENERGY GENERATED (MWile Nd) 414.179 634.302 90,872,164 13.
REACTOR SERVICE FACTOR (%)
100 29.3 74.4 14.
REACTOR AVAILAlf,LITY FACTOR (%)
100 29.3 74.4 15.
GENERATOR SER ilCE FACTOR (%)
100 27.6 71.2 16.
GENERATOR AVAILABILITY FACTOR (4) 100 13.5 71.2 17.
CAPACITY FACTOR (USING MDC Nd) (4) 74.4 18.9 58.0 18.
CAPACITY FACTOR (USING DER Nd)(%)
72.4 18.4 56.4 19.
FORCED OUTAGE FACTOR (%)
0 72.4 28.8 20.
SilUTDOWNS SCIIEDULED OVER Tile NEXT 6 MONTilS (Type, Date and Duration of Each)
NONE.
21.
IF SIlUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP N/A (p:\\pintmgr\\gfs93\\0015.93)
3.0 OPERATING DATA REPORT 3.2 OPERATING DATA REPORT - DRESDEN UNIT TIIREE DOCKET No.
050-249 DATE July 1,1993 COMPLETED BY P. K. Garrett and K. W. Sykes TELEPIIONE (815) 942 2920 OPERATING STATUS
- 1. REPORTING PERIOD: June,1993
- 2. CURRENTLY AUTIIORIZED POWER LEVEL (MWth): 2,527 MAXIMUM DEPENDABLE CAPACITY (MWe Net): 773 DESIGN ELECTRICAL RATING (MWe Net): 794
- 3. POWER LEVEL TO WillCII RESTRICTED (IF ANY) (MWe Net): N/A
- 4. REASONS FOR RESTRICTIONS (IF ANY): N/A REPORTING PERIOD DATA 5.
IlOURS IN PERIOD 720 4,343 192J60 6.
TIME REACTOR CRITICAL Oloun) 720 2.828.8 138,418.1 7.
TIME REACTOR RFSERVE SIIUTDOWN Oloun) 0 0
0 S.
TIME GENERATOR ON LINE Oloon) 720 2,771.8 133.045.2 9.
TIME GENERATOR RESERVE SilUTDOWN Olaun) 0 0
0 10.
TilERMAL ENERGY GENERATED (MWilt Gross) 1,690.324 6,377,261 273,906,716 11.
ELFCTRICAL ENERGY GENERATED (MWIIe Grou) 541,lR5 2,039,863 88,017,070 13.
ELFfTRICAL ENERGY GENERATED (MWile NH) 517,260 1,965,247 83.575.427 13.
REACTOR SERVICE FACTOR (%)
100 65.1 72.0 14.
REACTOR AVAILAlllLITY FACTOR (%)
100 65.1 72.0 5.
GENERATOR SERVICE FACTOR (4) 100 63.8 69.2 16.
GENERATOR AVAILA1111,ITY FACTOR (%)
100 45.4 69.2 17.
CAPACITY FACTOR (USING MIK' Nd) (%)
97.2 60.8 59.2 18.
CAPACITY FACTOR (USING DElt Nd)(%)
94.7 59.2 57.6 19 IT)RCED OUTAGE FACTOR (%)
0 36.2 30.8 l
20.
SHUTDOWNS SCHEDULED OVER THE NEIT 6 MONTHS (Type, Date and Duration of Each)
NONE.
31.
IF SHUTDOWN AT END OF REPORT PERIOD, ESTIMATED DATE OF STARTUP N/A (p:\\platmgr\\grs93\\0015.93)
3.3 AVERAGE DAILY UNIT POWER LEVEL DOCKET No.
050-237 UNIT Dresden 2 DATE July 1,1993 COh!PLETED BY P. K. Garrett and K. W. Sykes TELEPlIONE (815) 942-2920 h!ONTII: June,1993 DAY AVERAGE DAILY NET DAY AVERAGE DAILY NET POWER LEVEL (51We)
POWER LEVEL (5tWe) 1 479 18 353 2
480 19 530 3
484 20 414 4
481 21 374 5
509 22 602 6
501 23 678 7
616 24 663 8
578 25 663 9
627 26 637 10 633 27 619 11 628 28 664 12 597 29 635 13 595 30 680 14 640 15 642 16 617 17 620 l
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d 3.4 AVERAGE DAILY UNIT POWER LEVEL DOCKET No.
