ML20058M258
| ML20058M258 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 12/13/1993 |
| From: | Saccomando D COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| GL-89-04, GL-89-4, NUDOCS 9312200217 | |
| Download: ML20058M258 (11) | |
Text
) C:mmnnw dth Edison December 13,1993 e
O-1400 Opus Place i
Downers Grove, Illinois 60515 Dr. Thomas E. Murley, Director t
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Document Control Desk
Subject:
Braidwood Station Units 1 and 2 Inservice Testing Program Plan for Pumps and Valves Safety Evaluation Response dated September 14,1993 NRC Docket Nos. 50-456 and 50-457
References:
(a)
J. Dyer letter to D. Farrar dated September 14,1993, transmitting SER for Revision 5/5a IST Program (b)
T. Simpkin letter to T. Murley dated June 25,1992, transmitting Revision 5a IST Program (c)
R. Barrett letter to T. Kovach dated January 31,1992, transmitting SER for Byron Nuclear Power Station (d)
R. Pulsifer letter to CFCo dated October 10,1991, summarizing the October 1,1991 Meeting (e)
Generic Letter 89-04, " Guidance on Developing Acceptable Inservice Testing Program," issued April 3,1989.
Attachment A is Braidwood Station's response to Reference (a) which transmitted certain anomalies that were identified in the Safety Evaluation (SE) of the Inservice Testing Program Plan for Pumps and Valves. The Technical Evaluation Report (TER) which was transmitted in Reference (a) listed a total of eight (8) anomalies. Items 1,2,4,5 and 7 will be specifically addressed in this response. Items 3,6 and 8 will be addressed in the six-month written response required by Reference (a) to be submitted by March 14,1994.
Upon reviewing the NRC's reply in Reference (a), Item (5) and Braidwood's own re-review of this issue, we request that the NRC reconsider the five-year inspection frequency for VR-02, as submitted.in Reference (b). Additional information pertaining to this relief request is included in Attachment A, Item 5. However, in the interim, we propose to use the VR-02 relief request which is included in Attachment B for the Unit l' refuel outage which is scheduled to start March 5,1994. This relief request, which was-approved for Byron Station (Reference (c)) involves disassembly and inspection of one valve out of a group containing two valves, each refueling outage and is in accordance with GL 89-04, " Guidance on Developing Acceptable Inservice Testing Program," Position
- 2. In addition, Braidwood Station understands that per GL 89-04, we have approval to utilize this relief request until a reply is received concerning the dispositioning of the i
VR-02 relief request.
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br. Mdrley. December 13,1993 Please address any questions you may have regarding this matter to Denise Saccomando -
at (708) 663-6484.
Si.erely, i
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Denise M. Saccomando i
Nuclear Licensing Administrator i
Attachments I
cc:
R. R. Assa, Braidwood Project Manager - NRR S. G. Dupont, Senior Resident Inspector - Braidwood J. B. Martin, Regional Administrator - Region III Office of Nuclear Facility Safety - IDNS i
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Attachment A Response to SER Appendix A Revision 5a IST Program Anomalies Item r
(1)
The IST program does not include a description of how the components were selected and how testing requirements were identified for each component. The 3
review performed for this SEffER did not include verification that all pumps and valves within the scope of 10 CFR 50.55a and Section XI are contained in the IST program, and did not ensure that all applicable testing requirements have been identified. Therefore, the licensee is requested to include this information in the IST program. The program should describe the development process, such as a listing of the documents used, the method of determining the selection of components, the basis for the testing required, the basis for categorizing valves, and the method or process used for maintaining the program current with design modifications or other activities performed under 10 CFR 50.59.
Response
i The IST program at Braidwood was developed from Byron Station's IST program which had undergone the approval process of" questions and answers," followed by a site working meeting to resolve any additional outstanding issues. Braidwood was able to adopt Byron's IST program because the plants were built to the same design, except for the difference in heat sinks (lake versus cooling towers).
