IR 05000409/1985017
| ML20209H898 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 11/05/1985 |
| From: | Eng P, Guldemond W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20209H871 | List: |
| References | |
| 50-409-85-17, IEB-84-03, IEB-84-3, NUDOCS 8511110268 | |
| Download: ML20209H898 (9) | |
Text
{{#Wiki_filter:~ , .,, , U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Report No. 50-409/85017(DRS) Docket No. 50-409 License No. DPR-45 Licenseef Dairyland Power Cooperative 2615 East Avenue - South La Crosse, WI 54601 Facility Name: La Crosse Boiling Water Reactor-Inspection At: La Crosse Site, Genoa, WI Inspection Conducted: October 23 through November 1, 1985
- Inspector
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[ M Date , - -Approved By: W.G.G demon ! Chief Operational Programs Section Date .. Inspection Summary i Inspection on October 23 through November 1, 1985 (Report No_.
E _50-409/85017(DRS)) Areas Inspected: Routine, anncunced inspection by one regional inspector of-licensee action on previous inspection findings; followup of IE Bulletin 84-03; inservice testing program implementation; inservice testing of pumps; inservice testing of valves; inservice testing test instruments; inservice test performance and test records. The inspection involved a total of 38 . inspector-hours onsite.
Results: No violations or deviations were identified.
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' . . DETAILS 1.
Persons Contacted
- J. D. Parkyn, Plant Superintendent R. E. Gardner, Instrument and Electrical Supervisor
- L. Goodman Operations Engineer
- L. W. Kelley, Assistant to Operations Supervisor and Training
- R. R. We ry,
Quality Assurance Supervisor
- Denotes those attending the exit meeting on November 1, 1985.
Additional plant technical and administrative personnel were contacted by the inspector during the course of the inspection.
2.
Licensee Action on Previous Inspection Findings a.
(Closed) Violation (409/83013-02(DETP)): Instrument calibration program deficient. The item addressed the licensee's lack of calibration records and documentation of instrument accuracy acceptance criteria, as-found and as-left values, thereby complicating efforts to identify equipment found to be out of calibration. The inspector reviewed the ' latest revision to licensee procedures ACP 50.1, "LACBWR Preventive Maintenance Program", and ACP 13.1, " Control of Measuring and Test Equipment," and noted that acceptance criteria, as-found, and as-left data are now required for instrument and equipment calibration records.
Further investigation revealed that the licensee had not determined acceptance criteria for instrument accuracy for all plant instruments.
The licensee stated that, in some cases, accuracy requirements based on manufacturer's specifications were not available; therefore, accuracy requirements are assigned. The licensee stated that determination of acceptance criteria for all instruments or components requiring calibration was delayed due to the fact that some instruments can only be tested during reactor operation. Completion of acceptance criteria determination is expected by May 31, 1986. Review and verification of the validity of assigned acceptance criteria for instrument accuracy is considered to be an open item pending review in future inspections. (409/85017-01(DRS)) 3.
Followup of IE Bulletin 84-03: Refueling Cavity Water Seal On August 24, 1984, the NRC issued IE Bulletin (IEB) 84-03 to all power reactor facilities. The IEB, which described the events surrounding a refueling cavity water seal failure at the Haddam Neck facility, required licensees to evaluate the potential for and consequences of a seal failure and submit a sumary report supporting their conclusions.
On October 23, 1984, the licensee submitted the required report.
In this report, the licensee provided a description of the design of their seal system, the capacity of available makeup systems, a description of leak detection mechanisms employed, and an assessment of the potential for seal failure.
' . . , During the inspection, the inspector reviewed the licensee's response and supporting information in additica to the potential for loss of refueling cavity and/or Fuel Element Storage Well (FESW) water inventory by mechanisms other than seal failure with the following results: a.
The refueling cavity water seal is an expansion joint of 304 type stainless steel, single ply which is continuously welded to both the reactor vessel closure flange and a flange ring which is attached to the inside of the upper cavity liner. A 1/4 inch thick steel plate is installed above the seal area to protect the joint from damage due to falling objects. This design is considerably different than that at Haddam Neck, and the potential for a similar event does not exist.
b.
Drainage and/or seal leakage from the refueling cavity water seal is detected via a sightglass which is periodically inspected during refueling operations.
c.