050-249 UNIT Dresden 3 DATE July 1,1993 COMPLETED BY P. K. Garrett and K. W. Sykes TELEPIIONE (815) 942-2920 MONTII: June,1993 DAY AVERAGE DAILY NET DAY AVERAGE DAILY NET POWER LEVEL (MWe)
POWER LEVEL (MWe) 1 762 18 765 2
766 19 756 3
758 20 705 4
543 21 740 5
731 22 765 6
710 23 694 7
763 24 761 8
760 25 738 9
750 26 733 10 746 27 740 11 731 28 748 12 632 29 738 13 533 30 731 14 552 15 710 16 708 17 769 (p:\\pintmgr\\gfs93\\0015.93)
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Docket No.
050-237 UNIT NA3IE Dresden Unit 2 DATE July 1,1993 Completed Ily Paul Garrett &lbin Sykes Telephone (815) 942-2920 3.5 UNIT SIIUTDOWNS AND POWER REDUCTIONS REIN)RT SIONTil June,1993 No.
IMT15 TYPt 411 It' RAT > tN REA.WN (:)
METHOD OF trFNWI-DTNT SYSTIM col 1(J43 (T)htfUNINT CODt.(5)
CAUSF & CtikR!fTi\\ E NTION TO FRLTINT RECL'RRMdCE (WUR9
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"'U I Rt'P R.=haxd lead so Wirmmt draumg d oil f wn :A and 2B Recarc. pumps dw w mwn 4,a hish bet (1)
C Refuelity hierhod:
Entry Sheets for Licensee Event Reports (LER)
F: FORCED D Regulatory Restriction
- 1. h1anual File (NUREG-0161)
S: SCIIEDULED E Operator Training & Licensee Exam
- 2. hfanual Scram F Admir6strative
- 3. Automatic Scram (5)
(2)
G Operational Error
- 4. Other (Explain)
Exhibit i Same Source as abme.
Reasen:
11 Other (Explain)
- 5. l.oad Redaction A Equipment Failure (Explain)
(4)
B hiaintenance or Test (3)
Exhibit G Instructions fbr Preparation of Data i
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Docket No.
050-249 UNIT NAME Dresden Unit 3 DATE July 1,1993 Cornpleted By Paul Garrett &lbin Sykes Telephone (815) 942-2920 3.6 UNIT SilUTDOWNS AND POWER REDUCTIONS REPORT MONTil June,1993 i
vt Dart TYPun DURArioN REAMW (3 MLTHOD OF LkTNNIE IYIET
$1ST'M CODE (44 GMPONINF CODilS CAUSE & CORRt4%T ACTION lu l'9ExTNT RldritRFNCE (HOURS SHIITnNG DOWN RtPORT#
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(2)
(3)
(4)
F: IURCED Reason:
hiethod:
Exhibit G Instructions fbr Preparation of Data S: SCllEDULED '
Entry Sheets fi.r Licensee Event Reports (LER)
A. Equipment Failure (Explain)
- 1. hianual File (NUREG4)l61)
B hiaisuenance or Test
- 2. Atanual Scram C Refueling
- 3. Autornatic Scram (5)
D Regulatory Restriction
' 4. Other (Explain)
E Operator Trainity & Licensee Exam
- 5. Load Reduction Exhibit i Same Sotme as abme.
