ML021970132

NUREG/CP-0152, Vol. 4, (4:4) Page 3C-29 - End, Proceedings of Seventh Nrc/Asme Symposium on Valve & Pump Testing Held at Renaissance Washington, DC Hotel Washington, DC, July 15-18, 2002.

ML021970132
Person / Time
Site:	Palisades
Issue date:	07/31/2002
From:	Robinson M, Sagmoe R, Thomas Scarbrough Consumers Energy, K & M Consulting, Office of Nuclear Reactor Regulation, Division of Engineering
To:
References
-nr NUREG/CP-0152 V4
Download: ML021970132 (156)

Text

Check Valve Hinge and Disc Assembly Discovered Unassembled Roger Sagmoe, PalisadesNuclear PowerPlant Michael Robinson, K&M Consulting,Inc.

Background new testing was performed to provide this information. In almost all cases each test was On June 21, 2000, at the Palisades Nuclear backed up by multiple testing methods.

Power Plant, the High Safety Injection Pump In 2000, the NRC issued NRC Information P-66A failed to achieve its required flow Notice (IN) 2000-21, "Detached Check Valve reference value. Through evaluation, it was is Not Detected by Use of Acoustic and determined that the cause of this condition was Magnetic Nonintrusive Test Techniques." In that piston check valve number CK-ES3340 the summer of 2001, the Nuclear Industry located in the mini flow recirculation line was Check Valve Group (NIC) provided the stuck in a mid-stroke position. Check valve industry with guidance on this issue by CK-ES3340 has a safety function in both the developing an industry response to the open and closed positions.

IN. This response was sent out to all Once the cause was determined, a decision Vice Presidents and Managers at each nuclear was made to designate swing check site in the U.S.

valve number CK-ES3332 to provide the safety functions of open and closed. This Event Description determination was based on past Inservice On September 5, 2000, at 1820 hours0.0211 days <br />0.506 hours <br />0.00301 weeks <br />6.9251e-4 months <br />, a Testing (IST) and non-intrusive inspections radiography of check valve CK-ES3332 of CK-ES3332 in both the open and closed direction. in the train "A" common minimum flow recirculation line from high pressure safety In September 2000, radiography of the valve injection (HPSI) pump P-66A and low CK-ES3332 revealed that the valve internals pressure safety injection (LPSI) pump P-67A were not attached. The plant shut down and revealed that the check valve's disc/arm reviewed all of the safety-related check assembly was detached from the hinge pin and valves. This investigation verified that all was located in the bottom of the check valve the check valves had been tested such that body.

positive indication was provided for their Check valve CK-ES3332 was declared operational readiness. This was done by inoperable and technical specification 3.0.3 verifying that the test methods, had without was entered based upon the potential for a doubt, proven that the valve obturator is loose parts to affect additional components in intact and working correctly. If the testing the emergency core cooling system (ECCS).

method could not provide positive proof, then 3C-29 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing The plant was shutdown and depressurized further radiography of CK-ES3332 was no to 250 psi to effect repair-this isolated the longer considered immediately necessary.

mini flow lines to all ECCS pumps. Shutdown cooling was established with LPSI pumps in CK-ES3340 - Inspection Results their shutdown cooling (SDC) mode with flow through the PCS allowing for their mini flow. A Second Problem Arises The September 5, 2000, radiography of How Did We Get Here? CK-ES3332 was initiated with the intent of In June 2000, a problem with reduced HPSI increasing the knowledge of the condition pump P-66A recirculation flow had focused of the valve, based on a minimum amount attention on various check valves in the train of past data for it. The valve was not being "A" common minimum flow recirculation radiographed because it was suspected of line from HPSI pump P-66A and LPSI pump being failed.

P-67A. CK-ES3340 was determined to be A review of maintenance history, industry stuck in the mid-position. Note: CK-ES3332 operating experience, design and application is a significant valve due to its position in data revealed no problems.

the common flow path for recirculation flow from the right channel engineered safeguards The radiograph of CK-ES3332 performed on pumps, which include P-66A (HPSI), P-67A September 5, 2000, revealed that the check (LPSI), and P-54A (Containment Spray). valve's disc/arm assembly was detached from the hinge pin and was positioned in the bottom CK-ES3340 Nonintrusive Testing of the check valve body.

(NIT) Inconclusive CK-ES3332 Event Information The quarterly P-66-A HPSI pump test indicated reduced flow and the acoustic The initial supposition for the apparent NIT of CK-ES3340 was determined to be condition of CK-ES3332 was service induced inconclusive. Radiography of CK-ES3332 and failure. However, when CK-ES3332 was CK-ES3340 was parallel path with acoustic opened for inspection, it was discovered that analysis. the disc and hinge assembly, including the disk nut, disk washer and cotter pin, were First - Action on CK-ES3340 completely intact, laying in the bottom of the valve body and exhibiting no indication Radiography of CK-ES3332 was attempted of failure from service wear. Accordingly, it at that time to determine whether the valve was determined that the disc/arm assembly was contributing to the reduced recirculation had not been attached to the hinge pin. This flow. The radiography was inconclusive due condition has likely existed since original to inadequate radiation source strength used plant construction, dating back approximately for the radiography. Subsequently, upstream 30 years.

check valve CK-ES3340 was radiographed and was found to be partially open, (reference Safety Significance OE 11349) which explained the P-66A recirculation flow reduction symptom and CK-ES3332 has a safety function in the open direction to pass adequate minimum flow for NUREG/CP-0152, Vol. 4 3C-30

NRC/ASME Symposium on Valve and Pump Testing

2. No seat contact marks other than the initial HPSI Pump P-66A, LPSI Pump P-67A and CS bluing marks.

Pump P-54A. Observation over many years of pump operation and routine surveillance has 3. Because of the shape of the swing arm demonstrated that the as-found condition of casting, an interference was found to exist CK-ES3332 was not restricting recirculation between the hinge arm and valve body. If flow. the valve had previously seen any actual service, this interference would prevent Normally, CK-ES3332 has no safety function the valve disc from swinging open greater in the closed direction due to additional than 45 degrees and would have no upstream check valves CK-ES3340 and indication of any impacts on the hinge arm CK-ES3233 for HPSI Pump P-66A and LPSI back stop or valve body.

Pump P-67A, respectively. The upstream check valves are normally relied upon for 4. No indication of any impacts on the hinge closure in order to prevent the potential arm back stop or valve body.

over-pressurization of an idle pump's suction piping. 5. No evidence of rotation between the disc and disc stud.

Consequences of Taking Credit for 6. The cast side areas of the body hinge pin Closure of CK-ES3332 bosses were in the rough cast condition, no rub marks could be found as would be In the ten-day period between June 21, expected from the disc arm rubbing on this 2000, and July 2, 2000, CK-ES3332 was surface.

credited with the closed safety function when radiography identified that upstream HPSI check valve CK-ES3340 was stuck in Inspection of CK-ES3331 a mid-open position and, therefore, unable to provide the closed safety function. Prior to CK-ES3331 was inspected by boroscope and crediting CK-ES3332 with the closed safety found to be in excellent condition during the function, non-intrusive testing (acoustic and repair of CK-ES3332. The area where the dc magnetic testing) was performed, resulting tail-piece on the hinge arm contacts the body in "apparent" open and closed indications. was examined.

Based upon the as-found condition of It was clearly evident that they contact each CK-ES3332, it is apparent that the open and other-meaning that an interference fit does closed indications were caused by the disc/arm not exist. It was baseline tested with the assembly responding to changes in flow.

P-67B LPSI , using both dc magnetics and acoustic NIT methods.

Inspection of CK-ES3332 Detailed visual inspection of the condition of Risk Impact of Event all wear surfaces, conclusively determined The Palisades Probabilistic Safety Assessment that this valve had never been in service in a (PSA) was evaluated for the risk impact due fully assembled configuration. This conclusion to CK-ES3332 being unable to provide the is based on the following six inspection facts:

closed safety function. The only period during

1. No rotational indications on either the which the as-found condition of CK-ES3332 hinge pin or swing arm.

3C-31 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing would have caused potential concerns was was not ascertained. Though acoustic testing during the ten-day period when it was credited was performed, prior to discovering the with a closed safety function. Since neither disc laying on the bottom of the valve by CK-ES3332 nor CK-ES3340 were capable of radiograph testing (RT), results obtained closure during this time period, an evaluation from acoustic testing corresponded with the of possible operating and accident scenarios generically expected indications.

was performed to identify the maximum pressure that could be experienced in HPSI Lessons Learned Pump P-66A suction piping for comparison to design pressure ratings. The section of A review of industry experience for defi piping between HPSI Pump P-66A, upstream ciencies in the application of non-intrusive check valve CK-ES3183 and upstream branch testing (such as acoustic testing) was isolation valve CV-3071 was identified performed. Lessons learned from reviewing as having the potential to be pressurized the search material, in addition to conclusions beyond design pressure to a maximum of reached from this event, reinforce the need 1250 psi. While this section of piping is to use more than one confirmatory technique rated for 500 psi, and the aforementioned for valve condition when using nonintrusive valves are rated for 300 psi, evaluation has techniques.

concluded that the piping and valves would Another common theme noted is the need have maintained structural integrity under this for acquiring "good" baseline measurements increased pressure loading.

when using acoustic monitoring technology, i.e., the need to have reasonable assuredness Actions Taken of existing valve condition, that consistent test CK-ES3332 was inspected and reassembled, conditions are used, and that proper operation restoring it to its intended condition. An is established as part of the baselining process.

interference fit between the hinge arm and the body had to be corrected to allow the disk to Some points to make:

full open (some material was trimmed off the The level of baseline testing for each type of hinge arm).

check valve is not the same.

A restart review of all IST Program check " A one piece piston check would not loose valves was performed to ensure that an its disc into the system (generate loose adequate basis existed to conclude that each parts), whereas a swing check valve check valve is functioning properly. Where could loose the disc nut, hinge pin, hinge necessary, corroborating data was obtained arm, etc.

via additional testing. No other anomalies or degraded conditions were identified from this " Tilting disc checks generally do not have effort. a hinge arm that could move and impact the backstop, if the disc was dislodged Review of Previous Nonintrusive from the hinge pins, whereas if the disc Testing for CK-ES3332 separated from the hinge arm in a swing check, the hinge arm could still move and During two previous nonintrusive tests (11/97 hit the backstop.

and 6/00), the actual condition of CK-ES3332 NUREG/CP-0 152, Vol. 4 3C-32

NRC/ASME Symposium on Valve and Pump Testing When using NIT, it is also important for the Nonintrusives that are used should be test conditions to be repeatable so that the test selected to provide reasonable assurance results can be reviewed with prior tests. NIT of valve condition, such as when using acoustics, to collaborate impact data techniques need to be accurate and repeatable.

with backflow data (DP / leakage / flow), When NIT is to be used, it needs to be verified RT, UT, AC/DC Magnetics, or previous that the method being used will determine the inspection results. valve's function that is being detected. The qualification process may reveal that certain NIT techniques give inconclusive results for a Is There More Information We Are particular application.

Missing?

The NRC Information Notice (IN) 2000-21, The Palisades event was a very unique "Detached Check Valve Disc Not Detected By situation in that the disc was unassembled Use of Acoustic and Magnetic Nonintrusive from preservice days and that it had passed Test Techniques," concludes:

its operability flow test for 30 years. Having looked at the maintenance history (none found "If NIT techniques used to verify the opening except an acoustic NIT), industry operating or closing capability of safety-related check experience, design and application data, there valves are not properly qualified and a were no problems expected. The acoustic data baseline established for each individual valve by itself did not lead one to believe otherwise. when the valve is known to be operating acceptably, potentially inadequate valve Question-How many cases have actually performance may be undetectable in the been recorded where a single valve in a analysis of NIT results."

sample group degraded / acted drastically different from the group as a whole? To date An Industry Unified Response we have not come across any. Normally, the condition of one valve in the group is As a result of this IN, the Nuclear Industry representative of the whole. However, this Check Valve Group (NIC) provided an was not the case for the CK-ES3332 failure at industry response to this notice. A letter was Palisades. sent from NIC to all Site Vice Presidents and Managers. The letter was developed and voted NRC Concerns on by utility members that were present at the Summer 2001 Meeting.

