IR 05000250/1989047
| ML17347B476 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 12/12/1989 |
| From: | Blake J, Chou R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML17347B475 | List: |
| References | |
| 50-250-89-47, 50-251-89-47, NUDOCS 8912210239 | |
| Download: ML17347B476 (18) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-250/89-47 and 50-251/89-47 Licensee:
Florida Power and Light Company 9250 Mest Flagler Street Miami, FL 33102 Docket Nos.:
50-250 and 50-251 License Nos.:
.DPR-31 and DPR-41 Facility Name:
Turkey Point 3 and 4 Inspection Conducted:
October 23-27, 1989 Inspector:
C-Ric
.
u Approved by:
J,
.
lake, Chief M
e als and Processes Section gi eering Branch ivision of Reactor Safety (~-t~~9 Date Signed I> -t2-->>
Date Signed SUMMARY Scope:
This routine, announced inspection was conducted in the areas of previous open items, recent modifications, and the snubber-inspection and water-hammer-prevention programs.
Results:
In the areas inspected, violations or deviations were not identified.
The high corrosion areas had been painted to prevent or reduce the corrosion on piping systems and structures per the licensee's maintenance program.
The licensee took quick actions to respond to the inspector's concerns about the snubber irregular rotation and the Mej-It anchor bolt safety factor.
The safety factor less than two for two pipe supports installed with Mej-It anchor bolts was resolved on October 30, 1989 with the licensee's engineers working thru the weekend.
A Safety Evaluation Report was received on November 3, 1989.'912210235'91212 P13R ADOCK 0 0002
0 PDC
REPORT DETAILS Persons Contacted
Licensee Employees
- J.
W. Anderson, Quality Assurance (QA) Supervisor - Regulatory Compliance
- J. Arias, Jr., Technical Assistant to Plant Manager
"M. Blew, Inservice Inspection (ISI) Coordinator J.
Crockford, Operation Support Supervisor
- J.
E. Cross, Plant Manager
- D. J.
Feingold, System Engineer
- S.
M. Franzone, Lead Engineer R. Gavankar, Engineering Specialist/Consultant
- Juno Beach
"R.
D. Gil, Civil Engineering Manager - Juno Beach K.
H. Greene, Civil Engineer Supervisor - Juno Beach
- S. T. Hale, Engineering Project Manager
- D.
W. Herr in, Regulatory Engineer
- V. A. Kaminskas, Technical Supervisor
"L.
W. Pearce, Operation Superintendent
- D.
R. Powell, Regulatory Supervisor
- K. Remington, System Performance Supervisor
- F.
H. Southworth, Assistant to Site Vice President
"G.
A. Warriner, Quality Control (QC) Supervisor M. Wayland, Maintenance Superintendent Other licensee employees contacted during this inspection included craftsmen, engineers, mechanics, technicians, and administrative personnel.
Other Organizations Bechtel Power Corporation M. Shekmer, Senior Civil Engineer S.
Rao, Senior Mechanical Engineer - West Palm Beach NRC Resident Inspectors
- R.
C. Butcher, Senior Resident Inspector
- G. A. Schnebli, Resident Inspector T. McElhinney, Resident Inspector
- Attended exit interview 2.
Recent Modifications e
Main Steam Isolation Valve (MSIV) closure in five seconds or less could not be assured following a loss of instrument air.
This condition was a
violation of Technical Specification 3.3, 3.8. l.b and c requirements and
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was reported to the NRC as a design deficiency by License Event Report (LER) 85"020.
This could result in the eventual loss of main steam to drive the Auxiliary Feedwater Pump Turbines.
The MSIV Air Accumulator modification will provide a dedicated safety-related air reserve for each MSIV.
This modification will ensure that each MSIV will close in five seconds or less upon receipt of a
closure signal.
The system modifications were performed during the last refueling outage for Unit 4.
The work was done based on Plant Change/Modification (PC/M) No.88-245.
The inspector discussed the modification with the licensee's engineers and reviewed the pipe support calculations provided.
The support calculations were revised based on the new stress loads.
