IR 05000309/1993005
| ML20035C717 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 03/18/1993 |
| From: | Gray E, Lohmeier A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20035C715 | List: |
| References | |
| 50-309-93-05, 50-309-93-5, GL-92-01, GL-92-1, NUDOCS 9304090028 | |
| Download: ML20035C717 (10) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I-
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REPORT / DOCKET NOS. 50-309/93-05 LICENSE NOS.
DPR-36 LICENSEE:
Maine Yankee Atomic Power Company
FACILITY NAME:
Maine Yankee Atom.c Power Station i
INSPECTION AT:
Wiscasset, MA.
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INSPECTION DATES:
February 22-26,1993
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INSPECTOR:
VU> uu&
0*/f-73 A. Lohmeier, Sr. Reactor Engineer Date
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Materials Section, DRS/EB -
APPROVED BY:
E. Harold Gray, Chief, Materials Section Date
Engineering Branch, DRS
Areas Inspected: Engineering and technical support effectiveness in resolving the reactor _
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coolant pump vibration problem, responsiveness to Generic Letter 92-01, Revision 1, related
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to reactor vessel material integrity surveillance, and primary component transient operating cycle monitoring.
Results: Engineering and technical support staff effectively addressed these areas, providing
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for an engineered corrective action plan in the case of reactor coolant pump vibraf a, giving a comprehensive response to Generic letter 92-01, Revision 1, related to USNRC concerns for reactor pressure vessel material integrity surveillance, and properly recording and-evaluating the effect of plant operating cycles on remaining life of primary components.~
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DETAILS i
1.0 SCOPE OF INSPECTION (Inspection Procedure 37700)
The scope of this inspeedon includes engineering and technical support effectiveness in resolving the reactor coolant pump vibration problem, responsiveness to Generic Letter 92-01, Revision 1, related to reactor vessel material integrity surveillance, and primary component transient operating cycle monitoring.
2.0 FINDINGS r
2.1 Reactor Coolant Pump Vibration The inspector reviewed the issue of reactor coolant pump (RCP) vibration. Recently,' the licensee operators found that signals from the #3 RCP vibration monitors (with the signal obtained from proximity probes on the shaft) had increased from the normal level of 13-14'
mils to 16-17 mils double amplitude. Continued observation by the licensee noted further
increase in vibration to 18 mils, on which the operator reduced the plant power condition to hot standby while the vibration condition was investigated. The #3 pump was shut down, and inspection of the motor shaft indicated that the motor half coupling keyway had enlarged
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and the key became loose and shifted position along the shaft. After consulting with the
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pump manufacturer and motor vendor, the key was reinserted with appropriate shims to correct the key clearance. The motor was then restarted, the pump vibration returned to allowable levels, and the plant brought back to operation from the hot standby condition.
Licensee engineering, the motor vendor (Siemens), and the pump manufacturer (Byron Jackson) are presently reviewing the incident to de.termine the root cause of the failure.
Based on their findings, a technical evaluation will be written. The plant will be in a planned shut-down this summer, at which time the appropriate long term correcdve action will be taken on the basis of the technical evaluation.
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The inspector reviewed the vibration limits provided by Byron Jackson (BW/IP International, Inc. Pump Division - Byron Jackson Products. The normal levels of vibration allowed are 15 mils at steady state operation, at that level, an alarm is issued, and on continuous operation at 20 mils or transient operation for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 20-30 mils, it is specified that the pump be shut down. The level of pump vibration after the temporary repair returned to within the steady state operating limits.
The inspector noted that the corrective action taken by the licensee on discovery of the
change in vibration level, including response to alarms, controlled shut down, temporary l
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corrective action taken, initiation of technical evaluation, and planned subsequent permanent corrective action necessary, are consistent with sound engineering practice and commensurate i
with protection of the public health and safety.
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2.2 Reactor Vessel Structural Integrity - GL 92-01 The inspector reviewed the response of the licensee to Generic letter '(GL) 92-01, Revision 1, in a letter MN-92-65, July 2,1992, to the NRC. The licensee provided for either a direct response to each concern (question) of GL 92-01 or provided for a reference to prior. _.
correspondence with NRC that addressed the particular concem.' The inspector checked the
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responses to the letter together with the referenced correspondence and found the responses to be accurate for those areas inspected.- The GL 02-01 response is under review by.NRR.