The NRC considers NIT acceptable for in-service testing of check valves provided In addition to this letter, NIC continues to that the method used is qualified. If the owner move the industry forward on this and many should use NIT, they need to establish a other issues. Presently, NIC is conducting a performance baseline in both directions when Check Valve Performance Trending Initiative.

the check valve is in a known acceptable A recommended practice by the NRC and operating condition. A check valve's INPO is to trend check valve conditions, so performance can then be assessed against this that maintenance is performed prior to failure.

baseline. Both the NRC and industry have This NIC test initiative addresses this industry provided guidance on the use of NIT.

need. The scope of work proposed is to conduct testing to evaluate the capabilities of 3C-33 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve and Pump Testing various commercially available techniques and Part 10, paragraph 4.3.2.4, Valve Obturator technologies to trend parameters that would Movement and Paragraph 3.3, Reference reveal the internal condition of check valves. Values.

To effectively utilize these technologies, further verification of their capability to 4. ASME Inquiry OMI-00-08, June 2001 This inquiry is applicable to OMa-1988, trend parameters in detecting check valve Part 10, Paragraph 4.3.2.2, Exercising degradation is desired. The results of this Requirements; Paragraph 4.3.2.4, Valve initiative should allow utilities to demonstrate Obturator Movement; and Paragraph 3.3, that monitored and trended parameters are Reference Values.

repeatable, -reliableand defensible. Effective trending is expected to result in substantial 5. LER 50-225/00-04, "Discovery of reductions in both operation and maintenance Inoperable Check Valve Results in Plant costs. Shutdown," October 4, 2000 (Accession No. 9810270327).

References

6. NUREG-1482, "Guidelines for Inservice

1. NRC Information Notice 2000-21, Testing at Nuclear Power Plants," April "Detached Check Valve Disc Not Detected 1995.

By Use of Acoustic and Magnetic Nonintrusive Test Techniques." 7. "Evaluation of Nonintrusive Diagnostic Technologies for Check Valves (NIC-0 1),"

2. NRC Temporary Instruction 2515/110 Volume 1, February 1991, transmitted (Revision 1), Performance of Safety by a letter dated February 20, 1992, to Related Check Valves. Francis Grubelich, NRC, from the Nuclear Industry Check Valve Group (Accession

3. ASME Inquiry OMI-00-09, June 2001 No. 9205280219).

This inquiry is applicable to OMa-1988, NUREG/CP-0152, Vol. 4 3C-34

NRC/ASME Symposium on Valve and Pump Testing 900# Swing Check Valve SMILIFIW DRV FOR P-66A TEST DURING00-1EA HPSI pup 1 t st w CV-303 ES34 HPSI TRAIN2 SFW S140 MI-MT ES44 ESS3183

-I/ -,

HPSI pump 66A test 3C-35 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve and Pump Testing 2-inch T-pattern Piston Check Valve Radiograph of CK-ES3340 showing piston stuck open NUREG/CP-0152, Vol. 4 3C-36

NRC/ASME Symposium on Valve and Pump Testing View looking into cylinder / body showing fretting area at outlet port.

Radiograph of CK-ES3332 (side view) showing disk laying on bottom of valve. The seat side of the disk was facing up and the hinge arm was nearer the outlet port.

3C-37 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing cK-ES 3233 P-67A P.-SA SIIWT HPSI P-67B LPSI CK-ES3332 CK-ES3331 CKPES3330 EASTRPI "P

• srRML--I l--- HPS, CK-ES333S CK-ES3332 System Lineup Swing Check Valve CK-ES3332 Internals NUREG/CP-0152, Vol. 4 3C-38

tn PU

NRC/ASME Symposium on Valve andPump Testing CH -1: CK-ES3332 BACKSTOP w/ DC MAGNETIC OVERLAY - BP FILTER 1000 - 6500 HZ C4.-5 CHI -2: CK-ES3332 SEAT - BP FILTER 1000 - 6500 HZ

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Acoustic and DC magnetic traces of CK-ES3332 open test (credit for operability 6121100) using the P-66A, HPSI pump for flow. CK-ES3340 is stuck open at this time.

CH - 1: CK-ES3332 BACKSTOP w/ DC TRACE OVERLAY - BP FILTER 1000 - 5000 HZ 4 p Open Impacts 1

CH - 1: CK-ES3332 SEAT - BP FILTER 1000 - 5000 HZ

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s 'z Diu- .Wj, L.n, M:,) W Q W S -'o *14 Acoustic and DC magnetic traces of CK-ES3332 open test (credit for operability) using the P-66A, HPSI pump for flow. CK-ES3340 is stuck open at this time.

NUR* *G/CP-0152, Vol. 4 3C-40

NRC/ASME Symposium on Valve and Pump Testing CH -1: CK-ES3332 BACKSTOP - BP FILTER 1000 - 6500 HZ CH -2: CK-ES3332 SEAT - BP FILTER 1000 - 6500 HZ with DC TRACE OVERLAY Close Impact Acoustic and DC magnetic traces of CK-ES3332 close test (credit for operability) using the P-66A, HPSI pump for flow. CK-ES3340 is stuck open at this test.

CH -3: CK-ES3332 BACKSTOP - BP FILTER 1000 - 6000 HZ i7 ,

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.: 2)YA 2MA D-1 W- WD34ý'W S cW S Wax 6 00 6M0c 0- 7 S)S-- l9 '-3 CH -1: CK-ES3331 BACKSTOP - BP FILTER 1000 - 6000 HZ

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Open acoustic traces of CK-ES3332 and CK-ES3331 using the HPSI pumps for flow. The 11197 DAT data was downloaded into a newer version of software.

3C-41 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve and Pump Testing CH -3: CK-ES3332 SEAT - BP FILTER 1000 - 6000 HZ A_.W Seat impacting on pump stop CH -1: CK-ES3331 SEAT - BP FILTER 1000 - 6000 HZ 93

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• Close impact on pump stop IX.: " - 6 90*

=0 E.'Sm) VIM0 iil$ýl N les *:B 0**<

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.;. M 24,10 26 . TIN Close acoustic traces of CK-ES3332 and CK-ES3331 using the HPSI pumps for flow. The 11197 DAT data was downloaded into a newer version of software.

CH - 1: CK-ES3332 BACKSTOP - BP FILTER 1000 - 6000 HZ CH -2: CK-ES3332 SEAT - BP FILTER 1000 - 6000 HZ CH - 1: CK-ES3332 FFT OF BACKSTOP - BP FILTER 1000- 6000 HZ Acoustic and DC magnetic traces of CK-ES3332 opening after repair (baseline test) using the P-67A, LPSI pump for flow.

NUREG/CP-0 152,r Vol. 4 3C-42

NRC/ASME Symposium on Valve and Pump Testing CH - 1: CK-ES3332 BACKSTOP - BP FILTER 1000 - 6000 HZ CH -2: CK-ES3332 SEAT - BP FILTER 1000 - 6000 HZ Acoustic and DC magnetic traces of CK-ES3332 closing after repair (baseline test) using the P-67A, LPSI pump for flow.

Acoustic and DC magnetic traces of CK-ES3332 closing after repair (baseline test) using the P-67A, LPSI pump for flow 3C-43 NUREG/CP-0 152, Vol. 4

NRC/ASME Symposium on Valve and Pump Testing Nuclear Industry Check Valve Group June 7, 2001 Site VP

SUBJECT:

The Nuclear Industry Check Valve Group (NIC) Response to:

NRC INFORMATION NOTICE 2000-21 On December 15, 2000, the Nuclear Regulatory Commission (NRC) issued the subject Information Notice to Licensees. Although the Notice did not require response, the issues raised are of sufficient importance that NIC chooses to inform the members of its perspective.

NIC supports the continued use of nonintrusive testing. NIC performed, in the early 1990's, Phase 1, 2, & 3 studies that evaluated technologies that have been successfully demonstrated to assist in determining check valves operational readiness. and reliably Since then the NIC has successfully continued to demonstrate, improve, and refine the applications of these technologies.

NIC has provided various technical documents (Analysis Guide, Phase 1 through 3 reports, Flowtest, etc.) to help owners use and qualify nonintrusive technologies.

These reports strongly recommend the use of multiple technologies (in combination) to provide as much information as possible about the check valves operational readiness. When multiple technologies possible (or results are not conclusive), then the test should be augmented are not with other corroborating information. This information may be in the form of indications of proper operation, past disassembly and inspection, etc.

Part of the basis for determining operational readiness is having a baseline test when the valve is known to be operating acceptably. Establishment of a baseline requires supporting information to determine the capability of the valve to perform its intended function(s).

Application of these principles when using nonintrusive testing should help improve the ability of the nuclear industry to demonstrate check valve operational readiness.

Tony Maanavi NIC Chairman Exelon Nuclear Corporation Byron Nuclear Station CC: NIC Members & Associates Francis Grubelich, US NRC Joseph Colaocino. US NRC The Nuclear Industry Check Valve Group P.O. Box 4 Crum Lynn, PA 19022 NUREG/CP-0152, Vol. 4 3C-44

Valve Performance Solutions Determining Frictional and Dynamic Loads from In Situ Test Evaluations John Holstrom Altran Corporation Abstract all positions of travel are directly comparable.

This improves the ability of the investigator The main focus of this presentation will be to identify normal and abnormal loads, the methods that can be employed to separate anomalies and to quantify the effects of the frictional loads from dynamic loads in the load and operational changes.

typical industry static and dynamic testing Resulting data can be analyzed to obtain methods and how this data can be used to dimensionless engineering parameters predict operating loads at other dynamic to better predict such effects as flow conditions. A secondary focus will be on the characteristics, side loading and fluidynamic side loads and fluidynamic lift found in large lift (or torque).

diameter, angle pattern, balanced and sleeved globe valves. In-Situ test data from a large diameter sleeved and balanced globe valve will be used to The industry generally excepted test show how frictional loads, flow coefficients, information has been to show valve and system time history data and benchmarking side loading, and fluidynamic lift can be determined from observed data.

specific events in the output traces such as a zero transition, seat contact, peak seating and Some observed operational problems and unseating loads, torque switch trip and final solutions would also be provided.

output.

When these data are converted to position history rather than time history all events at 3C-45 NUREG/CP-0152, Vol. 4

Power Up-Rate Solutions MSIV Dynamic Stroke Time Evaluation John Holstrom Altran Corporation Abstract MSIVs to be used in the ASME overpressure analysis. This analysis created a mathematical The extended power uprate of two power model of the double-acting, spring assisted stations owned by Exelon Nuclear (Dresden actuator which included the hydraulic and Quad Cities stations) involved a speed control damper to calculate a realistic reanalysis of the ASME overpressure event to relationship of valve position, and time.

determine the ability to maintain the Technical The MSIV internal design was analyzed to Specification Safety Relief Valve Setpoint establish the flow area at each valve position.

Tolerance of +/-1%. This transient assumes that the reactor is operating at 102% of full These products were finally combined to power when Main Steam Isolation Valve establish a refined flow area versus time (MSIV) closure occurs. Anticipatory scrams relationship that could be used in the existing associated with MSIV closure is not assumed transient analysis model.

to occur. This results in a reactor scram on high reactor flux. Reactor pressure relief This presentation will explain the conditions occurs via lifting of the safety relief valves. that lead to the need for this approach, the methods of determining probable benefits, the When traditional analysis was applied to basic engineering methodology, and the results uprate conditions, the +/-1% safety relief valve from the analysis.

tolerance was found to be challenged. A review of the existing model found that the The model can be benchmarked against the MSIV closure profile used on the existing static test stroke time data. The dynamic transient model may have been overly conditions and loads can then be added to conservative. predict stroke time changes.

