ML20058A264
| ML20058A264 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 08/17/1990 |
| From: | Kovach T COMMONWEALTH EDISON CO. |
| To: | Davis A NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| ID191, NUDOCS 9010250362 | |
| Download: ML20058A264 (24) | |
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August 17,1990
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J Mr. A. Bert Davis l
Regional Administrator i
U.S. Nuclear Regulatory Commission '
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'799 Roosevelt Road r
Glen Ellyn,IL 60137 4
Subject:
Quad Cities Stations Units 1 and 2-Response to Notice of Violation i
Contained in Inspection Report 50-254/89024 and 50-265/89024 i
NRC Docket Nos. 50-254 andEI).265_
Reference:
H.J. Miller to Cordell Reed letter dated June 15,1990.
Mr. Davis:
The referenced letter transmitted Inspection Report 50-254(265)/89024.
which contained two (2) Notices of Violation. Attachment A to this letter,provides 4
our response to the violations. Commonwealth Edison understands the significance of the NRC's concern regarding the performance of containment isolation valves and a
penetrations, and the need to implement timely and effective corrective actions.
n In addition to the Notices of Violation, the referenced letter requested that Commonwealth Edison provide a res >onse to three (3) Unresolved Issues, a
Attachment B provides the requestec responses. Finally, Attachment C contams
..information to assist in your staffs review of the responses.
Conunonwealth' Edison would like to correct a statement contained in i
Notice ofViolation 50-254(265)/89024-04b. In the Notice of Violation, the inspector indicated that one of the O-rings (with the new material) failed during the-exceeded leakage limits durm, cycle. Of the three Unit 2 feedwater check valves 1989-1990 Unit 2 operating g testing, only valve 2-220-62B contained the new
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Parker 0-ring material. - An inspection of valve 2-220-62B, which was performed to identify the cause of the leaka e, revealed that the 0-ring was intact and did not.
appear to have deteriorated, l(he cause of the failure is believed to be inadequate l
. seating of the disc during testing and not failure of the 0-ring.-
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,sv Mr. A. Bert Davis ' -
August 17,1990 An extension to the response due date was requested on July 5,1990 and granted by G. Wright of the Region III staff.
If there are any questions or comments to this response, please direct them -
to R, Stols at 708/515-7283.
l Very truly yours, mr T.J. Kovach Nuclear Licensing Manager cc:
L. Olshan, Project Manager - NRR J. Hind, Section Chief-Region III F. Maura, Regional Inspector T. Taylor, Senior Resident Inspector l
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- ATTACHMENT A t
RESPONSE TO NOTICE OF VIOLATION.
50-254(265)/89024-02 1
50-254(265)/89024-04 t
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_j BESPONSE TO NOTICE OFVIOIATION uu 50-25426DB9024-02 l
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10 CFR Part 50, Appendix B, Criterion XVI requires in part, that in the case of significant conditions ac verse to quality measures s hall be taken to assure that the' cause of the condition is determined and corrective action taken to preclude j'
repetition. 10 CFR Part 50,' Appendix B, Criterion II requires, in part, that the licensee implement a quality assurarre program through plant life. For the Quad Cities site, this program'is specified : the Commonwealth Edison Company Quality Assurance Program T ical Report CE-1-A, Revi. ion 57. Section 2.2 of the Topical Report states that the uad Cities Nuclear Station commits to comply with
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Regulatory Guide 1.33 ated November 1972. This regulatory guide requires,in part, that the licensee comply with ANSI Standard 18.7-1972.- Paragraph 6.4 of this ANSI standard requires that a surveillance test program be described to ensure that
. safety related structures, eystems, and components will operate to keep parameters.
1 withm normal bounds, or act to place the plant in a safe condition. It further.
requires that the testing frequency be established as prescribed by paragraph 5.1.7.3 1
The latter prescribes a test frequency that is related to the results of rehability analyses, frequency and type of service, or age of system. Paragraph 5.1.6.1 requires that a maintenance system be develo ad to maintain safety-related equipment at x
quality required to perform its intenc ed function; that experience with failed.
-I eqttipment be reviewed to determine whether a replacement of the same type can be octed to perform its Lnetion reliably;by appropriate performance testing, and that a suitable level of confidence in 1 ex1,hsystems or components be attained suc Contrary to the above, the licensee's corrective measures have not been Ladequate to assure that' the causes of excessively leaking containment isolation valves and penetrations were determined and that corrective action was taken to preclude repetition. As a consequence, the combined leakage for valves and penetrations subject to Type ld and C tests have exceeded the 10 CFR 50, Appendix J y
Criterion of 0.6 La for every refueling outage of Unit 1 since 1979, and for Unit 2 since 1974. In addition, the licensee has failed to adjust its surveillance test and.
preventive maintenance pro trams to reflect the unreliability of these containment
' isolation valves to perform t teir safety function throughout an entire fuel cycle.
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~Inituduction Commonwealth Edison recognizes that valve performance of several I
containment pathways during the conduct of the local leak rate tests (LLRT) has repetitively exceeded acceptance criteria. Commonwealth Edison has implemented actions to correct those specific deficiencies and acknowledges that, in some cases, it has taken longer than expected to achieve resolution of certain valve deficiencies.
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, y, Since 1986, Commonwealth Edison has demonstrated a more aggressive approach to the resolution of these issues. Quad Cities Station has replaced various 4
va. ves that have a:cceded acceptance criteria values, improved maintenance and testing technique., and developed modifications to resolve these issues. For example, prior to 1987, the Umt 1 Main Steam Isolation Valves were experiencing a L.
failure rate of a pproximately 65% The failure rate on Unit I has been reduced to approximately 10E Also improved leakage performance has been achieved for the drywell head gasket which had exceeded leakage requirements from 1979 to 1986.
