IR 05000341/1988005
| ML20149M790 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 02/23/1988 |
| From: | Danielson D, Gavula A, Liu W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20149M776 | List: |
| References | |
| 50-341-88-05, 50-341-88-5, NUDOCS 8802290082 | |
| Download: ML20149M790 (8) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-341/88005 (DRS)
Docket No. 50-341 License No. NPF-43 Licensee: -Detroit Edison Company 2000 Second Avenue Detroit, MI 48224 Facility Name:
Fermi 2 Inspection At:
Monroe, Michigan Inspection Conducted:
February 2-5, 1988 Inspectors-
. C. Liu 23 //
Date 2. A M
. A. Gavula Date~
W 2 Ah Approved By:
D. H. Danielson, Chief Materials and Processes Section Date Inspection Summary Inspection on February 2-5, 1988 (Report No. 50-341/88005(DRS))
Areas Inspe'cted:
Special announced safety inspectics of licensee action on previous inspection findings (92702), onsite followup of events at operating reactors (93702), onsite followup of written reports (92700) and non-license training (41400).
Results:
One apparent violatior, was identified (inadequate design control -
Paragraph 2.b).
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DETAILS e
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Person Contacted-Detroit Edison Company (Deco)
W. Orser, Vice President, Nuclear Production
- R. S. Lenart, General Director, Nuclear. Engineering
- S. G.Catola, Chairman, NSRG
- A. A1chalabi, Supervisor
- A. I. Hassoun, Lead Engineer
- J.
W.' Contoni,-Lead Engineer
- A. K. Lim, Lead Engineer
- L. D. Burr, Plant. Engineer
- P. Anthony, Compliance
- Denotes those attending the exit ir.terview on February 5, 1988.
2.
Licensee Action on Previous Inspection Findings a.
(Closed) Violation (341/87033-01):
Expansion anchor spacing violations were not evaluated in accordance with the specified
methodology.. Deco's letter of response dated September 17, 1987,
was reviewed and determined to be acceptable.
The NRC inspector
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reviewed the documentation associated with the prescribed corrective
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actions and concluded that the followup actions were adequate.
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i Design calculation DC-638 was revised to reflect the appropriate methodology given in Specification 3071-266.
The revised
calculation confirmed that the alternate methodology used for the
specific installation resulted in more conservative interactions.
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addition, six recent calculations were reviewed and found to reflect
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the appropriate anchor spacing violation methodology.
i The requirement to utilize the specified methodology was reiterated in a letter issued August 14, 1987, from C. R. Gelletly to Stone and
Webster, Sargent and Lundy, and Deco Engineering.
The training and documentation records were reviewed and indicated that each engineer in the Architect / Civil, Electrical, Instrumentation and Controls, Mechanical and Fluid Systems, and Engineering Technology groups had completed the mandatory reading of this letter.
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Based on the documentation reviewed, this item is considered closed.
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b.
(Closed) Violation (341/87033-02):
Design bases were incorrectly
translated into the following design documents
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(1) Calculation DC No. 974
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(2) Drawing SC 721-2002
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(3) Specification No. 3071-226
(4) Calculation DC No. 4497 l
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l DECO's letter of response dated September 17, 1987, was reviewed and determined to be acceptable by Region III.
The NRC inspectors held discussions with licensee's representatives and examined the corrective actions as stated in the letter of response.
It was concluded that DECO had determined the full extent of the subject violation, performed the necessary survey and followup action to correct the present conditions, and developed the necessary corrective actions to preclude recurrence of similar circumstances.
The corrective actions duentified in the letter of response have been implemented.
During the course of the above inspection, the analysis for the long term block wall evaluation was also reviewed.
Design Calculation No. 841, Volume I, Revision F, "Seismic Analysis of Block Walls in Reactor /Auxilary Building", was reviewed to determine whether it complied with NRC requirements and licensee commitments.
On page 15 for typical column base plate connections, the calculation calls for 3/ s inch fillet weld on a 8/ inch base plate.
The American a
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Institute of Steel Construction (AISC) Specification requires a minimum fillet weld size of 1/4 inch for this thickness of plate, and of 5/18 inch for a material over 3/ inch in thickness.
In addition,
no calculation was performed to demonstrate the adequacy of the specified weld.
Also, on the following drawings thirteen locations were designated as requiring a fillet weld, but no fillet weld size was specified.
- 6C721-2608 Revision H Design-Reinforcement for Existing Block Walls
6C721-2609 Revision K Design-Reinforcement for Existing Block Walls
6C721-2610 Revision H Design-Reinforcement for Existing Block Walls The above instances are examples of a violation of 10 CFR 50, Appendix B, Criterion III, in that design bases were incorrectly translated into the above design calculations and design drawings.
(341/88005-01A&B)
Subsequent actions by the licensee included performance of field walkdowns to verify fillet weld sizes and an analysis of the weld configuration to determine the size adequacy.
