IR 05000458/1993022
| ML20056C960 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 07/15/1993 |
| From: | Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | |
| Shared Package | |
| ML20056C959 | List: |
| References | |
| 50-458-93-22, NUDOCS 9307300162 | |
| Download: ML20056C960 (25) | |
Text
- .
_ ,
f r
.. .
!
. !
APPENDIX f U.S. NUCLEAR REGULATORY COMMISSION
REGION IV
i Inspection Report: 50-458/93-22 ;
i Operating Licenses: NPF-47 .
Licensee: Gulf States Utilities :
P.O. Box 220 St. Francisville, Louisiana 70775
.
Facility Name: River Bend Station i
Inspection At: St. Francisville, Louisiana, and NRC Region IV offices, Arlington, Texas Inspection Conducted: June 7-11 and 21-25, 1993 at RBS June 14-18, 1993 at Region IV Offices ;
Inspectors: C. J. Paulk, Reactor Inspector, Engineering Section ,
Division of Reactor Safety ;
P. A. Goldberg, Reactor Inspector, Engineering Section Division of Reactor Safety 1
,
Approved: /wo T. F. Westerman, Chief, Engineering Section 2-/P Date l Division of Reactor Safety
!
Inspection Summary Areas Inspected: Routine, announced inspection of engineering and technical .
support activities, including the design change proces Results:
- The licensee was observed to have implemented a very good design change process. Governing procedures for design changes and temporary ;
modifications were comprehensive and well written. However it was
'
observed that some temporary modifications (Prompt Modification' '
Requests) were three years old. (Section 3.1, and 3.2) :
t
'
- In general, the engineers interviewed were knowledgeable in their areas of direct responsibility; however, their knowledge of system .i interrelations appeared to be weak. (Section 3.3) ,
i e From the review of condition reports assigned to engineering, it was :
noted that more than half were not being completed within-the time- frame !
}
9307300162 930723 p i PDR ADOCK 05000458 ry !
G PDR he
. .,
- k
.
-2-assigned. Those with extensions were generally completed after the extended date. Based on discussions with responsible engineers, it appeared that individual problems were being corrected well, however, the condition reports were found to be generally weak in documentation '
of problem, cause, and corrective action required and take (Section 3.3.1)
- The licensee was monitoring their engineering backlog. It was noted that there was a backlog of modification requests and a significant backlog of drawing changes. In addition the number of open engineering i condition reports and vendor information requests have been increasing since mid 1992. Increased contract engineering services were planned to be added to supplement the current staff. (Section 3.3.5)
- The licensee indicated that increased emphasis was being placed on improving plant performance and reliability. It was recognized that this will result in additional engineering work and increased man power needs. The current unplanned outage was affecting manpower availability since it was delaying engineering activities in order to address '
emerging issues. (Section 3.3.5)
- The inspector's review of the design bases sizing calculations for the safety-related air accumulators associated with the main steam safety '
relief valves indicated that they were well prepared and reflected conservative engineering practices. (Section 3.3.4) ;
- A number of observations were made with regard to issues related to !
safety and relief valve sizing and testing. (Section 3.3.2)
-
The inspectors identified a number of discrepancies between the valve vendor specification data and the -
sizing calculations. No immediate operability concerns were identifie The set point testing methodology utilized for the main steam safety relief valves did not appear to take into account the temperature profiles the valves experience during normal operatio As-found testing of the main steam safety valves was not ;
performed after removal since all valves were replaced each .
refueling outag ;
-
Over pressure protection of the safety-related portion of the air system for the main steam safety valves was provided by relief valves in the nonsafety-related portion of the main steam safety relief air syste . ._ _
.. .
.
.
-3- j
- Relief valves in the emergency diesel generator system were found to be mounted in a horizontal positio The manufacturer was contacted and they recommended r that the relief valves be installed in a vertical ,
position. .The manufacturer, however, indicated that this was more significant for larger valves. The
'
licensee was reviewing the relief valve orientatio j
-
The inspectors also noted that a relief valve had been installed in the horizontal position in the nonsafety i portion of the instrument air system. A field change l has been implemented to reorient the relief valv l
-
Two ASME III pressure relief valves were noted during 5 a walkdown by the. inspectors to have broken and/or missing seal wires. The licensee demonstrated that l condition reports were being written at the time of i testing for similar conditions, but the licensee did not appear to have been proactive to conditions in the plan Summary of Inspection Findings: ,
!
'
= Unresolved Item 458/9304-05 was closed (Section 2.1)
- Violation 458/9222-01 was reviewed but not closed (Section 2.2) [
- Inspection Followup Item 458/9322-01 was opened (Section 3.3.2)
- Inspection Followup Item 458/9322-02 was opened (Section 3.3.2)
- Inspection Followup Item 458/9322-03 was opened (Section 3.3.2) ,
i
- Inspection Followup Item 458/9322-04 was opened (Section 3.3.2) {
- Inspection Followup Item 458/9322-05 was opened (Section 3.3.2) l
,
- Inspection Followup Item 458/9322-06 was opened (Section 3.3.2)
'
- Inspection Followup Item 458/9124-01 was reviewed but not closed (Section 3.3.4). ,
Attachments: I F
= Attachment I - Personnel Contacted and Exit Meeting :
,
- Attachment 2 - Documents Reviewed !
!
- Attachment 3 - Valve Body Temperatures
!
,
s
_ _
. .
.
-4- l i
i DETAILS 1 PLANT STATUS During this inspection period, the plant was shutdow PREVIOUSLY IDENTIFIED ISSUES i
2.1 Followup (92701) j
(Closed) Unresolved Item 458/9304-05: Operability of Reactor Plant Closed Cooling Water Containment Isolation Valve ICCP*M0V138
'
During a previous inspection, inspectors noted that Valve ICCP*MOV138 exhibited a negative load sensitive behavior in that the thrust at control switch trip during the dynamic test of the valve was much higher than the '
thrust at control switch trip under static test conditions. The inspectors also found that the apparent valve factor was 1.8 when the assumed value ,
was :
The licensee retested the valve under identical conditions and obtained the *
same results. The licensee determined that an additional test was warranted using a different sensing unit. The test with the different sensing unit -;
mounted on the inside of the yoke produced results similar to its sister
-
valve. The new results produced a dynamic thrust at control switch trip less !