This analysis was performed to establish a realistic, yet bounding, closure profile for the 3C-47 NUREG/CP-0152, Vol. 4

Session 4 Regulatory Activities Update Session Chair Thomas G. Scarbrough U.S. Nuclear Regulatory Commission

Air-Operated Valve Performance and Inservice Testing Issues James Strnishaand Joseph Colaccino Mechanicaland Civil EngineeringBranch Office of Nuclear Reactor Regulation U.S. NuclearRegulatory Commission Abstract regard to AOVs has been focused on AOV performance.

This paper discusses current regulatory

Background

activities involving the inservice testing (IST) of air-operated valves (AOVs) in nuclear For the past several years, the NRC staff power plants. The paper addresses the scope has been working with industry groups and of AOVs to be included in IST programs, consensus bodies to monitor the development AOV-related Code cases approved by the of design basis verification and inservice NRC staff, and the status of current licensing testing programs for AOVs. In 1999, the reviews of risk-informed AOV programs. NRC met with the Joint Owners' Group on Air-Operated Valves (JOG-AOV) to discuss Introduction a voluntary industry program to address AOV issues. The JOG-AOV, which was AOVs are used in all U.S. light-water facilitated by the Nuclear Energy Institute reactor plants. They are used in a variety of (NEI), developed a risk-informed program applications, and the population of AOVs (Refs. 2 & 3) that established guidance for in each plant varies widely. The number of verifying AOV performance at design-basis AOVs in a plant can be over a thousand, and conditions and for performing long-term the number of safety-related AOVs per plant periodic verification of safety-related AOVs can be several hundred. Many plants have categorized as high-safety significant. The a number of AOVs that have an important JOG-AOV program also provided guidance role from a risk perspective but are not for a less-rigorous verification of AOV designated as "safety-related." The major functionality for those AOVs determined to safety concern identified as a result of a recent be low-risk significant. Although the NRC NRC study (Ref. 1) from a risk perspective staff did not formally review nor approve the is the simultaneous common-cause failure of JOG-AOV program, it did provide feedback AOVs which could disable redundant trains comments on the JOG-AOV program of a system important to safety. Most of the document in a letter to NEI dated October 8, recent NRC staff and industry attention with 1999 (Ref. 4).

This paper was prepared by staff of the U.S. Nuclear Regulatory Commission. It may prescnt information that does not currently represent an agrecd-upon NRC staff position. NRC has neither approved nor disapproved the technical content.

4-1 NUREG/CP-0 4 152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing On March 15, 2000, the NRC staff issued "Throughout the service life of a boiling Regulatory Issue Summary (RIS) 2000-03, or pressurized water-cooled nuclear power "Resolution of Generic Safety Issue 158: facility, pumps and valves which are classified Performance of Safety-Related Power as ASME Code Class 1, Class 2, and Class 3 Operated Valves Under Design Basis must meet the inservice test requirements ...set Conditions," (Ref. 5). The RIS discussed the forth in the ASME OM Code." ASME Code staff's intent to close out Generic Safety Issue Class 1 valves include all valves within the (GSI) 158 (Ref. 6) on the basis that current reactor coolant pressure boundary. Regulatory regulations provide adequate requirements Guide (RG) 1.26 (Ref. 10) provides guidelines to ensure verification of the design-basis for establishing the quality group classification capability of AOVs (and other power (and ASME Code classification) for water-,

operated valves) and that no new regulatory steam-, and radioactive-waste-containing requirements were needed. The RIS also components in nuclear power plants other than noted that the NRC staff would continue to those in the reactor coolant pressure boundary work with industry groups and to monitor (i.e., ASME Code Class 2 and 3 components).

licensees' activities to ensure that safety In 10 CFR 50.55a(b)(3), the NRC incorporates related AOVs (and other power-operated by reference the ASME OM Code, 1995 valves) will remain capable of performing Edition with the 1996 Addenda. ISTC 1.1 of their specified functions under design-basis the 1995 OM Code with the 1996 Addenda conditions and to provide a timely, effective, further defines the scope by stating that IST and efficient resolution of the concerns programs shall include active or passive regarding AOV performance. valves that are required to perform a specific function in shutting down the reactor to a safe Inservice Testing Program Scope for shutdown, in maintaining the safe shutdown Air-operated Valves (Aovs) condition, or in mitigating the consequences of an accident. The scope of the OM Code In establishing an IST program in accordance also covers pressure relief devices used for with the American Society of Mechanical protecting systems or portions of systems that Engineers (ASME) Boiler and Pressure perform a required safety-related function.

Vessel Code,Section XI (ASME Code) Therefore, the scope of valves to be included (Ref. 7) or the ASME Code for Operation in IST programs must include ASME Code and Maintenanceof NuclearPowerPlants Class 1, 2, and 3 valves that are covered (OM Code) (Ref. 8), a question that arises in ISTC 1.1 of the ASME OM Code. In frequently is, "What is the scope of AOVs addition, NUREG-1482, Section 2.2 (Ref. 11) that should be included in an IST program?" provides guidance for selecting valves for the This question becomes more complex when IST program.

a licensee is establishing a risk-informed IST program for AOVs. Based on the above requirements and guidelines, the licensee establishes the scope The requirement for the scope of valves to be of its IST program. The NRC retains the included in an IST program is addressed in option to verify the licensees' IST program Title 10 of the Code ofFederalRegulations scope by inspection. Many licensees also (10 CFR) (Ref. 9) in Section 50.55a(f). include augmented AOVs in their IST Specifically, 10 CFR 50.55a(f)(4) states, programs. Augmented AOVs in a licensees' NUREG/CP-0152, Vol. 4 4-2

NRC/ASME Symposium on Valve and Pump Testing and approval, provided all conditions listed IST program are AOVs which are outside in the regulatory guide are followed. Use of the scope of the program but are included in ASME Code Cases are voluntary. However, the IST program for testing purposes. These once they are implemented, they become valves are not required to meet ASME Code regulatory requirements with the same force testing requirements.

of law as ASME OM Code requirements When developing a risk-informed IST and NRC regulations. The draft DG-1089 program for AOVs using Code Case OMN-12, was published in the FederalRegister for "Alternative Requirements for Inservice public comments, and the 90 day comment Testing Using Risk Insights for Pneumatically period ended on March 25, 2002. The final and Hydraulically-Operated Valve Assemblies regulatory guide will be given a new number in Light-Water Reactor Power Plants," and is scheduled to be issued in September of (Ref. 12), a clear understanding of the 2002.

program scope is needed for successful Included in DG-1089 are two risk-informed implementation of the program. When using Code cases of particular interest to the this approach, the RI-IST program scope for AOV IST programs: Code Case OMN-3, AOVs is similar to the scope of current IST "Requirements for Safety Significance programs except that licensees must include Categorization of Components Using Risk non-ASME Code AOVs that are categorized Insights for Inservice Testing of Light Water as high-safety significant (HSS). Non-ASME Reactor Power Plants," (Ref. 14), and Code Code AOVs that are categorized as low-safety Case OMN- 12, "Alternative Requirements significant (LSS) components are not required for Inservice Testing Using Risk Insights for to be included in the RI-IST program, but if Pneumatically- and Hydraulically-Operated the licensee does choose to include these valve Valve Assemblies in Light-Water Reactor for testing purposes, they should be identified Power Plants." Code Case OMN-3 establishes to the NRC to avoid confusion at a later date if the methodology and process to categorize questions arise whether they must meet ASME components that are part of an ASME Code Code testing requirements.

risk-informed IST program into HSS and LSS components. Code Case OMN-12 establishes NRC Draft Regulatory Guide alternative AOV test strategies used in DG-1089 conjunction with Code Case OMN-3 risk On December 28, 2001, the NRC staff informed categorization.

issued a notice in the FederalRegisterof Revision 0 to Code Case OMN-3 was the availability of Draft Regulatory Guide published in the 1998 Edition of the ASME DG-1089, "Operation and Maintenance OM Code. In DG-1089, the NRC proposed Code Case Acceptability, ASME OM Code,"

four conditions on the use of Code Case (Ref. 13). DG-1089 is a new proposed OMN-3. Condition 1, which relates to regulatory guide that endorses ASME OM program scope, specifies that HSS components Code Cases that have been determined by must include non-ASME components the NRC to be acceptable alternatives to categorized as HSS (this is similar to the requirements of the ASME OM Code.

categorization of non-ASME components in Licensees may use the approved Code Cases Ref. 3).

without submitting a request for NRC review 4-3 NUREG/CP-0 152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing Revision 0 to Code Case OMN-12 was for air-operated valves to the NRC staff in published in 2001 Edition of the ASME a letter dated September 11, 2000 (Ref. 17).

OM Code. In DG-1089, the NRC proposed The B&W Owners' Group Topical Report eight conditions on the use of Code Case BAW-2359, "Demonstration Project to OMN-12. The conditions ensure technical Apply Risk-Informed Inservice Testing to philosophy consistent with Code Case Air-Operated Valves," (Ref. 18), which was OMN-1, "Alternative Rules for Preservice referenced in the Davis-Besse risk-informed and Inservice Testing of Certain Motor AOV program was submitted to the NRC staff Operated Valves Assemblies in Light-Water on July 14, 2001.

Reactor Power Plants," (Ref. 15), developed The Davis-Besse risk-informed AOV for motor-operated valves (MOVs). The conditions proposed in DG-1089 for HSS program was reviewed in detail by the staff AOVs would require licensees to (1) include and underwent several iterations. Due to a multitude of complications that arose a mix of static and dynamic testing that may in the review including higher priorities be altered when justified by evaluation of test data, (2) evaluate within five years or three both at Davis-Besse and at the NRC, the refueling outages adequacy of diagnostic test completion of the review was delayed. In the interval, (3) evaluate potential increases in meantime, the staff issued Draft Regulatory core damage frequency (CDF) and risk of Guide DG-1089 as previously discussed that interval extension to ensure consistency with proposed to approve Code Case OMN-12 NRC Regulatory Guide 1.174 (Ref. 16), and for RMST of AOVs with certain conditions.

Because the final version of DG-1089 is (4) evaluate degradation rate and capability scheduled to be issued in September 2002, margin to ensure AOVs remain capable of performing their design-basis functions until rather than continue with the risk-informed the next scheduled test. The conditions AOV review, the staff and licensee mutually proposed in DG-1089 for LSS AOVs would agreed that the most efficient and effective require that (1) AOVs remain capable of approach at this time was to withdraw the submittal and implement Code Case OMN-12 performing their design basis function until when DG-1089 is issued as a final regulatory the next scheduled test, (2) setpoints are based on direct dynamic test information, a guide. In this manner, the licensee may test-based methodology, or grouping with implement Code Case OMN- 12 without the need for NRC staff review and approval.

dynamically tested valves, (3) initial and periodic diagnostic tests are performed to The status of Topical Report BAW-2359 is verify setpoints, and (4) the operability of an uncertain at this time. The report may be AOV is evaluated if the valve does not satisfy overtaken by approval of Code Case OMN-12, the acceptance criteria. and the need for staff review of the report may be reassessed.

Status of Risk Informed Air Operated Valve Program Reviews References Two risk-informed AOV programs have 1. U.S. Nuclear Regulatory Commission, been formally submitted to the NRC staff for NUREG-1275, "Evaluation of Air review and approval. The licensee for the Operated Valves at U.S. Light-Water Davis-Besse nuclear power plant submitted Reactors," Volume 13, February 2000.

its proposed risk-informed testing program NUREG/CP-0152, Vol. 4 4-4

NRC/ASME Symposium on Valve andPump Testing Testing at Nuclear Power Plants," Section

2. Letter from D. Modeen to E. Imbro, 2.2, "Criteria for Selecting Pumps and "Joint Owners' Group Air Operated Valves for the IST Program," April 1995.

Valve Program Document, Revision 0,"

July 19, 1999. 12. ASME/American Nuclear Standards Institute (ANSI), Codefor Operation

3. Letter from D. Modeen to E. Imbro, and Maintenance of Nuclear Power "Joint Owners' Group Air Operated Plants,2001 Edition, Code Case Valve Program Document, Revision 1,"

OMN-12, "Alternate Requirements for March 27, 2001.