The support calculations were reviewed for qualification of the welds, U-bolt, straps, member size, clamps, base plates, and expansion anchor bolts.
The reviewed calculations were listed below and all were acceptable.
Support Calculation No.
P-SJ 169-01, Rev.
Support Nos.
H-848-01 to H-848-14 Support Nos.
H-849-01, 02, 03, 04, 05, ll, and 14 Snubber Inspection Program and Procedures (70370)
The snubber inspections and tests for Inservice Inspection (ISI) are required per American Society of Mechanical Engineers (ASME)
Code Section XI.
The snubber inspections are to check on fluid levels, hot settings, damage, obstructions for the moving parts, and corrosion.
The snubber tests, also called functional tests, are to verify that the maximum drag force is within required limits; activation is achieved within the specified range of velocity or acceleration (in both tension and compression);
and that the snubber bleed, or release rate, where required, is within the specified range in compression or tension.
Per ASME Code Section XI, Article IWF-5400(b),
a 'representative sample of 10 percent of the total number of nonexempt snubbers whose load rating is less than 50 kips shall be tested each inspection period.
The inspector discussed the snubber inspection program with the licensee's engineer and reviewed the information provided.
The Technical Specification, Section 4.14, snubber, includes inspection types, visual inspections, visual inspection acceptance criteria, functional tests, functional test failure analysis, mechanical snubbers functional test acceptance criteria, and snubber service life monitoring.
The visual inspections include a
subsequent visual inspection schedule based on number inoperable snubbers of each type (on any system)
per inspection period per unit.
For the functional tests, snubbers which failed th'
previous functional test shall be retested in addition to the regular sample during the next test period.
The following procedures were reviewed to check the adequacy of the program:
Administrative Procedure 0190.83, Mechanical Shock Arrestor Surveillance Program
Administrative Procedure 0190.85, Functional Testing of Mechanical Shock Arrestors Operating Procedure 0209.9, Visual Examination of Mechanical Shock Arrestors Maintenance Procedure 0707.33, Snubber Removal and Replacement Examination Procedure NDE 4.3, Visual Examination (VT-3/VT-4)
The above procedures were reviewed and determined to be acceptable and adequate.
Table A-1 for Unit 3 and Table A-2 for Unit 4 attached to Administrative Procedure 0190. 83 list all snubbers for the surveillance program.
Those tables listed the drag force acceptaance criteria as five percent for all except for four snubbers for each unit, which have a two percent criteria.
The manufacturer's recommended drag force was one percent, but due to the difficulty of maintaining the drag force within one percent range, the licensee contracted to Gibbs and Hills, Inc.,
San Jose, California for a study of the use of five percent drag force.
Pressurizer Spray Piping Snubber Drag Force Study Completion Report performed by Gibbs and Hill, Inc.
dated November 1985, concluded that the five percent drag force could be used except for four supports.
Therefore, the licensee does have program and procedures for the snubber inspections and tests.
Water Hammer Prevention Program (70370)
One form of internal impact force is due to the propagation of pressure waves produced by sudden changes in fluid momentum.
This phonomena is often called water or steam hammer.
It may be caused by the rapid opening or closing of a valve in the system.
The location and magnitude of water hammer are hard to predict.
A lot of water hammers occurred due to the wrong sequences of valve opening and closing.
Most water hammers were noticed after the damages was found.
Since water hammer is hard to predict, the licensee does not have any program to control it.
The inspector discussed with the licensee engineer and reviewed the procedures provided.
The licensee did experience three major water hammer problems on Feedwater Lines:
a.
P rob 1 em Original design on steam generator feed ring and supply line was inadequate.
Solution The original design allowed the horizontal feedwater piping and feed ring to drain.
Upon a loss of feed and -subsequent re-initiation there was some damage.
This problem was resolved with a feed ring redesign and a feedwater piping reroute.'
b.
Probl em System operation with warm-up lines and the recirculation valve on the steam generator feedwater pump resulted in water hammers in the case of the plant trips and a loss of power.
Solution Section 5.0, Startup/Normal Operation of Procedure Nos.