'
than the static thrust at control switch trip and a valve factor of 0.3 The inspectors were informed that the procedures for testing and analysis would provide for a better evaluation to identify similar results in the future. This item is closed on the basis that the valve was operabl .
2.2 Followup on Corrective Actions for Violations (92702)
(0 pen) Violation 458/9222-01: Fqilure to Initiate Condition Reports for l Inservice Test Failures During a previous NRC inspection, a violation was identified for the failure t to generate condition reports for multiple failures of ASME Section III Code, Class 2, main steam safety relief valve accumulators and their check valve The licensee responded by letter dated October 19, 1992, and provided its corrective action i During this inspection, the inspectors reviewed test data for relief valves that were tested in 1991; however, the surveillance completion forms were not '
completed until August,1992. The inspectors noted that 13 of the 26 relief ;
valves tested had failed their as-found tests. The as-found set pressures of the 13 valves exceeded the allowable ASME Section III set pressure tolerance of 3 percent. The licensee had not initiated any condition reports for ,
these failures even though they were identified after the violation was ;
l
. .
a b-5-identified for th e main steam safety relief valves and their check valve The inspectors considered this to be an additional example of the violation previously cite i Also during this inspection, the inspectors found that the licensee had ,
identified additional examples where surveillance failures were not being identified on condition reports. This was documented in Quality Condition Report 0-93-04-002. The recommended corrective actions for this quality condition report included placing a note in all surveillance test procedures to reinforce and remind the technicians of the requirement to write condition reports when an as-found condition would cause the failure of the surveillanc This violation will remain open pending the licensee's completion of its corrective actions for the violation and the quality condition repor DESIGN CHANGES, MODIFICATIONS, ENGINEERING AND TECHNICAL SUPPORT (37700)
3.1 Modification Requests (MRs)
The inspectors examined eight design modifications (MRs) to verify that the changes were in conformance with the requirements of the Technical Specifications, the FSAR,10 CFR Part 50.59, and the applicable codes and standards.
, The inspectors reviewed the licensee's process associated with plant i modifications. The governing procedure for permanent plant design changes at the River Bend Station was RBNP-010, " Design and Modification Control,"
Revision 4. The licensee established the overall direction for processing and control of design modifications from request through design, review, approval, '
implementation, testing and final closecut in this site directive. The *
licensee also specified the responsibilities of the various groups associated with the design modification In addition to Procedure RBNP-010, the i inspectors reviewed Procedure ENG-3-006, " River Bend Station Design and Modification Request Control Plan," Revision 8. The preparation, review, approval, and revision of design change packages were described in this procedure. The inspectors found the procedures to be comprehensive and well writte . MR 85-1493, Wiring Changes The licensee initiated MR 85-1493 May 29, 1985, in response to a quality control inspection repor The quality control inspection identified wires landed that were supposed to have been spared by a previous modification. The purpose of this MR was to change plant drawings to agree with the as-found -
condition. The licensee initially cancelled this package on July 17, 1986, but approved it for implementation January 19, 199 The inspectors noted that the licensee completed the drawing changes on April 4, 1993. The inspectors considered the time to correct this finding to
__ ._.
.
L-6-have been excessive. The inspectors were concerned that the licensee would elect to implement a drawing change to compensate for a previous failure to j
'
install and inspect a modification. The inspectors were informed that a management decision had been made to correct the deficiency in this manner because of the large backlog of work in the maintenance organizatic The inspectors reviewed the initial safety and environmental evaluation for ;
'
this modification. The inspectors found that the licensee had performed the evaluation in accordance with the procedure ,
3.1.2 MR 86-0088, Revise Setpoint for Reactor Core Injection Cooling High ['
Steam Flow Isolation
,
The licensee initiated MR 86-0038 February 25, 1986, to change the setpoint of the reactor core injection cooling high steam flow isolation as a result of startup test results. The inspectors reviewed the MR and the initial safety and environmental evaluatio No concerns were identified. The inspectors noted that the licensee closed this MR on June 22, 199 ,
3.1.3 MR 90-0064, Mark Number Identification and Drawing Discrepancy .
'
The licensee initiated MR 90-0064 March 29, 1990, to change the mark number of non-safety-related components that were the same as safety-related components '
!
except for a " " or "*" in the number. The inspectors noted that the field work was completed on April 12, 1991; however, the package was open at the end of this inspectio ;
3. MR 92-0040, Motor-operated Valve Circuit Modifications (Hot Shorts) '
The licensee initiated MR 92-0040 March 31, 1992, to address NRC Information i
'
Notice 92-18. The information notice was related to potential hot shorts in the control circuits of motor-operated valves as the result of a fire. The inspectors reviewed the package and the initial safety and environmental evaluation and found that the licensee did a good job on this modificatio !
The modification, which required rewiring some motor-operated valves, was *
fully implemented August 30, 199 .1.5 HR 93-0018, Change Fuse Size from 30A to 35A
.
The licensee initiated MR 93-0018 April 6,1993, to replace the 30A fuses, for the silicon controlled rectifier power controllers in the offgas heater control circuits, with 35A fuses. This modification was initiated because the !