1 We believe that with further enhancements to our test program and maintenance i
techniques, the failure rate of containment isolation valves and penetrations will continue to decrease.
In addition to the actions implemented on specific valves at Quad Cities, i
Commonwealth Edison has developed a corporate program to improve the containment testing program at each of our nuclear stations. An engineer that is i
assigned to the Corporate Engineering Department has been working with each of our nuclear stations to ensure that tests are performed consistently and to ensure lessons learned are effectively communicated to each station. Detailed l
self assessments are being performed to assure that the station program is effectively implemented. Tie engineer has been working closely with the Region III Staff to ensure effective enhancements are developed.
In this response, Commonwealth Edison has addressed the overall corrective actions to monitor the performance of all containment valves and penetrations. This should improve the performance of the as found Integrated Leak Rate Testing program. In addition, Commonwealth Edison has specifically addressed the corrective action which will be im piemented for the containment isolation valves and penetration specifically addressed in the NRC Inspection Report.
i Quad Cities Imcal Leak Rate Testing Progre.a Quad Cities Station recognizes that the actions impleraented to resolve the specific containment isolation valve or penetration deficiencies,in the past, focussed on the individual valves and may not have fully addressed the performauco of the as found ILRT. To address that weakness, Quad Cities Station has developed the
,r following corrective actions:
1.
A comprehensive history of the performance of all Type B and C tested containment isolation valves and penetrations will be developed and maintained. This history willinclude the leakage results and any corrective actions taken. Utilizing this history, predictive / preventative maintenance programs will be developed, as appropriate. This will be completed by June,1991. This schedule for development will ensure that the program will be available for the Unit 2 Fall,1991 Refueling Outage.
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A post outape review will be conducted on the results of the Local Leak Rate Test.
Fhe review participants will include members of Co. porate i
Engineering, Quality Programs and Station personnel. The purpose of this review is to assess the results of the local leak rate tests, compare the results,to previous tests and develop an action plan for the following test campaign.
1 We believe that these actions will be effective and willimprove as found ILRT results. The comprehensive review of the as found LLRT results will ensure that effective corrective actions are developed for valve failures and that good valve
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performance is maintained. This review will also ensure the effectiveness of the i
preventative / predictive maintenance program.
NRC IDENTIFIED CONTAINMENTVALVE8 In the inspection Re > ort, the NRC specifically discussed the following eight (8) l containment isolation pathways in support of the Notice of Violation:
a) Main Steam Isolation Valves b) Feedwater Check Valves c)
Main Steam Line Drain Valves d) Cleanu i Suction Valves e) Drywel Purge f)
Drywell Exhaust i
g) Drywell Floor Drain Sump Valves h) HPCI Steam Exhaust Check Valves The following discussion arovides a brief background and planned corrective e
action for each of the identificc valves / penetrations.
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L MAINRIEMLISOMTION_YALYES
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The recent test performance on the MSIVs for both units has significantly improved when compared to previous years. During the period of19791986, the l
Unit 1 valves failed twenty six (26) times during the conduct of forty (40) tests.
Similarly, during the period of 1974 to 1985, the Unit 2 valves failed twenty (20) times during the conduct of fifty six (56) tests. Significant improvements have been achieved with only two valve failures on Unit 1 (smee 1987) and one valve failure on Unit 2 (since 1986); The total number of tests conducted during this priod for both units is thirty (30). The improvements in the valve performance can >e attributed to better maintenance practices and testing techniques.
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i In 1986, the Mechanical Maintenance Department acquired a special lapping 1
tool which provides a more uniform seating surface to allow for better leak tightness. Since the use of the lapping too., MSIV valve performance has improved.
A so the MSIV air operators are rebuilt every other Refueling Outage to maintain good performance of the closing motive force.
1 Improvements to the testing technique have also been developed. To accommodate testin g, three of the MSIVs are closed with steam flow. Also the MSIVs are tested while the steam lines are still warm. These steps are believed to improve the LLRT results by simulating the thermal and dynamic conditions that would exist when the valves would be required to perform their safety function.
CometiveActiestto "rment_Eutursfaihms The MSIV performance will be monitored through testing; however, due to the restrictive acceptance criteria for valve performance (11.5 SCFH), it is difficult to aredict valve failures. A predictive and/or preventative maintenance program will
>e developed for MSIV components to minimize future valve failures.
EEEDWNI11R.CHECILVEYES
Background
The Feedwater Check Valves have a history of exceeding the acceptance criteria during the LLRT Various modifications, use of more resilient 0 ring materials, improved maintenance techniques and tests to determine valve failure mechanisms have been applied.
valves on the A loop (1-220-58A,pplied to the Unit 1 inboard and outboard che The corrective measures a 1220-62A) have been effective as demonstrated by successful test results since 1986, While the same maintenance techniques were applied to the inboard and outboard check valves on the B loop (1-220 58B and 62B),
these valves continue to exceed the acceptance criteria. As a result, the B loop valves were extensively refurbished durmg the 1989 Refueling Outage using techniques which have been effective at Dresden and LaSalle Stations. The valve improvements included:
- Hinge pins and bushings were replaced with a more wear resistant material.
- Seat rings were trued to minimize distortion and interference with the valve bore.
- The seats were ground and the seat and disc were bored for alignment.