A combination of containment sump level, FESW 1evel, and radiation monitoring alarms is available to alert the operating staff to leakage from either the refueling cavity or the FESW.
d.
Procedures are in place directing that fuel which is not located in either the reactor core or in fuel storage racks be placed in an appropriate location to prevent uncovery. These actions can be completed before damage occurs as refueling areas are continuously manned whenever fuel is being transferred or suspended from the refueling crane.
Based on the above, it is concluded that system design is such that the probability of catastrophic seal failure is acceptably low.
In the event that such a failure did occur, fuel damage is not anticipated based on existing procedural requirements associated with fuel handling.
It is concluded that the licensee has adequately addressed the issues identified in IEB 84-03.
During the inspection, a review was conducted to detennine if other potential mechanisms for loss of water from the refueling system existed.
The inspector noted that LACBWR currently stores fuel in a two tier rack configuration located in the FESW.
With the exception of the FESW cooling water injection line, well penetrations are located above the active fuel elevation.
Review indicated that the most vulnerable section , of the cooling water injection line is approximately 10 feet in length, and located between the bottom of the FESW and two check valves installed in series which, in the event of earthquake damage to the cooling piping, prevent drainage and back flow of FESW inventory into the FESW closed cooling system. Short of an extremely violent seismic event or sabotage, no credible mechanism was identified for failure of the cooling water injection line.
It is concluded that the issue of loss of refueling system water inventory is adequately resolved.
No violations or deviations were identified.
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Inservice Testing of Pumps The inspector reviewed the Inservice Testing program and associated relief requests for pumps. The licensee is conducting inservice testing of pumps using approved test procedures and in accordance with the schedule defined by Section XI of the American Society of Mechanical Engineers' Boiler and Pressure Vessel Code (ASME Code),1974 Edition and associated addenda up through Sumer 1975. Criteria for determining pump operability is clearly defined in each test procedure facilitating determination of the status of the pump imediately upon completion of the test. Pump acceptance criteria and associated reference values are incorporated into the LACBWR Operating Manual.
It was noted that the licensee is measuring pump suction pressure with the pumps both idle and running. Bearing temperatures and vibration readings are obtained in accordance with the frequency and methods delineated in the ASME Code.
With respect to vibration measurements, the licensee indicated that an evaluation of the points used for obtaining vibration data was planned.
Due to the age of various plant components, some of the marks on the pumps indicating the location where vibration data was to be obtained were faint although still discernible. The licensee stated that confirmation that the points currently used are still valid and to insure that the points are clearly marked would ensure confidence and continued consistency with respect to vibration data. This evaluation is expected to be complete by December 31, 1985. Completion of the licensee's evaluation of -he vibration measurement locations and remarking said points, if necessary, will be tracked as an open item.
(409/85017-02(DRS)) No violations or deviations were identified.
5.
Inservice Testing of Valves The inspector reviewed the licensee's inservice testing program for valves.
It was noted that the licensee had specified maximum leak rates for those valves requiring leak testing and was analyzing leak test data per the requirements of IWV-3420f and g.
The inspector noted that almost all valves in use at LACBWR which fall under the auspices of the inservice testing program are air operated; also, that the stroke time method was not explicitly stated. Discussions with all operating personnel indicated that full stroke time for valves is defined as the time from initial movement to the cessation of movement of the valve stem, as locally observed. The inspector noted that the 1974 Edition of Section XI does not explicitly define full stroke time; however, later editions of the Code state that " Full stroke time is that time interval from initiation of the actuating signal to the end of the actuating cycle." The licensee acknowledged the observation of the inspector and agreed that obtaining stroke times as defined in later editions of the ASME Code would provide indication of problems associated with operator air supply, actuator diaphragm or bleed ports. The licensee agreed to evaluate their valve strcke time method and revise the method if deemed necessary. The licensee stated that evaluation would be complete by June 30, 1986. Completion of the licensee's review of valve stroke timing methods employed at the plant and subsequent procedure revisions, if any, will be tracked as an open item. (409/85017-03(DRS)) i
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l . . . - During the procedure review, the inspector noted that while the licensee used calibrated stopwatches for stroke timing, stopwatch identifying numbers were not recorded on the data sheets. Calibration records revealed that all stopwatches currently used in the plant were calibrated. The inspector stated that should one of the stopwatches be found to be out of calibration, present plant practice would necessitate evaluation of all time data taken since the last calibration of the stopwatch. The licensee acknowledged the observation of the inspector and stated that procedure revisions to explicitly identify which stopwatch was used to obtain test data would be initiated. The licensee stated that such procedure revisions would be complete by November 30, 1985. This will be tracked as an open item. (409/85017-04(DRS)) No violations or deviations were identified.