30A fuse blew frequentl The inspectors noted that the initial MR indicated a 35A replacement. This was supported by a facsimile received from the vendor, dated April 6,1993, :
that authorized the 35A size fuse. .The inspectors also found that a final t design, issued April 7,1993, specified a 40A replacement and was also supported by a facsimile, dated April 7,1993, from the same vendor '
representative. The inspectors found that the initial size of 35A was l
,
I
- . - . - -
. . -
-7-selected because the licensee engineers thought that the physical size change of the fuses occurred at the 40A rating. The licensee engineers discovered, however, the physical size change occurred at the 30A rating and, therefore, decided to use the 40A rating size. The licensee made this decision because modifications to the fuse mounting would be required for either size. The inspectors noted that the circuitry and wiring were adequately protected with the 40A fuse installe .1.6 MR 93-0028, Remote Shutdown Switches for Division 1 LSV Air Compressor 4 The licensee initiated MR 93-0028 May 21, 1993, to install a remote shutdown switch and a manual start switch in the circuitry for the pressure vessel ,
leakage control system air compressor, ILSV*C3A. This modification was determined to be necessary to perform a shutdown from outside the control room in the event of a control room evacuation. The inspectors reviewed the package and initial safety and environmental evaluation for this modification and did not identify any concerns. The inspectors noted that the modification was released for field work May 23, 199 ,
3.1.7 MR 89-0240, Remove Dead Legs and Install Cap on Reactor Water Cleanup System :
The licensee initiated MR 89-0240 October 23, 1989, to replace carbon steel piping with stainless steel piping in the reactor water cleanup system. The implemented the modification to reduce the internal piping corrosion buildup in the carbon steel piping and minimize the potential dose rates from the piping. In addition, dead legs on the reactor water cleanup system, along with associated pipe supports and valves, were removed to reduce the dose ;
rates. In-service inspection of welds, as well as snubber inspections, were required in these areas, so substantial man-rem exposure would be received during the inspection if the dead legs were left in place. In addition, permanent decontamination connections with integral drains were added to the piping. The piping was safety-related and designed and procured to the requirements of ASME Section III Class .
The inspectors found that the assertions and assumptions were well documented in the modification and reflected conservative engineering practices. The '
modification field work was completed and signed off in September,1992. The inspectors noted that the package was in design engineering for final review and closeout during the inspectio :
3.1.8 MR 91-0126, Add Isolation Valves on Standby Cooling Tower Return Lines l The licensee initiated MR 91-0126 December 16, 1991, to add manually-operated ;
isolation valves on the standby cooling tower return lines. This' licensee l initiated this modification because, with the new closed-loop service water system, a potential existed to drain 'the service water system when the standby cooling tower return valves, ISWP*MOV55A and B, were opened when normal service water was operating with standby service water in standby mod During performance of Procedures STP-256-3301 and 3302, " Standby Service Water j i
j l
l
- - - - - -
..
l
!
,
l-8-Valve Operability Test," Valves ISWP*MOV55A and B would be stroked. This l would cause normal service water to discharge into the standby cooling tower .
basin at approximately 10,000 gallons per valve cycle. The manual butterfly i valves, ISWP*V3302 and V3303, were added downstream of the motor-operated J valves. During performance of the service water valve operability test, the manual valves would be closed to prevent loss of water inventory during valve testin During normal operation, the manual valves would be locked ope ,
The manual valves installed were 76.2 cm (30 inch) ASME Section III Class 3 butterfly valve .
The inspectors walked down the design modification and observed that the manual valves were installed in accordance with the design package and the i valves were locked open as required. The inspectors reviewed Change Notice j No. 93-0008, January 18, 1993, to STP-256-3301 and noted that Valve ISWP*V3302 was required to be closed before performance of the stroke test for Valve ISWP*MOV55A which was in agreement with the modification package. The inspectors considered the modification package to be thorough and to reflec conservative engineering practice .2 Prompt Modification Reauests (PMRs)
The inspectors reviewed the three PMRs (temporary modifications). The inspectors found the PMRs to have been well prepared and in accordance with procedures. The inspectors had the following comments:
3. PMR 89-0006, Field Change Notice 5, Standby Diesel Air Compressor The licensee determined that a standby diesel air compressor was needed to '
supply the instrument air system because the alternate source of instrument air must be shut down every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for scheduled maintenance. The licensee initiated Field Change Notice 5 to PMR 89-0006 June 2,1990, to install the standby diesel air compresso The inspectors found that the licensee had purchased a diesel air compressor $
to be installed permanently in 1991. The licensee, however, did not complete ,
the design package until 1993 and will not implement the modification until 1994. The inspectors were informed that the lengthy time to develop the package was due to the backlog of engineering work and the need to provide engineering support for the outage .
The inspectors found the temporary modification to have been developed in -
accordance with plant procedures. The inspectors noted that this temporary ;
change had been in effect for three year . PMR 89-0006, Field Change Notice 9, Relief Valve Undersized The licensee initiated Field Change Notice 9 January 24, 1991, to temporary l modification PMR 89-0005 to replace an undersized pressure relief valve, Tag ;
No. WTS-RV128, on the air accumulator tank. The valve did not have adequate pressure relief capacity to accommodate the diesel air compressors. The
.
F
-
.,
P
l
.. .
_g_
pressure relief valve was rated at 3.6 cubic meters per minute (cmm) (127 cfm) j at 930.8 kPa (135 psig) and the output of the diesel air. compressors was .
41.1 cmm (1451 cfm) at 930.8 kPa (135 psig). The valve was replaced with a l 1.27 cm (0.5 inch) inlet valve rated at 45.3 cmm (1600 cfm) at a set pressure of 944.6 kPa (137 psig). The new valve tag number is IAS-RV128. This '
temporary modification was considered Q-Class I .