Inservice Testing Using Risk Insights for

4. Letter from E. Imbro to D. Modeen, Pneumatically and Hydraulically Operated "Comments on Joint Owners' Group Air Valve Assemblies in Light-Water Reactor Operated Valve Program Document," Power Plants, OM Code 1998, Subsection October 8, 1999. ISTC," New York, NY.

5. Regulatory Issue Summary (RIS) 2000-03, 13. U.S. Nuclear Regulatory Commission, "Resolution of Generic Safety Issue 158: Draft Regulatory Guide DG-1089, Performance of Safety-Related Power "Operation and Maintenance Code Operated Valves Under Design Basis Case Acceptability, ASME OM Code,"

Conditions," March 15, 2000. December 2001.

6. Generic Safety Issue (GSI) 158, 14. ASME/ANSI, Codefor Operationand "Performance of Safety- Related Power Maintenance of Nuclear PowerPlants, Operated Valves Under Design Basis 1998 Edition, Code Case OMN-3, Conditions." "Requirements for Safety Significance Categorization of Components Using Risk

7. American Society of Mechanical Insights for Inservice Testing of LWR Engineers (ASME), Boiler and Pressure Power Plants," New York, NY Vessel Code,Section XI, 1995 Edition and 1996 Addenda. 15. ASME/ANSI, Codefor Operationand Maintenanceof NuclearPower Plants,

8. ASME, Codefor Operationand 1999 Addenda, Code Case OM7N-1, Maintenance of Nuclear Power Plants, "Alternative Rules for Preservice and 1995 Edition and 1996 Addenda. Inservice Testing of Certain Motor Operated Valves Assemblies in Light

9. United States Code of Federal Water Reactor Power Plants," New York, Regulations, Title 10, "Energy," Part 50, NY.

"Domestic Licensing of Production and Utilization Facilities." 16. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.174, "An Approach

10. Regulatory Guide 1.26, "Quality Group for Using Probabilistic Risk Assessment In Classifications and Standards for Risk-Informed Decisions on Plant-Specific Water-, Steam-, and Radioactive-Waste Changes to the Licensing Basis," July Containing Components of Nuclear Power 1998.

Plants," March 1972.

17. Letter from FirstEnergy to U.S. Nuclear

11. U.S. Nuclear Regulatory Commission, Regulatory Commission, "Request to NUREG- 1482, "Guidelines for Inservice 4-5 NUREG/CP-0 152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing Implement a Risk-Informed Inservice 18. Babcock & Wilcox Topical Report BAW Testing Program," Docket No. 50-346, 2359, "Demonstration Project to Apply September 11, 2000. Risk-Informed Inservice Testing to Air Operated Valves," July 14, 2001.

NUREG/CP-0152, Vol. 4 4-6

Validation Approach for Valve Performance Prediction Methodologies Stephen G. Tingen and Thomas G. Scarbrough Mechanicaland Civil EngineeringBranch Division of Engineering Office of Nuclear ReactorRegulation U.S. Nuclear Regulatory Commission used in supporting the validation of those Abstract methodologies, and identifies key attributes to Since 1989, the NRC has reviewed several be addressed in presenting a well-supported programs established by nuclear power plant validation of a valve performance prediction licensees in response to Generic Letter (GL) methodology.

89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance," GL 95-07, I. Introduction "Pressure Locking and Thermal Binding of Nuclear power plant licensees, industry Safety-Related Power-Operated Gate Valves,"

groups and consultants develop methodologies GL 96-05, "Periodic Verification of the to provide a generic approach to address Design-Basis Capability of Safety-Related specific technical issues. The staff of the Motor-Operated Valves." During these U.S. Nuclear Regulatory Commission (NRC) reviews, the NRC has evaluated several may review these methodologies as part of methodologies developed by industry groups, evaluations of plant-specific activities or individual licensees, and consultants for the industry-wide programs. Since 1989, the prediction of the performance of valves under NRC has reviewed programs established by various system and ambient conditions. These nuclear power plant licensees in response to methodologies predicted valve performance in Generic Letter (GL) 89-10, "Safety-Related areas such as thrust and torque requirements Motor-Operated Valve Testing and to open and close motor-operated valves under Surveillance," GL 95-07, "Pressure Locking differential pressure and flow conditions, and Thermal Binding of Safety-Related uncertainty in those predicted operating requirements, and the thrust required to open Power-Operated Gate Valves," GL 96-05, "Periodic Verification of the Design-Basis a valve under pressure-locking conditions.

Capability of Safety-Related Motor-Operated This paper provides examples of the types Valves." During these reviews, the NRC of methodologies for predicting valve has evaluated several methodologies for the performance that have been reviewed by prediction of the performance of valves under the NRC, indicates the various approaches that does not currently represent an This paper was prepared by staff of the U.S. Nuclear Regulatory Commission. It may present information neither approved nor disapproved the technical content.

agrced-upon NRC position. NRC has 4-7 NUREG/CP-0152, Vol. 4

Testing Pump and Valve onSymposium NRC/ASME NRC/ASME Symposium on Valve and Pump Testing various conditions. These methodologies Institute (NEI) submitted Electric Power predicted valve performance in areas such as Research Institute (EPRI) Topical Report thrust and torque requirements to open and TR-103237, "EPRI MOV Performance close motor-operated valves (MOVs) under Prediction Program," to the NRC for its differential pressure and flow conditions, review and acceptance. EPRI developed the uncertainty in those predicted operating MOV Performance Prediction Methodology requirements, and the thrust required (PPM) for use by licensees in predicting to open a valve under pressure locking the thrust and torque required to operate conditions. In this paper, the authors discuss gate, globe, and butterfly valves under various methodologies for predicting valve dynamic flow conditions. The EPRI MOV performance that have been reviewed by PPM program included the development of the NRC. The paper includes the various improved methods for prediction or evaluation approaches used in supporting the validation of system flow parameters; gate, globe, and of those methodologies, and the key attributes butterfly valve performance; and motor to be addressed in presenting a well-supported actuator rate-of-loading effects (load sensitive validation of a valve performance prediction behavior). EPRI also performed testing to methodology. evaluate parameter separately (separate effects testing) to provide information for refining II. Electric Power Research Institute the gate valve model and rate-of-loading Mov Performance Prediction methods; and conducted numerous MOV Methodology tests to provide data for model and method development and validation, including flow In response to weaknesses in MOV loop testing, parametric flow loop testing performance, the NRC issued GL 89-10 on of butterfly valve disk designs, and plant June 28, 1989, to request that licensees ensure in-situ MOV testing. EPRI integrated the the capability of MOVs in safety-related individual models and methods into an overall systems to perform their intended functions methodology including a computer model and by reviewing MOV design bases, verifying implementation guide.

MOV switch settings initially and periodically, testing MOVs under design-basis conditions EPRI developed the PPM from fundamental where practicable, improving evaluations engineering principles related to MOV design of MOV failures and necessary corrective and operation including consideration of action, and trending MOV problems. The fluid and friction forces. EPRI based specific NRC requested that licensees complete their aspects of the MOV PPM (such as valve GL 89-10 programs within approximately internal friction coefficients) on the results three refueling outages or 5 years from the of separate effects testing. EPRI validated issuance of the generic letter. Subsequently, the individual models of the MOV PPM the NRC issued GL 96-05 to provide more (system, gate valve, globe valve, and butterfly detailed recommendations for the establish valve models) using applicable data from ment of long-term programs to verify the MOV flow tests. EPRI made adjustments design-basis capability of safety-related to the MOV PPM where determined to be MOVs on a periodic basis. appropriate based on MOV flow tests, such as including a 5% margin factor for gate In support of the effort by the nuclear industry valves manufactured by Borg Warner. EPRI to respond to GL 89-10, the Nuclear Energy performed an assessment of the integrated NUREG/CP-0152, Vol. 4 4-8

NRC/ASME Symposium on Valve and Pump Testing of its methodology in a technically defensible MOV PPM using flow loop and plant manner.

in-situ test data. EPRI provided detailed documentation of the development and III. EPRI Thrust Uncertainty assessment of the methodology. Method The NRC with its contractor (Idaho National EPRI has developed a supplemental Engineering and Environmental Laboratory methodology (referred to as the Thrust (INEEL)) evaluated the development of Uncertainty Method) in an effort to address the models used in the EPRI MOV PPM, potential conservatisms in the valve operating the application of test data to validate those requirements predicted by the EPRI MOV models, and the overall PPM assessment PPM. EPRI has presented the methodology conducted by EPRI. The NRC discussed to the NRC for approval in Addendum 2 the MOV PPM with EPRI in detail and to Topical Report TR-103237-R2, "EPRI provided written questions to EPRI on the Motor-Operated Valve (MOV) Performance development and application of the PPM. Prediction Program." The Thrust Uncertainty On March 15, 1996, the NRC issued a safety Method establishes an average conservatism evaluation (SE) finding that the EPRI MOV in the thrust predicted by the EPRI MOV PPM is an acceptable methodology with PPM to be necessary to operate gate valves certain conditions and limitations to predict under dynamic flow conditions. The the thrust or torque required to operate gate, Thrust Uncertainty Method then treats the globe, and butterfly valves within the scope conservatism as a random uncertainty that is of the program, and to bound the effects of statistically combined with other uncertainties.

load sensitive behavior on motor-actuator In this effort, EPRI compared the thrust thrust output. On February 20, 1997, the NRC required to operate sample gate valves during issued a supplement to the SE that accepted flow loop tests conducted as part of the EPRI methods developed by EPRI for two unique MOV Performance Prediction Program to gate valve designs to predict their operating the thrust requirement predicted by its MOV thrust requirements with certain conditions PPM. EPRI calculated an average prediction and limitations. ratio from the sample gate valves operated The application of solid engineering principles under either cold or hot water conditions.

with directly applicable test data represents EPRI specifies that the Thrust Uncertainty an effective manner in which to justify a Method is only applicable for predicting the methodology. In this case, the justification for thrust required to close gate valves.

the MOV PPM by EPRI reflected a technically At the outset of the review of the Thrust sound approach through the application of Uncertainty Method, the NRC noted several first engineering principles with separate areas of concern regarding the acceptability effects test data used to establish reasonable of the method during a public meeting on values for performance parameters. By the September 20, 2000. First, if the valves used use of first principles, EPRI was able to in calculating the conservatism of the EPRI present a clear description of its approach and MOV PPM as part of the Thrust Uncertainty resulting methodology to licensee personnel Method were not fully preconditioned, and the NRC. The valve performance data the thrust required to operate those valves obtained from specifically designed flow might increase with age. If so, the Thrust tests enabled EPRI to support the precision 4-9 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing Uncertainty Method might become inadequate of test data used in establishing an average to ensure the capability of those valves over prediction ratio.

time and service. Second, in that the EPRI MOV PPM was developed as a first-principles In an NEI submittal dated December 6, 2001, model rather than a statistical database model, EPRI indicated that several actions had it was not clear that sufficient test data are been taken to help support its development available to determine in a reliable manner and validation of the Thrust Uncertainty the conservatism of the EPRI MOV PPM for Method. For example, EPRI limits the Thrust a wide range of gate valve types and their Uncertainty Method to only cold water service conditions. Third, the validation of applications up to 150'F. Further, EPRI will the Thrust Uncertainty Method as described apply the median value of the prediction in Addendum 2 to the EPRI topical report ratios in predicting a nominal value for the did not provide a clear indication that the thrust required to close a gate valve under MOVs included in the validation effort would cold water conditions as part of the Thrust continue to be able to perform acceptably if Uncertainty Method. EPRI also presented their torque switches were set using the Thrust additional analysis regarding the Thrust Uncertainty Method. Uncertainty Method to address the remaining NRC concerns. The NRC is continuing its In an NEI submittal dated January 5, 2001, interaction with NEI and EPRI to complete the EPRI provided further information on its review of the Thrust Uncertainty Method.