6.
Inservice Testing Instruments During the review, the inspector noted that there was no objective evidence that the licensee had evaluated test instruments for compliance with the range and accuracy requirements delineated in IWP-4110 through IWP-4115. The licensee acknowledged the inspector's observation and stated that an evaluation would be performed by December 31, 1985.
Comparison of test instruments against the range and accuracy specifications delineated in the ASME Code and resolution of any identified variances is considered to be an Unresolved item (409/85017-05(DRS)). No violations or deviations were identified.
7.
Inservice Test Performance The inspector cbserved the performance of three valve stroke tests and the testing of the Alternate Core Spray (ACS) Diesel Driven Pump 1A, Demineralized Water Transfer Pumps 1A and IB, and the Component Cooling Water Pumps 1A and 18.
It was noted that testing of said components did not require removing one pump from service, and that test data was obtained efficiently and in a timely manner.
During testing, the inspector noted that the calibration status of the permanently installed tachometer on the ACS diesel driven pump was indeterminate.
Further investigation revealed that the licensee had not included the tachometer in the instrun.ent calibration program. The licensee stated that the pump was periodically inspected by the manufacturer, but they were unsure if the inspection included calibration of the tachcreter. The licensee stated that calibration status of the tachometer and addition of the tachometer to the calibration program would be complete by May 31, 1986. Determination by the licensee of the calibration status of the permanently installed tachometer and addition of the tachometer to the plant instrument calibration program is consideredanunresolveditem.(409/85017-06(DRS)) No violations or deviations were identified.
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, y,.,. , 8. - Inservice Test Records Pump acceptance criteria and associated reference values are incorporated into the LACBWR Operating Manual. As such,-any changes to these values requires a formal review by plant management. The inspector noted that engineering evaluations supporting changes to reference values and acceptance criteria were appropriately documented.
Inservice test records are appropriately stored and controlled. The inspector noted that original test records are kept and microform is not used. Records of post' maintenance and post modification test requirements and test data are acceptable.
9.
Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action on the part of the NRC or licensee or both. Open items disclosed during the inspection are discussed in Paragraphs 2, 4 and 5.
d 10. Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items, open items, deviations, or violations. Unresolved items disclosed during the inspection are discussed in Paragraphs 6 and 7.
11. Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1) on November 1, 1985, to discuss-the scope and findings of the inspection.
The licensee acknowledged the statements made by the inspectors with respect to-items discussed in the report. The inspector also discussed the likely informational content of the inspection report with regard to documents or processes reviewed by the inspectors during the inspection.
The licensee did not identify any such documents / orocesses as proprietary.
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, . XI 7742
/ . Interpretation: XI-77 02 Subject: Section XI, Division 1 Subsection IWV Date Issued: February 18,1977 File: BC-16 433 i Question 1: The reactor coolant pressure boundary extends to the second valve. These are two check valves in series, two manual valves, two auto isolation valves, and the pressurizer pressure control valve backed up by a remote manualisolation valve. Is it the intention of IWV-3420 to measure leakage of each of these valves including the pressurizer pressure control valve and safety relief valves? Reply 1: In accordance with Subarticles IWV-1100 and IWV 1300, all Code Classes 1,2, and 3 valves as stated therein must be tested.The intent of the Code is to test all valves required to be in operational readiness.
Question 2: Reg. Guide 1.26, page 1.26 2 footnote implies that check valves on all influent lines do not have high leaktight integrity and that with additional valve with high leaktight integrity the line can be classified as group D. It is likely that RCPB check valve leakage will exceed leakages allowed in IWV-3420 Table I and may cause costly shutdown time for repairs. These valves are not isolatable from the primary coolant loops for testing or repair. Is it permissible to avoid testing these check valves and extend the RCPB to the next leaktight valve for category A leakage measurement? Reply 2: Same as Reply 1.
Question 3: IWV 3420 allows testing using pressure differentials less than the functional pressure.
