ML20045D604
| ML20045D604 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 06/22/1993 |
| From: | Colburn T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9306290206 | |
| Download: ML20045D604 (35) | |
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c i E UNITED STATES 7s7'/!
NUCLEAR REGULATORY COMMISSION
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WASHINGTON. D C 20555-0001 June 22, 1993 LICENSEE:
Detroit Edison Company FACILITY:
Fermi-2 Nuclear Plant
SUBJECT:
MEETING
SUMMARY
FOR JUNE 14, 1993 A meeting was held with Detroit Edison Company (DECO) representatives and-NRC management and members of the technical staff on June 14, 1993, to discuss proposed changes to the Fermi - 2 Technical Specifications (morning session) and update management on the status of turbine control valve problems at Fermi (afternoon session).
The proposed Technical Specification changes and related request for exemption to the requirements of 10 CFR Part 50, Appendix J, were l
submitted for staff review by application dated May 24, 1993. The proposed changes would modify the local leak rate (LLRT) testing requirements for the low pressure coolant injection isolation valves of the residual heat removal (RHR) system.
During the afternoon session, the licensee updated the management on the status of turbine control valve problems experienced.since modifications were completed to the valves during the last refueling outage to support power uprate. The licensee is currently operating at 93 percent power in order to minimize turbine control valve oscillations which were experienced at higher power levels upon restart from the: refueling outage. The licensee plans to continue to operate at the current power level until the next refueling outage-at which time it will fix the turbine ~ control valve problems. The licensee'.s vendor is currently conducting modeling and analysis to determine the exact i
fix for these valves.
l It was previously viewed by the staff that the turbine control valve problems i
and resultant vibration and system pressure pulse were related to two extraction steam line failures that occurred at Fermi in December 1992 and April 1993. The licensee stated that it has now ruled out the link between l
these two steam line breaks and the' turbine control valve problems and no longer believes them to be related. The licensee believes that the extracticn i
steam line breaks were the inevitable result of poor. initial system design and j
The licensee also described plans for completion of the power uprate testing program once the turbine control valves are fixed and it is able to proceed to 100 percent power.
j 9306290206 930622 PDR ADOCK 05000341 P
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i Enclosed is a copy of the licensee's slides used in the presentation and a list of attendees.
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- k. WAw Timothy G. Colburn, Sr. Project Manager Project Directorate III-I Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
- 1. List of Attendees
- 2. Presentation Slides cc w/ enclosures:
See next page l
1
. 4 Enclosed is a copy of the licensee's slides used in the presentation and a list of attendees.
Original Signed By:
Timothy G. Colburn, Sr. Project Manager 1
Project Directorate III-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosures:
- 1. List of Attendees
- 2. Presentation Slides cc w/ enclosures:
See next page l
0FFICE LA:PDIII-l n PM:PDIII-1 (A)PD:PDIII-1 AD:HIlf53Is NAME CJamersonChh'TColburn 70 JHall)3fehl JZwolinski DATE 06/1*/93 b/21/93 6 /2k93 6/**/93
/kES)40 8E')/NO
[fhS/N0 YES/ff0 )
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4 Fermi-2 Detroit Edison Company l
cc:
John Flynn, Esquire Senior Attorney Detroit Edison Company i
2000 Second Avenue Detroit, Michigan 48226 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N. Logan Street P. O. Box 30195 Lansing, Michigan 48909 Mr. Wayne Kropp U.S. Nuclear Regulitory Commission Resident Inspector Office 6450 W. Dixie Highway Newport, Michigan u8166 Monroe County Office of Civil Preparedness 963 South Raisinville Monroe, Michigan 48101 Regional Administrator, Region 111 U.S. Nuclear Regulatory Commission 799 Roosevelt Road Glen Ellyn, Illinois 60137 Mr. William E. Miller Director - Nuclear Licensing i
Detroit Edison Company Fermi-2 1
6400 North Dixie Highway Newport, Michigan 48166 Mr. Douglas R. Gipson Senior Vice President Nuclear Generation Detroit Edison Company l
6400 North Dixie Highway Newport, Michigan 48166 l
l
ENCLOSURE I MEETING ATTENDANCE SHEET
~
JUNE I4, 1993 MEETING WITH DETROIT EDIS0N COMPANY AND THE NRC STAFF RELATED TO PROPOSED TECHNICAL SPECIFICATION CHANGES (MORNING SESSION) AND UPDATE ON THE TURBINE CONTROL VALVE PROBLEMS AT FERMI-2.