The inspectors walked down the temporary modification and noted that the.new pressure relief valve was mounted in a horizontal position instead of vertical as recommended by the valve manufacturer. However, the licensee had prepared a field change notice, FCN 12, June 22, 1993, the day before the walkdown to remount the valve in a vertical position. The licensee stated that a >
permanent modification, MR-91-0081, was in design engineering and was 80 percent complete. The permanent modification would install a diesel driven 1 air compressor and an air dryer to replace the temporary modificatio . PMR 93-0003, Annunciator 601-19A-H06 in Alarm Due to High Level in Standby Liquid Control Storage Tank The licensee initiated PMR 93-0003 January 23, 1993, to raise the high level alarm setpoint for the standby liquid control storage tank because operations . i had performed an incorrect valve line-up and filled the tank approximately )
200 gallons. The licensee determined that the tank level could not be lowered by draining without reducing the amount of enriched Boron-10 below the Technical Specification limit The inspectors found that the Technical Specifications did not require the standby liquid control system be operable when the plant was shutdown. The ;
licensee decided that, since the Boron-10 levels were within the Technical Specification limits, the tank level could be lowered by evaporation and still maintain the Boron-10 within specifications. The inspectors reviewed the ,
surveillance results on the chemistry samples for boron in the tank. The
'
inspectors found that the sodium pentaborate concentration and the average weight of the Boron-10 were increasing slowly with the tank level remaining constant over the 4-month period between February and May 199 .3 Engineering and Technical support The inspectors reviewed condition reports, design basis for pressure relief ;
valve sizing, pressure relief valve testing, accumulator sizing, and motor- *
operated valve undervoltage calculations for the quality of engineering's >
support and evaluation of these topics. The inspectors also held discussions '
with many engineers and noted that, while the engineers were knowledgeable in their areas of responsibilities, their knowledge of system interrelations was considered wea . Condition Reports The inspectors reviewed the 23 condition reports listed in Attachment 2 for the quality and timeliness of the engineering responses. The inspectors noted
. -
.
-10-that more than half of the condition reports were not completed within the timeframe assigned. In some instances, the request for extension was not initiated until the day the response was due. Additionally, when extensions were granted, the responses were generally completed after the extended dat The inspectors noted that the condition reports were not well written and did not contain sufficient information for a knowledgeable engineer to review the document and understand the problem, identify the root cause of the problem, and identify what corrective actions were required and taken. On the basis of discussions with responsible engineers, the inspectors found that the licensee corrected the individual problems well. The inspectors considered the quality of the condition reports to have been constant over the timeframe of the documents reviewe . Safety and Relief Valve Sizing and Testing
During the inspection, the inspectors requested that the licensee personnel provide their design basis documents for sizing safety-related pressure relief valves for review. The licensee supplied the General Electric Design Specification No. 22A4622, " Nuclear Boiler System," Revision 7, and its design specification data sheet, No. 22A4622AT, which described the design requirements for the main steam safety relief valves. In addition, the licensee supplied a pressure relief valve purchase specification and technical data sheets prepared by Stone & Webster Engineering Corporation, which listed the required capacities for the ASME Section III valves. The licensee stated that, for the most part, the purchase specifications were considered the design basis. The inspectors reviewed the following calculations which dealt with pressure relief valves:
- 12210-IA-13, " Analysis to Verify Orifice Size and Capacity of Safety and Relief Valves," Revision 2
,
Backpressure," Revision 0
- 12210-PN-062, " Safety Relief Discharge Line Sizing and Bubble Pressure Calculation," Revision 1
- 12210-PH-121, " Main Steam Safety Relief Valves Vent Line Backpressure,"
-
Revision 0 Calculations 12210-PH-ll2, PN-062, and PH-121 determined backpressures in the discharge piping for line sizing but did not address how the capacity requirements were determined for the pressure relief valves. The purpose of Calculation 12210-IA-13 was to verify that the valves selected by the valve vendor met the capacity requirements specified in the valve technical data
, sheets which were part of the valve procurement specification. The valve vendors' sizing calculations were used to verify the valve orifice size.
,
During review of the calculation, the inspectors noted that there were a
- _ _ - . .
.. .
} !
!
-11- ,
,
!
,
number of discrepancies between the capacities determined by the valve vendors, which were listed in the procurement specification data sheets, and i
~
the capacities determined in the calculation. The inspectors did not identify any immediate concerns with the pressure relief valves, however, the licensee's resolution of the discrepancies was identified as Inspection l Followup Item 458/9322-0 .
The inspectors reviewed the licensee's evaluations of NRC information notices :
relating to safety relief valve The inspectors concluded that the :
licensee's evaluation of the notices was deficient. The licensee did not ;
consider any of the notices applicable to the River Bend Station even though some of the valves discussed were made of the same materials as the valves installed in the plan Information Notice 93-02 discussed the temperature effects of a safety valve with a carbon steel body and stainless steel internals, which are the same materials as the River Bend valves. The licensee also did not recognize the importance of knowing the temperature ,
profiles of the valves installed in the plant so that the testing would i simulate the installed configuration. The inspectors reviewed safety relief r valve test results conducted in February 1992. The inspectors noted that the body temperatures measured during the set pressure tests were not maintained at specific temperature. The only criteria for body temperature for thermal equilibrium during the testing was to insure that the body temperature did not change by more than 2.2 C (4*F) in a 15-minute time period. (See Attachment-3 '
for test results and valve body temperatures.)
The licensee reevaluated the response to the notices and concluded that some action was needed to demonstrate that the testing methodology adequately simulated the installed configurations. The licensee had not made any decisions on this by the end of the inspection. The issue of testing safety valves was identified as Inspection followup Item 458/9322-0 During the inspection, the inspectors determined that the licensee had not performed any as-found testing on their main steam safety relief valves to determine if the set pressure was within the Technical Specification tolerance ;
requirement of +0, -2 percent. The licensee stated that River Bend had a total of 95 safety relief valves, 32 of which were originally purchased for River Bend and 63 purchased from another license In addition, the lic.ensee stated that during each outage, the valves that had been in service were removed and store New valves, which had never been in-service, were sent to ,
the valve vendor for refurbishment and testing and then installed at River Bend. The replacement of all safety relief valves without conducting as-found '
testing met the letter of the ASME Code applicable to River Bend. The inspectors found that the licensee intended to perform as-found testing on one previously installed valve which would be installed during Refueling 5, and on 16 previously installed valves for Refueling 6. The review of the as-found test results was identified as Inspection Followup Item 458/9322-0 .
The licensee identified four emergency diesel generator skid mounted lube oil j relief valves were installed in a horizontal position. The relief valves were >
installed on both the suction and discharge sides of the two lube oil pump ,
.