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y The effectiveness of these corrective actions will be determined during the upcoming I
Fall,1990 Unit I test campaign.
i During the Unit 21990 LLRT, three out of the four feedwater check valves exceeded leakage requirements. The valve improvements discussed above were applied. Also, the laLeral clearance between the disc and seat ring of the valves was nunimized which will further improve the seating ability of the valve disc. Te ting of the valve has demonstrated excellent leak tightness after repair. Shop tests of the refurbished trim were performed and demonstrated the disc's ability to close when the dise was pulled to either side along the axis of the hinge pins, which appeared to be the failure mechanism of the valve during testing. The ef Fectiveness of these actions will be determined during the next Unit 2 Refueling Outage.
Finally, the current test configuration does not duplicate the conditions which the valves would experience under accident conditions. In order to test the dves, the feedwater lines are drained of water through both check valves to a drain line inside containment. This test method requires the valves to be lifted off their seat and sufficient external force is not provided to resent the valves. The volume is pressurized using air; however, due to the large volume, it is postulated that the slow pressurization does not assi8t in seating the valve. During the test, the valves are, therefore, not properly seated. Under accident conditions, the valves are closed and seated by reactor pressure.
CorrectivitActions.io_PrnmsmLEuturstEmihues 1.
Additional drain lines on the Feedwater check valve volume will be installed to allow the drainin g of the volume without opening the check valves. This test configuration will improve the test methodology and potentially the valve merformance during the LLRT. The drain lines will be installed to test the feedwater check valves during the Fall,1990 Unit 2 Refueling Outage.
2.
An investigation of feedwater check valve maintenance techniques utilized by L
other utilities will be conducted to identify if any additional maintenance i
repair techniques have been effective, In the event that the most recent maintenance repairs are not effective lon 3.
Commonwealth Edison will notify the Commission of our corrective act plans for these valves within ninety days after startup of Unit 1 following the fall,1990 Refueling Outage. The Corporate Engineering Department is l
currently reviewing the potential alternatives to resolve the concerns.
MADUlTEAKLINE.DRMNJALYES i
Backsnmad Of the two units, the Main Steam Line Drain Valves on Unit I have experienced the greater number of repetitive failures. Due to the unacceptable valve performance in 1987, modification M-4-1-85-48 was installed. This modification replaced the inboard and outboard drain valves which were 900 pound class Crane gate valves with new 1500 pound class Anchor Darling split wedge gate valves. The l
valves successfully passed the LLRT in 1989.
The Unit 2 Main Steam Line Drain Valves exceeded leakage criteria during the 1989 LLRT. The original Crane gate valve was replaced with the new Anchor Darling valve. Since that replacement, the valve successfully passed the LLRT in 1990.
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CorrectiveActiona.tefamentfuturelaihme l.
A predictive maintenance program will be developed based upon the valve performance in future testing.
REACTOR WATER _ CLEAN-UP SYETEKIRW_CU)
SUCHON_YALYES l
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Background
During the period from 1976 to 1986, the Unit 1 RWCU suction valves (1-12012 and 1 1201-5) successfully passed the as found LLRT. As a result of the failures, the valves were disassembled to investigate the failures. The investigation revealed cracking of the stellite material on the disc. The stellite material was removed and new stellite material was fie.iJ welded on the valve disc.
The valves again exceeded leakage requirements during the Fall,1989 Unit I s
Refueling Outage. The cause of the failure was again due to cracking in the stellite material. Outboard valve 1-1201-5 was replaced with a new Anchor-Darling split wedge gate valve that has been used successfully as a replacement for the Mam Steam Line Drain isolation valve. This replacement is expected to result in an 4
acceptable minimum pathway leakage rate for this penetration.
The corresponding valves on Unit 2 have successfully passed the as found LLRT for the past nine operating cycles. Due to the experience of stellite material cracking on Unit 1, the 2-1201-5 valve was replaced during the Spring,1990 Unit 2 Refueling Outage with the Anchor Darling split wedge valve.
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assure its effectiveness. placement valve will be tested following one cycle of use to The 1(2)-1201-5 re The valve leakage will be trended to identify the need for additional maintenance activities.
Generic Letter 8910 and I.E. Information 90-04 identified a potential problem with RWCU containment isolation suction valves. The NRC sponsored, full flow MOV testing results raised concerns about the ability of these valves to close under accident conditions. It is Commonwealth Edison's understanding that the Commission plans to issue further correspondence concerning these and other containment isolation valves.
Conunonwealth Edison has evaluated the available information on this issue and has determined that the motor operator on valve 1(2)-12012 is potentially undersized and would require replacement.~ The larger motor operator would require a valve replacement. Actions are being taken to size and procure f
replacement valves. Due to the lead time, this valve will not be available during the upcoming Unit 1 outage.
The station had contemplated the replacement of the 1-12012 RWCU suction isolation valve during the upcoming outage. The 1(2)-1201-2 valve is located in an area of high radiation dose and, therefore,in light of required resolution of the NRC's MOV concerns, the valve replacement was delayed so that the NRC MOV concerns and valve performance could be addressed simultaneously. The valve will be replaced in accordance with the schedule dictated by the Commission or the Spring,1992 Refueling Outage, whichever is shorter, DRYWEMIDRUSEURGilLSYRIEM
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The Drywellfrorus Purge System consists of the three subsystems; Purge supplyilfrorus Purge Spurge exhaust, and 'l orus to Reactor Building vacuum breakers. The
' Pratt butterfly valves, ystem is isolated from primary containment by eight (8) lar ge Drywe two (2)large check valves amf four (4) small gate or butterfly
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valves (Reference attached diagrams). The unacceptable LLRT performance of this system is typically the result ofleakage from the large Pratt butterfly valves. The valve's rubber seatin g material, which has a limited useful life cannot be readily removed from the va ve body, and therefore, when the valves fail, the entire valve is replaced.