Braidwood and Byron Stations are currently working together to perform a detailed scope review of Technical Specifications, Updated Final Safety Analysis Report, Safety Related Component List, Emergency Operating Procedures, etc.
This review is being done in three phases: 1) Valves not currently in the IST program; 2) Valves currently in the program for safety function direction; and 3)
Categorizing, establishing test requirements and frequency of testing. The review is also using the scope statements found in ASME/ ANSI OM-1988, Parts 6 and 10.
OM-6 and OM-10 are being used so that the review can be used for Byron's ten-3 year update. The scope review process being performed has to be done around -
scheduled refueling outages at both stations (Fall of'93 for Byron, Spring and Fall of '94 for Braidwood), and is scheduled to be done in June of 1994.
a Byron is required to submit a ten-year IST Program update in the Spring of 1995 (six months prior to their starting date of the second ten-year interval). Due to the joint Braidwood/ Byron effort in scope review, Braidwood would also like to submit our IST Program Revision at that time (Spring of 1995, but no later than I
April ~1,1995), which will reflect the Scope Review described above. - This would represent a six-month extension beyond the time granted in the SER, but also involves a larger effort than that described in the SER. Braidwood requests a six.
month extension to Item (1) of Appendix A of the TER, but will go beyond describing the process used for epe selection in that an actual scope review will be performed.
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Attachment A (continued) ltem (2)
'PR-05 requests relief from the instrument full-scale range requirements of Section XI for the CC and essential service water pumps. The licensee proposes to measure pump flow rate using ultrasonic flowmeters with digital readouts whose accuracy is *4% of the indicated.' reading and is independent of the full-scale range. Subsection IWP does not specifically address digital instruments; however, they are addressed in OM-6 which is covered by rule-making effective September 8,1992. The note for Table 1 of OM-6 specifies that digital instruments must be accurate to *2% over their calibrated range. Since the licensee's ultrasonic instruments are less accurate than the OM-6 requirement and the repeatability of these portable instruments are unknown, there could be significant data scatter of the tests measurements. Therefore, the licensee should either demonstrate that these instruments provide adequate repeatability, or they should develop a method to compensate for the additional 2% inaccuracy wh-a evaluating these pumps.
One possible method of accounting for the additional uncertainty would be to add 2% onto measurements above the reference value v.d subtract 2% from measurements below the reference value when comparing to the allowable ranges l
of flow rate. The proposed alternative should be authorized pursuant to 5 50.55a 1 (a)(30(ii), with the provision that the licensee either demonstrate that the instruments provide indication that is sulliciently accurate and repeatabic to detect degradation and permit the use of the allowable ranges of Table IWP-3100-2, or develop a method to compensate for the additional 2% inaccuracy when evaluating these pumps.
Resnonse i
The PR-05 relief request will be deleted from the program. Special wet flow calibrations now being performed on all ultrasonic flowmeters used to test ASME XI pumps makes the instrument accurate to within the OM-6 required *2% of measured flow. This program change will be incorporated in the next revision of the IST program, scheduled for March 14, 1994.
Item (4)
Some valves can only be tested during cold shutdowns or refueling outages. In VR-20 the licensee requests relief from the prescribed corrective actions for valves in this category whose stroke time measurements enter the Alert Range. This proposal could allow valves that are possibly serio.usly degraded to remain in i
service for extended periods without additional testing to determine their state of degradation or operational readiness. Therefore, the licensee's proposed corrective L actions are not acceptable and relief should not be granted as requested.
^
ASME/ ANSI OMa-1988, Part 10, was included in f50.55a rulemaking, and-Subsection 4.2.1.9 does not require an increased test frequency for valves that do
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not meet the stroke time acceptance criteria. The licensee should measure stroke times in accordance with Part 10 and comply with the corrective actions of Subsection 4.2.1.9 (Refer to Section 3.1.1.1 of this report).