3-OP-074 and 4-0P-074, Steam Generator Feedwater Pump, have been revised to minimize this condition by:
1) opening warm-up bypass valves 20'30 minutes prior to starting feed pump; 2) verifying the feedwater pump recirculation valve control switch to be in the
"close" position; 3) closing warm-up bypass valves after pump was started.
The licensee has not experienced a water hammer from this type event in a number of years after taking the above control steps.
c, Probl em-Auxiliary feedwater check valve backleakages resulted in a water hammer.
Solution Attachment 2, Nuclear Power Operations Logsheets of Procedures Nos.
3-0SP-201.3 and 4-0SP-201.3, Nuclear Power Operation Daily Logs requires every shift to check auxiliary feedwater pump discharge piping temperature with hands to be able to hold more than three seconds which equals to approximately 120'F.
If pipe can not be held three seconds, the inspector should. notify the Plant Supervisor - Nuclear for the possible leakage.
Those checks have helped identify
- the irregular condition prior to it being a problem that could result in a water hammer.
5.
Independent Inspection The resident inspector during the normal walkdown found that a snubber located at an outside and back area of plant had a
sag in the middle portion of the snubber.
The inspector inspected this snubber, No. 3-1019 (Support No.
PRWH-1) of the Safety Injection and Residual Heat Removal System near Refueling Water Storage Tank (RWST) Unit 3 and found it to be 7.5 out of alignment.
This is out of required tolerance which is 6'.
This snubber is located at one side of a major walkway with very heavy traffic.
People might carry heavy equipment and step on it.
The licensee gC inspector checked the snubber condition against the support drawing and found four additional discrepancies between the as-installed condition and as-built drawings such as rear bracket attachment, snubber unit attachment, snubber installed 180 from specified orientation and double nuts.
Nonconformance Report (NCR)
No. 89-0367 was issued.
The licensee's engineers evaluated this NCR and concluded that the discrepancies found had no safety impact to the plant operation.
The disposition for this NCR is to rotate clamp and snubber within the tolerance, to revise drawings to resolve the discrepancies, and to
e determine the root cause a'nd long term fix later.
Plant Work Order (PWO)
No.
WA 891025 161848 was initiated and the 7.5 'nubber mis-alignment was corrected in the field.
A Design Change Request (DCR) will be issued for the drawing change.
6.
Action on Previous Inspection Findings (92701)
e a ~
(Closed)
Inspector Followup Item 50-250,251/87-52-02, Maintenance Procedures for Piping Systems The inspector held discussions with the licensee's engineers concerning the progress being made on developing a
program and schedule for the maintenance of piping systems based on Procedure No.
O-ADM-718, "Component Supports Preventive Maintenance Inspection Program."
The above procedure was written and approved on September 6,
1988 in response to the previous inspection results where the piping systems were found to lack maintenance such as heavy, corrosion, missing parts, member deformation, spring can setting, etc.
The licensee has contracted with Williams Power Company to remove rust, clean and paint supports.
All the heavy corrosion areas had been cleaned and repainted.
The project is approximately 50 percent completed.
The licensee estimates that the first cycle through the plant wi,ll= be completed in about 30 months where one cycle is about four to five years.
To control good quality painting, the licensee imposes a written procedure which requires each painter to pass a test before performing the work.
The Technical Department (ISI Group) is developing an Inspection Schedule which should be completed in about three months.
The Technical Department is also currently developing a program for the inspection of the secondary side turbine plant supports on the main condensate, feedwater, standby steam generator feed pump, main steam, and extraction steam systems.
This program will consist of the following:
Phase
"A" -
Acquiring the isometric drawings.
To be completed by October 31, 1989.
Phase "B"-
Phase
"C" '-
Walkdown of the isometric drawings to verify piping configuration, hanger location and identification of the hangers.
To be completed by January 31, 1990.
Development of a computer program for tracking and scheduling of the above identified systems.
To be completed by February 15, 1989.
This is the first phase of walkdown inspection for the piping systems to follow the Procedure No.
0-ADM-718 and test its adequacy.