Thrust Uncertainty Method that was discussed at a public meeting on October 18, 2001. At IV. Pressure Locking and Thermal the end of the meeting, the NRC stated that Binding Thrust Prediction several significant concerns remain regarding Methodologies the establishment and validation of the Thrust Uncertainty Method. For example, the data On August 17, 1995, the NRC issued used in the Thrust Uncertainty Method to GL 95-07 to request that licensees perform, establish an average prediction ratio for or confirm that they had previously determining a nominal value for the thrust performed, (1) evaluations of the operational required to close a gate valve represented a configurations of safety-related, power very small sample of the total population of operated gate valves for susceptibility to safety-related motor-operated gate valves in pressure locking and thermal binding; and the nuclear industry. Further, the non-normal (2) further analyses, and any needed corrective distribution of the prediction ratios of the actions, to ensure that safety-related power actual thrust required to close the sample gate operated gate valves that are susceptible valves under cold water conditions to the to pressure locking or thermal binding are EPRI MOV PPM thrust prediction reflected a capable of performing the safety functions median value higher than the mean value used within the current licensing basis of the for the average prediction ratio in the Thrust facility. Pressure locking can occur in Uncertainty Method. The NRC also noted flexible-wedge and double-disk gate valves that a significant concern existed regarding when pressure in the bonnet is higher than the viability of the Thrust Uncertainty Method the line pressure on both sides of a closed for gate valves operated under hot water disk and the valve actuator is not capable of conditions because of the minimal amount overcoming the additional thrust required as a result of the differential pressure. Thermal NUREG/CP-0 152, Vol. 4 4-10

NRC/ASME Symposium on Valve and Pump Testing pressure to be applied across the valve binding is generally associated with a solid disks. The total stem force required to open or flexible-wedge gate valve that is closed at a valve during pressure locking conditions is high temperature and is allowed to cool before determined from the unwedging load, vertical reopening is attempted such that mechanical pressure load, and pressure-lock load based interference occurs because of contraction of on total contact load minus the stem rejection the valve body on the disk wedge. load.

In response to GL 95-07, many licensees used The NRC review of the ComEd pressure a pressure-locking methodology developed by locking methodology focused on the test Commonwealth Edison Company (CornEd),

results that were used to validate the pressure which is now a member of Exelon, to locking methodology. The NRC verified that demonstrate that flexible wedge gate valves the quality of the testing accomplished by are capable of operating under pressure CoinEd to validate its methodology provided locking conditions. In a letter to the NRC meaningful and accurate test results. Actual dated May 24, 1996, CoinEd provided the pressure locking test results indicated that as test results from a 4-inch (1500-pound) the differential pressure between the bonnet Westinghouse valve; a 10-inch (900-pound) and the downstream (or upstream) side of Crane valve; and a 10-inch (300-pound) the valve increased, the stem thrust required Borg-Warner valve that were used to validate to open the pressure locked valve increased.

its pressure-locking methodology. A public The NRC verified that the ComEd pressure meeting was conducted on April 9, 1997, locking methodology results trended with to discuss the CornEd flexible wedge gate actual pressure locking test results. The valve pressure locking analytical method NRC also verified that actual coefficients of and validation testing. In a letter to the friction obtained during testing were used to NRC dated May 29, 1996, ComEd provided validate the methodology. The NRC and its additional information on its pressure-locking contractor (INEEL) tested a flexible wedge methodology. After May 29, 1996, the NRC gate valve under pressure-locking conditions, issued a number of safety evaluations on and used the test results to verify that the GL 95-07 submittals finding that the ComEd CornEd pressure-locking methodology methodology provides a technically sound accurately predicted the thrust required to basis for assuring that valves susceptible to open the valve. The results of this testing are pressure locking are capable of performing documented in NUREG/CR-66 11, "Results of their intended safety-related function.

Pressure Locking and Thermal Binding Tests The CornEd pressure-locking thrust prediction of Gate Valves." The NRC concluded that methodology is based on the Sixth Edition the ComEd pressure-locking methodology of Roark s Formulasfor Stress and Strain is acceptable for use provided that minimum (Young, Warren C., McGraw-Hill Book margins are applied between calculated Company, New York, NY, 1989). The valve pressure-locking thrust and actuator capability disk is assumed to act as two ideal disks and that diagnostic equipment accuracy and connected by the hub. The differential methodology limitations are applied. The pressure between the bonnet and the upstream NRC accepted reduced margins between side of the valve is averaged between the calculated pressure-locking thrust and actuator bonnet and the downstream side of the valve capability when using an enhanced version of to determine a pressure locking differential the ComEd methodology.

4-11 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve and Pump Testing In response to GL 95-07, several licensees are capable of performing their intended used a modified industry gate valve thrust safety-related function.

equation to predict the thrust required to open flexible wedge and double disk gate Pressure-locking tests sponsored by the NRC valves during pressure-locking conditions. were also conducted by INEEL on a double In this methodology, the total required force disk gate valve (NUREG/CR-66 11). Test to operate the valve during pressure-locking data demonstrated that the modified industry conditions is the sum of the vertical forces gate valve thrust equation underestimated resulting from differential-pressure loads the thrust required to open a pressure-locked across the two valve disks. Although a double disk gate valve; however, the results of number of licensees used this methodology in the equation properly trended with actual test their GL 95-07 submittals to the NRC, none of results. The NRC issued a number of safety the licensees validated the methodology with a evaluations on GL 95-07 submittals finding test program. For flexible wedge gate valves, that sizing the power actuator to satisfy the one licensee demonstrated that results of the modified industry gate valve thrust equation modified industry gate valve thrust equation provides reasonable assurance that double disk were more conservative than the results gate valves susceptible to pressure locking obtained from the CornEd pressure locking are capable of performing their intended methodology. In its GL 95-07 submittal to safety-related function provided that there the NRC, the results of the ComEd pressure is an appropriate margin between predicted locking-methodology were compared to the pressure-locking thrust and actuator capability.

results of modified gate valve methodology It would have been very difficult for the for the same valve and pressure-locking NRC to approve use of the modified industry conditions. gate valve thrust equation as an acceptable corrective action for pressure locking of Pressure locking tests sponsored by the NRC double disk gate valves without the use of the were conducted by INEEL on a flexible wedge test results in NUREG/CR-6611.

gate valve (NUREG/CR-661 1). Test data demonstrated that the modified industry gate In response to GL 95-07, several licensees valve calculation conservatively estimated proposed the use of a pressure locking the thrust required to open a pressure-locked thrust prediction methodology that the flexible wedge gate valve. Test data from a NRC was unable to approve. The NRC 4-inch Westinghouse valve and a 10- inch review of the test data used to validate the Crane valve were used by the NRC to acceptability of the proposed methodology demonstrate that the modified industry gate indicated that in some instances the proposed valve methodology conservatively estimated methodology underestimated the amount that thrust required to open a pressure-locked of thrust required to open several different flexible wedge gate valve. The NRC issued types of flexible wedge gate valves during a number of safety evaluations on GL 95-07 pressure-locking conditions. Validation of submittals finding that sizing the power the proposed pressure-locking prediction actuator to satisfy the modified industry gate methodology became further complicated valve thrust equation provides a technically because the actual disk friction factor was sound basis for assuring that flexible wedge not used to validate the methodology. The gate valves susceptible to pressure locking NRC believes that the disk friction factor is a critical parameter when validating any NUREG/CP-0152, Vol. 4 4-12

NRC/ASME Symposium on Valve and Pump Testing valve performance methodology, and it was during thermal-binding or pressure-locking conditions. However, adequate test data not clear to the NRC why a generic disk were not available to the NRC to evaluate the friction factor was used in lieu of the actual licensee's thrust prediction methodologies.

disk friction factor to validate the proposed Methods other than the proposed thermal pressure-locking methodology. Further, actual binding or pressure locking methodology pressure locking test results indicated that as were used to demonstrate that valves were the differential pressure between the bonnet capable of opening during thermal-binding or and the downstream (or upstream) side of pressure-locking conditions.

the valve increased, the stem thrust required to open the pressure locked valve increased.

V. Conclusion The proposed pressure-locking methodology predicted that the opposite would occur in The application of solid engineering principles that, as the differential pressure between the with directly applicable test data represents bonnet and downstream (or upstream) side of an effective manner in which to justify a the valve increased, the stem thrust predicted methodology. Actual test valve parameters to open the pressure locked valve decreased. such as disk friction factor, packing load, stem It was not apparent to the NRC why the thrust, test pressures and valve characteristics results of the proposed methodology were not should be used in the validation process consistent with the actual test results. Several whenever possible. Any inconsistencies or public meetings were conducted to discuss the anomalies between actual test results and proposed pressure-locking thrust prediction the methodology should be understood and methodology, and additional information on thoroughly explained. Typically, it is not the proposed pressure locking method was feasible for the NRC to review methodologies provided in several letters to the NRC. As as part of plant inspection activities because a result, the NRC was unable to approve the methodologies are generally too complex proposed pressure-locking methodology, and to perform a sufficiently detailed review licensees used other methods to demonstrate during the time period allotted for inspection that valves were capable opening during activities unless prior arrangements are made.

pressure-locking conditions. Licensees should work with their owners groups or NRC project manager to determine In response to GL 95-07, other licensees the most efficient approach in obtaining NRC proposed the use of a thermal binding acceptance of methodologies developed to or pressure-locking thrust prediction address specific technical issues.

methodologies that were developed to calculate the thrust required to open valves 4-13 NUREG/CP-0152, Vol. 4

Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves Thomas G. Scarbrough Mechanicaland Civil EngineeringBranch Office of Nuclear ReactorRegulation U.S. Nuclear Regulatory Commission that safety-related power-operated gate valves Abstract susceptible to pressure locking or thermal binding are capable of performing their safety Many fluid systems at nuclear power plants functions. Licensees of all active operating depend on the successful operation of reactor units have completed their programs to motor-operated valves (MOVs) in performing verify initially the design-basis capability of system safety functions. As a result of safety-related MOVs in response to GL 89-10, problems identified in the 1980s with MOV and to address potential pressure locking performance at nuclear power plants, the and thermal binding of safety-related power NRC issued Generic Letter (GL) 89-10, operated valves in response to GL 95-07.