PurgeEupply. Subsystem The Unit 1 purge supply exceeded the as found LLRT acceptance criteria valve.g the 1987 Refueling Outage due to leakage through the 1-160122 butterfly durin l
The 18" Pratt valve was replaced during the 1987 Outage and the volume passed the as found LLRT during the test campaign conducted in the Fall,1989 3efueling Outage.
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9 During the 1990 Refueling Outage, the Unit 2 volume exceeded the as found LLRT acceptance criteria. The investigation of the failure revealed that valves 2-160121 and 2-160156 leaked and the valves were subsequently replaced. The 21601-22 valve was visually inspected using a soap bubble solution which revealed no visible seat leakage. The 2-1601-22 valve was previously replaced in 1986.
PurseExhausuhaberstess During the 1989 test campaign, the Unit 1 purge exhaust volume exceeded the as found LLRT acceptance critena. Investigation of the failure revealed leakage through valves 1 160124 and 1-1601-60 which are 18" butterfly valves. As a result, the valves were replaced.
The three Unit 218" butterfly valves were re placed in 1985. Since the replacement, the volume has successfully passed the LLRT for three consecutive operating cycles.
l Torne-to_ReactorEnildingXocuanLBreakers Performance of the as found LLRT on the Unit I vacuum breakers has been good with no failures since ti, Spring of 1979.
The Unit 2 vacuum breaker Fnes have had good past performance with successful as found LLRT results from 1978 to 1990. In 1990, seat leakage on valve 2-1601-20B (20" butterfly valve) resm'ed in exceeding the acceptable linuts on one line. A yacking leak on the 20" check ulve caused the other line to exceed leakage limits, borrective measures included th 9 replacement of valve 2-1601-20B and l
packing repairs on the check valve.
l CorrectiveActices_toImma Juturefailtrwe Based on the performance of these valves in the past, it is clear that the valves have ourall demonstrated good performance and suggests that a valve replacement program be implemented on a periodic basis to prevent failures. Consequently, a preventative maintenance program has been developed to replace these valves as follows:
1 1.
R( placement of valves which have not been previously changed out will be co n aleted during the Fall,1990 Refueling Outage on Unit 1 and the Fall, 199:. Refueling Outage on Unit 2.
2.
A preventative maintenance program has been develo ad and requires l
that the purge butterfly valves be replaced following taree operatmg cycles. Phase 1 of this program consists of replacing one half of the valves during the Unit 1 Spring,2992 Refueling Outage and Unit 2 Fall,1991 L
Refueling Outage. The preventative maintenance " time clock" will commence following the replacement..
DRYWELLILOOR.DEARLSUMPlSOMTION.YALVE L
acceptance criteria $r the first time in 19b. Isolation valves exceeded LLRTAt th The Unit 1 D cll Floor Drain Sum these valves revealed a build up of crud on the valve internals, due to particulates in the water contained in the sump. The valves were cleaned, repaired, and the l
orientation of the valves was modified to obtain better seating. The valves, however, continued to exceed leakage limits during the subsequent two test cycles.
t In 1988, the Unit 2 Drywell Floor Drain Sump Isolation valves exceeded the leakage allowance due to build up of crud. Corporate Engineering performed a study in an effort to improve valve performance. The study was commleted in 1989 and recommended several different types of valves that are more suitable for this application.
l Cet79C%iMLAC%ichtL10In 1 w.Afutut1LEtih4RS Modification M 4-1(2) 89 52 has been developed to replace the gate valves with plu g valves which have been successfully utilized m this application at Braidwood -
and Byron Stations. The modification will be installed durmg the Unit 1 Fall,1990 and the Unit 2 Fall,1991 Refueling Outages. The effectiveness of this modification will be monitored through testing after the operating cycle.
HIGRERESSURE 0QOIANT. INJECTION.SYHTEM SIZAR.EKHAUSr CHECKYALYE MWi-a In 1984, Corporate Engineering performed a study to investigate the cause of the unacceptable leakage of the Unit 1 HPCI steam exhaust check valve failures.
The study concluded that the valve seating material (Buna-N rubber) was being cut by steam which caused the valve to exceet leakage requirements. To correct the deficiencies, Modification M-41-85 68 was implemented in March,1986. This modification replaced the existing valve with a C&S valve. The new valve was believed to be more resilient, in that the Nordel elastomer seat was more resistant to high temperatures and low pressure steam. The valve exceeded the LLRT acceptance criteria in the subsequent cycle.
In 1987, the C&S valve was replaced with a Marlin check valve, The Marlin valve was identical to the C&S valve; however, the Marlin seat was inlaid to prevent a
steam from directly impacting the seating surface. This valve exceeded the LLRT acceptance criteria in t be subsequent cycle and was replaced with an identical valve in 1989.
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Unlike +.he Unit 1 HPCI steam exhaust check valve, better LLRT performance of the Unit 2 valve has been achieved. In 1985, the Unit 2 HPCI steam exhaust check valve was replaced with a Mission check valve, rebuilt by Clow Corporation.
The rebuilt valve included stainless steel springs and bushings which was intended to reduce the wear of the components. The valve successfully passed the LLRT in 1986 and 1988, however, it exceeded leakage requirements during the 1990 campaign. An inspection of the valve was performed in 1990 and revealed that the damage to the elastomer materialin the valve seat was present. The seating j
material appeared to have been cut by steam flow through the valve. The hmge i
pins, bushmgs springs and disc were in acceptable condition, having only slight 1
wear. The valve was replaced with a Marlin check valve similar to that installed on Unit 1.
ShortTerm..CorrectimActions To address the cause of the failure based on the investigation results of the last Unit I failure, the seating material for the check valve will be replaced with a more suitable material based on recent vendor and test data. The material change out i
will be accomplished dming the upcoming Unit 1 Refueling Outage (November, 1990).