F Administrative G Operational Error H Other (Explain) l j
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3.7 COMMONWEALTil EDISON COMPANY DRESDEN NUCLEAR POWER STATION MAXIMUM DAILY ELECTRICAL LOAD FOR TIIE MONTil OF June,1993 Dresden 2 Day llour Endine KWe 1
1100 521,800 2
0800 523,500 3
0600 521,800 4
2200 535,200 5
1000 551,300 J
6 0200 537,400 l
7 1500 713,500 8
1600 684,400 9
0100 677,200 10 1400 690,800 11 0100 683,700 12 2100 689,900 13 1400 699,300 14 0100 688,000 15 0700 696,300 16 1200 762,500 17 1100 734,500 18 0100 578,300 19 1900 647,300 20 0100 639,500 i
21 1900 467,800 22 2200 762,900 23 0100 752,400 24 0400 705,600 25 1400 707,500 26 1100 716,400 27 1400 721,700 28 0100 711,900 29 1700 733,000 30 0100 728,400 l
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3.7 COMAIONWEALTIl EDISON COMPANY DRESDEN NUCLEAR POWER STATION MAXIMUM DAILY ELECTRICAL LOAD FOR TIIE MONTII OF June,1993 Dresden 3 1)hty.
IIour Endine KWe 1
1200 803,900 2
1600 802,300 3
2100 802,200 4
2400 701,500 5
1500 797,100 6
1400 800,300 7
1800 801,700 8
0100 801,000 9
0400 787,400 10 0400 786,200 11 0200 780,800 12 1000 739,700 13 1200 625,100 14 0600 584,500 15 1100 806,900 16 1500 806,500 17 1100 812,100 18 0500 811,700 19 1100 809,700 20 0100 808,100 3
21 1000 804,300 22 0700 805,800 23 2100 800,400 24 0700 801,500 25 1100 804,100 26 1400 796,400 27 0700 784,900 28 1300 801,100 29 1500 804,000 30 0100 778,600 (p:\\pintmgr\\gfs93\\0015.93) t
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4.0 UNIQUE REPORTING REQUIREMENTS 4.1 MAIN STEAM RELIEF VALVE Ol'ERATIONS None i
l 4.2 OFF-SITE DOSE CALCULATION MANUAL (ODCM) CIIANGES I
None
-1 (p:\\pintmgr\\gf393\\0015.93)
4.3 MAJOR CIIANGES TO TIIE RADIDALTIVE WASTE TREATMENT SYSTEMS DURING June,1993 Current Status of Radioactive Waste Treatment System Upgrade Project:
No significant change to report for the month of June,1993.
Floor Drain Collection Pump replacement reported as being 95% complete in last month's report is still in progress. The replacement is being performed under modification number M12-2/3-87-002N (NWR #78778).
4.4 FAILED FUEL ELEMENT INDICATIONS 4.4.1 Unit 2 Unit 2 fuel perfonnance during June,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 2 had excellent fuel perfonnance.
4.4.2 Unit 3 Unit 3 fuel perfonnance during June,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 perfonnance.
(p:\\pintmgr\\gfs93\\0015.93)
5.0 ItANT OR I'ROCEDURE CilANGES, TESIS, EXPERIMENTS, AND SAFETY RELATED MALNTFRANCE 5.1 Amendments to Facility License or Technical Specifications during June,1993.
None.
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5.2 Additional Changes to Procedures Widch are Described in the FSAR (Units 2 and 3) during June 1993. Only those procedures that required a new or an additional 10 CFR 50.59 review of changes are included.
None t
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5.4 Safety Related Maintenance (Unit 2 and 3)
Safety related maintenance activities for June, 1993, are summarized in the attached tables.
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5.5 Completed Safety Related Modification (Units 2 and 3)
Modifications reported to Regulatory Assurance in June 1993, which were authorized for operation during April, May and June 1993 are listed. For ease of reference, the changes have been identified by their design change control modification number.
Previously, only modifications that had been completely closed out were reported.
Modification No.
Description M12-3-89-011E Unit 2/3 Diesel Fire Pump Controller This modification provides electrical power supervision for the Unit 2/3 diesel fire pump controller to meet the requirements of NFPA Code 72D-1975.