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' Attachment A (continued)
Response
I'ha cwrective actions described in relief request VR-20 will be changed to reflect ASME/ ANSI OMa-1988, Part 10 requirements for valves that can only be tested in cold shutdown or refueling. These actions are planned to be completed by the next scheduled refueling outage of Unit 2 (A2R04) in September of 1994, t
The fixed alert ranges described in relief request VR-20 will be implemented over a period of time; as the fixed alert ranges are established and the procedures are -
revised to conform with VR-20, the current Section XI requirements and station procedures will be followed in regards to alert range testing requirements.
Item (5)
VR-02 requests relief from the test method and frequency requirements of Section XI for the check valves in the line between the spray additive tank and the CS educator,1(2)CS020A and B. The licensee proposes to verify valve full-stroke capability by disassembly and inspection at least once every five years, and follow this by exercising the valves open with flow during the TS educator full flow test.
GL 89-04 states that the use of disassembly to verify full-stroke capability of check valves is an option only where full-stroke exercising cannot practically be performed by flow or by other positive means. Since TS required educator flow can be verified through these valves periodically, this exercising should be used to meet the IST requirements instead of a GL 89-04, Position 2, dis-assembly program.
ASME/ ANSI, OMa-1988, Part 10, Subsection 4.3.2.2 permits deferral of full-stroke exercising until refueling outages when this exercising is not practical during plant operation or cold shutdowns. However, performing this testing once every five years, as proposed by the licensee, is a significant departure from the Code-allowed testing frequencies that may not be justified. Check valve exercising is generally a " pass or fail" test that does not provide any objective means to i
evaluate valve condition or detect the presence of degradation. Supplementing the r
exercise test with a diagnostic evaluation of the valve's condition, such as the proposed disassembly and inspection, would provide much useful information about the valve's condition. This information might be used to justify extending the test interval for these valves. To support extending the test interval, a diagnostic evaluation should be performed on each valve in the group and the results documented in detail. The evaluations should provide objective evidence that the valves do not suffer significant degradation over time, and that they -
retain a high assurance of operational readiness.
Jased on the detr nedon that more information is needed to support extending f
the exercising frequency for these valves from once each refueling outage as 1
approved by rulemaking effective September 8, ~1992, relief should not be granted
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as requested. (Refer to Section 3.2.1.1 of this report.) '
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Attachment A (continued)
Resnonse i
It was Braidwood's original intent to test these valves in accordance with the relief request approved for Byron Station per Reference (c). This relief request was approved to use the GL 89-04 sample disassembly plan in Position 2, at a frequency of every other refueling outage. The VR-02 relief request actually submitted for Braidwood was a departure from the Byron relief request due to a meeting held October 1,1991, with NRR. During that meeting, it was our understanding that based on the particulars associated with the valves addressed
.n VR-02, NRR was receptive to extending the frequency for disassembly and inspection for these valves to once every five years in conjunction with the educator flow test. Upon reviewing your reply _in Reference (a) and our own re-review of this issue, we are requesting the five-year inspection frequency, and provide additional supporting information below.
r In addition, the following information is provided to support the proposed five-year j
frequency for the 1/2CS020A and B check valves:
The Technical Specification educator flow test is required to ensure a.
adequate NaOH concentration of the containment spray system during an accident. This test has to be performed at least once every five years. Due to the hardships involved with this flow test as described in the proposed i
relief request, it is not practical to perform this test every refueling outage in lieu of disassembly and inspection. This is due to: 1) The potential to generate mixed waste that has to be stored on site. There are no commercial disposal sites available. 2) The risk associated with reducing the safe shutdown margin for the fuel below acceptable levels by the -
i relatively large amounts of pure water necessary to_ accomplish educator flow testing, which is directly recirculated back to the RWST. This reduces the boron concentration in the RWST, which is used as the make-up supply for the fuel pool and reactor refueling cavity. 3) Inconsistent with the ALARA concepts of the newly revised 10 CFR 20 regulation, due to the -
l higher exposure levels involved for the educator flow test versus inspection.