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Corporate En gineering and the Station are currently reviewing the failures of the HPCI check valve. The following actions are being pursued:
1.
HPCI Surveillance Testin The Station is conducting diagnostic testing of the valve during the HPC.g: testing. General Electric (the manufacturer of I
the turbine) has indicated that the low speed operation of the turbine may have a detrimental effect on the valve and piping. Quad Cities has conducted testing which verified the chattering of the valve during the low speed operation of the turbine, Turbine vibration data will be collected to assure that no detrimental effects to the turbine are created with the deletion of the 2000 RPM pause during startup. If appropriate, surveillance procedures will be revisec to restrict the time period for the operation of the turbine at low speeds. This will be comple'ed by October 1,1990.
2.
Turbine exhaus+ steam pressure may not be sufficient to force water from the wrtion of the exhaust pi ping which terminates below the water line in the corus. This water quenches the steam inside the line creatin g a vacuum at a rate fast enough to cause the water in the exhaust d owncomer i
to rise. This reverse flow causes the check valve to slam shut before the vacuum breakers allow sufficient air to enter the downstream side of the valve. Diagnostic testing, described in (1) above, used acoustic monitors to verify that the check valve does chatter during this " chugging" phenoir.enon. An analysis is currently underway to quantify the extent of the chuggirg ar d is scheduled for completion by October 1,1990. If the analysis mdicates that spargers are required for the underwater exhaust piping, the spargers will be mstalled during the Spring,1992 Unit 1 Refuel l
Outage and the 1993 Unit 2 Refuel Outage.
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The check valve is currently installed in the vertical orientation close to an elbow of pi pe. This configuration is not ideal for the valve and therefore, a alping mod ification is bemg reviewed. The feasibility and associated costs
'o mount the valve in a horizontal orientation will be evaluated. This option, however, does not appear to be practical since the exhaust piping is located in a congested area of equipment and it is unclear if the new pipe run could be installed effectively. Evaluation of this option will be completed by October 15,1990.
4.
Mid cycle testing of the HPCI steam exhaust valves will be conducted at nine month intervals to verify the leak tightness of the valves.
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Commonwealth Edison has implemented corrective actions which have im roved the performance of various containment isolation valves and penetrations at und Cities Station. Since 1986, Commonwealth Edison has enhanced the per rmance of the Main Steam Isolation, Main Steam Line drain, RHR Suppression Chamber Spray, RCIC Steam supply and exhaust, and the oxygen analyzer suction valves and drywell head gasket. While the individual actions implemented on these valves and component have been effective, we acknowledge that additional actions are warranted to improve the performance of the overall containment, during the as found Integrated Primary Containment Leak Rate Testing (IPCLRT).
o, Commonwealth Edison believes that the corrective actions proposed by.this response will improve the performance of the remaining valves that exceed leakage requirements. In the Inspection Report, the NRC Inspector recognized the efforts that Quad Cities Station is taking to improve the performance of these valves, For l
example, the Inspector noted the ' positive and aggressive approach" to the resolution of the Feedwater Check valve leakage.
Quad Cities Station recognizes that in order to further improve the valve 4
~ leakage performance, a more inte trated and predictive approac h is required. To achieve this goal, a predictive anuor preventative maintenance program, including L
periodic valve replacement, will be developed for containment isolation valves.
Of the eight (7) systems which are specifically discussed in the Inspection Report, six (6) systems have either been modified and successfully tested or modifications are planned during the upcoming outage. Commonwealth Edison, therefore, does not currently plan to decrease the interval between the conduct of l
Type B test on the identified valves (with the exce ation of HPCI steam exhaust l
check valve). Instead of decreased testing interva:s, we have proposed to develop a continuing program focusing available resources on predictive and preventative L
maintenance. This should minimize future valve failures. Furthermore, we believe l>
- that the specific actions proposed will be effective in future Appendix J test j
campaigns.
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BESEDNSElRNODCE_OE.Y10LNI1ON 50_-254(366Y89024:44
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10 CFR 50.59, Section (bXI) requires in part that the licensee maintain records of changes in the facility... to the extent that these changes constitute changes in the facility as described in the safety analysis report... It goes on to state 1
that "These records must include a written safety evaluation which provides the bases for the determination that the change, test, or experiment does not involve an unreviewed safety question."
Ca/ -y to the above:
(a)
A safety evaluation performed in 1984 to address addition of an 0 ring,to the feedwater check valves did not provide en adequate basis for determinmg that no unreviewed safety question existed. L'pecifically, the safety evaluation did not address the consequences of failure of 'he 0-ring on the valve's ability to maintain leak tightness, a safety function of the valve described in USAR section 5.2.2. Failure of the 0 rmg was continued to occur since 1984 and has resulted in failures of the valves' local leak rate tests which in turn have resulted in failures of the containment integrated leak rate tests.
BeWOR80 Commonwealth Edison, through the investigation of the cause for the Notice of Violation, identified that modification M41(2)-83 43 did not involve the feedwater check valves which provide the containment isolation function.
Commonwealth Edison notified the NRC Inspectors of this finding Modification M 41(2)-74-15, which modified the seat ring to valve body seal from a metal seal ring gasket to a rubber 0-ring, was provided to the Inspectors. The Inspectors performed a review of the safety evaluation associated with this modification and concluded that the 10 CFR 50.5'9 evaluation was acceptable. The Inspectors indicated that this violation is retracted.