(authorized for operation June 6, 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 Final Safety Analysis Report has not increased. By adding the electrical power supervision, this modification has upgraded the Unit 2/3 diesel fire pump controller as described above.
The current margin of the plant will not be affected due to the installation of the supervisica Tne modification will have no bearing on the probability or consequence of an acci6 tat or malfunction of equipment important to safety as previopoly evaluated in the FSAR/UFSAR, since analyt s a take no crudit for supervision cf the Unit 3 f 3 diesel fire pump controller.
2.
The pos ibility for an accident or malfunction of a different type than auf previously evaluated in the Final Safety Analysis Report has not been created. This modification is classified as non-safety related and regulatory-related. The single failure events and design basis accidents as analyzed in the FSAR/UFSAR have been reviewed and are not affected by this modification. Thus, the pes,1bility for an accident or malfunction of a different type than any previously evaluated in the Units 2&3 FSAR/UPSAR will not be created.
3.
The margin of safety, as defined in the basis for any Technical Specification, has not been reduced. The Technical Specifications are unaffected by this partial modification. The Dresden Unit 2&3 Technical Specifications do not contain any reference to the fire protection system.
Section 3/4.12, which originally (p:\\pintmgr\\gfs93\\0015.93)
contained the system, was deleted from the Unit 2&3 Technical Specifications via Amendments 106 and 101, respectively.
M12-2-87-42 LPCI Loop Select Logic The purpose of this modification is to provide torus cooling capability through unselected LPCI Loop via manual override through the use of the 316/117 switches. (authorized for operation April 15, a!*3)
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 FSAR has not increased. Torus cooling capabilities are improved.
2/3 core coverage via the selected loop's injection logic remain unchanged. After the operator verifies that directing LPCI injection flow from the' core will not compromise core cooling, torus. cooling capabilities can be improved.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. The torus cooling safeguards currently in place to respond to a Design Basis Accident event maintain their integrity. This change does not interface with other safety systems.
3.
The margin of safety, as defined in the basis for any Technical Specification, has not been reduced. The margin of safety is not reduced by allowing Torus cooling by the unselected loop.
The mod allows additional operator flexibility.
After the operator verifies that core cooling is not compromised the operator may divert flow from the core during LPCI initiation to allow for Torus cooling.
M12-2-88-22C Upgrade Panel 902-54, -55, -56, -65 and 923-4,
-5A, -7 Annunciator System to Meet DCRDR.
This partial modification added visual and audible ringback to the annunciator systems of the control room service panels 902-65, -54,
-55,
-56 and 923-4,
-5A, -
7.
(authorized for operation May 18, 1993)
Safety Analysis 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 increased. The annunciator system is not discussed in the accident analysis section of the FSAR.
This (p:\\pinungr\\gfs93\\0015.93)
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system is not required for accident mitigation.
j Since the annunciator system is electrically isolated from the safety related systems, the failure of the non-safety related annunciator system will not affect the operation of any of the plant's safety-related systems.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created. No change has been made which affects any of the bounding conditions of the FSAR accident i
anaylsis. All bounding conditions remain the same, no new accidents are introduced by this modification.
3.
The margin of safety, as defined ir4 the basis for any Technical Specification, has not been reduced. When applicable, LCO 3.12.F and the Surveillance Requirement 4.12.F for the Fire Protection System's fire barriers will be adhered to for the installation of the cables.
No other LCOs, Surveillance Requirements or their bases will be affected by the installation, operation or failure of the modified annunciator system.
M12-2-90-029A Hardened Wetwell Vent Hodification The Augmented Primary Containment Vent System is intended to relieve Primary Containment pressure to avoid overpressurization and possible breach of the Primary Containment which may occur due to events which are beyond the design basis of the plant.
(authorized i
for operation May 21, 1993)
Safety Evaluation 1.