Will you provide a formula that can be used to interpret the paragraph? Reply 3: In accordance with paragraph IWV 3420(c)(5), the adjusted leakage equals observed leakage times (function pressure over test pressure)8 Question 4: Is it the intent of Section XI to include all containment isolation valves or only those lines Classed 2 or 3 that extend beyond the containment boundary? Reply 4: Same as Reply 1.
Question 5: If containment boundary valves are tested according to Appendix J,is it the intent to also test these valves in accordance with Section XI? Reply 5: Yes. Specific permissible leakage rates for individual valves are required to be determined by the plant owner in accordance with IWV-3420(f).
Question 6: Does Section XI accept reasons, such as high temperatures, airborne activity, lack of complete visibility of masks, and lack of oxygen, for not performing tests on valves inside containment dur-ing operation ifit is necessary to enter in order to perform the test? Reply 6: No.
Question 7: (a) Does a manual valve used for containment boundary fall into Category A and require leak testing? (b) Does a check valve on the discharge of a pump require functional testing? 146 a
. . ' . !~ XI-7742. XI 77-03. XI 7744 (c) Can sample line (~ 3/8 in.) remote operated valves be excluded? (d) Can test lines with remote operated valves be excluded since they are not directly related to operation? Reply 7: Same as Reply 1.
Question 8: Since specific max' mum leakage for individual valves are not listed in technical specifica-tions should they be disregarded altogether? If not please provide examples.
Reply 8: Specific individualleak rates shall not be disregarded. Specific permissible leakage rates for individual valves are required to be determined by the plant owner in accordance with IWV-3420(f).
Question 9: Does IWPinclude turbine driven auxiliary feedwater pumps or only motor driven pumps? Reply 9: Determination of whether turbine driven auxiliary feedwater pumps is included in the scope of IWP is contained in the words of IWP 1100,"... are provided with an emergency power source." The type of driver is not a consideration.
Interpretation: XI 77 04 Subject: Section XI, Division 1, Article IWP 4000 Date Issued: March 21,1977 File: BC-76469 Question: What kind ofinstrumentation should be procured and used by the owner of the power plant in order to conduct the periodic inservice testing of pumps to measure the differential pressure and bearing temperature? Reply: Instruments that satisfy the requirements of Article IWP-4000 may be used to measure differential pressure and bearing temperature.
Interpretation: XI I 7918 Subject: Section XI, Division 1, Testing inaccessible Valves, IWV 3300 Date issued: December 12,1979 File: BC 79130 Question: Is it the intent of Section XI, Division I to require that those valves which are accessible be directly observed at each valve exercising to confirm that remote valve indications accurately reflect valve operations? Reply: It is the intent of Section XI, Division I to require that all valves, accessible and inaccessible, that have remote valve indicators be visually checked at least once every 2 years to verify that remote valve indications accurately reflect valve operatio. , . . , . 'i ' xt... xi.i = Interpretation: 'XI t 7919 Subject: . Section XI, Division 1, Operability IJraits of Pumps,IWP-3210 . Date issued: December 12,1979 File: BC-79150 Question: IWP 3210 of the 1977 Edition of Section XI, Division I states that "In the event these ranges (specified in Table IWP 3100-2 for the differential pressure across pump) can not be met, the Owner shall specify in the pump record the reduced range limits te allow the pump to fulfill its function in lieu of the ranges given in Table IWP-3100-2".
Do the Alert Ranges specified in Table IWP 3100-2 refer to pump test data that falls outside the specified range,or does it refer to a system analysis which may indicate that the required Action Range be less restrictive than those ranges specified in the Table? Reply: IWP 3210 refers to Table IWP 3100-2 which specifies three ranges - Acceptable Range, . Allowable Range,and Required Action Range. The limits within each of these ranges refers to the pump and not to the system,that is, the ranges are for the pump test data. If these ranges cannot be met, the Owner can specify new range limits (e.g., from a range of 0.93 to 1.02 to a range of 0.89 to 1.03 for AP).
Using the less conservative ranges the Owner shall show that the overall pump performance has not degrad-ed from its intended function.
Interpretation: XI4141 Subject: Section XI, Division 1,IWV-2200(a) Categorization of Containment Isolation Valves Date issued: April 17,1981 File: BC.819 Question: Is it the intent of Section XI, Division I that Category A as defined in IWV 2200(a),1977 Edition, Summer 1978 Addenda, applies to containment isolation valves only, or is this definition intended to be applied to pressure isolation valves? Reply: IWV.2200(a) defines Category A valves.The definition applies to the use of the valve by the Owner, not the type of valve involved. If an Owner requires that containment isolation valve leakage is to . be limited to a specific amount, then by definition, containment isolation valves would be classified as Category A. This would also apply to pressure isolation valves,1f the Owner specifies a specific limit to the . amount ofleakage of that valve.