-12-During the inspection, the licensee contacted valve manufacturer (Crosby) was contacted and stated that they did not recommend mounting the valve ir a horizontal position. The manufacturer did indicate that this recommendation was more significant for larger valves. Horizontally mounted Crosby relief valves could be subject to cocking open and have been found to lift prematurely if jolted or jarred. The licensee performed a preliminary seismic analysis to determine if the horizontally mounted valves would open prematurely under a seismic event. The preliminary analysis concluded that the valver would remain closed. The licensee stated that they were going to review their maintenance activities to determine if there had been any maintenance problems with the horizontally mounted valves and they were going to review the horizontal valve applicatio Evaluation of the River Bend actions concerning the horizontally mounted relief valves was identified as Inspection Followup Item 458/9322-0 During a walkdown in the plant, the inspectors noted two ASME Section III pressure relief valves with broken seals and wires. The seals and wires had been installed after testing to insure that the valve set pressure could not be changed and the adjusting ring setting remained in the as tested positio alve IRHS-RV3B was found with the seal broken on the seal wire for the adjusting ring set screw. Valve RV-3A, the emergency diesel generator fuel oil booster pump relief valve, was found with no seals and wires for either the set pressure or ring setting. River Bend Procedure CMP-9166,
" Safety / Relief Valve Testing," Revision 4, required that a new lead seal be installed once the valve successfully completed the test The licensee provided the inspectors Condition Reports CR 89-0314, CR 90-0999, and CR 92-0433, in which the licensee had identified broken seal wires on pressure relief valves. The broken seal wires had been identified while performing surveillance testing or installing the valves. The licensee's corrective action was to inspect the valves for tampering and to replace the wires and seals. The licensee did not take any corrective action to identify the cause of the seals being broken, or to identify this as a problem to the plant personnel to be cognizant, or to examine the other relief valves in the plant. Evaluation of the licensee's actions concerning broken seals and wires on pressure relief valves was identified as Inspection Followup Item 458/9322-0 The inspectors found that relief valves in the nonsafety-related portion of the main steam safety relief air system or steam valve venting (SVV) were providing overpressure protection for safety-related portion of the main steam safety relief valve air system. The normal SVV air system pressure is 4 approximately 965.3 kPa (140 psig). The main steam safety relief valves were Crosby (20.32 cm (8 inch) inlet by 25.4 cm (10 inch) outlet) dual function i spring / pneumatic operated relief valves. The SVV air system is nonsafety-related up to the main steam safety relief valve pneumatic check valves prior to the air accumulators. The latter portion of the system was considered safety-related. A solenoid valve actuates the pneumatic function of the main steam safety relief valves. There was no pressure regulation or relief valve protection in the safety-related portion of the air supply to the main steam
.
,
-13-safety relief valves. The inspectors found that the maximum air pressure was limited to 1392.7 kPa (202 psig) by the relief valves on the SVV air compressors and air dryers. The design pressure for the main steam safety relief valves pneumatic system was stated to be 1723.7 kPa (250 psig).
The safety-related portion of the main steam safety relief valve air system was protected from over pressurization by the nonsafety-related SVV air system relief valves. These relief valves had not been tested nor were they included in the in-service testing program. There have been problems with Target Rock safety relief valves and solenoid leakage. There is the potential for over pressurizing the pneumatic solenoid which could result in inadvertent actuation of the safety relief valves or over pressurization of the safety-related portion of the safety relief valve air system. This issue was identified at the end of the inspection and was being evaluated by the licensee. Review of the licensee's evaluation is identifies as Inspection Followup Item 458/9322-0 .3.3 Accumulator Sizing As part of the design basis review, the inspectors reviewed the accumulator sizing for safety-related equipment. The following documents were reviewed:
- Calculation 12210-PN-255, "HS SRV Pneumatic Piping Design Verification,"
Revision 3
- Calculation 12210-PB-315, " Air Accumulator Tanks Sizing for Category I Air Operated Dampers," Revision 2 ,
- Calculation G13.18.2.0, " Minimum Pressure Required in MSIV Accumulator Tanks," Revision 1 The licensee prepared Calculation 12210-PN-255 to demonstrate that the safety relief valve automatic depressurization system was capable of performing in accordance with Technical Specification requirements and that the automatic depressurization system accumulators were appropriately sized. The licensee performed Calculation 12210-PB-315 to determine the capability of the air accumulators to supply Category I SVV air after a design basis accident and for achieving safe shutdown of the heating ventilation and air conditioning systems in the auxiliary, control, and fuel buildings. The licensee performed Calculation G13.18.2.0 to determine the minimum pressure required in the inboard main steam isolation valve air accumulator tanks to ensure that sufficient closing force was available to close the main steam isolation valves against the peak design basis accident drywell pressure within 3 to 5 seconds. The actuator sizing calculations were well prepared and reflected conservative engineering practice The inspectors had requested accumulator sizing calculations for safety-related air-operated valves. The licensee searched their data base and
. .
l l
'
-14-
!
Jetermined that there were no additional air-operated safety-related valves in the plan ;
3.3.4 Motor-0perated Valve Undervoltage Calculations
'i During an electrical distribution system functional inspection (NRL Irispection Report 50-458/90-200) and an inspection of the licensee's program for footor- ,
operated valves (NRC Inspection Report 50-458/91-24), the licensee's method of ,
determining the voltage available at the motors under degraded voltage conditions was evaluated. During this inspection, the inspectors reviewed the revised methodology for determining the voltage at the motors. The inspectors noted that the licensee was using an industry procedure for these calculation The inspectors also reviewed the licensee's actions taken to address the !