Based on conversations with licensee representatives, the fillet weld sizes
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observed in the field were 1/4 inch for both the unspecified weld 3/ c inch had been specified.
locations and for the locations where t
c.
(0 pen) Unresolved Item (341/87025-01):
RHR pump motor termination l
box mounting failures.
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The NRC inspector reviewed Calculation No, ME-946 "Seismic Stress Analysis of 2000 HP Vertical Motor".
Page 21 of this calculation qualified the larger conduit box mounting for the specific motor.
The analysis showed that the stress margin for the existing 1/ inch
diameter mounting bolts was greater than a factor of five for allowable stresses.
The initial design appeared to have more than sufficient allowance to account for any vibration induced stresses.
Engineering design package EDP-7400, Revision A, June 10, 1987, was also reviewed at this time.
The 1/2 inch bolts on Pump B were changed to 1 inch bolts.
Also, the mounting bolts were changed from SAE Grade 5 to SAE Grade 7.
Steel spacers were installed in lieu of the rubber gasket at the mounting bolts for all four RHR pumps and minimum torques were specified for all termination mounting bolts.
Calculation DC No. 0367, Volume II, Revision C, October 6, 1987, was reviewed, as well.
Actual field measured vibrations were used as load inputs to this calculation.
Accelerations were assumed to be out of phase in order to maximize the stresses.
A 2.25 g acceleration was applied in all three directions simultaneously.
Resulting bolt stresses were with code allowables.
As a result of the multiple vibration inJuced fatigue failures of components located on or near the RHR pumps, a special vibration assessment program has been implemented.
The program includas:
(1) Small piping and tubing attached to the pump and suction /
discharge piping using established piping vibration acceptance criteria.
(2) Components connected to the pump and motor including all conduit boxes, seal water heat exchangers and sea / water cyclone separators.
(3) Hydraulic snubber connections between the fluid reservoir and hydraulic unit.
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(4) Lugs for pipe supports.
The actions taken to date appear to adequately address the problem.
No violation or deviations were identified within the areas inspected.
However, pending the results and conclusions of the above assessment program, this item will continue to be considered open.
3.
Actions on Licensee Event Reports a.
(Closed) LER (341/87-013-00):
Material failure of terminal box adaptors on reactor building cooling fan motors.
This item is currently being tracked under an Unresolved Item (341/87014-02(DRP))
and is discussed further in NRC Inspection Report No. 50-341/87025.
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Corrective actions initiated as a result of this event included the following:
(1) A field survey to identify other Reliance Electric motors that might use the adaptors.
(2) Temporarily _ replace the broken adaptors with new adaptors of the same design.
(3) Design and install replacement adaptors made from aluminua.
(4) Revise Specification 3071-128-EG controlling the installation of QA-I motor to include specific torque values for each motor.
(5) Design external support frames for the terminal boxes and specific Reliance Electric motors.
The NRC inspector reviewed the calculations associated with.
EDP-7278, Revision B, September 17, 1987, and EDP-7588, Revision 0, September 4, 1987.
These design packages pertained to the design of the aluminum adaptor plates and the external terminal box supports, i
respectively.
Also, ABN 7554-1, Revision C was reviewed by the NRC inspector.
This revised the motor mounting specification 3071-128-EG to include specific torque valves.
No adverse comments were made during these reviews.
Pending additional reviews of concerns delineated in NRC Inspection Report No. 50-341/87025, the Unresolved Item associated with this will remain open, b.
(Closed) LER (341/87-030-00) Overpressurization of the High Pressure Coolant Injection (HPCI) System Piping due to application of inappropriate type of check valve.
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This event was reviewed in detail during the inspection documented in NRC Inspection Report No. 50-341/87033.
It was ultimately, concluded that the HPCI System was capable of withstanding the loads developed during this event.
No violations or deviations were identified within the areas inspected.
4.
Vibration Testina - Results Evaluation a.
During startup vibration testing accelerometer A-601 on the RHR head spray line appeared to have exceeded its Level 1 acceptance criteria.
Data taken at approximately 70% steam flow indicated that the
vibration level of 12.8 mils had exceeded the allowable limit of 12.0 mils.
Evaluations of these results concluded that the accelerometer was defective and did not produce accurate readings.
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Based on the flexibility of the line configuration it was determined that the other sensor readings on the same line could not substantiated the indicated vibration.
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The NRC inspector reviewed the documentation associated with the above situation and concurred with the licensee's assessment.
No violations or deviations were identified.
b.
During startup vibration testing, accelerometer A-006 on the main steam "8" line appeared to have exceeded its acceptance criteria.
Data taken at approximately 80% steam flow indicated that the vibration level of 77 mils had exceeded the allowable limit of 50 mils.
Using configuration specific frequencies, pipe lengths and accelerometer locations, a revised acceptance criteria was calculated in Sargent and Landy Calculation EMD-064413.