2he probability of an occurrence or the con equence of an accident, or malfunction of equ: rnent important to safety as previously eval. +ed in the PSAR has not increased. The addition and the operation of the Augmented Primary Containment Vent System is outside the design basis of the plant and does not increase the probability of an accident previously evaluated in the PSAR.
The Augmented Primary containment Vent System (APCVS) is designed such that its addition and operation will not affect the function or operation of any existing equipment. The Pressure suppression System and Primary Containment Isolation Systems will function as required prior to the modification.
The existing Containment Isolation Valve (s) circuitry is modified, but existing valve function, redundancy and separation criteria are maintained with the addition of the new circuitry. Double isolation is included to separate class lE power from non-class IE components. Essential bus integrity is (p:\\pinbugr\\grs93\\0015.93)
maintained through proper current fault trip coordination. Additionally, the mode switch must be turned to the APCVS mode before the Group II Primary Containment Isolation Signal override is possible.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created.
The APCVS is designed and shall be operated such that the impact on existing systems or functions does not create the possibility of an accident or malfuction of a type different from those evaluated in the UFSAR/FSAR.
3.
The margin of safety, as defined in the basis for any Technical Specification, has not been reduced. No Technical Specification requirement, action item, surveillances, or bases is associated with this modification. The operation of the APCVS deals with events which are outside the design basis of the plant.
M12-2-88-22A Control Room Annunciatur System Upgrade, Cable and Equipment Installation.
This partial modification installed the conduit, conduit supports, cables and piggyback terminal blocks with isolation diodes, additional horn logic relays and AN-159 power supplies. It also installed the SER peripherals and the dual electronic horns at the panel groups 902-3 &
-4,
-5 & -6,
-7 & -8, and 923-1 &-5.
(authorized for operation May 18, 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 FSAR has not lucreased.
The annunciator system is not discussed in the accident analysis section of the FSAR.
This system is not required for accident mitigation.
The failure of the non-safety related annunciator system will not affect the operation of any of the plant's safety-related systems.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created.
No change has been made which affects any of the bounding conditions of the PSAR accident analysis. All bounding conditions remain the same, no new accidents are introduced by this modification.
3.
The margin of nafety, as defined in the basis for any Technical Specification, has not been reduced. When applicable, the LCO 3.7.C for (p:\\pintmgr\\gfs93\\0015.93)
the Secondary Containment System's penetration seals and LCO 3.12.F and the Surveillance l
Requirements 4.12.F for the Fire Protection System's fire barriers will be adhered to for the installation of cables. No other LCOs,-
Surveillance Requirements for their basis will be affected by the installation, operation or failure of the modified annunciator system.
M12-2-90-028 HPCI MO2-2301-10 Replacement.
The purpose of this modification is to eliminate erosion problems resulting from high differential pressure across the valve when throttling for system
)
testing. The replacement valve is designed specifically for this application.
(authorized for operation May 26, 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 FSAR has not increased. The longer stroke time of the new valve in itself cannot create an accident. Although the longer stroke time can potentially delay full flow of HPCI injection when the system is in test mode, the probability of such an occurrence is very low.
The probability of a malfunction of' equipment in the HPCI system remains the same.
The valve which was replaced had increased erosion as a result of cavitation; the replacement valve provides a design to reduce cavitation. The consequences of an equipment malfunction remain the same.
The system functions remain the same and, thus, the failure modes of the HPCI system remain the same.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created.
The modification replaced the existing valve with a more reliable valve specifically designed for the application. The change does not adversely impact any system or function.
Also, the longer stroke time, in itself, does not affect any plant operation. This is significant only whan the system is in test mode.
The worst case, complete failure of the HPCI system, has been evaluated and is still the bounding failure for the system. this longer stroke time of the HPCI Test valve does not result in the HPCI system failing completely. The effects of the longer stroke time during this period on LOCA and transient events have been found to be negligible.
Therefore, the longer stroke time of the HPCI (p:\\pintmgr\\gfs93\\0015.93)
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i valve does not result in having an accident or malfunction of a different type.
3.
The margin of safety, as defined in the basis for any Technical specification, has not been reduced.