NAME ORGANIZATION / TITLE MORNING SESSION (10:30-11:30?
James R. Hall NRC/NRR/ Acting Director, PD III-I Timothy G. Colburn NRC/NRR/ Project Manager, PD III-I James Pulsipher NRC/NRR/ Containment Systems and Severe Accident Branch-Bruce Sheffel Detroit Edison Company l
Len fron Detroit Edison Company Paul Fessler Detroit Edison Company William E. Miller Detroit Edison Company-Gopal K. Sharma Detroit Edison Company AFTERM00N SESSION (I:00-3:00)
John A. Zwolinski NRC/NRR/ Assistant Director for Region III Reactors James R. Hall
_NRC/NRR/ Acting Director PD III-I Timothy G. Colburn NRC/NRR/ Project Manager PD III-I-Robert Stransky NRC/NRR/ Power Uprate Lead Project Manager David Robare General Electric Co. Licensing Kathy Berry General Electric Power Uprate Project Paul Fessler DECO / Technical Manager l
William E. Miller Deco / Director - Nuclear Licensing Glen Ohlemacher Deco / Licensing i
l Tom Dong Detroit Edison Company ten Fron Detroit ~ Edison Company L
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ENCLOSURE 2 l
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Log No.93-013 Revision No. 2 ISI-NDEllST-PROGRA.\\1 EVALL'ATION Date: June 9.1993 IST Prog. l N l ISI-NDE Prog. l l
Prepared by: Rands BresmaierFA Other l
l Leakrate Prog. l X l Approsed b : Bruce J. ShelTel GENERAL SUPERVISOR ISI/ PEP 3
U
Subject:
10CFR50 Annendis "J". T3 0e "C" LLRT Testine of Valves E1100-F050A & B. E1100-F610A & B and Ell 50-F015A & B Page1 of5
================================================-.
- = = = = = = = = - - - - = = =
CURRENT TESTING REOUmWN:
Ell 50-F015A -
Ts pe "C" air test. Pressure Isolation Yah e i eat Tec (PlV). Full Stroke Test Open and Closed. PIT Ell 50-F015B -
Tspe "C" air test. Pressure Isolation Valve Leak Test (PlV). Full Stroke Test Open and Closed. plt (E1100-F050A -
Tspe "C"2ir test. Pressure Isolation Yah e 1 eak Test (PlV). Full Stroke Test q
Open and Closed. PIT (El100-F0508 -
Type "C" air test. Pressure Isolation Yah e Leak Test (PlV). Full Stroke Test Open and Closed. PIT E1100-F610 \\ -
hre "C" air test. Pressure Isolation Valve I.eak Test (PlV). Full Stroke Test Closed PIT F. I 100-F610B -
Type "C" air test. Pressure Isolation Valve Leak Test (PlV). Full Stroke Test Closed PIT PROPOSED ALTERMATIVE TESTING & ASSOCIATED JUSTlFICATION:
1.
Deleting the requirement for a 10CFR50, Appendis J air LLRT for valves Ell 50-F015A
& Ell 50-F015B.
For a 10CFR50. Appendix J. Type "C" tested valve to be tested with a water medium, instead of an air medium, the installed isolation vah e seal water system fluid inventory must be sufficient to assure the sealing fbnction for at least 30 days at a pressure of 1.10 Pa l
Containment isolation Yahes Ell 50-F015A & B are assured a 30 day water seal based on the j
following O
a)
LOOP Seal Configuration & Water Seal-The inboard section of piping runs horizontally for approximately 20 feet and then rises venically for an additional 20 feet thereby creating a " loop seal" configuration
Page 2 of 4 b)
Penetration & System Piping Normally Filled With Water-During normal operation, this section of piping is filled with water from the El 150-F015 A & B vr.lves back to the Reactor Pressure Vessel Piping outboard of valves El150-F015A & B is maintained filled with water through a keep fill system cl 30 Day Water Seal Maintained Inboard of E1150-F015A & B-The Design Basis Accident Scenario which is the bounding accident scenario and results in the highest containment pressure rise and has the potential for the largest amount of radioactive release, is a guillotine rupture of the recirc suction line in Division I with a concurrent loss 6 c. v p r "a / /"
s of off-site power assumed Under these conditions, calculations show that the water seal
~
will be maintained against the inkaard side of El150-F015A & B for 30 days ( Reference EFA 02-015. Revision 1 )
The water seal outboard of the El150-F015A & B vahes will remain for at least 30 day s based on the followinu During the first 30 da,s s, at least one RHR pump will be in operation and will pressurize both disisions of RHR based on the divisional cross tie vah e (El 150-F610) being open This will pressurize the downstream sides of El150-F017A &
B and El150-F611 A & B to pump discharge pressure ( 115 psig) which is greater than Containment Design Accident Pressure of Pa Based on the above. any leakage past the El150-F015A & B would go into the piping section between the E l 150 F015 A & B and El 100-F017A & B and El100-F611 A & B and no further.