L Contrary to the above:
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(b)
In 1987 the licensee failed to perform a safety evaluation prior to changing the O ring material. As discussed above, failure of the 0 ring can result in failure L
of the valve to perform its intended containment isolation function. One of the O rings failed during the 1989 - 1990 Unit 2 operating cycle.
BeePonse Commonwealth Edison's Quality, Assurance Procedure QP 3-51 defines the reguirements for the substitution of materials. The Program requires that a i
smtability for application evaluation be performed for material replacement. This technical evaluation must consider seismic requirements, environmental qualifications and the need to perform a 10 CFR 50.59 evaluation. The Quality Procedure, therefore, does not require that all material substitutions be evaluated under the provisions of10 CFR 50.59. This requirement is supported by NSAC 125 L
GuidelintLfor 10 CFR 50.59 Evaluations. This guideline states that mamtenance L
activities (which includes replacement parts that have been demonstrated and documented to be equivalent) do not require a 10 CFR 50.59 evaluation.
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5 The basis for the substitution of material for the feedwater check valve O ring was based on a recommendation from the Corporate Engineering Department. The reconunendation was supported by extensive testing of various compounds conducted by Wyle Laboratories. The material testing consisted of a
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baseline functional test, exposure to radiation levels of 2.2 X 10 rads (gamma), post radiation functional testing and 1% years of equivalent thermal apng. The testmg program concluded that the Parker seal compound was best suitet for the feedwater check valve application. The Station performed a " Classification of Spare Parts" evaluation (QTP 500-S12). Quad Cities Station, however failed to document the consideration of the 10 CFR 50.59 evaluation, as required by the Quality Assurance Manual. Quad Cities Station, therefore, did not violate the provisions of10 CFR 50.59 (as discussed in NSAC 125) but rather, did not meet a il the requirements of the Quality Assurance Program.
Since 1987, the program for the evaluation of material substitutions has evolved. Quad Cities Station Procedure QTP 600-16 " Suitability of Application Evaluation" includes the considerations required by the Quality Assurance Manual l
and is currently applied to all material substitutions. A Nuclear Operations Directive on the conduct of10 CFR 50.59 has been developed and provides the screening criteria for determining the need for a 10 CFR 50.59 evaluation. This Directive is currently under review and will be issued by October 1,1990. The proposed screening criteria for the 10 CFR 50.59 evaluations provides a more conservative approach to 10 CFR 50.59 evaluations than that defined in the NSAC 125 document, for maintenance activities.
Conclusimo Commonwealth Edison accepts the Notice of Violation on the basis that there is no evidence that the need to perform a 10 CFR 50.59 evaluation (as required by the Quality Assurance Procedure) was considered during the acceptance of this material substitution. We believe, however, that the techml evaluation (acceptance throu h testing) was acceptable and within the guidelines that are desenbed in NSA 125.
CorrectitutActiOIllOE,u 30CutITA00 As previously discussed, Qand Cities Station Procedure QTP 600-16
. " Suitability of Application Evaluatic n"is being implemented for all material substitutions. The arocedure requires that the consideration for the need to conduct a 10 CFR 50.59 eva untion be docu.nented. Quad Cities Station has developed interim screening criteria for the conduct of 10 CFR 50.59 evaluations, which is based on the requirements contained in NSAC 125. The Station's 10 CFR 50.59 L
evaluation screening procedure will be revised to incorporate the requirements of the directive.
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ATTACHMENT B RESPONSE TO UNRESOLVED ITEMS 50-254(265)/89024-03 50-254(265)/89024-05 50-254(265)/89024-06 l
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UNRESOLVED ITEM liG-25K266VB9024-08 "The licensee has in their technical specifications and procedures that the i
Main Steam Isolation Valve (MSIV) leakage results do not need to be added to the Type B & C totals. The inspectors requested that the licensee produce the exemption from the requirements of 10 CFR Part 50, Appendix J, which allows them to remove the MSIVs from the totals. The licensee was unable to produce this exemption by the end of the inspection. This has been made an Unresolved Item pendmg the production of the exemption or the correction of the technical specifications."
RESEORSE i
On April 13,1981, Commonwealth Edison proposed changes to the Technical Specifications for Quad Cities Station Units 1 and 2. The proposed amendment revised the primary containment inte prated leak rate test requirements
- and schedules to conform with the re-quirements oL Appendix J to 10 CFR 50 and
^
modified the associated Limiting Condition for Operation to include the definitions of the nomenclature that are used to identify specific leakage limitations required by Appendix J. In that amendment, Commonwealth Edison proposed a combined lea kage rate for all penetrations excluding the MSIV leakage. Specifically, the Technical Specification requirement was proposed as follows:
"A combined leakage rate of 3: 0.60 L for all penetrations and valves, a
except for main steam isolation valves subject to Type B and C tests when pressurized to P "
a "With the measured combined leakage rate for all penetrations and valves, except for main steam isolation valves, subject to Type B and C tests exceedmg 0.60 L, restore the combined leakage rate for all a
penetrations and valves, except for main steam isolation valves, subject to Type B and C tests to 0.60L "
a The NRC approved the proposed Technical Specification amendment (Amendment 82 and 76 for DPR-29 and 30, respectively) on November 2,1982. In the Safety Evaluation, the Nuclear Reactor Regulation staff specifically acknowledged the proposed Limiting Condition for Operation discussed above and concluded that the Technical Specification meets the requirements of 10 CFR Part 50, Appendi:: J. (Reference attached Safety Evaluation).
Also NUREG-0123 (revision 3) Standardlechnical. Specification 1or General Electric _ Boiling Water _Reaciara contains the identical presentation for the requirement of measured combined leakage,i.e. MSIV leakage is excluded. The NUREG does not indicate that an exemption to 10 CFR Part 50 Appendix J is required in order to adopt the proposed requirement.