"Safety-Related Motor-Operated Valve Testing Licensees are currently implementing their and Surveillance," and GL 96-05, "Periodic long-term MOV programs in response to Verification of the Design-Basis Capability GL 96-05. The NRC staff has completed its of Safety-Related Motor-Operated Valves,"

review of GL 96-05 programs established requesting that nuclear power plant licensees at individual nuclear plants through verify initially and periodically the design significant reliance on licensee commitments basis capability of MOVs in safety-related to implement the JOG program on MOV systems. In response to GL 96-05, the nuclear periodic verification. This paper discusses power plant owners groups developed an NRC staff activities regarding the periodic industry-wide Joint Owners Group (JOG) verification of the design-basis capability of program for periodic verification of the safety-related MOVs, and monitoring of the design-basis capability of safety-related nuclear industry's activities to ensure proper MOVs. In a safety evaluation, the NRC performance of safety-related MOVs.

accepted the JOG program as an industry-wide response to GL 96-05 with respect to age I. Introduction related valve degradation. The NRC issued GL 95-07, "Pressure Locking and Thermal Many fluid systems at nuclear power plants Binding of Safety-Related Power-Operated depend on the successful operation of Gate Valves," requesting that licensees ensure It may present information that does not currently represent an This paper was prepared by staff of the U.S. Nuclear Regulatory Commission.the technical content.

agreed-upon NRC staff position. NRC has neither approved nor disapproved 4-15 NUREG/CP-0 152, Vol. 4

NR C/ASME Symposium on Valve andPump Testing motor-operated valves (MOVs) in performing (OM Code) to supplement the quarterly their system safety functions. MOVs must MOV stroke-time testing specified in the be capable of operating under design ASME Code with a program to verify MOV basis conditions, which may include high design-basis capability on a periodic basis.

differential pressure and flow, high ambient temperature, and degraded motor voltage. Operating experience at nuclear power plants The design of the MOV must apply valid in the 1980s and 1990s revealed weaknesses engineering equations and parameters to in many activities associated with MOV ensure that the MOV will operate as intended performance. For example, some engineering during normal plant operations and design analyses used in the original sizing and basis events. Manufacturing, installation, setting of MOVs did not adequately predict preoperational testing, operation, inservice the thrust and torque required to open and testing (IST), maintenance, and replacement close valves under design-basis conditions.

must be conducted by trained personnel using Both regulatory and industry research proper procedures. Surveillance must be programs later confirmed the weakness performed and testing criteria must be applied in the initial design and qualification of on a soundly based frequency in a manner that MOVs. For example, the NRC Office of suitably detects questionable operability or Nuclear Regulatory Research sponsored degradation. Moreover, these activities must an extensive program at the Idaho National be monitored by a strong quality assurance Engineering and Environmental Laboratory program. (INEEL) to study the performance of MOVs under various flow, temperature, and voltage The regulations of the U.S. Nuclear conditions. In addition, the nuclear industry Regulatory Commission (NRC) require that sponsored a significant program by the components that are important to the safe Electric Power Research Institute (EPRI) to operation of a U.S. nuclear power plant develop a computer methodology to predict be treated in a manner that ensures their the performance of MOVs under a wide performance. Appendix A, "General Design range of operating conditions. Poor MOV Criteria for Nuclear Power Plants," and performance also resulted from shortcomings Appendix B, "Quality Assurance Criteria for in maintenance programs, such as inadequate Nuclear Power Plants and Fuel Reprocessing procedures and training. Further, testing of Plants," to Part 50 of Title 10 of the Code MOVs to measure valve stroke times under of FederalRegulations (10 CFR Part 50) zero differential-pressure and flow conditions contain broadly based requirements in this was shown not to detect certain deficiencies regard. In 10 CFR 50.55a, the NRC has that could prevent MOVs from performing required U.S. nuclear power plant licensees their safety functions under design-basis to implement provisions of the American conditions.

Society of Mechanical Engineers (ASME)

Boiler & Pressure Vessel Code (B&PV Code) II. Verification of MOV Design-Basis for testing of MOVs as part of their IST Capability programs. On September 22, 1999, the NRC revised 10 CFR 50.55a to require licensees In response to weaknesses in MOV implementing the 1995 Edition with the 1996 performance, the NRC staff issued Generic Addenda of the ASME Codefor Operation Letter (GL) 89-10 (June 28, 1989), "Safety and Maintenance of Nuclear PowerPlants Related Motor-Operated Valve Testing and NUREG/CP-0 152, Vol. 4 4-16

NRC/ASMAE Symposium on Valve and Pump Testing total thrust required to operate the valve; Surveillance." In GL 89-10, the NRC staff (6) torque, thrust, and motor operating requested that licensees ensure the capability parameters were needed to fully characterize of MOVs in safety-related systems to perform MOV performance; and (7) reliable use their intended functions by reviewing MOV of MOV diagnostic data requires accurate design bases, verifying MOV switch settings equipment and trained personnel. The NRC initially and periodically, testing MOVs under provided detailed test results in NUREG/

design-basis conditions where practicable, CR-5406 (October 1989), "BWR Reactor improving evaluations of MOV failures and Water Cleanup System Flexible Wedge Gate necessary corrective action, and trending Isolation Valve Qualification and High Energy MOV problems. The NRC staff requested that Flow Interruption Test;" NUREG/CR-5558 licensees complete their GL 89-10 programs (January 1991), "Generic Issue 87: Flexible within approximately three refueling outages Wedge Gate Valve Test Program;" NUREG/

or 5 years of the issuance of the generic letter.

CR-5720 (June 1992), "Motor-Operated In support of the regulatory activities to Valve Research Update;" and NUREG/CR ensure MOV design-basis capability, the 6100 (September 1995), "Gate Valve and NRC Office of Nuclear Regulatory Research Motor-Operator Research Findings." The identified areas in which research and analysis NRC summarizes some of the results of the were required to assist in evaluating MOV MOV research program in NRC Information programs at nuclear power plants. For Notice 90-40 (June 5, 1990), "Results of example, the NRC performed research to NRC-Sponsored Testing of Motor-Operated evaluate (1) performance of MOVs under Valves." Additional examples of MOV pump flow and blowdown conditions; research sponsored by the NRC are discussed (2) output of ac-powered and dc-powered later in this paper.

MOV motor actuators; (3) the increase in To assist nuclear power plant licensees in friction of aged samples of valve materials; responding to GL 89-10, EPRI developed the (4) methods to determine appropriate values MOV Performance Prediction Methodology for stem friction coefficient; (5) pressure (PPM) to determine dynamic thrust and torque locking and thermal binding of gate valves; requirements for gate, globe, and butterfly and (6) the effect of ambient temperature on valves based on first-principles of MOV stem lubricant performance. For example, the design and operation. EPRI described the NRC sponsored flow testing of several MOVs methodology in Topical Report TR-103237 by INEEL under normal flow and blowdown (Revision 2, April 1997), "EPRI MOV conditions. The testing revealed that (1) more Performance Prediction Program." The EPRI thrust was required to operate gate valves MOV PPM program included the development than predicted by standard industry methods; of improved methods for prediction and (2) some valves were internally damaged evaluation of system flow parameters; gate, under blowdown conditions and their globe, and butterfly valve performance; and operating requirements were unpredictable; motor-actuator rate-of-loading effects (load (3) static and low flow testing might not sensitive behavior). EPRI also performed predict valve performance under design-basis separate effects testing to provide information flow conditions; (4) during valve opening for refining the gate valve model and rate-of strokes, the highest thrust requirements might loading methods; and conducted numerous occur at unseating or in the flow stream; MOV tests to provide data for development (5) partial valve stroking did not reveal the 4-17 NUREG/CP-0152, Vol. 4

NRC/A SME Symposium on Valve and Pump Testing and validation of the models and methods, test results. The industry expended significant including flow loop testing, parametric flow resources to resolve the deficiencies in the loop testing of butterfly valve disk designs, design, qualification, and application of and in-situ MOV testing. EPRI integrated the safety-related MOVs that led to the issuance individual models and methods into an overall of GL 89-10. The results of the GL 89-10 methodology including a computer model and programs and their implementation include implementation guide. On March 15, 1996, (1) MOV sizing calculations and switch the NRC staff issued a safety evaluation (SE) settings have been revised to reflect actual accepting the EPRI MOV PPM with certain valve performance; (2) improved valve conditions and limitations. On February 20, performance prediction methods have been 1997, the staff issued a supplement to the SE developed; (3) valve internal dimensions on general issues and two unique gate valve are being addressed to provide assurance of designs. On April 20, 2001, the staff issued predictable gate valve performance under Supplement 2 to the SE addressing an update blowdown conditions; (4) friction coefficients of the computer model. in new or refurbished gate valves have been found to increase with service until a NRC Information Notice (IN) 96-48 plateau reached; (5) MOV output prediction (August 21, 1996), "Motor-Operated Valve methods have been updated; and (6) personnel Performance Issues," alerted licensees to training and maintenance practices have been lessons learned from the EPRI MOV program. improved. The NRC staff has evaluated Among the lessons learned were: (1) the thrust the MOV program at each nuclear plant requirements to operate some gate valves through onsite inspections of the design-basis under pump flow and blowdown conditions capability of safety-related MOVs. The NRC were higher than predicted by the valve staff has closed its review of GL 89-10 for manufacturers; (2) a potential exists for gate each active U.S. nuclear power plant.

valves to be damaged when operating under blowdown conditions such that the thrust III. Long-term Aspects of MOV requirements can be unpredictable; (3) the Performance effective flow area in some globe valves can be larger than expected and can cause thrust On September 18, 1996, the NRC staff requirements to be higher than predicted; and issued GL 96-05, "Periodic Verification (4) the friction coefficients for sliding surfaces of Design-Basis Capability of Safety in gate valves can increase with service before Related Motor-Operated Valves," to provide reaching a plateau. In IN 96-48, the staff recommendations for assuring the capability noted that some of the EPRI information is of safety-related MOVs to perform their applicable to gate, globe, and butterfly valves design-basis functions over the long term.

regardless of the type of actuator operating the In GL 96-05, the NRC staff requested that valve. licensees establish a program, or ensure the effectiveness of their current program, to Nuclear power plant licensees implemented verify on a periodic basis that safety-related the recommendations of GL 89-10 through a MOVs continue to be capable of performing combination of design-basis reviews, revision their safety functions within the current of MOV calculations and procedures, static licensing basis of the facility. The guidance and dynamic diagnostic testing, industry in GL 96-05 supersedes the guidance in sponsored research programs, and trending of GL 89-10 on long-term MOV programs.

NUREG/CP-0 152, Vol. 4 4-18

NRC/ASME Symposium on Valve andPump Testing Joint Owners Group (JOG) Program on MOV In GL 96-05, the NRC staff noted five Periodic Verification to obtain benefits from attributes of effective programs for periodic sharing information between licensees on verification of safety-related MOV design MOV performance. The participating owners basis capability at nuclear power plants:

groups are the Boiling Water Reactor Owners (1) A risk-informed approach may be used Group (BWROG), the Babcock & Wilcox to prioritize valve test activities, such as Owners Group (B&WOG), the Combustion frequency of individual valve tests and Engineering Owners Group (CEOG), and selection of valves to be tested. the Westinghouse Owners Group (WOG).

Elements of the JOG program include (1) an (2) The valve test program provides adequate "interim" MOV periodic verification program confidence that safety-related MOVs will for applicable licensees to use in response remain operable until the next scheduled to GL 96-05; (2) a 5-year dynamic testing test. program to identify potential age-related increases in required thrust and torque to (3) The importance of the valve is considered in determining an appropriate mix of operate gate, globe, and butterfly valves under exercising and diagnostic testing. In dynamic conditions; and (3) a long-term MOV establishing the mix of testing, the benefits diagnostic program to be based on information (such as identification of decreased thrust from the dynamic testing program. On output and increased thrust requirements) October 30, 1997, the NRC staff issued an and potential adverse effects (such as SE accepting the JOG Program on MOV accelerated aging or valve damage) Periodic Verification with certain conditions are considered when determining the and limitations. Most licensees committed to appropriate type of periodic verification implement the JOG program as part of their testing for each safety-related MOV. response to GL 96-05.