2.3 Transient Operating Cycle Monitoring 2.3.1 Background
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The primary system components are designed to meet the requirements of Section III of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels. The Code requires a " design by analysis" approach to evaluating not only whether the components can sustain the prescribed steady state pressure and thermal loadings, but also the cyclic application of these -
loadings in view of the fatigue strength of the component materials.
The utility (owner of the components) specifies the types and circumstances of loadings which are anticipated during the plant lifetime. ' Components are dcsigned in accordance with these specifications. Therefore, in the case of cyclic loading, the specification will state the numbers and types of transient operation that can be anticipated threughout the plant life.
These transients are described in the Final Safety Analysis Report (FSAR) for the nuclear power plant. Operation beyond the specified numbers of cycles is outside the design bases described in the FSAR.
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Since primary system components are designed to sustain limited numbers of transients, the plant technical specifications (TS) reflect the requirement that records and documents relating l
to the cyclic operation of the plant must be maintained through the plant lifetime. These -
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data identify critical areas of the components subject to the operating transients for
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monitoring to determine whether the design life of the component has been expended.
The criteria for exhaustion of fatigue life are reflected in a cumulative usage factor.(CUF),
j which is an integrated summation of the ratio of expected numbers of cycles at the applied '
y strain range to the cycles at that strain range necessary to cause fatigue failure. An -
appropriate factor of safety in terms of strain level or cycles is utilized in the same sense as a.
l factor of safety for stress level in relation to fracture stress.
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2.3.2 Final Safety Analysis Report and Technical Soecification Reauirements l
The inspector reviewed the FSAR for Maine Yankee Power Station (MYNPS), which J
specifies the number and types of reactor coolant system (RCS) transients for which each
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component has been designed over the duration of its 40 year operating license.' These l
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values are listed in " Design Basis - Design Cyclic Loads," FSAR Table 4.2.2.-
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applicable Code requirements are described in this table. RCS components include the I
reactor vessel, steam generators, pressurizer, RCP casing, quench tank, pressurizer safety system, loop isolation valves and piping. These components are designed to Section III of -
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the ASME Boiler and Pressure Vessel Code - Nuclear Vessels with the exception of piping, which is designed to ANSI B31.1.
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Review of the TS by the inspector indicated that Section 5.10.2.f requires that records of operational cycles for the reactor pressure vessel and the reactor coolant system be retained for the duration of the facility operating license.
As a result of this review, the inspector found that the licensee is. required to operate the RCS within the limits of the design basis expressed in the FSAR and that the records of r
cyclic operation be retained for the life of the license. Operation of the primary system
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components and piping must be within these transient limitations.
2.3.3 Retention of Ooerational Cycle Records The inspector examined the system used by the licensee to collect, retain, and disseminate operational data records. It was found that the licensee had established a comprehensive system for the collection, retention and dissemiaation of operational records. The
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responsibilities and controls for the collection, processing, storage and retrieval of quality assurance records is described in licensee Procedure No. 0-17-1.
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The operational records (on supervisory instrumentation charts) are generated in the control
room under guidelines of MYAPC Section XVII - Quality Assurance Records. The
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operations department provides for temporary storage of records under procedures 0-17-1 and 1-200-8 (Operations Department Surveillance Schedule, Implementation, and Records).
Records are transferred to the Technical File Center in accordance with procedure 0-17-2.
l Receipt of the records is performed under procedure 16-250-1, temporarily stored under
procedure 16-250-2, and permanently stored in accordance with procedure 16-250-3. The l
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records management system and records control procedures are consistent with the requirements of Regulatory Guide 1.88 and ANSI N45.2.9.
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A " Maine Yankee Monthly Statistical Report" is sent to the NRC, which is consistent with the TS requirements to retain records of operational transients for the lifetime of the license.
As a result of the inspection of record retention procedures, it was found that the licensee
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activity is consistent with the TS requirement expressed in 5.10.2 (f) for the retention of.
records of transient or operational cycles. Procedures provide for monthly reporting of plant
operation statistics to the NRC.
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2.3.4 Encineerine Review of Transients
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The inspector noted that, although MYAPS had no formal procedure for comparison of actual operational cycles with the cycles for which the plant was designed, MYAPC
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engineering had shown alertness to this issue on the basis of generic information related to fatigue usage concerns at operating nuclear power plants.