Since that time, hand held vibration measurement were taken in the field during comparable operating conditions.
At that time, a very good correlations was found for adjacent accelerometers A-005 and A-008.
However, for A-006 the hand held data was found to be lower by more than a factor of six.
On this basis, accelerometer A-006 was considered defective and no longer considered in evaluation of piping vibration for the main steam "B" line.
The NRC inspector reviewed the documentation associated with the above situation and concurred with the licensee's assessment.
No violations or deviations were identified.
c.
During startup vibration testing, instrumentations 0-015 and 0-016 on the main steam high pressure loop piping inboard from the turbine leads appeared to have exceeded its acceptance criteria.
Data taken at approximately 85% steam flow indicated that the vibration level of 69 mils and 59 mils had exceeded the allowable limits of 34 mils and 37 mils, respectively.
Hand held vibration reading gave adequate correlation to substantiate the level of vibration.
Revised allowable displacements were calculated using the results of the transient analyses and detual weld configuration.
Calculations contained in EMD-061696 Addendum B showed that the acceptance limit could be increased to 102 mils and 105 mils respectively.
DECO reviewed this new acceptance limit under Safety Evaluation No. 88-0012 Revision 1, January 18, 1988.
The NRC inspector reviewed portions of the above documentation and had no adverse comments.
No violations or deviations were identified.
5.
Review of Licensee's DER 88-0045 for the feedwater Pumps a.
Background On January 19, 1988, the licensee initiated Deviation Event Report (DER) 88-0045 regarding potential overpressurization of the reactor feedpump section housings.
This action was prompted by a letter from the pump manufacturer (DeLaval) on January 18, 1988 that stated:
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"It has come to DeLaval's attention... that Fermi II is running the subject pump with a suction pressure of 700 psig.
DeLaval considers this pressure of 700 psig to be an unsafe condition.
The pumps were hydro tested at the factory to 600 psig with cold water.
The rated design condition for suction pressure is 395 psig."
The pump purchase specification 3071-5, September 15, 1969, "Reactor Feed Pumps", stated that the manufacturer shall hydrostatically leak test the pump at a pressure equal to 1.5 of the design pressure.
The design pressures were given as 1299 psia maximum discharge pressure and 410 psia maximum suction pressure.
Due to the pump housing configuration DeLaval was able to isolate the suction housing from the discharge housing and hydro tested the suction side to 600 psig and the discharge side to 2000 psig.
This practice is reportedly not common to most pump manufacturers.
Furthermore, this appears to conflict with the Hydraulic Institute recommendation to hydro test the pump to a pressure of 1.25 times the pressure that would result if the pump discharge valve were closed.
On this basis, it appears that the entire pump should have been hydro tested to approximately 1600 psig.
It should be noted, however, that the hydro described test is basically for leak detection and does not necessarily prove design adequacy.
If the pump design pressure is to be empirically determined, the burst pressure is determined using an actual burst test and the pressure rating is designated as one fifth of that pressure, b.
Corrective Actions The iemediate remedial actions taken by the plant were to maintain the suction pressure as low as reasonably possible, restrict personnel access to the reactor feed pump room, and monitor the pump bearing vibration for any signs of distress.
For a long term solution a parallel path approach was taken.
First, the existing feedpump casing will Le analyzed to oetermine if the suction operating pressure can be raised to approximately 800 psig.
Second, determine what range of operating and test conditions would require reductions in the available feed pump suction pressure and for how long.
Third, define the methods which could be employed to affect the pressure regulation needs as defined in the above second task.
Finally, define the cost and schedule for a replacement pump casing
installation.
From a generic viewpoint, a review of other plant
systems based on an overpressurization protection study done in 1986
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will also be performed.
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On February 3, 1988 a letter from Hopper and Associates concluded
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that the feed pumps are safe to operate at temperature with an I
internal pressure as high as 820 psig.
This information was based i
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b on "an accurate stress analysis, conservative material property assumptions and consideration of reserve design safety margin." The long term design capability assessment will be completed after detailed pump configuration information is collected during the upcoming LLRT outage.
Based on a review by one of DECO's lead engineer of the actual finite element analysis, the Nuclear Engineering department issued a memo concurring with the Hopper conclusion that suction pressures up to 820 psig can be safely sustained during plant operation and field verification in the form of strain gages and ultrasonic testing in the LLRT outage is essential.
In the opinion of the NRC inspector, the prog am implemented in response to the OER will adequately address the problem, No violations or deviation were identified within the areas inspected.
6.
Exit Interview The Region III inspector met with the licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on February 5,1988.
The inspector summarized the purpose and findings of the inspection.
The licensee representatives acknowledged this information.
The inspector also discussed the likely informational content of the inspection report with regard to documents nr processes reviewed during the inspection.
The licensee representatives did not identify any such documents / processes as proprietary.
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