(4) All safety-related MOVs covered by the The NRC staff meets periodically with JOG GL 89-10 program are considered in the to discuss the status and results of the JOG development of the periodic verification program. General observations to date from program. The program includes safety the JOG program include (1) the dominant related MOVs that are assumed to be influence for valve factor increase in gate capable of returning to their safety valves is disassembly and reassembly position when placed in a position that of valves prior to testing; (2) for non prevents their safety system (or train) from disassembled gate valves, initially low valve performing its safety function; and the factors tend to increase and high valve factors system (or train) is not declared inoperable remain stable or decrease; (3) bearing friction when the MOVs are in their nonsafety degradation was not identified for butterfly position. valves with bronze bearings in treated water, or with non-bronze bearings in treated or (5) Valve performance and maintenance are untreated water systems; (4) significant evaluated and monitored, and the periodic verification program is periodically variation was found in bearing friction for adjusted as appropriate. butterfly valves with bronze bearings in untreated water systems; (5) balanced disk In response to GL 96-05, nuclear power plant globe valves demonstrated stable valve owners groups developed an industry-wide factors; and (6) unbalanced disk globe valves 4-19 NUREG/CP-0152, Vol. 4

NRCI/ASME Symposium on Valve andPump Testing demonstrated only small changes in valve efficiency might not be maintained at "run" factor. The JOG dynamic test program is efficiency published by the manufacturer; scheduled to be completed in October 2002, (2) degraded voltage effects can be greater but a few dynamic tests will be conducted than predicted by the square of the ratio after that date. JOG plans to submit a revised of actual to rated motor voltage; (3) some topical report describing the long-term MOV motors produce more torque output than periodic verification program following its predicted by their nameplate rating; and evaluation of the MOV dynamic test program (4) temperature effects on motor performance results. The NRC staff intends to prepare a appeared consistent with the Limitorque supplement to the SE on the JOG program guidance. The NRC study of ac-powered upon review of the revised topical report. MOV output is described in NUREG/CR 6478 (July 1997), "Motor-Operated Valve Licensees are applying risk insights in (MOV) Actuator Motor and Gearbox implementing their long-term MOV programs. Testing." The nuclear industry also eval In Topical Report NEDC 32264, "Application uated the output capability of ac-powered of Probabilistic Safety Assessment to Generic MOVs at several plants. In response to Letter 89-10 Implementation," BWROG the new information on ac-powered MOV describes a methodology to rank MOVs performance, Limitorque provided updated according to their relative importance to core guidance in its Technical Update 98-01 (May damage frequency and other considerations 15, 1998) and Supplement 1 (July 17, 1998) to be applied by an expert panel. On for the prediction of ac-powered MOV February 27, 1996, the NRC staff issued an SE motor actuator. The NRC alerted licensees accepting the BWROG methodology for risk to the new information on ac-powered MOV ranking MOVs with certain conditions and output in Supplement 1 (July 24, 1998) to limitations. On June 2, 1997, WOG submitted Information Notice 96-48. In its technical Engineering Report V-EC- 1658 (Revision 1) update, Limitorque also indicated that updated describing an MOV risk-ranking approach guidance for predicting the output capability for Westinghouse-design nuclear plants. On of dc-powered motor actuators would be April 14, 1998, the NRC staff issued an SE issued.

accepting the WOG methodology for risk ranking MOVs with certain conditions and Following the NRC review of ac-powered limitations. MOV performance, the NRC sponsored research at INEEL to study the performance As the JOG program focuses on potential of Limitorque dc-powered MOV motor increases in MOV operating requirements, actuators under various temperature and licensees address potential degradation in voltage conditions. For the Limitorque the output of MOV motor actuators by their dc-powered motor-actuator combinations plant-specific programs. In the late 1990s, tested, the research indicated that (1) ambient the NRC sponsored research at INEEL to temperature effects were more significant than study the performance of ac-powered MOV predicted; (2) use of a linear voltage factor motor actuators manufactured by Limitorque needs to consider reduced speed, increased Corporation, under various temperature motor temperature, and reduced motor output; and voltage conditions. For the Limitorque (3) stroke-time increase is significant for some ac-powered motor-actuator combinations dc-powered MOVs under loaded conditions; tested, the research indicated that (1) actuator and (4) actuator efficiency may fall below the NUREG/CP-0 152, Vol. 4 4-20

NRC/ASME Symposium on Valve andPump Testing In support of the NRC review of the JOG published "pullout" efficiency at low speed program, the NRC has sponsored studies at and high load conditions. The research results INEEL and Battelle Institute in Columbus, are provided in NUREG/CR-6620 (May Ohio, of the effects of aging on Stellite 6 1999), "Testing of de-Powered Actuators for which is used on sliding friction surfaces Motor-Operated Valves." in valves. The tests of specimens in On June 23, 2000, the BWROG forwarded environments of temperature, pressure, and Topical Report NEDC-32958 (March 2000), water chemistry typical of BWR nuclear "BWR Owners' Group dc Motor Performance plants were intended to determine the effects Methodology - Predicting Capability and of film buildup on seating surfaces and the Stroke Time in dc Motor-Operated Valves," impact of the film on valve performance. The to the NRC staff for information. On test results indicated that friction coefficients October 2, 2000, the BWROG recommended continue to increase with film thickness an implementation schedule of 12 months or and that friction coefficients decrease with the first refueling outage (whichever is later) subsequent valve strokes. For one selected for first priority MOVs (those with one- or test, specimens subjected to prior periodic two-cycle JOG static test frequencies), and strokes demonstrated a lower trend in the two refueling outages for second priority friction coefficients than those specimens MOVs (remaining GL 96-05 MOVs) with a that were not subject to periodic strokes.

start date of when the NRC acknowledged An independent evaluation of test results the methodology. On August 1, 2001, the indicated that the trends were valid, but NRC issued Regulatory Issue Summary that more data are needed to obtain precise (RIS) 2001-15, "Performance of dc-powered conclusions. The test results are provided Motor-Operated Valve Actuators," that in INEEL/EXT-99-00116 (April 1999),

informs licensees of the availability of "Summary and Evaluation of NRC-Sponsored improved industry guidance for predicting Stellite 6 Aging and Friction Tests." The NRC dc-powered NIOV actuator performance. is conducting limited additional research to In RIS 2001-15. the NRC staff stated that, verify the overall program results.

based on a sample review, the BWROG To provide additional support for the NRC methodology represents a reasonable approach review of long-term MOV programs, the NRC to improvement of past industry guidance is sponsoring an ongoing study at INEEL of for predicting dc-powered MOV stroke time the aging of stem lubricants and the effects and output. The staff considers the BWROG of ambient temperature on their lubricating methodology to be applicable to Boiling properties. Results to date have indicated Water Reactor (BWR) and Pressurized Water that the stem friction coefficient for some Reactor plants because of similarity in the lubricants can increase significantly under design and application of dc-powered MOVs.

high ambient temperature conditions. The With the availability of the new BWROG resulting increased stem friction coefficient methodology, the staff considers that the can cause a loss in the thrust delivered by the regulatory issue of adequate prediction of dc MOV motor actuator. The NRC summarizes powered MOV performance can be effectively the current results of the research in NUREG/

resolved through implementation of improved CR-6750 (October 2001), "Performance industry guidance.

of MOV Stem Lubricants at Elevated Temperature."

4-21 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve and Pump Testing Each U.S. nuclear power plant licensee Register notice (64 FR 51370) issuing the submitted a description of plans for periodic rule, the NRC discusses the implementation verification of the design-basis capability of MOV programs in response to GL 89-10 of safety-related MOVs in response to and GL 96-05 at nuclear power plants, and the GL 96-05. The NRC staff reviewed the requirement to supplement MOV stroke-time licensee submittals and conducted inspections testing.

of GL 96-05 programs at a sample of nuclear plants. The staff prepared an SE to document In response to concerns regarding the its review of the response to GL 96-05 by adequacy of MOV stroke-time testing, the each licensee. Where a licensee committed ASME Operations and Maintenance Code to implement the JOG program, the NRC Committee developed performance-based staff relied to a significant extent on that ASME Code Case OMN- 1, "Alternative Rules commitment in preparing the SE without the for Preservice and Inservice Testing of Certain need for plant-specific inspection activity Electric Motor Operated Valve Assemblies in in most instances. The NRC staff reviewed LWR Power Plants, OM Code 1995 Edition; GL 96-05 programs of licensees that did not Subsection ISTC." As an alternative to commit to the JOG program by a separate quarterly stroke-time testing, ASME Code process of submittals and inspections, as Case OMN-1 allows periodic exercising of all appropriate. The NRC has completed its safety-related MOVs once per refueling cycle review of GL 96-05 programs for each active and periodic diagnostic testing under static U.S. nuclear power plant. The NRC will or dynamic conditions, as appropriate, on a monitor the long-term MOV programs at U.S. frequency determined by MOV performance nuclear plants using Inspection Procedure in terms of margin and degradation rate.

62708, "Motor-Operated Valve Capability," as In GL 96-05, the NRC staff noted that the part of the NRC reactor oversight program. method in ASME Code Case OMN-1 could be used as part of a licensee's response to the IV. ASME Code Improvements for generic letter.

MOV Inservice Testing In the regulations, the NRC endorsed the use The ASME Code specifies that stroke-time of ASME Code Case OMN-1 as an acceptable testing of MOVs be conducted as part of the alternative to the quarterly MOV stroke-time IST programs of nuclear power plants on a testing specified in the ASME OM Code with- .

quarterly frequency where practical. The certain conditions. The NRC stated that, NRC and the industry have long recognized where a selected test interval for an MOV the limitations of stroke-time testing as a under ASME Code Case OMN-1 exceeds means of assessing the operational readiness 5 years, the licensee must evaluate information of MOVs to perform their design-basis obtained from valve testing during the initial safety functions. In the most recent revision 5-year period to validate assumptions made in to 10 CFR 50.55a, the NRC requires U.S. justifying the longer test interval. The NRC nuclear power plant licensees implementing also specified that licensees must evaluate the 1995 Edition with the 1996 Addenda of the the potential increase in risk associated with ASME OM Code to supplement the quarterly extending the quarterly exercise frequency MOV stroke-time testing specified in the Code for MOVs identified as having a high with a program to verify MOV design-basis safety significance. In the FederalRegister capability on a periodic basis. In the Federal notice, the NRC indicated that, as part of NUREG/CP-0 152, Vol. 4 4-22

NRC/ASME Symposium on Valve and Pump Testing capable of overcoming the additional thrust implementing ASME Code Case OMN-1, required as a result of the differential pressure.

licensees need to consider the benefits (such Thermal binding is generally associated with as identification of decreased thrust output a solid- or flexible-wedge gate valve that is and increased thrust requirements) and closed at high temperature and is allowed potential adverse effects (such as accelerated to cool before reopening is attempted such aging or valve damage) when determining that mechanical interference occurs because appropriate testing for each MOV. Also, the of contraction of the valve body on the NRC noted that the provisions of ASME Code disk wedge. On August 17, 1995, the NRC Case OMN-1 would satisfy the regulatory issued GL 95-07, "Pressure Locking and requirements for supplementing quarterly Thermal Binding of Safety-Related Power MOV stroke-time testing with the conditions Operated Gate Valves," to request that specified in the rule.

licensees perform, or confirm that they had The NRC staff has granted requests from previously performed, (1) evaluations of the several nuclear power plant licensees to apply operational configurations of safety-related, performance-based ASME Code Case OMN-1 power-operated (including motor-, air-,

as an alternative to the quarterly MOV stroke and hydraulically operated) gate valves for time testing in their particular ASME Code susceptibility to pressure locking and thermal of record. The NRC staff is completing a binding; and (2) further analyses, and any regulatory guide that proposes to accept on needed corrective actions, to ensure that a generic basis the use of ASME Code Case safety-related power-operated gate valves that OMN- 1 as an alternative to the MOV stroke are susceptible to pressure locking or thermal time test provisions of the ASME Code with binding are capable of performing their safety certain conditions. The regulatory guide functions within the current licensing basis of also proposes to accept ASME Code Case the facility.