Based on the Technical Specification amendment Safety Evaluation and the language proposed by the Standard Technical Specification, Commonwealth Edison has concluded that the current Technical Specification LCO meets the requirements of10 CFR 50 Appendix J and no further action is required.
ID162
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t UNRESOLVED ITEM f
fdh25436tW89024-06 "The inspectors examined a replacement valve for one of the 24" butterfly h
valves and noted that the flanges and packing on the containment side of the upstream valves were not testable when the penetration is pressurized between the valves. The requirement to perform Type B testing of the fTanges and packing was t
being met for the "as left" condition by the performance of a CILRT every outage.
However, when the licensee returns to a normal CILRT test schedule, then Type B testing of the flan ges and packing will have to be performed in accordance with Appendix J. The Licensee was asked to determine how the testing requirements for these flanges would be met."
RESE0NSE I
The Drywell/I'orus Purge Supply and Exhaust lines are isolated in part by severallarge butterfly valves (see attached). These wafer style butterfly valves are bolted to the piping between flanges and are scaled by a gasket on each side of the valve.
During the Local Leak Rate Test (LLRT) on the butterfly valves, the volume between the inboard and outboard valves is pressurized to 48 psid (P,L The measured leakage rate consists of the leakage of the seats of both valves, the packing i
on the outboard valve, the flange on the outboard side of the inboard valve, and the side and the packing leakage of the inboard valve are excluded dun,e on the inbo flange on the inboard side of the outboard valve. The flange leakag ng the performance of the test. During the conduct of the Type A test, however, the flanges and packing are challenged by pressurization in the proper direction; therefore, the intent of the Type C test for these components has been met by the Type A test.
p To meet the 10 CFR 50 Appendix J testing requirement (without conducting the Type A test), Quad Cities Station is developing a modification that f
will assure that the flange and packing are pro wrly challenged. The modification willinclude removable blank flanges which wil: allow the pressurization of the valves from inside containment.
, his LLRT method has been used successfully at Monticello S% tion. The modification will be implemented before the accelerated Type A test hedule is terminated (following the successful completion of 2 consecutive Type A tests).
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l UNRESOLVED ITEM J
50-364(366VB902446 l
"The inspectors also had questions as to how minimum and maximum pathway leaka purge volumes.ges were being determined, especially for the four and six - valv A meeting was held in the region on February 13,1990. During this meeting, the licensee discussed the test methodology for each penetration. Neither-I l
the individual penetration nor the generic test procedures contained the necessary tidance to make such determinations. The guidance given in the CILRT procedure TS 150-8 was flawed and was deleted prior to the Umt 2 outage. This issue is an nresolved Item nding the licensee's revision ofits LLRT procedures and reviews by the inspectors.p' RESEONSE A meeting was conducted with the Region III Staff on 7 30 90 to discuss i
the Inspector's concerns regarding the Local Leak Rate Test (LLRT) Procedures. A l
draft of the Nuclear Operations Directive (NOD) on containment leak rate testing
)
was the subject of the discussion. The draft NOD clarified the methodology for correcting the Type A test results with the LLRT results. The Region III representatives concurred with the changes to the directive.
The guidance contained in the Directive that concerns the calculation of minimum and maximum pathway leakage for multi valve pathways was also discussed. Since the calculations for these pathways are complex,it is difficult to provide general guidance that is appropriate for all cases. The NRC Staficoncurred that the guidance could be deleted from the directive. Plant specific guidance would be includ ed in the Station procedure. Examples of the plant specific guidance were discussed and agreed upon at the meeting.
The revision to the Nuclear Operations Directive on containment leak rate testing will be issued in December,1990. The plant specific puidance for minimum and maximum pathway leakage for multi valve pathways wi.1 he incorporated in the Quad Cities Station procedures prior to the next Refuelmg Outage which is currently scheduled for November,1990.
I 1 ID162-
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ATTACHMENT C SAFETY EVALUATION SUPPORTING AMENDMENT NO. 82 (DPR-29)
AND AMENDMENT NO. 76 (DPR-30)
PURGE INLET SKETCH DRYWELL EXHAUST SKETCH e
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Specifically the proposed changes better define the Limiting Conditions 4
for Operation (LCO) for primary containment (PC) leakage rates by i
specifying:
- 1) an.overall integrated leakage rate for the PC,
'2) a combined leakage rate for all penetrations and valves, except main steam isolation valves (MSIV),
- 3) an acceptable leakage rate for any one MSIV, and
- 4) an acceptable leakage rate for any one air lock.
The proposed LCOs also describe actions to be taken when the measured Finally the p'oposed changes leakage rates are not.within specifications.
r delete the prescriptive surveillance requirements fr lemonstrating con-tainment leakage rates, and replace these with the,(
' aria specified in Appendix J of 10 CFR 50, using the methods and provin.pns of ANSI N45.4 (1972).
In reviewing the licensee's proposed changes submitted April 13 and December 2, 1981 we find that they are consistent with the.BWR Standard Technical Specifications, NUREG-0123 Revision 3, which served as the basis in assessing the conformance of the licensee's proposed Technical Specification changes to Appendix J requirements. The Standard Technical Specifications, pages 3/4 6-2 through 3/4 6-4, pertaining to primary containment leakage testing re-quirements (and the associated Bases) are recognized by the staff as an ac-ceptable implementation of the applicable requirements of ADpendix J.
There-fore, we conclude that the Technical Specification changes pertaining to containment integrated leakage testing meet the requirements of 10 CFR Part 50, Appendix J. and are acceptable.