- 4) Impedes the optimization of the preventive maintenance program, as
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implied by 10 CFR 50.65 (a)(2), by not utilizing the information and data through monitoring and trending as gained through visual inspection for these valves, b.
The design of these valves consists of a wafer style body and dual disk plates, commonly referred to as a duo check. They are simple check valves,=-
requiring no internal disassembly of their internals in order to perform a thorough visual examination when removed from the piping system.
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Attachment A (continued)
The application of these valves, with regards to size, proximity to turbulent c.
sources in the piping system, and orientation has been reviewed and found d
acceptable by recent industry guidelines. They are not subject to high concentrations of NaOH due to the tpstream isolation valve being closed i
and the internals of the valve being on the downstream cide of the disk plates. They are subject to stagnant borated water from the Refueling Water Storage Tank (RWST), which is not as severe an environment as the NaOH stagnant line. This should eliminate the accelerated corrosion rate concern discussed in the TER.
d.
The failure history for these valves at both Byron and Braidwood is non-existent. Their are a total of 12 inspections documented for these valves, with no degradation mechanisms identified (e.g., erosion, corrosion, fouling, wear, binding, loose parts, and fatigue failure). This represents ten outages worth ofinspection data or 15 valve-years without any problems associated with valve condition. The failure data for the industry is insufficient to determine any type of preventive maintenance frequencies which could factor into the proposed five year test frequency. Additionally, these valves have been reviewed in detail, using the check valve inspection program optimization methodology, which evaluated each valve as having a preventive maintenance (PM) inspection frequency of 7.5 years. This detailed evaluation process reviewed: inspection data, surveillance test data, maintenance history, vendor / industry information, NRC information, safety significance, system service media, operating conditions, and system application / design information, Diagnostic non-intrusive techniques for these valves would require flow e.
through the line in order to detect valve opening. However, design flow would be obtained before disc full-stroke, based on the critical flow l
calculation for these valves. Acoustic testing, which Commonwealth Edison employs at it's nuclear units for non-intrusive testing, on similar duo type check valves has not been successful in proving full-stroke of both disk plates. The smaller the valve the more difficult it is to detect and evaluate -
full-stroke. Based on this, the design flow through the valve would be used in lieu of acoustics for meeting the Section XI full-stroke open test requirement, if the educator flow test is imposed to prove valve operational readiness.
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Attachment A (continued) f.
The water in the RWST is processed reactor coolant, and therefore potentially contaminated (it contains low level activity). During refueling.
i outages the RWST is used as a source of water for filling +.he reactor refueling cavity. When the reactor refueling cavity is drait:ed it is sent back.
to the RWST as well. If there is any leakage back to the Na0H Mnk, then it has the potential to contain low levels of radioactivity, making it mixed.
hazardous waste.
g.
GL 89-04, Position 2 allows for a maximum of six years between inspections i
for an individual valve out of a sample group of four valves. Braidwood's proposal to inspect any single valve at a frequency of no greater than'five years (three refuel outages) is bounded by the six year acceptable interval established in the GL. Also, the full flow versus the partial flow testing after disassembly will meet the GL requirement to partial flow the valve after reassembly. The expansion criteria requirement will also be followed if the valve has a problem related to full-stroke operability, per Position 2.
While waiting for NRC response to the inspection frequency proposed above, we will use the VR-02 relief request approved for Byron Station for the Braidwood Unit I refueling outage scheduled to start March 5,1994. This involves disassembly and inspection of one valve out of a group containing two valves, each refueling outage. The VR-02 relief request to be used is Attachment B to this response. As this is in accordance with GL 89-04, we will assume we have approval to utilize this relief request immediately unless we receive a reply otherwise.