elevated temperature effects on alternating current motors (Inspection '
Followup Item 458/9124-01). The inspectors noted that the licensee had completed two runs of the program, but had not made any conclusions. This ;
item will remain ope .3.5 Engineering Workload / Backlogs i The inspectors found that the licensee was trending engineering workloads on a [
monthly basis. There were 172 open MRs both released to the field for installation and being worked by design engineering. There were another 322 ;
" betterment" MRs for which 145 had been prioritized by system engineerin t There were approximately 36 design engineering complete MRs that had been >
removed from the approved list. The licensee indicated those requests may be deleted altogether. The number of prompt MRs was running around 30 on a monthly basis since the beginning of 1993. There were eight older than 3 years. The average age was indicated to be about 225 days. The current :
management directions and design goals were to implement plant improvements
'
and enhancements for improved plant reliability and increased capacity facto This was indicated to result in additional modifications and engineering work not previously budgeted. The licensee management has indicated that increased contract engineering services were planned to be added to provide increased manpower for the increased work load and for reduction of backlogs. The present unplanned outage was also indicated to have impacted the engineering workload The licensee ha divided their drawing changes into three levels with the changes affecting control room drawings designated as Level 1. Drawing changes to control room drawings are entered within 15 days. There were no delinquent control room drawing changes. There were, however, over 4000 backlogged Level 2 and 3 drawing changes. The present stafting level was estimated to be able to incorporate 250 changes per month. The licensee indicated that additional contract drafting services were to be added to reduce the drawing change backlo i e
i
L
- . l l-15- i
'
Open engineering condition reports per month have shown an increasing trend of from around 40 in July 1992 to 70 in June 1993. Vendor information requests 1 also represented an increase of from around 35 in July 1992 to 50 in June-1993. Both of these increasing trends were also being considered in the plans
-
to increase contract engineering services. The open engineering requests have !
remained at around 12 over the past 12 month ;
i
[
t
'
b
,
!
i P
f
!
r
,
- - - - - - - . - - - - - _ _ _
_ wa,
, '!
ATTACHMENT 1 1 PERSONS CONTACTED 1.1 Licensee Personnel
- Biggs, Supervisor, Quality Control
- J. Blakely, Assistant Plant Manager, System Engineering J. Booker, Manager, Safety Assessment and Quality V--ification
- R. Buell, Supervisor, Nuclear Systems Design .
- J. Burton, Supervisor, Nuclear Safety Evaluation Group
- A. Bysfield, Supervisor, Electrical and Special Projects, Plant Engineering
- T. Crouse, Director, Engineering Support Services
- W. Curran, Site Representative, Cajun Electric
- J. Davis, Senior Quality Assurance Engiu er S. Desai, Principle Engineer
- P. Freehill, Assistant Plant Manager, Outage Management
- K. Gardner, Engineer, Nuclear Licensing
- R. Gaylor, Director, Computer Systems K. Giadrosich, Director, Quality Assurance C. Glass, Technical Specialist
- P. Graham, Vice President, River Bend J. Ham, Senior Mechanical Engineer, Balance of Plant, Plant Engineering
- J. Hamilton, Manager, Engineering T. Hoffman, Supervisor, Civil / Structural Design J. Hurst, Supervisor Engineering Support Training L. Leatherwood, Supervisor, Core Engineering
- J. Leavines, Supervisor, Nuclear Safety Assessment Group
- D. Lorfing, Supervisor, Nuclear Licensing R. Lundholm, Supervisor, Mechanical Process Systems
- J. Mead, Supervisor, Control Systems 1
- D. Melear, Senior Engineer, Nuclear System Design
- C. Mermigas, Section Head, Engineering Administration ,
C. Miller, Supervisor, Maintenance Modification / Construction
- J. Miller, Director, Nuclear Engineering !
- A. Soni, Supervisor, Environmental Qualification
- M. Stein, Director, Plant Engineering '
- J. Thompson, Supervisor, Balance of Plant, Plant Engineering
- R. Vachon, Senior Engineer
- D. Wells, Supervisor, Operational Experience 1.2 NRC Personnel
- D. Loveless, Resident Inspector W. Smith, Senior Resident Inspector In addition to the personnel listed above, the inspectors contacted other personnel during this inspection perio * Denotes pertonnel that attended the exit meetin ___
.. ,
,
l
-2-
2 EXIT MEETING An exit meeting was conducted on June 25, 1993. During this meeting, the inspectors reviewed the scope and findings of the report. Although .
proprietary information was reviewed by the inspectors, no proprietary information is contained in the repor !
,
i s
I
~
I
'
.
ATTACHMENT 2
DOCUMENTS REVIEWED
'
Condition Reports 90-0307 EPA Circuit Breaker C71*5003H Tripped Open :
90-0327 EPA Circuit Breaker C71*5003H Tripped Open
!
90-0383 Turbine Component Cooling Relief Valve Leaking 90-1242 Instrument Air System Compressor Overheated 91-0176 Instrument Air System Compressor Relief Valve 91-0230 Violation of Technical Specification Section 4. Limit Switch Wire Disconnected IC11*ACT0001 92-0037 MSIV Opening Stroke Time 92-0329 Rubber Seat Ring Torn on Valve t
!
92-0572 Surface Indications Found on Feedwater Check Valve Seat 92-0594 ECCS Relief Valves 92-0637 ASCO Solenoid Valves 92-0720 Service Water Check Valve Failed Operability Test 92-0739 Forward Start Valve Stuck Open EGA*SOVYllB ,
93-0022 Rosemount Trip Units Failed E51*N651 and E51*N659 - 93-00S Service Water Sulfuric Acid Pump
,
'
93-0068 ' A' Recirc Pump Suction Valve Drifted Shut B33*MOVF023A 93-0072 Frequent Failures of Instrument Air Vaives IIAS-TRPIA,B,C and IIAS-TRP3A,B,C 93-0020 Low Pressure Core Spray Minimum Flow Valve Suction to Suppression r Pool Failed to Close Against System Pressure E21*MOVF011 93-0159 Borg Warner Check Valve Covers Installed Without Sealant I
Inboard 'C' Main Steam Line Isolation Valve Failed Local Leak Rate
'
93-0233 Test 1B21*A0VF022C l
93-0220 Low Pressure Core Spray Minimum Flow Valve to Suppression Pool l l
)
l
-
l
,
- a l
F
<
f-2-
.