The modification does not have an.
impact on any Technical specification.
M12-2-91-003 Control. Rod Drive (CRD) Return Line Removal This modification removed the existing CRD return line between the reactor vessel and the drywell penetration to eliminate the IGScc susceptible welds.that were a part of the piping.
(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 FSAR has not increased. The subject accident considers LOCA due to small.
and large pipe line breaks.
Since this modification removes a section of small bore piping, the probability of the accident will not increase. The probability of a. malfunction of any safety-related equipment will not increase. The new caps added to the system
-were designed and qualified to the same-requirements as the existing system. The consequences of a malfunction of equipment important to safety will not increase. The modification did not affect the functional design of the remaining system. The consequences of a failure of the remaining portion of the system is the same as the existing system.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the PSAR has not been created.
The primary function of the existing system is to provide a high pressure water return flow path of the reactor vessel. Presently, high pressure supply is available via leakage past the control rod drive seals through the charging water supply line.
Since the original intent of the system is still performed.with the CRb return line removed, there will'be"no..
accidents or malfuncticus of a type different from those evaluated in the SAR.'
3.
The margin of safety, as defined in the basis for any Technical Specification, has not been reduced. The penetration was capped after removal of the line, thereby o'_'rinating the need for isolation. Drywell integrity is improved with one less penetration.
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M12-2/3-87-002N Floor Drain Collector Pump Replacement This modification replaced the single Floor Drain collector Pump and associated piping with two 100%
capacity pumps.
(authorized for operation June 15, 1993)
Safety Evaluation 1.
The probability of an occurrence or the conseqvence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR has not increased. The subject pumps are not part of and do not interact with any equipment or system important to safety.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created.
Failure or malfunction or the subject pumps cannot create any type of event that can interact with equipment important to safety to create an accident or possibility for malfunction of a different type than already evaluated in the FSAR.
3.
The margin of safety, as defined in the basis for any Technical Specification, has not baen reduced. The subject pumps are not addressed in the basis for any Technical Specification and, therefore, has not reduced Technical Specification safety margins.
M12-2(3)-82-034 SPDS Computer Points This modification installed computer inputs as required for the Safety Parameter Display System, including display screens in the control room. (authorized for operation June 23, 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 FSAR has not increased. The SPDS is not evaluated in the FSAR and.is not needed for the operation of the plant safety systems.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR has not been created.
The SPDS is part of the plant process computer peripherals. Where interface with safety systems is required, suitable isolation will be provided.
3.
The margin of safety, as defined in the basis (p:\\pintmgr\\gfs93\\0015.93)
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for any Technical-Specification, has not been reduced. The installation of the SPDS is not'a basis for any Technical Specification safety margin. The addition of this modification will indirectly increase the safety margin in that-'
it will aid the plant. operating personnel to assess the safety status of the plant.
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s 5.6 Temporary System Alterations Installed (Unit 2 and Unit 3)
A " Temporary System Alteration" refers to electrical jumpers, lifted leads, removed fuses, fuses turned to non-conducting position, fuses moved from normal to reserve 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 system design, and hoses installed as part of a venting or draining process are not included.
THE POLLOWING INPORMATION REPLECTS TEMPORARY ALTERATIONS WHICH WERE INSTAILED DURING'@ifFIMONTH OP June f993_
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Log Number Description II-12-93 Disable open damper 2-5772-47.
Reason: High exhaust air temperatures; need the outside air intake damper full open to avoid equipment damage due to high heat.
Cooler outside air cools down the system. (Installed June 11, 1993)
Safety Evaluation - None required.
II-29-93 This temporary alteration installed a cable from penetration I-205B to the Klethly Electrometer.
This temporary alteration utilizes the ECP probes that were installed under temporary alteration II-40-91.
Temporary alteration II-40-91 was cancelled and this new temporary alteration (II-29-93) is controlling the entire ECP monitoring system. Reason To determine ECP values for the Recirculation System.