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APR I 5 1985 W ~ Isf- ~ Pi~ > --- Docket No.: 50-341 Q y[[ - wti , - F1th:%'ia MEMORANDUM FOR: Richard L. Spessard, Director Division of Reactor Safety Region III FROM: Hugh L. Thompson, Jr., Director Division of Licensing Office of Nuclear Reactor Regulation SUBJECT: RESPONSE TO REQUEST FOR TECHNICAL ASSISTANCE REGARDING MAXIMUM STROKE TIME TESTING FOR IST OF VALVES We have reviewed the infonnation submitted in your request for technical assistance dated November 14, 1984 regarding testing of the maximum stroke time as part of the in-service testing (IST) program at the Fenni-2 facility.
Our basic position on this request is that the applicant has comitted to comply with the requirements of the ASME Code and has not requested specific relief from the applicable portion of the ASME Code. Our response is directed towards the third concern outlined in your letter (i.e., the acceptability of baseline data established for valve testing in accordance with Section XI of the ASME Code) since the first two concerns were previously resolved.
Acceptability of Baseline Data Established for Valve Testing per Section XI j With respect to the applicant's procedures for measuring valve stroke times, l as described in your letter dated November 14, 1984, the staff agrees that these procedures are not in accordance with the requirements of Section XI, Subsection IWV-3A17 of the ASME Code (the Ccde). The use of such procedures would require prior written relief by the staff from the specific requirements of the Code.
The specific applicable Code requirements are: IWV-3417 Corrective Action (a) If, for power operated valves, an increase in stroke time of . 25% or more from the previous test for valves with full-stroke times greater than 10 sec or 50% or more for valves with full-stroke times - _less than or equal to 10 sec is observed, test frequency shall be increased to once each month until corrective action is takert, at which time the original test frequency shall be resumed.
In any case, any abnonnality or erratic action shall be reported.
(Emphasis added).
- Contact: M. Lynch, 492-7050 g 1 a 1985 l l l& ' (h Aqp,
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,/ APR y y,gg 'R. L. Spessard, Director-2- . (b) If a valve fails to exhibit the required change of valve stem or disk position or exceeds its specified limiting value of full-stroke time by this testing, then corrective action shall be initiated insnediately.
If the condition is not, or cannot be, corrected within 24 hours, the valve shall be declared inoperative. When corrective e:: tion is required as a result of tests made during cold shutdown, the condition shall be corrected before startup. A retest showing ' acceptable operation shall be run following any required corrective action before the valve is returned to service.
As cited above, each in-service test valve stroke time is required to be com- ' pared to the previous in-service test valve stroke time and is not related in any way to the design or purchase specification for a valve. Additionally, the staff does not interpret a corrective action to be the acceptance of the
i new stroke time measu,ed on the first monthly test. When a valve has exceeded this criterion on one in-service test, the monthly frequency must be maintained , until maintenance is performed on the valve so that it will not become inoperable.
It appears that the applicant's practice for establishing maximum limiting stroke times for valves is also inconsistent with the staff's interpretation of the Code. Subsection IWV is specifically a " component" test code and, therefore, requires that the owner specify the maximum limiting stroke times ' l.
for each power operated valve (IWV-3413).
It is the staff's position that these limiting values of full stroke time are required to be based on reason-able engineering judgement of component (valve) operability, not minimum system ' requirements. System (or component) response time limitations, as stated in the applicarit's FSAR or in the plant Technical Specifications, are also time limitations placed on each subcomponent of that system (or component). How-ever, the staff's position is that these response time limitations should rarely take precedence over a component-oriented limiting valve stroke time.