93-0245 Inboard Main Steam Isolation Valves Exhibited High Leak Rates l
!
IB21*A0VF022A and IB21*A0VF022D Test Reports !
!
Crosby Test Report Valve Serial No. N63800-00-0039 ;
Crosby Test Report Valve Serial No. N63800-00-0040 Crosby Test Report Valve Serial No. N63800-00-0047 Crosby Test Report Valve Serial No. N63800-00-0081
,
Crosby Test Report Valve Serial No. N63800-00-0095 ;
Crosby Test Report Valve Serial No. N63800-00-0097 Crosby Test Report Valve Serial No. N63800-00-0098
i Crosby Test Report Valve Serial No. N63800-00-0100 'l Crosby Test Report Valve Serial No. N63800-00-0106 ,
Crosby Test Report Valve Serial No. N63800-00-0107 I
Crosby Test Report Valve Serial No. N63800-00-0109 Crosby Test Report Valve Serial No. N63800-00-Olll Crosby Test Report Valve Serial No. N63800-00-Oll2 Crosby Test Report Valve Serial No. N63800-00-0120 [
Crosby Test Report Valve Serial No. N63800-00-0121 ;
Crosby Test Report Valve Serial No. N63800-00-0123
,
i l
I
_ ,
.
ATTACHMENT 3 VALVE BODY TEMPERATURES Serial No. N63800-00-0039 (Setpoint 8032.4 kPa (1165 psig))
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE February 5, 107.200 8025.5 kPa February 29, 118.9 C 7991.0 kPa 1992 (225 F) (1164 psig) 1992 (246 F) (1159 psig)
107.2 C 8025.5 kPa 121. loc 7984.1 kPa (225 F) (1164 psig) (250oF) (1158 psig)
108.3oC 8025.5 kPa 121. loc 7977.2 kPa (227oF) (1164 psig) (250 F) (1157 psig)
110.0 C 8025.5 kPa 122.8 C 8011.7 kPa (230 F) (1164 psig) (253oF) (1162 psig)
Serial No. N63800-00-0040 (Setpoint 8032.4 kPa (1165 psig))
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE February 12, 118.3oC 7887.6 kPa February 26, 100.0 C 7963.4 kPa 1992 (245 F) (1144 psig) 1992 (212 F) (1155 psig)
120.0oC 7901.4 kPa 102.2 C 7894.5 kPa (248 F) (1146 psig) (216 F) (1145 psig)
120.0 C 7901.4 kPa 104.4oC 7922.1 kPa (248oF) (1146 psig) (220oF) (1149 psig)
121. loc 7935.9 kPa 104.4oC 7922.1 kPa (250oF) (1151 psig) (220oF) (1149 psig)
Serial No. N63800-00-0047 (Setpoint 8204.8 kPa (1190 psig))
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE February 6, 89.4 C 8177.2 kPa March 4, 121.1 C 8094.4 kPa 1992 (193 F) (1186 psig) 1992 (250 F) (1174 psig)
90.6oC 8142.7 kPa 122.20C 8135.8 kPa (195oF) (1181 psig) (252 F) (1180 psig)
-
87.8oC 8163.4 kPa 124.4 C 8128.9 kPa (190oF) (1184 psig) (256 F) (1179 psig)
87.8 C 8094.4 kPa 126.7 C 8122.0 kPa (190oF) (1174 psig) (260oF) (1178 psig)
_ _ _ .
t I
}
,
-2-
,
!
Serial No. N63800-00-0081 (Setpoint 8135.8 kPa (1180))
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE I February 7, 111.7oC 8073.8 kPa March 7, 97.8oC 7997.9 kPa 1992 (233of) (1171 psig) 1992 (2080F) (116e sig)
113.9 C 8060.0 kPa 99.40C 8018.6 kPa (237 F) (1169 psig) (211 F) (1163 psig) i 115.6 C 8115.1 kPa 101.7 C 8053.1 kPa (240oF) (1177 psig) (215 F) (1168 psig)
115.60C 8060.0 kPa 103.9 C 8060.0 kPa 1 (240oF) (1169 psig) (219 F) (1169 psig) i Serial No. N63800-00-0095 (Setpoint 8135.8 kPa (1180))
DATE OF TEST B0DY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE February 11, 110.0 C 8066.9 kPa February 21, 126.l*C 7970.3 kPa ,
1992 (2300F) (1170 psig) 1992 (2590F) (1156 psig) l 111.1 C 8080.6 kPa 126. loc 8046.2 kPa b (232 F) (1172 psig) (259 F) (1167 psig) !
111. loc 8101.3 kPa 127.2 C 8025.5 kPa (232oF) (1175 psig) (261 F) (1164 psig)
112.8oC 8128.9 kPa 126.7 C 8018.6 kPa !
(1163 psig)
'
(235of) (1179 psig) (260oF)
Serial No. N63800-00-0097 (Setpoint 8135.8 kPa (1180)) ;
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE l
. ,
'
February 15, ll2.8oC 8039.3 kPa March 16, 112.8 C 8025.5 kPa !
,
1992 (235oF) (1166 psig) 1992 (235oF) (1164 psig)
112.8 C 8046.2 kPa 113.9 C 8011.7 kPa '
(235oF) (1167 psig) (237 F) (1162 psig)
'
114.4oC 8060.0 kPa 115.6 C 8101.3 kPa (238 F) (1169 psig) (240oF) (1175 psig) j
'
ll5.0oC 8135.8 kPa ll5.6oC 8053.1 kPa !
(239of) (1180 psig) (240oF) (1168 psig)
'
,
l
,
, - ,
'
.. .
,
~3-
!
,,,,,.<.c Serial No. N63800-00-0098 (Setpoint 8135.8 kPa giltb);
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE
'
February 14, 116.1 C 8101.3 kPa March 6, 107.2 C 8025.5 kPa 1992 (241*F) (1175 psig) 1992 (225 F) (1164 psig) ,
117.2 C 8108.2 kPa 105.6oC 8066.9 kPa (243 F) (1176 psig) (222oF) (1170 psig) ;
ll8.3oC 8108.2 kPa 107.2aC 8025.5 kPa i (2450F) (1176 psig) (225 F) (1164 psig) i
!