IV ENVIRONY. ENTAL CONSIDERATIONS We have determined that the amendments do not authorize a change in effluent types or total amounts nor an increase in power level and will not result i
in any sfgnificant environmental impact.
Having made this determination, we have further concluded that the amendments involve an action which is insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of these amendments.
V CONCLUSION We have concluded, based on the considerations discussed above, that:
(1) because the amendments do not involve a significant increase in the probability or consequences of an accident previously evaluated.
'do not create the possibility of an accident of a type different from any evaluated previously, and do not involve a significant reduction in a margin of safety, the amendments do not involve a significant hazards consideration. (2) there is reasonable assurance that the health
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UNITED STATES
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORT *NG AMENDMENT NO 82 TO FACILITY OPERATING LICENSE NO. DPR-29 AMENMENT N3.76 TO FACILITY OPERATING LICENSE NO. DPR-30 COMMONWEALTH EDI.SDN COMPANY AND IOWA-ILLINOIS GAS AND ELECTRIC COMPANY QUAD CITIES STATION UNIT NOS. 1 AND 2
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DOCKET NOS. 50-254 AND 50-265 1..
.lNTRODUCTION By letter dated April 13, 1981 and supplement dated December 2, 1981 Common-i[
wealth Edison Company (licensee) proposed changes to the Technical Specifi.
cations for Quad Cities Nuclear Power Station, Units 1 and 2 to: revise _
the primary containment integrated leak rate test requirements and schedules to conform with the requirements of Appendix J to 10 CFR Part 50;- modify 3
the associated Limiting Condition for Operation to include the definitions of the nomenclature used and identify specific leakage limitations as re-1 I '
quired by Appendix J;- and modify the surveillance requirements to provide direct references to Appendix J methodology and termino < logy.
11.
BACKGROUND i
Beginning in August 1975, the NRC staff requested licensees to review their containment leakuge testing programs and.the associated Technical Specifi-cations for compliance.with the requirements of Appendix J. to 10 CFR Part
- 50.. Recognizing.at that time that there were already many operating plants l
and a. number more in advanced stages of design or construction, we requested I
licensees to propose design modifications and Technical Specifications changes and, as necessary, request exemptions to attain conformance with the l-regulations. As part.of that respont,e, the licensee requested a number of exemptions to the provisions of Appendix J.
Those requests are under review l
and are not addressed in this Safety Evaluation, 111 EVALUATION By letter dated April 13, 1981 and supplement dated December 2, 1981 the licensee proposed amending the Quad Cities Nuclear Power Station, Units 1 and 2 Technical Specifications (TS) to modify the primary containment inte-1 grated leakage testing requirements and schedules to conform witb 10 CFR i
L Part 50, Appendix J requirements. The proposed changes also provided for
' direct references and use of Appendix J methodology and terminology.
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Q-and safety of the public will not be end6ngered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.
Date:
November 2,1982 Principal Contribu' tor: Roby Bevad
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.b UNITED STATES' NUCLEAR REGULATORY CO'dISSION
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_ DOCKET NOSI 50-254 A':D 50-265 COMMDNWEALTH EDISON COMPANY NOTICE OF-ISSUANCE OF. AMENDMENTS TO, OPERAilfG LICEliSE5
.The U. S. Nuclear Regulatory Comission (the Commission) has issued
'?
Amentnent: Nos. 82 and 76 Facility Nerating License Nos.,DE.
R-29 and POR-30, issued 'to Coxnonwe$ith Edison Company, And IowE Illinois Gas and
ElectricCompanywhichrevised}theTechnicalSpecitifationsforoperation oftheQuadCitiesNuclearPowerSt5 tion, Unit's1and2 loc 5tedinRock Island County, Illinois. The amendments are effective as of the date of issuance.
The$hanges to the Technical Specifications provide for primary containment integrated leak rate te'st requirements and schedules consistent
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with Appendix 0 to 10 CFR Part 50.
The changes also provide for direct
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references and use of Appendix 0 methodology and termirfology.
The application for the amendments complies with the standards and
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re:virements cf the At:mic Er.ergy Act of 1554, as amended (the Act)., and the Corrr.ission's rutes and regulations.
The Corenission has made appropriate j
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' ' findings as recuired by the Act and the Co:r;nission's rules ant' regulations in 10 CFR Chapter I, which are set forth in the license amendments.
Prior 1
l public notice of these amendments was not required s'ince the amendment $ do L
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n:t inv:1ve a significant hazards c:nsideration.
lhe C:r.,issi:n has ceter.ined that the issuan:e of :P.ese amendments will not.resuit in any significant envirenrental it;ac and that pursuant to 10 CFR 51.5(d)(4) an environmental impact statement or negative declaration e
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2-7590-01 i
and environmental impact appraisal need not be prepared in connection with issuance of these amendments.
For further details with respect to this action, see (1) the application i
for amendments dated April 13, 1981 and letter dated December 2,1981(2) i Amendment No. 82 to License No. DPR,-29 and Amendment No. 76to, License No.
DPR-30,and(3)the' Commission'srd75tedSafetyEvaluation.' tall 6fthese items are available for public knspection at the Commission's Public Document Room, 1717 H street, N. W., Washington, D. C., and at the Moline Public Library I
504-17th Street, Moline. Illinois.
A copy of items (2) and (3) may be ob-tained upon request. addressed to the U. S. Nuclear Regulatory Commission, Washington, D. C. 10555, Attention:' Director Div'ision of Licensing.
Dated at Bethesda, Maryland, this 2nd day of November 1982.
FOR THE NUCLEAR REGULATORY COMMISSION i
Domenic B. Vassallo, Chief
' Operating Reactors 9 ranch #2 Division of Licensihg k
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