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The original fatigue evaluation of primary components was performed in the period 1967-70.
In 1987, the fatigue life usage was evaluated on the basis of information related to the lack of transient monitoring at the Vogtle station. In 1988, fatigue life usage was evaluated as a
result of an operation request to review the effect of loop isolation. In.1989, fatigue life usage was reevaluated as a result of a CE information bulletin 88-01 indicating that a nonconservative usage factor evaluation existed in the original fatigue evaluation. In 1991, a
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200*F cooldown incident motivated another review of fatigue life usage. In late 1991, pressurizer surge line stratification issues also initiated a review of fatigue life usage.
This plant is presently operating at the mid-point of the 40 year life for which the primary components have been designed. It is now in its 14th operating cycle. The primary component fatigue life usage factors at this point are below half the values of transients for which the plant was designed. At present, there appears to be every indication that the plant will operate with primary component transients at a level under the number for which the components were designed. As a consequence, the component fatigue life usage will remain under 1.0.
r The licensee has recognized the lack of a procedure for counting and recording of transient operating cycles and comparison with design operating cycles. The inspector was informed
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by the licensee that funding for improvement of procedure documentation is being considered in the coming year. As the plant approaches end of life, changes in numbers of operating l
cycles may increase, the records provide guidance in focusing on component areas where
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approaching the fatigue life usage limit is a problem.
2.3.5 Component Stress Report Review l
In order to verify the current status of primary component fatigue life usage, the inspector i
reviewed summary sheets of primary component fatigue life usage. The inspector also reviewed equipment specifications, stress analyses, and fatigue evaluation reports of the
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equipment manufacturers to verify the areas reviewed in the stress reports in relation to the areas specified in the FSAR. The licensee summarized the significant transient cycle expenditure as follows:
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COMPARISON OF ACTUAL CYCLES AT MID-LIFE WITH DESIGN CYCLES Design Actual Cycles Cycles Plant Imading/ Unloading 15,000 484 F
Plant Heatup/Cooldown 500
Primary Hydrotest 3125 psi
1 Primary Leak Test 2500 psi 200
Sec Hydro 1250 psi
4 Sec Leak Test 1000 psi 200
React Trips fr 100% Pwr 400
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Manway Boltups 500
RCP Starts / Stops 4,000 200 Injection at 120F - 50 gpm 200
Seismic 1,000
It was noted by the inspector that the numbers of transients for which the primary components have been designed to sustain the consequent fatigue cycling have not been exceeded. Furthermore, should the cycling continue, at the previously experienced rates, for the remainder of the license lifetime, the design cycles will not be exceeded. Therefore, the primary system components have been operated within their design bases.
The inspector reviewed the licensee summarization of design cumulative usage factors for primary system components having significant levels. They are listed as follows:
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CALCULATED CUMULATIVE USAGE FACTORS (40 Year Life)
i REACTOR PRESSURE VESSEL STEAM GENERATORS
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Head Flange
.1073 Secondary Shell
.0472
Closure Studs
.3907 Steam Outlet Nozzle
.0089 Vessel Flange
.0100 Secondary Handhole
.0269 Control Rod Housing
.1339 Feedwater Sparger
.0208
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Instrumentation Nozzle
.0826 Tubes and Tube Supports
.2917 Vent Tube
.0826 Tube to Tubesheet Weld
.3907 Inlet Nozzle
.0646 Tubesheet and Structure
.9139 l
Outlet Nozzle
.1392 Support Skirt
.1295-Vessel Wall Transition
.0053 Primary Nozzles
.2546 Holder and Brackets
.0940 Primary Manway Cover
.1424 Vessel Shell Bottom Head
.0009 Primary Manway Stud
.2500 Bottom Head Inst Nozzle
.0000 Divider Plate
.1231
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Flow Baffle
.0125 PRLMARY PIPNG PRESSURIZER i
Cold leg
.0158 Bottom Head -Support Skirt
.3621 Safety Injection Nozzle
.3589 Surge Line Nozzle
.1355 Surge Nozzle
.2317 Spray Nozzle
.3000
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Charging Inlet Nozzle
.0294 Temperature Nozzle Welds
.4648 Stop Valve -
.0292 Manway Assembly
.0114 Fill Nozzle
.3625 Heater Sleeve
.0078 i
Sampling Nozzle
.4527 Review of the list of cumulative usage factors (CUF) computed for the 40 year lifetime of the components verifies the licensee estimate that the CUF limit of 1.0 will not be exceeded unless the number of transient operating cycles increases. Those transients with significant
lifetime CUF (>.3) have been shown in bold print for emphasis. Should changes occur in
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the material fatigue curves (such as environmental effects on fatigue strength), these values may approach or exceed 1.0 during the component lifetime.