119.4 C 8101.3 kPa 109.40C 8122.0 kPa (247 F) (1175 psig) (229 F) (1178 psig)
,
Serial No. N63800-00-0100 (Setpoint 8135.8 kPa (1180)) '
l DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE
l February 7, 114.4 C 7970.3 kPa 1992 (238oF) (1156 psig) ,
115.0 C 7977.2 kPa (239 F) (1157 psig)
24bF 165p g)
(2iF 6bp g) j
"
i Serial No. N63800-00-0106 (Setpoint 8032.4 kPa (1165 psig))
'
i DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE I
"
February 13, 97.8 C 7970.3 kPa 1992 (208oF) (1156 psig)
97.8oC 7963.4 kPa (208 F) (1155 psig)
98.9oC 7956.5 kPa (210oF) (1154 psig)
98.9 C 7929.0 kPa (210*F) (1150 psig)
- . __
- . _ _ . ._
. - - . .
.
-4-i Serial No. N63800-00-0107 (Setpoint 8032.4 kPa (1165 psig))
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE :
February 12, 115.6oC 7935.9 kPa February 25, 121.1 C 7894.5 kPs 1992 (240oF) (1151 psig) 1992 (250oF) (1145 psig) ;
115.0 C 7949.6 kPa 120.0 C 7915.2 kPa !
(239 F) (1153 psig) (248 F) (1148 psig) j 115.6oC 7997.9 kPa 121.7 C 7991.0 kPa ,
(240*F) (1160 psig) (251oF) (1159 psig)
117.20C 7997.9 kPa 123.3oC 7949.6 kPa !
- (243 F) (1160 psig) (254oF) (1153 psig) ,
'
_
!
Serial No. NS3800-00-0109 (Setpoint 8032.4 kPa (1165 psig))
{
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT l TEMP PRESSURE TEMP PRESSURE :
February 10, 126.1 C 8018.6 kPa '
1992 (?59 F) (1163 psig)
128.9 C 7915.2 kPa (264 F) (1148 psig)
131. loc 7963.4 kPa (268oF) (1155 psig) i 131.1 0 7956.5 kPa .
(268oF) (1154 psig) l
!
Serial No. N63800-00-Olli (Setpoint 8032.4 kPa (1165 psig)) !
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE , TEMP PRESSURE ;
February 4, 90.6 C' 7997.9 kPa March 2, 112.8oC 7963.4 kPa f 1992 (1950F) (1160 psig) 1992 (235 F) (1155 psig) ,
93.3oC 7977.2 kPa 113.3 C 7984.1 kPa
'
(200oF) (1157 psig) (2360F) (1158 psig)
96. loc 7956.5 kPa 115.0oC 7991.0 kPa i (205 F) (1154 psig) (239 F) (1159 psig)
96.1 C 7956.5 kPa 117.2 C 7977.2 kPa
'
(205 F) (1154 psig) (243 F) (1157 psig) j
l l
P
. -
.
"
-- . _
.
.
-5-
Serial No. N63800-00-Oll2 (Setpoint 8032.4 kPa (1165 psig))
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP PRESSURE February 19, 123.9oC 7970.3 kPa March 13, 120.0oC 8004.8 kPa 1992 (255 F) (1156 psig) 1992 (248oF) (1161 psig) ;
126.1 C 8004.8 kPa 120.6 C 8004.8 kPa (259*F) (1161 psig) (2490F) (1161 psig) !
126.7 C 7963.4 kPa 121.1 C 8032.4 kPa (260oF) (1155 psig) (250oF) (1165 psig) :
127.2 C 8004.8 kPa 121.1 C 7991.0 kPa "
(261oF) (1161 psig) (250oF) (1159 psig) l
.
Serial No. N63800-00-0120 (Setpoint 8204.8 kPa (1190 psig)) ,
i DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT [
TEMP PRESSURE TEMP PRESSURE February 11, 125.0oC 8108.2 kPa February 22, 107.80C 8080.6 kPa ;
1992 (257 F) (1176 psig) 1992 (226oF) (1172 psig) l
'
1125.6C 8108.2 kPa 110.0oC 8135.8 kPa (258 F) (1176 psig) (230 F) (1180 psig) i I
127.8 C 8184.1 kPa 108.96C 8108.2 kPa (2620F) (1187 psig) (228oF) (1176 psig)
129.4 C 8135.8 kPa 111.10C 8177.2 kPa l
_
(265 F) (1180 psig) (232 F) (1186 psig) :
!
Serial No. N63800-00-0121 (Setpoint 8204.8 kPa (1190 psig))
.
DATE OF TEST BODY LIFT DATE OF TEST B0DY LIFT TEMP PRESSURE TEMP PRESSURE l February 14, 126.1 C 8156.5 kPa i 1992 (259 F) (1183 psig) l b ipsg) _
127.8oC 8122.0 kPa (262oF) (1178 psig)
128.9oC 8135.8 kPa (264 F) (1180 psig)
l
_ . . _ _ _ _ _ _
.
l-6-Serial No. N63800-00-0123 (Setpoint 8204.8 kPa (1190 psig)) !
DATE OF TEST BODY LIFT DATE OF TEST BODY LIFT TEMP PRESSURE TEMP
'
PRESSURE l
February 28, 97.2oC 8135.8 kPa 1992 (207 F) (1180 psig) ;
96.1 C 8191.0 kPa l (205 F) (1188 psig) ;
97.2 C 8073.8 kPa (207oF) (1171 psig) ,
99.4*C 8122.0 kPa (211 F) (1178 psig)
,
t
-
'.
T t
?
'
,
.
I h
?
.
-
l
- . . _ . -. __ ___ _ ________ - . _ . _ _ _ _ . _ _ _ _ _ _ _