because the downstream side of El100-F017A & B and El100-F611 A & B would be at a higher pressure 11 is likely that some leakage would pass through the E I 150-F015-\\ & B into the containment (Reference EFA 92-015. Res 1) d)
Closed System Outside of Containment at Operating Pressure > Pa-The RHR piping is a closed system outside of containment designed and built to ASME Section 111, Class 2 requirements and constitutes an isolation barrier The system remrins filled with water and operates at a pressure greater than Pa post LOCA The system is included in the ASME Section XI Inservice Inspection Program and receives the required NDE examinations for Class 2 piping The Inservice Inspection Program requires the sy stem to be inspected at pressure with any visible leakage requiring repah in addition.
the RHR System is included in the NUREG-0737 Leakage Reduction Program and is periodically inspected at normal operating pressure for external leakage with an acceptance criteria of 40 ml/ min. for each division.
l l
Page 3 cf 5 2.
Proposed Deletion of 10CFR50, Appendix "J" Leakage Testing for Valves E1100-F050A
& B and E1100-F610A & B.
For containment isolation integrity purposes, valves El100 F050A & B and valves El100-F610A
& B are not required for the following reasons:
a)
GDC Criteria - 10CFR50, Appendix "A", Criterion 55 - Reactor coolant pressure boundary penetrating containment, states that "each line that is part of the reactor coolant pressure boundary and that penetrates primary reactor containment shall be prosided with containment isolation valves - unless it can be demonstrated that the containment isolation provisisons for a specific class oflines - are acceptable on some other defined basis" b)
Meeting GDC Criteria -The penetrations containing valves El 100-F050A & B and El100-F610A & B are X-13A and X-13B Without these valves being used for containment isolatior> :redit, these penetrations satisfy the GDC 55 based on an j
alternative basis as allowed in GDC 55 That basis is as follows bl)
First Containment Isolation Boundary - Lines that must remain in senice following an accident and lines which should remain in senice during normal operation for safety reasons are provided with at least one isolation valve This l
requirement is satisfied by contamment isolation valves Ell 50-F015A & B.
l b2)
Second Containment Isolation Boundary - A second isolation boundary is l
formed in accordance with NUREG-0800, Standard Review Plan, Section 6.2.4 l
" Containment Isolation System" Page 6.2.4-4 ofNUREG 0800 states that containment isolation provisions that differ from explicit requirements of GDC 55 and 56 are acceptable if the basis for the difference is justified It further describes the acceptable alternate containment isolation provisions Paragraph "e" on page 6.2 4-4 of NUREG 0800 is applicable to penetrations X-13A and X-13B The following statementsjustify deleting the inboard set ofisolation valves.
l b2a-The containment isolation valves on RHR are different from other containment isolation valves as they do not receive a containment isolation signal This is because these valves may be required for low pressure coolant injection.
b2b - The single containment isolation valve can accommodate any single active failure and will still provide the containment isolation feature by preventing t
the release of fission products that may result from a postulated accident.
I This is because:
- 1. The RHR piping inside containment will have a water seal for a period of 30 days following a DBA.
~
Page 4 cf 5
- 2. The RHR pumps develop pressure in the RHR piping so that any leakage will be into the primary containment-b2c - RHR is a closed system outside containment that is protected from any postulated missiles and pipe whip, designed to seismic Category I standards, classified Quality Group B or better and is designed to meet or exceed the maximum temperature and pressure of the containment.
b2d - The piping from the inboard check valves El100-F050A and B to valves El150-F015A and B conforms to AShE Section III, Class 1 requirements No pipe breaks or cracks are postulated as this piping fulfills the design requirement stipulated in NRC Branch Technical Position MEB 3-1.
b2e - The system is included in the AShE Section XI Insenice Inspection
)
Program and receives the required NDE examinations for Class 2 piping.
which includes the requirement for the system to be inspected at pressure, with any visible leakage requiring repair. In addition, the RHR system is included in the NLREG 0737 Leakage Reduction Program and is periodically inspected at normal operating pressure for external leakage with an acceptance criteria of < 40 ml/ min Conclusion - The RHR System meets or exceeds all of the above requirements and therefore qualifies as a second isolation boundary. It should be noted that other plants have applied this logic and had it accepted by the NRC.