Inasmuch as the IST program requirements become applicable when Detroit Edison , declares that the Fenni-2 facility has gone " commercial " you should bring this l ' ! matter to its attention so that it can be properly resolved.
f dw2d, ('- E Hugh L. Th5npson, Jr., Director i [4 Division of Licensing Office of Nuclear Reactor Regulation ' l " . . i I ! - - ... -- . . , _ _ _ _ _ _ _
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- . NUCLEAR REGULATORY COMMISSION - , wasumann, o. c.aosss s ,p- ,... . ), October 19, 1984 - . - ,. - .' MEMORANDUM FOR: R. L. Spessard, Director Division of Reactor Safety, Region III
, FROM: Darrell G. Eisenhut, Director Division of Licensing, NRR , SUBJECT: RELIEF REQUESTS 'FROM LE/.K TESTING REQUIREMENTS AS STATED IN ' SECTION XI, SUBSECTION IWV-3420 0F THE ASME CODE TIA 84-62 , !' REFERENCE: R. L. Spessard memorandum to D. G. Eisenhut dated . July 19, 1984; Request.for Technical Assistance - Relief - Requests from Leak Testing Requirements as Stated in Section XI, i , Subsection IWV-3420 of the ASME Code (AITS F03043684) ' ! Your July' 19, 1984 memo noted that the Commission has granted relief from leak rate testing requirements of IWV-3420 for containment isolation valves and . permitted 10 CFR Part 50, Appendix J, type C testing as an alternative.
This i' practice has led to two questions: , 1.
Does granting such relief exempt licensees from specifying discreet or weighted leak rates for Category A valves addressed by the relief request? ' ,. 2.
Does granting such relief exempt licensees from leak rate analysis and corrective action requirements as stated in'IWV-3426 and 3427, respectively as well as those requirements stated in IWV-3420 through IWV-3425? As requested, we have reviewed the questions and the implications of the granting of exemptions from Section XI, IWV-3420 of the ASME Code. Section XI of the ASME Code requires individual testing for each component in the IST program, including individual acceptance criteria. Containment Isolation ' Valves (CIVs) are required to be individually included in the IST program because of their accident mitigation service requirements. However, since licensees are required to perform leak rate testing of CIVs in accordance with 10 CFR Part 50, Appendix 'J. NRR has routinely granted relief from the leak rate test requirements of the ASME Code for these components.
For cases where this relipf is granted the staff requires that the licensee still meet the Analysis of Leak Rates and Corrective Action requirements of the Code, , paragraphs IWV-3426 and IWV-3427 of the 1980 Edition, respecti,vely.
l The staff believes that a " weighted" approach is the most appropriate method of assignino allowable leak rates. This method is based on the existence of ' a linear relationship between valve sizes with respect to allowable leakage ' (i.e., a 6" valve would be allowed twice the leakage of a 3" valve).
Additionally, when the allowable leak rates are added up for all type C tested CIVs, the total should not exceed 0.6 L. This allows a certain amount of A . OCT29 894
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2- . f7exibility since the 0.6L value specified by Appendix J is the maximum allowed for the combined chmulative leak rates of type C tested CIVs and containment penetrations as determined by type B testing.
' This completes NRR review pursuant to TIA 84-62.
- -gsk' . UC , Darrell F Eife'nhu, Director 3. - i Division of Licensing Office of Nuclear Reactor Regulation , cc: R. Wessman, NRR ' C. E. Norelius, RIII T. T. Martin, RI J. A. Olshinski, RII R. Denise, RIV T. W. Bishop, RV J. M. Taylor, IE ' J. G. Partlow, IE R. J. Bosnak, NRR - F. C. Cherny, NRR J. D. Page, NRR ' > . ( ) . O ! ! . . e e , g s ? ,. -...., _ _ _ _ _ _ _ -.. - _... _.... ,.. _,, _.,,, _ _, _. - _ _ _ _ _..,,... _ _ _.. - _., _ _, _, _. _. . - _ _ _, _ ._.-,.m_,.__._,--_,
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NUCLEAR REGULATORY COMMISSION WAsNeafG700s. O. C. 20006 ' ' $. . % s g i.... ,I ~ APR 18.1985 - Docket No.: STN 50-483 - .- . DEMORANDUM 'FOR: Richard L. Spessard, Director Division of Reactor Safety - Region III ' , FROM: Hugh L. Thompson, Jr., Director . Division of Licensing , ( Office of Nuclear Reactor Regulation i i ,
SUBJECT: CLOSURE VERIFICATION OF NORMALLY CLOSED CHECK VALVES ~ DURING PREOPERATION TESTING AT CALLAWAY (TIA 83 117) , l REFERENCE: Letter from R. L. Spessard to D. G. Eisenhut on the above subject, dated November 8, 1983.
l i The referenced letter requested the staff position regarding testing of l normally closed check vgives for closure capability during preoperational testing and during plant life. The staff position is that normally closed , check valves, that have a safety function in the closed position, should > have the closure function verified both during preoperational testing and In addition, the staff verifies < periodically throughout the plant life.