(Installed June 24, 1993)
Safety Evaluation 1.
The probability of an occurrence or the consequence of an accident, or malfunctio.as of equipment important tc cafety as previously evaluated in the Final Safety Analysis Report did not increase aa a result of this temporary alteratior A Lost-Parts Analysis was conducted for the previour installation of three similar probes in tbo D2 "A"
Recirculation loop.
That analysis remains valid for this "B"
loop instalation because a single failure is assumed and there is not a common mode of failure.
The probes are installed in a recessed area inside the recirculation piping, approximately 12" away from the recirculation flow stream Therefore, the flow induced forces will be limited and breakago of the probes from their supports is not expected.
However, in the (p:\\pintmgrigfi93\\0015.93)
2 highly unlikely event that the ECP probes become dislodged, the broken pieces will fall into the flow stream and be discharged through the jet pumps into the vessel lower plenum region.
Since the largest possible broken piece is still rather small, the result of the analysis indicates that the consequences of postulated blockage will be acceptable.
Furthe rmore, since the probes assemblies are made mostly of stainless steel, there should be no concern for corrosion or chemical reaction with other reactor internal materials.
The most likely break is postulated to occur at the Swagelok connection. A failure of this type is highly unlikely because the Swagelok fittings have been analyzed at pressure and temperature conditions exceeding the design of the recirculation system. A break of all three connections outside the blind flange would result in a break area of less than 0.0368 square inches. This area is within the spectrum of small breaks assumed in the Dresden UFSAR, therefore its consequences are bounded by the current analysis. A small break of this size has an estimated leak rate of approximately 15 gpm of saturated water. During normal plant operation, this loss can be replaced by the normal coolant makeup sources, and is within the range of the leak detection capability inside the drywell. Therefore, the additional effect of such a small break to the Design Basis Loss of Coolant Accident is inconsequential with respect to the calculated accidental consequences in the Dresden UFSAR.
A less likely break to occur can be postulated at the weld connection to the flange. The resulting break area is 0.636 square inches, with a corresponding flow rate of approximately 252 gpm.
This break is also within the spectrum of small breaks assumed in the Dresden UFSAR.
However, this postulated leakage rate exceeds the limit of Technical Specifications 3.6.D and will require plant shutdown.
If required, makeup capability and reactor water level control (p:\\pintmgr\\gfs93\\0015.93)
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can be provided by the feedwater and HPCI systems.
Since the blind flange is located downstream from valve 102-4B, the affected loop can be eventually isointed from the reactor vessel to alnimise the loss of reactor coolant.
Because the modification to the blind flange will meet the applicable codes and standards, there is no reason to assume that a break will occur. Therefore, the-probability of occurrence of any accident considered in the UFSAR is not increased.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR was not created as a' result of this temporary alteration. The proposed addition of three ECP electrodes does not introduce any chances of malfunctioning of any previously analuzad or unanalyzed safety related systems or components because the consequences of any potential blockage from a postulated broken electrode is acceptable. Furthermore, if the broken pieces create jet streams and drywell missiles, the small jet forces and projectiles can only impinge an interior concrete wall and would not cause damage.
3.
The margin of safety, as defined in the basis for any Technical Specification, has not been reduced. There are no Technical Spacification requirements associated with tLe electrochemical measurements in the recirculation system, nor are there any specific leakage requirements for the recirculation system.
II-34-93 This temporary alteration installed pressure monitoring devices at the Low Pressure Switch PSL J-1501-57B sensing line and at the 3-1501-87B drain valve for M03-1501-20B and -38B.
Reasons To investigate pressure / flow spiking during cycling of the HO3-1501-21B and -38B valves.
(Installed June 14, 1993)
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Safety Evaluation 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR did not increase as a result of the temporary.
alteration. There were no FSAR affected accidents.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR was not created as a result of the temporary alteration. The function of the low system pressure alarm in the LPCI Loop II was not altered during the temporary alteration because it used the existing calibration taps.