3.
Proposed Substitution of a Pressure Isolation Valve (PIV) Test for Ell 50-F015A & B In j
Lieu of the Required 10CFR50, Appendix "J" Water Test.
a)
Through Seat Leakage - As stated above in Section I., it is not possible to have through seat leakage past E1150-F015A & B due to system design and outboard operating pressure being greater than Pa b)
External Leakage -With through seat leakage being eliminated from consideration, the only remaining leakage path of consideration is external leakage through the shaft packing and bonnet gasket areas of El150-F015A & B.
c)
Test Pressure - The required Appendix "J" water test pressure is 1.1 Pa, or specifically for Fermi 2,62.2 psig The test pressure for the PIV test is 1045 psig Conclusions - For purposes of determining external leakage for El150-F015A & B, the PIV test is a more consenative test since any external leakage would increase with pressure and the PlV test pressure is much greater than the Appendix "J" test pressure (1045 psig vs 62.2 psig) Since the only possible leakage path is externalleakage, bas:d on system design, the PIV test at a pressure of 1045 will satisfy all Appendix "J" testing objectives.
Page 5 cf 5 4.
Proposed Raising of the Allowable Leakage of the PIV Test from I gpm to 10 gpm for Valves E1100-F050A & B and E1100-F610A & B. Propose Lowering the Allowable Leakage of the PIV Test from I gpm to 0.4 gpm for Valves El150-F015A & B.
a)
Purpose of test -The PlV test is performed to ensure the low pressure pipe is protected from the reactor pressure and prevent intersystem LOCA.
b)
Overpressure Protection - Each division of RHR is protected from overpressurization by a safety relief valve with a capacity of approximately 290 gpm at a pressure of 450 psig Conclusions - With the relief capacity of 290 gpm for each division of RHR, engineering design calculation supports changing the existing PlV Tech. Spec. test criteria from 1 gpm to 10 gpm for valves El100-F050A & B and E100-F610A & B with no adverse consequences. Calculations show that there will be adequate protection of the low pressure RHR piping. This is also consistent with other plants Lowering of the PlV leakage test criteria from I gpm to 0.4 gpm in conjunction with the establishment of a 5ml/ min externalleakage is necessary to assure an adequate water seal on the inboard side of Ell 50-F015A & B for the 10CFR50, Appendix J required 30 day period to allow for classification as a water sealed penetration.
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STARTUP TEST PROGRAM eTests originally planned for 100% power were evaluated and performed at 98% power eHigh Pressure Core Injection and Reactor Core Injection Cooling Operated well within the design limits at uprated power eFeedwater and Pressure Control Systems-Well behaved and highly stable
- Radiation Levels and Water Chemistry Levels were as expected for uprated conditions i
- Startup test report Complete and in management review l
Startup Test Program Following Modifications To Achieve 100% Power 1
- Feedwater and Pressure Control system testing
- Collect and trend data for projection to higher power Conditions l
- Collect and analyze data to confirm the effectiveness of the modifications
1 Update - April Outage Restart Testing Current Plant Power Level April 1993 Outage e
Restart Program Focus of Restart Testing Test Data Points and Hold Points Analysis & Acceptance Criteria o
Comparison to March 1992 Pre-Power Uprate; Power Uprate and Restart Data Data Review - GEC, GE, Hopper Associates s
o Operations at 93% Power o
Turbine Control Valve Oscillation o
Vibration and Concerns About Attachments between TCV and HP o
Damage of Small Components o
Two Phenomenon o
TCV oscillations l
o 52~ Manifold Pressure Pulse l
1 2
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Program:
o GEC Valve Testing i
o Duplicate Valve Performance Curve 1
o Provide Large Diameter Sleeve o
Open Throat of Valve from 13.7" to 15" I
o General Electric o
Independent Overview - Valve Testing l
o Support in Data Review 1
o 52" Manifold Pressure Pulse i
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Target Schedule -
June 1993 Complete Testing Sept 1993 Hardware On Site j
r-l July 1993 Large Sleeve -
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Date:
April 21, 1993 Ib To:
L. Fron SupervisoF f fff) 3 Turbine From:
B. Myers Reactor Engineer
Subject:
Restart Plan for Outage 93-03 Startup Abstract 4
o Selected plant data will be recorded at various points in the power ascension. As data is being obtained between the hold 1
points, it should be reviewed. Any abnormal readings should be brought to the attention of the Shift Test Director. The Shift Test Director will then inform others as appropriate.