. ' ' that closure verification testing is specified in their nonnal review of the IST program, and if not, measures are taken to see that the program is revised.
l.; ; In an attempt to have the ASME clarify ambiguities within Section XI of the ASME Code regarding valve categorization and leak testing, the staff submitted an inquiry to the society. The response time from the society (approximately Enclosed one year) was a major factor in the delay of this response to you.
is a more detailed staff evaluation of the subject.
b: ! 12f*2 i - Hugh L.khompso, Jr.. Director }[DivisionofLicensing I - l Office of Nuclear Reactor Regulation i Enclosure: As stated cc: T. Martin , P. Semis . R. Denise T. Bishop P. Wohld A QPelke1e.,Riside t Inspector ,..P. , , Contact: T. Alexion. L8fl (FTS492-8929) $%MSN[NO
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. . I " ' ~ ENCLOSURE ~ STAFF EVALUATION REGARDING TESTING
., l 0F NORMALLY CLOSED CHECK VALVES
~ FOR CLOSURE CAPABILITY ~ s i-Reference: November 8, 1983 memorandum from R. L. Spessard, Region III, ' to Darrell G. Eisenhut, NRR , i ' ,i, The referenced memorandum states that there is no apparent requirement in .Section XI of the ASME Code to verify closure of normally closed check valves
that are classified Category C in accordance with Subsection IWV of the code.
, It correctly points out that there are normally closed check valves, other than Containment Isolation Valves (CIVs) and Pressure Isolation Valves (PIVs), l' . that have a safety related function in the closed position. An example of such valves is given for the Callaway plant and stated to be a normally closed
- ECCS suction line check valve between the Refueling Water Storage Tank and the l
ECCS pumps.
- i It is stated that verification of the closure function of normally closed check
, valves is a generic safety concern to the extent that surveillance is never , i- - done after construction to verify the closure function. Specifically, the staff position was requested regarding: 1) Testing of normally closed check valves for closure capability , during preoperational testing.
, 2) Testing of normally closed valves for closure capability during plant life.
In response, the staff position is, and has been, that normally closed check - valves, that have a safety function in the closed position, other than CIVs and PIVs, should have the closure function verified both during preoperational
testing and periodically throughout the plant life.
In the staff'.s normal
reviewofISTprogramswheneveravalveofthistypeisidenti{ied,thestaff ! veriffes that closure verification testing is specified in the IST program, l and if not, the staff either requires that the program be revised to so specify ' or the staff specifies in the IST SER that closure testing must be perforced.
. Even though that is the position that the staff has been implementing, the . , staff does believe that there is some ambiguity within ASME Section XI re- > ' , Check garding closure verification testing of normally closed check valves.
> valve testing is specified in paragraph IWV-3520 of the 1983 Edition of Section XI. Paragraph IWV-3522 " Exercising Procedure" requires that check valves be periodically exercised to the position required 'a fulfill their function.
< ' Testing intervals required vary from a minimum of every three months to each I Cold Shutdown.
(Earlier Section XI editions are essentially the same.)
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. . . -2- , 'i ' a The' Code ambiguity arises from the fact that paragraph IW-3522(b) "Normally Closed Valves discusses in detail the performanct of tests for periodic verification of the valve opening function but does not specifically mention periodic closure verification' tests.
In an attempt to have ASME clarify the ambiguity in the Code paragraphs, the staff submitted a related inquiry to the society. The response to the inquiry, , recently received, unfortunately was inconclusive. The staff inquiry was written in broad terms to cover both check and gate valves used in applications where the valve disk in the closed position was essential to the fullfillment of the valve's safety related function. The inquiry asked whether such valves sh'ould be categorized as A or AC and leak tested in accordance with paragraph ' IW-3420.
If the ASME response was "no" to the A or AC categorization and leak tests, it was hoped that the reply would at least confirm that the intent of IW-3522 for check valves and IW-3412 for gate valves was that some kind of closure verification test was to be performed. Unfortunately the response that was received is concentrated on the categorization aspect of the inquiry , and simply states that categorization is the Owner's (licensee's) responsibility.