Based on the licensee's actions to date and their future plans, this item is close b, (Closed) IFI 50-250, 251/89-01-04, Spent Fuel Pool Cooling System This system was upgraded from nonsafety-related system to safety-related system to ensure that its cooling function will be maintained throughout a seismic event.
The inspector discussed this system with the licensee engineers and reviewed stress and support calculations provided.
The licensee performed the following stress calculations for upgrading the system:
Unit 3 Calculation No. M08-595-01, Rev.
3 (Iso, 5610-P-690, Rev. 1),
From Pool to Pump Suction Calculation No. M08-595-02, Rev.
2 (Iso, 5610-P-691, Rev. 0),
From Pump Discharge to Heat Exchanger Calculation No. M08-595-03, Rev.
1, From Heat Exchanger Back to Pool Calculation No. M08-595-04, Rev.
1, Drainage Piping Unit 4 Calculation No. M08-595-05, Rev.
1, From Pool to Pump Suction Calculation No. M08-595-06, Rev.
2, From Pump Discharge to Heat Exchanger Calculation No. M08-595-07, Rev.
0, From Heat Exchanger Back to Pool The inspector randomly selected stress calculation Nos.
M08-595-01 and 02 for Unit 3 for review.
These two calculations are considered as the main calculations.
P8ID Drawing No. 5-610-T-E-4515, Rev.
52, ISO Drawing No. 5610-P-690, Rev.
and-691, Rev.
0 also were reviewed to confirm the computer input for the stress calculations.
The stress calculations contain:
1) stress analysis summary; 2) support load summary; 3) input file listing; 4) mathematical model; 5) output; 6) miscellaneous information; and 7) isometric drawings.
American National Standard Institute (ANSI) Code B31. 1 Power Piping and Bechtel Standard Computer Program ME101 Version K3 were used.
The analysis conditions include two cases which consisted of (i) design temperature of 212'F and pressure of 150 psi and (ii) operating temperature of 143'F and pressure of 70 psi.
The pipe diameters included 10", 8", 4",
and 3".
The material was SA312 TY304.
The inspector checked the stress input data and reviewed the stress output data.
The maximum stress ratio was 0.27 for the primary bending and 0.98 for thermal for stress calculation No. M08-595-01.
The maximum stress radio was 0.39 for the primary bending and 0.56 for thermal for stress calculation No. M08-595-0 The calculated stresses were within the code allowables.
Due to the, stress reanalyses, modification is required to add new supports and modify the existi ng supports.
During the review, temperature and pressure conditions used in the analyses could not be found in Reference 3.
This reference should be replaced by the reference from Westinghouse Line List Description.
Also Reference 4 should be replaced with Westinghouse Material Classification for Pipe, Specification No. G-569866, Rev.
4.
The licensee engineers agreed to revise the calculations to refer to the proper references.
The following pipe support calculations were also reviewed due to the new stress loads.
The review included computer input, computer output, mathematical model, member size, lug, U-bolt, weld, stanchion, sway struts, deflection, base plates, anchor bolts, and change request notice evaluation and disposition.
'alculation No. P-456-04, Rev.
Support Nos.
HlA, H1B, H2A, H2B, H3A, H3B, H4A and H4B.
Calculation No. P-456-05, Rev.
Support Nos.
H5A and H5B Calculation No. P-456-11, Rev.
Support Nos.
H-694-05 and
Calculation No. P-456-12, Rev.
Support Nos.
H-695-01, 02, 03, 04, 05, 06, and
Calculation No. P-456-13, Rev.
Support Nos.
H-695-01, 02, 03, 04, 05 and
The inspector considered that the stress calculations and support calculations were of good quality and acceptable.
The Inspection Report No. 89-29 documented a physical inspection to verify the field modification.
Discrepancies in two supports were found during that walkdown.
Discrepant Field Condition (DFC)
No.
DFC-89-192 for Support No. H-694-05 and DFC-89-193 for Support No. H-694-06 were issued by the licensee.
The licensee transferred the above two DFC into NCR No. N-89-0876 for tracking, evaluation, and disposition.