Also, the function of the MO3-1501-20B and -38B valves remained.
operable during troubleshooting.
3.
The margin of safety, as defined in the basis, for any Technical Specification,was not reduced. There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were af fected.
II-35-93 This temporary alteration installed a nitrogen freeze seal in line 2-4328-3/4"-L at the inlet and outlet of the 2-1301-31 valve. Reason: To repair valve 2-1301-31.
Safety Evaluation 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR did not increase as a result of the temporary alteration.
2.
The possibility for an accident or malfunction of a different type than any previously evaluated in the FSAR was not (p:\\pintmgr\\gfs93\\0015.93)
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s created as a result of the temporary alteration. As long as the water level of the loop seal remained constant for the duration of the freeze seal, the probability of an accident involving secondary containment was not increased.
3.
The margin of safety, as defined in the basis, for any Technical specification,was not reduced.
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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.
Item Description Exempt Change (NWR #D12403)
This Exempt Change upgrades the mechanical seal cooling system for 3B Reactor Feed Pump (RFP).
The seal cooling heat exchanger and tubing are being lowered to below the centerline of the pump.
The Turbine Building closed cooling Water System (TBCCW) Jdping to the heat exchangers is being upsized to one inch.
In addition, the condensate Booster Pump seal injection lines to the RFP will be cut at the supply header. The piping from the header to the pump including two solenoid operated valves will be removed.
Safety Evaluation 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR will not increase as a result of the Exempt change. 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 will not be created as a result of the Exempt Change. This Exempt Change merely removes an unused portion of the condensate booster System and upgrades the existing seal cooling system for the 3B RFP.
In addition, the TBCCW supply to the mechanical seals is upsized to maximize cooling to the seals. The removal of the unused 125 VDC loads (solenoid operated seal injection valves) has no adverse impact on the (p:\\pintmgr\\gfs93\\0015.93)
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system which could introduce new types of failures.
3.
The margin of safety, as defined in the basis, for any Technical Specification,was not reduced.
There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were affected.
Exempt Change (NWR #D12404)
This Exempt change upgrades the mechanical seal cooling system for 3C Reactor Feed Pump (RFP).
The seal cooling heat exchanger and tubing are being upsized to 3/4".
The TBCCW system piping to the heat exchangers is being upsized to 1" and a vent is being provided on the TBCCW side of both seal heat exchangers.
In addition, the condensate Booster Pump seal injection lines to the RFP will be cut at the supply header. The piping from the header to the pump including two solenoid operated valves will be removed.
Safety Evaluation 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR will not increase as a result of the Exempt Change. There are no FSAR affected accidents.
2.
The possibility for an kccident or malfunction of a different type than any previously evaluated in the FSAR will not be created as a result of the Exempt Change. The Exempt Change merely removes an unused portion of the condensate Booster System and upgrades the existing seal cooling system for the 3C RFP.
In addition, the TBCCW supply to the mechanical seals is upsized to maximize cooling to the seals. The removal of the unused 125 VDC loads (solenoid operated seal injection valves) has no adverse impact on the system which could introduce new (p:\\pinungr\\gfs93\\0015.93)
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types of failure.
3.
The margin of safety, as defined in the basis, for any Technical Specification,was not reduced.
There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were affected.
Mixed Waste Storage A safety evaluation was performed to address the impact of storing radioactive waste and/or material in the station's new Mixed Waste Building (MWB).
This is being done to comply with Federal and State Regulations which require on-site storage of mixed waste.
Safety Evaluation 1.
The probability of an occurrence, the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the FSAR will not increase. 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 will not be created as a result of the new storage arrangement. The use of the MWB does not interact with or impact any SSC or SSC function. Although radiological release from outside the power block is not specifically addressed in the FSAR, all associated environmental release limits are; these release limits are based on 10 CFR requirements.
3.
The margin of safety, as defined in the basis, for any Technical Specification,was not reduced.
There are no Technical Specifications where the requirement, associated action items, associated surveillances, or basis were affected.
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