Item (7)
VR-19 requests relief from the test frequency requirements of Section XI for the -
AFW pump suction check valves. The licensee proposes to verify closure of these valves using acoustic techniques every other refueling outage (one valve will be.
tested every refueling outage on a sampling basis). ASME/ ANSI OMa-1988, Part 10, Subsection 4.3.2.2, permits deferral of full-stroke exercising until refueling outages when this exercising is not practical during plant operation or cold shutdowns. However, performing this testing on a sampling basis, one valve every refueling outage, as proposed by the licensee, is a departure from the Code testing frequencies that may not be justified. The licensee has not provided an. adequate justification for not exercising both of these valves each refueling outage.
Therefore, relief should not be granted to this test frequency as requested. (Refer to Section 3.7.1.1 of this report.)
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B Attachment A (continued)
R'esponse This relief request will be revised to test both valves every refueling outage using acoustical means. The Reliability Centered Maintenance (RCM) failure modes and effects analysis (FMEA) performed on the AFW system was the impetus for the test frequency specified in relief request VR-19. FMEA determined that a 3-year test frequency would provide a commensurate level of valve operability and function to the ASME Section XI Code.
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l Attachment B i
RELIEF REQUEST VR-02 1.
Valve Number:
ICS020A 2CS020A 1CS020B 2CS020B l
2.
Number of Items:
4 3.
ASME Code Catecorv:
C 4.
ASME Code.Section XI Requirements:
Exercise check valves to the position required to fulfill their function (open=Ct; i
closed =Bt), unless such operation is not practical during plant operation, per d
IWV-3522.
5.
Basis for Relief:
The check valves in the spray additive system cannot be stroked 'without introducing Na0H into the CS system.
G.
Alternate Testinc:
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The A and B train valves are of the same design (manufacturer, size, model number, and materials of construction) and have the same service conditions, including orientation; therefore, they form a sample disassembly group.
l GROUP 1 GROUP 2 ICS020A 2CS020A 1CS020B 2CS020B One valve from each group, on a per unit basis, will be disassembled and examined each refueling outage. If the disassembled valve is not capable of being full stroked exercised or there is binding or failure ofinternals, the remaining.
j valve on the affected unit will be inspected. In addition, the disassembled valve -
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will receive a partial flow test upon reinstallation.
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Attachment B (continued) 7.
Justificat_ ion:
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1 Full flow testing of these valves cannot be accomplished without posing a serious threat to the safety of equipment and personnel. It is impractical to either full or partial stroke exercise these valves since flow through them would result in the introduction of NaOH into the CS system. Full flow testing would require a special test hook-up and necessitate flushing the system.
The alternate test frequency is justifiable in that maintenance history and previous inspections of these valves at both Byron and Braidwood stations has shown no evidence of degradation or physical impairment (this is to be expected since.these valves see very limited operation). Industry experience, as documented in NPRDS, showed no history of problems with these valves. A company-wide check valve evaluation addressing the "EPHI Application Guidelines for Check Valves in Nuclear Power Plants" revealed that the location, orientation and application of these valves are not conducive to the type of wear or degradation correlated with SOER 86-03 type problems. However, they still require some level of monitoring to detect hidden problems.
- t The wafer type design of the valve body for these valves make their removal a simple process, with little chance of damage to their internals. Also, there is no disassembly ofinternal parts required; all wear surfaces are accessible by visual examination. After inspection and stroke testing, the valve is reinstalled into the line and post maintenance testing is performed. The valve inspection procedure requires post-inspection visual examination of the check valve to insure that the pin is orientated properly and that the flow direction is correct.
The alternate test method is sufficient to ensure operability of these valves and is consistent with Generic letter 89-04.
8.
Applicable Time Period:
This reliefis requested. once per quarter during the first inspection interval.'
9.
Approval Status:
- a. Relief granted per Generic Letter 89-04..
- b. R.esubmitted per Revision 5a SER Appendix A, Item (5) response.
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