The licensee will revise dr awings to reflect the field conditions.
Based on the above review of the stress and support calculations and the field walkdown on the modification, this item is considered closed.
(Open) IFI 50-250, 251/89-29-02, Evaluation of Design Capacities of Installed Wej-It Concrete Anchor Bolts The inspector held discussions with the licensee's engineers and reviewed the Safety Evaluation Report provided.
Safety Evaluation SE No. JPN-PTN-SECJ-89-113,
"Operability Assessment of Pipe Supports with Reduced Capacity Wej-It Concrete Anchor Bolts",
Rev.
0, dated
October 25, 1989 was reviewed.
The licensee decided to use the same design capacity which was used by Florida Power Corporation for Crystal River based on their test results.
The reasons are that the test results at Crystal River and Turkey Point were very close and the test concrete conditions, especially the aggregate, were the same.
After applying the Crystal River design capacity reduction factor of 0.4 or 0.6 for both tension and shear allowables, the licensee evaluated all supports involved Wej-It anchor bolts and listed
supports with safety factors between 1.6 and 2.6 based on the reduced capacity on Table 3 of the above safety evaluation report.
In the safety evaluation report, the licensee concluded that those critical supports with safety factors between 1.6 and 2.6 met the IE Bulletin 79-02 requirements for plant continued operation.
The inspector pointed out that IE Bulletin 79-02 requires a safety factor of two or greater for the expansion anchor bolts for plant continued operation unless other justification can be made.
The inspector requested the licensee reevaluate the supports which had a safety factor below two.
After the licensee removed the conservative loads and reevaluated the stress and support calculations, the safety factor for Support Nos.
3-MSH-20 and SR-612 were still below two by the time of exit meeting.
The licensee continued to try to resolve those two supports.
The catalog capacities in Table 1 of the Safety Evaluation Report are different from those in Bechtel Calculation No.
C-7754-14, Rev.
0 and Specification No. 5177-098-C-OOl, Rev.
which were used for the Wej-It anchor bolt capacities in Turkey Point.
Based on the comparison between Crystal River Test Results and.Capacity listed in the above calculation and specification, the capacity reduction factor for 1"g and l>4"g may be reduced to 0.5 or 0. 55 instead of 0. 6 which Turkey Point plans to use.
The licensee continued their justification and evaluation through the weekend and informed the NRC via telephone about their results on October 30, 1989.
Support No.
SR-612 has a safety factor over two by using the sum of ratio square which was permitted by NRC.
Support No.
3-MSH-20 has a safety factor below two and was temporarily removed from the system analysis.
The reanalyses for this system in Piping Stress and Supports without Support No.
3-MSH-20 met the continuing operation requirements of IE Bulletin 79-02.
The inspector through telecopy received and reviewed a revised Safety Evaluation Report for Wej-It Anchor Bolts (Rev. 1), dated November 3, 1989.
The resolution did provide the basis for the plant continuing operation.
Therefore, Turkey Point met the interim operation requirements.
Pending the licensee continuing their evaluation or modifications if required to meet the long term operation requirements per IE Bulletin 79-02, this item remains ope.
The inspection scope and results were summarized on October 27, 1989, with those persons indicated in paragraph 1.
The inspector described the areas inspected and discussed in detail the inspection results.
Proprietary information is not contained in this report.
Dissenting comments were not received from the licensee.
By the exit interview, the licensee still had two pipe supports installed with Wej-It anchor bolts with a safety factor below two.
The safety factor of two or greater for the expansion anchor bolts is required for the interim (or continued) operation per IE Bulletin 79-02.
The licensee continued their evaluation effort since two supports still had a safety factor below two.
Another possibility of the safety factor below two could exist during the licensee detail evaluation of about 270 pipe supports.
Therefore, the licensee stated during the exit interview that they will complete a modification within two weeks if any pipe support with Wej-It anchor bolts has a safety factor below two and could not be justified with other methods.
The management promised to resolve the safety factor below two for two supports by the following Monday and inform the NRC about the results.
The NRC did receive results on October 30 and November 3, 1989, as stated in Paragraph 6.
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