o Ascension to full power (985) will proceed as per a normal startup, with the exception that 2-3 hour " Hold Points" will occur at 50%, 70%, 865, and 965 to allow for offline plotting of GETARS data, evaluation of plant data against specified limits or expected noras, and management concurrenoe.
A.
conference meeting will be held at 50%, 70%, 865, 965 and 985 i
power. The Shift Test Director will chair the meeting. Leads responsible for evaluating data should report their results at this time. The data forms should be signed by the responsible individual and any comments noted. A conference call will then be made between the Shift Test Director, L. Fron, P.
Fessler, the NSS and R. McKeon to discuss further power increases.
o Key Engineering experts will maintain around the clock shift coverage from the time of turbine roll until 985 power is achieved in order to coordinate data collection and provide a real time assessment of the data as it is collected to ensure that the plant is performing as expected. Special emphasta will be placed on evaluating the changes in the data related to the pressure pulse phenomenon, turbine control valve movements and positions, and turbine / HP Unitized Actuator vibration as we ascend in power.
Any changes to this restart plan need to be approved by o
B. Myers.
N
l.
R: start Plan fcr Outage 93-03 Startup Page 3 1
o In order to better understand the pressure pulse phenomenon l
occurring in the system, 6 pressure transducers have been installed and connected to. the above recording equipment. The j
pressure transducers will sonitor pressure in the following locations:
4 M/S extraction steam pressure l
4 north heater shell pressure j
East MSR hot reheat to center LP West MSR hot reheat to center'.,*
3 north extraction steam pressure j
52 inch manifold 1
j o
All available GETARS points will be recorded to magnetic tape 1
-and will be available for offline plotting and analysis at-i each hold point. Particular points of interest that will be 2
recorded are:
Reactor Pressure APRM First Stage Pressure
. Core Flow
~
-52" Manifold Pressures
)
Control Valve Positions l
GETARS data should also be recorded above 10% power but prior to Main Turbine roll in order to observe 52" Manifold. pressure oscillations while operating on the Bypass Valves.
[
o Beginning at 20% power, the Shift Technical Advisors will obtain a print-out of special-logs for BOP performance data.
The data points contained in this log bound all of the essential data collected during the Power Uprate Test Program in the Steady State Data Collection test procedures of SOE 92-01.
o Three accelerometers have been installed on the unitized actuator for the #3 Turbine control Valve in order to monitor unitized actuator vibration. The output from these accelerometers will be fed to'the same recording equipment being used to collect the strain gage, accelerometer, and pressure transducer data. In addition, hand held vibration data will collected on all Four Turbine Control Valve' unitized actuators at each hold point.
o Main Turbine vibration will be sonitored from the vibration shack throughout the power ascension.
q i
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i Power Operations at 93% Power Level o
Power Operation Above 93% Results in Greater Dynamic Loadings i
o Turbine Control Valve Oscillation 1
o Pressure Pulsation in 52" Steam Manifold Area of Concern a
i o
increased Vibration Associated with Unitized Actuator o
Significant increase in Turbine Control Valve Oscillation
(> - ) %-3 o
Pressure Fluctuation in Steam Manifold intensified o
increased Megawatt Swings 5
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o Two chest pressure taps (braided hoses) o Two bottom loop pipe pressure tap root valves (i.e.,
packing, loose spindle) o Rangers on cold reheat line o
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Two failures of the #4 Extraction Steam Line o
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DISTRIBUTION FOR MEETING
SUMMARY
OF JUNE 14, 1993 DISTRIBUTION w/encls. I & 2 Docket File NRC & Local PDRs PD 3-1 Reading File WShafer, RIII TColburn DISTRIBUTION w/ encl. 1 T. Murley/F. Miraglia (12-G-18)
J. Partlow (12-G-18)
J. Roe J. Zwolinski W. Dean C. Jamerson OGC J. Hall J. Pulsipher R. Stransky E. Jordan (MNBB-3701)
G. Grant (17-G-21) i l
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