Nevertheless, the staff personnel that participate in ASME Section XI standards writing activities were present at some of the meetings when the inquiry was The impression received at the meetings was that the intent of the discussed.
Code for both check valves and gate valves was that periodic verification of closure function is required for valves, whether normally open or closed, if they perform a safety function in the closed position. Verification would also be required during preoperational testing by paragraph IW-3100 " Pre-service Tests" which requires that all tests to be performed periodically ' during plant life per IW-3000 also be performed after installation and prior l to service.
One additional item that supports the requirement for periodic closure verification testing is that the latest draft of ANSI /ASME OM-10 " Inservice Testing of Valves" specifically requires that check valves be exercised or examined in a manner which veriffes obturator travel to the closed, full open, or partially open position required to fulfill its function. ANSI /ASME 0M-10 is generally expected to be an eventual replacement for subsection IW of,ASME Section XI.
In sunmary, the staff position is that normally closed check valves that have > a safety related function in the closed position should be tested for closure l capability during preoperational testing and periodically during plant life ! I I in accordance with the intervals specified in IW-3520.
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MAR 171980 ROI Rdg.
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. , 551?:5 tio. 6025Mp M.90RMDLM FOR: R. C. Lewis, Acting Chief. RO&tS Branch, Region !! , FR0n: Sanuel E. Brysn A/0 for Field Coordination OR01, IE SU5 JECT: OPERABILITY REQUIREMENTS FOR PW5 (AITS NO. F02-700028-N07) -
.. As we understand then, the questions in yet r february 1 memo are: . 1.
Do the Technical Specification ACTION statement time per10s run consecutive or concurrently with the data evaluation time (96 hours) given in IWP-3220 cf Section XI of the ASE Soller and pressure Vessel Code,1974 Edition with Addenda thru the Suurer 1975, and 2. 1;henshouldthetestresultsbersElewedand,ifout-of-specification, the associated puup declared inoperable? The answer to the first goestion is the Technical Specification ACTION state-
- ent tim period starts after the detensination is made that the pump is inoperable as defined inW1on XI SIP-3230fc). If the data is within the Required Action Ranoe of Table IW-3100-2 and it is decided to reca11brate the ir.strunents and rerun the test, as provided for in IW-3230(b), the Technical Specification ACTION statement time starts when the detamination is cade that the data is within the Required Action Ranfe. The reasoning behind the preceeding statement is that once the detem nation is made that the data is within the Required Action Range the pure must be declared inoperable. The provisions in IW-3230 to recalibrate and rerun the test to show the pwp is still capable of fulfilling its function are interpreted by us as an alternative to replacement or repair, not an additional actlen that can be taken before declaring the pump-inoperable.
The answer to the second question is that*as soon as the data is recognized as being withlo the Required Action Range the pop sust be declared inoperable.
state Nst the test Ian shall SectionXI,StP-6230,"InserviceTestPlans*bt)IsSubsection" include "The reference values ' Table IW -310 limits of Pg and 'Tatie IWP-3100-2). and any other values required by th Thi statement . " D
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J. C. Stone, IE me-e , ...... l 4p ........ i .. . .... i , , ~ .
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. . . . c e R. C. Lewis-2-MAR 171980 l l l l then requires the acceptance criteria to be included in the test plan.
With that infomation available, the shift supervisor should be able to make the detemination as to whether or not the data meets the requirements.
The important point is that once the data becomes available that shows the l pop cannot meet the inservice inspection requirements and by definition ) cannot fulfill its function then the pump must be declared inoperable.
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. . We have discussed the above interpretations with DOR personnel and the ' . Standard Technical Specification Group and they agree. If you have any - further questions, please call.
S auel E. Bryan - ,
Assistant Director . for Field Coordination Division of Reactor l Operations Inspection. IE l cc: II. C. Moseley IE J. 5. Wetmore. STS < G. Johnson, EB J. C. Stone, IE ~ - F. J, notan, IE J. I. Riesland IE 8. R. Nessitt, RII i E. J. Brunner, RI .
i i R. F. Heishman RI!! G.L.Madsen,d!Y J. L. Crews, RV i . . [..\\ (.! ? FC:R01:!E,h ADFC:ROI 00R :E[' ? f, , .. JC5 tone:LD.
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