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The foregoing values of CUF were determined on the basis of stress analysis of the stmeture.
with assumed histograms for transient operational cycles over the design lifedme. The licensee reviewed with the inspector the consequence of additional transients applied to the structure which were not considered in the original design of the components. The transient
'7 for which additional consideration was given by the licensee was that due to flow stratification in the pres;urizer surge line piping. This analysis was reviewed by the inspector and the following results noted:
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LIFETIME CUMULATIVE USAGE FACTORS (Including stratified flow effects)
Surge Line U
.937 Reactor Coolant Pipe Surge Line Nozzle U
.630 Pressurizer Surge Line Nozzle U
.424 The inspector noted the effect of considering flow stratification in piping and nozzles is to increase the rate of fatigue life usage of the component materials to significant levels. As the component approaches end of 40 year lifetime, attention must be given those component.
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areas nearing a CUF of 1.0.
The licensee maintains a procedural system in which the Maine Yankee Atomic Power Station (MYAPS) primary system operating transient data is recorded and transferred to record storage facilities. A " Maine Yankee Monthly Statistical Report" is sent to the NRC, which is consistent with the TS requirements to retain records of operational transients for the lifetime of the license. The records show that the operation of the primary system components under the specified types of design transients has been in accordance with the design parameters described in the FSAR. Operation under considerations not described in
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the FSAR (stratified flow in the pressurizer surge line system) were evaluated and the
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lifetime fatigue life usage was within Code limits.
3.0 SUMMARY OF FINDINGS i
Corrective action taken by the licensee on discovery of the change in vibration level of a reactor coolant pump including response to alarms, controlled shut dow11, temporary corrective action taken, initiation of technical evaluation, and planned subsequent permanent corrective action necessary, are consistent with sound engineedng practice and commensurate with protection of the public health and safety.
The licensee response to the concerns of NRC expressed in GL 92-01, together with the referenced correspondence were found to be accurate for those areas inspected. The response to GL 92-01 is under review by NRR.
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The license, maintains a procedural system in which the MYAPS primary system operating transient data is recorded and transferred to record storage facilities and a monthly statistical report is sent to the NRC. The records show that the operation of the primary system components under the specified types of design transients has been in accordance with the i
design parameters described in the FSAR. Operation under transients not described in the FSAR (stratified flow in the pressurizer surge line system) were evaluated and the estimated lifetime fatigue life usage was within Code limits.
4.0 MANAGEMENT MEETINGS
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An entrance meeting was held on February 22,1993, with Maine Yankee Atomic Power Company personnel. An exit meeting to discuss inspection findings with licensee personnel was held on February 26,1993. Attendees are listed in Attachment A.
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ATTACHMENT A
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The following persons were contacted by the inspector at the entrance meeting on February 22,1993, at the exit meeting on February 26,1993, and during the course of the -
inspection:
Maine Yankee Atomic Power Comoany-
- R. Blackmore, Plant Manager - MYAPS
- C. Eames, Principal Engineer.- MSCED
- J. Frothingham, Manager - MY Quality Programs
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H. Gilpatrick, Mechanical Section Head - Corporate Engineedng
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- J. Hebert, Manager - Licensing and Engineering Support
- W. Henries,12ad Mech Engr - YNSD
- H. Jones, Jr., Principal Nuclear Engineer - MY
- R. Jordan, Senior Licensing Engineer, MY
- R. Nelson, Manager - Corporate Engineering W. Schubert, MY Plant Engineering
- C. Shaw, Manager - MY Plant Engineering E. Soule, Section Head - Mechanical Engineering G. Whittier, Vice President - MY Licensing and Engineering State of Maine
- P. Dostie, State Nuclear Safety Inspector - State of Maine
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U.S. Nuclear Reculatory Commission
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C. Marschall, Senior Resident Inspector - MYNPS
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- W. Olsen, Resident Inspector - MYNPS j
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- Attended Exit Meeting on February 26,1993
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