OMN- 11, "Risk-Informed Testing of Motor NUREG/CR-6611 (May 1998), "Results of Operated Valves," with certain conditions that, Pressure Locking and Thermal Binding Tests when implemented in conjunction with Code of Gate Valves," describes testing sponsored Case OMN-1, provides emphasis on high-risk by the NRC Office of Nuclear Regulatory MOVs with relaxation of the test provisions Research at INEEL to study pressure locking for low-risk MOVs. Over the longer term,-___ and thermalbi-nding of gate-valves. The test ASME is preparing a mandatory appendix valves included a six-inch Walworth flexible to replace the quarterly MOV stroke-time wedge gate valve and a six-inch Anchor/

testing specified in the ASME Code with Darling double-disc gate valve. Both valves performance-based provisions similar to those were determined to be susceptible to pressure in ASME Code Case OMN-1.

locking. During the INEEL testing, heatup of the valve caused the bonnet to pressurize V. Pressure Locking and Thermal slowly until leakage was overcome and then Binding of Gate Valves to pressurize rapidly. Air pockets were found One typical method that "pressure locking" to remain trapped in the valve bonnet after can occur in flexible-wedge and double-disk both heatup and subsequent cooldown. No gate valves is when pressure in the bonnet is significant increase in thrust requirements was higher than the line pressure on both sides found during thermal binding tests for these of a closed disk and the valve actuator is not valves. A previous test program had revealed 4-23 NUREG/CP-0152, Vol. 4

NRC/A SME Symposium on Valve and Pump Testing a significant increase in unseating load under reactor units have completed their programs to thermal binding conditions. verify initially the design-basis capability of safety-related MOVs in response to GL 89-10, In reviewing the response of each licensee to and to address potential pressure locking GL 95-07, the NRC staff determined whether and thermal binding of safety-related power the licensee had performed appropriate operated valves in response to GL 95-07.

evaluations of the operational configurations Licensees are currently implementing their of safety-related power-operated gate valves long-term MOV programs in response to to identify valves that are susceptible to GL 96-05. The NRC staff has completed its pressure locking or thermal binding. The review of GL 96-05 programs established staff then determined whether the licensee at individual nuclear plants through had taken, or was scheduled to take, the significant reliance on licensee commitments appropriate corrective actions to ensure that to implement the JOG program on MOV these valves are capable of performing their periodic verification. In its regulations, the intended safety functions. As part of its NRC has directed licensees implementing review, the staff evaluated methodologies the ASME OM Code to supplement the developed by licensees to predict the thrust quarterly MOV stroke-time testing in their required to open flexible-wedge gate valves IST programs with a program to periodically under pressure locking conditions. The NRC verify MOV design-basis capability. The staff has completed its review of licensee NRC staff has granted requests from several responses to GL 95-07 through issuance of an licensees to apply performance-based ASME SE addressing each active U.S. nuclear power Code Case OMN-1 as an alternative to the plant. quarterly MOV stroke-time testing in their ASME Code of record. In its regulations, the VI. Conclusions NRC has accepted the use of ASME Code As a result of problems identified in the 1980s OMN-1 as an alternative to MOV stroke-time with MOV performance at nuclear power testing for licensees implementing the ASME plants, the NRC issued GLs 89-10 and 96-05 OM Code. The NRC staff is preparing a requesting that licensees verify initially and regulatory guide that proposes to accept on a periodic-qlly the design-basis capability of generic basis ASME Code Cases OMN- 1 and MOVs in safet'y-rclated systeim at nuclear ,.-i I foi- prforman-baed pproaches to power plants. In response to GL 96-05, the MOV testing together with the application of nuclear power plant owners groups developed risk insights. The NRC continues to monitor an industry-wide JOG program for periodic licensee activities related to the performance verification of the design-basis capability of of safety-related MOVs through the reactor safety-related MOVs. The NRC accepted the oversight program.

JOG program as an industry-wide response to GL 96-05 with respect to age-related valve VII. References degradation. The NRC issued GL 95-07 ASME, Boiler and Pressure Vessel Code.

requesting that licensees ensure that safety-related power-operated gate valves ASME, Codefor Operationand Maintenance susceptible to pressure locking or thermal of NuclearPower Plants.

binding are capable of performing their safety functions. Licensees of all active operating NUREG/CP-0152, Vol. 4 4-24

NRC/ASME Symposium on Valve andPump Testing NRC, Code of FederalRegulations, Title 10, ASME, Code Case OMN-1, "Alternative Part 50.

Rules for Preservice and Inservice Testing of Certain Electric Motor Operated Valve NRC, Generic Letter 89-10 (June 28, 1989),

Assemblies in LWR Power Plants, OM Code "Safety-Related Motor-Operated Valve 1995 Edition; Subsection ISTC." Testing and Surveillance."

ASME, Code Case OMN- 11, "Risk-Informed NRC, Generic Letter 95-07 (August 17, 1995),

Testing of Motor-Operated Valves." "Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves."

BWROG, Letter dated October 2, 2000, on BWR Owners' Group dc Motor Performance NRC, Generic Letter 96-05 (September 18, Methodology (ADAMS Accession # 1996), "Periodic Verification of Design-Basis ML003758535). Capability of Safety-Related Motor-Operated Valves."

BWROG, Topical Report NEDC-32264A (Revision 2, September 1996), "Application NRC, Information Notice 90-40 (June 5, of Probabilistic Safety Assessment to Generic 1990), "Results of NRC-Sponsored Testing of Letter 89-10 Implementation." Motor-Operated Valves."

BWROG, Topical Report NEDC-32958 NRC, Information Notice 96-48 (August 21, (March 2000), "BWR Owners' Group dc 1996), "Motor-Operated Valve Performance Motor Performance Methodology-Predicting Issues," and Supplement 1 (July 24, 1998),

Capability and Stroke Time in dc Motor "Motor-Operated Valve Performance Issues."

Operated Valves."

NRC, Inspection Procedure 62708, "Motor Electric Power Research Institute, Topical Operated Valve Capability."

Report TR 103237 (Revision 2, April 1997), "EPRI MOV Performance Prediction NRC, NUREG/CR-5406 (October 1989),

Program," non-proprietary version. "BWR Reactor Water Cleanup System Flexible Wedge Gate Isolation Valve Idaho National Engineering and Qualification and High Energy Flow Environmental Laboratory, INEEL/EXT Interruption Test."

99-00116 (April 1999), "Summary and Evaluation of NRC-Sponsored Stellite 6 NRC, NUREG/CR-5558 (January 1991),

Aging and Friction Tests." "Generic Issue 87: Flexible Wedge Gate Valve Test Program."

JOG, Topical Report MPR- 1807 (Revision 2, July 1997), "Joint BWR, Westinghouse NRC, NUREG/CR-5720 (June 1992), "Motor and Combustion Engineering Owners' Group Operated Valve Research Update."

Program on Motor-Operated Valve (MOV)

NRC, NUREG/CR-6100 (September 1995),

Periodic Verification."

"Gate Valve and Motor-Operator Research Limitorque Corporation, Technical Update Findings."

98-01 and Supplement 1 (July 17, 1998),

NRC, NUREG/CR-6478 (July 1997), "Motor "Actuator Output Torque Calculation."

Operated Valve (MOV) Actuator Motor and Gearbox Testing."

4-25 NUREG/CP-0 152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing NRC, NUREG/CR-6611 (May 1998), NRC, Safety Evaluation Supplement "Results of Pressure Locking and Thermal dated February 20, 1997, on EPRI MOV Binding Tests of Gate Valves." Performance Prediction Methodology (NUDOCS Accession # 9704300106).

NRC, NUREG/CR-6620 (May 1999),

"Testing of dc-Powered Actuators for Motor NRC, Safety Evaluation Supplement 2 dated Operated Valves." April 20, 2001, on EPRI MOV Performance Prediction Methodology (ADAMS Accession NRC, NUREG/CR-6750 (October 2001), # MLO11100121).

"Performance of MOV Stem Lubricants at Elevated Temperatures." NRC, Safety Evaluation dated October 30, 1997, on JOG Program on MOV Periodic NRC, Regulatory Issue Summary 200 1 Verification (NUDOCS Accession #

15 (August 1, 2001), "Performance of dc 9801160151).

powered Motor-Operated Valve Actuators."

Westinghouse Electric Company, Engineering NRC, Safety Evaluation dated March 15, Report V-EC- 1658-A (Revision 2, July 1998),

1996, on EPRI MOV Performance Prediction "Risk Ranking Approach for Motor-Operated Methodology (NUDOCS Accession # Valves in Response to Generic Letter 96-05."

9608070280).

NUREG/CP-0152, Vol. 4 4-26

Rulemaking Activities on Inservice Testing David Terao and Stephen G. Tingen Mechanicaland Civil EngineeringBranch Division of Engineering Office of NuclearReactor Regulation U.S. Nuclear Regulatory Commission the revised NRC approach for referencing Abstract ASME Code Cases; and NRC endorsement of The U.S. Nuclear Regulatory Commission significant new Code Cases.

(NRC) regulations in Section 50.55a of Title 10 of the Code of FederalRegulations I. Incorporation By Reference A (10 CFR 50.55a) establishes requirements Later Edition and Addenda of ASME for the application of codes and standards in Code the performance of inservice inspection and On August 3, 2001(66 FR 40626), the testing of components used in U.S. nuclear NRC published a proposed rule in the power plants. The NRC periodically updates FederalRegister that presented an amendment 10 CFR 50.55a to incorporate by reference to 10 CFR Part 50, "Domestic Licensing of recent editions and addenda to the American Production and Utilization Facilities," that Society of Mechanical Engineers (ASME) would have revised the requirements for Codefor Operation and Maintenance of construction, inservice inspection (ISI), and NuclearPower Plants (OM Code) for inservice testing (IST) of nuclear power plant inservice testing of pumps and valves used components. For construction, the proposed in U.S. nuclear power plants. The NRC amendment would have permitted the use is currently updating 10 CFR 50.55a to of the 1997 Addenda, 1998 Edition, 1999 incorporate by reference a recent edition to Addenda, and 2000 Addenda of Section III, the ASME OM Code. Further, the NRC is Division 1, of the ASME Boiler and Pressure revising the previous approach in referencing Vessel (BPV) Code for Class 1, Class 2, ASME Code Cases for use by nuclear power and Class 3 components with no new plant licensees as acceptable alternatives modifications or limitations. For ISI, the to the provisions of the ASME OM Code.

proposed amendment would have required This paper will present the status of current licensees to implement the 1997 Addenda, rulemakings and future rulemaking plans 1998 Edition, 1999 Addenda, and 2000 related to inservice testing of pumps and Addenda of Section XI of the ASME BPV valves; key aspects of recent rulemakings to incorporate by reference the ASME Code; It may present information that does not currently represent an This paper was prepared by staff of the U.S. Nuclear Regulatory Commission.the technical content.

agrced-upon NRC staff position. NRC has neither approved nor disapproved 4-27 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing Code, for Class 1, Class 2, Class 3, Class MC, licensees regarding the intent of the regulatory and Class CC components with modifications requirement for MOVs. However, to avoid and limitations. For IST, the proposed any potential confusion in the future, 10 CFR amendment would have required licensees to 50.55a(b)(3)(ii) is being revised to clarify that implement the 1997 Addenda, 1998 Edition, licensees must comply with the provisions of 1999 Addenda, and 2000 Addenda of the the ASME OM ISTC Code for testing MOVs.

ASME OM Code for Class 1, Class 2, and Class 3 pumps and valves with one new 10 CFR 50.55a(b)(3)(vi) in the proposed rule modification. would have required an exercise interval of 2 years for manual valves within the scope of Interested parties were invited to submit the ASME OM Code in lieu of the exercise written comments for consideration on the interval of 5 years specified in the 1999 proposed rule. Comments were received Addenda and the 2000 Addenda of the ASME from 17 separate sources on the proposed OM Code. The 1998 Edition of the ASME rule. These sources consisted of 10 utilities, OM Code specified an exercise interval of 4 service organizations, and 3 individuals. 3 months for manual valves within the scope In consideration of the public comments, of the Code. The 1999 Addenda to the ASME the NRC deleted or revised a number of OM Code revised ISTC-3540 to extend the modifications and limitations that were in the exercise frequency for manual valves to proposed rule in this final rule. The following 5 years, provided that adverse conditions public comments on the proposed rule pertain do not require more frequent testing. A to the ASME OM Code. number of commenters stated that 10 CFR 50.55a(b)(3)(vi) in the proposed rule should Comments on OM Code be withdrawn because sufficient justification exists to allow the extension of the exercise Although the technical requirements in interval for manual valves to 5 years. The 10 CFR 50.55a(b)(3)(ii) were not revised justification for the 5-year frequency is the in the proposed rule, several commenters simplicity of manual valves (limited number stated that the reference to motor-operated of failure causes) and that the ASME OM valve (MOV) stroke-time testing in Code allows other valves (safety and relief the existing 10 CFR 50.55a(b)(3)(ii) is valves) to be tested on a 5-year or longer confusing because there are other MOV frequencies. The NRC believes there is a test requirements in the ASME OM Code lack of operational data or experience to (such as position indication and seat leakage allow extending the exercise interval for testing) that are applicable in addition manual valves to 5 years. The NRC review of to stroke-time testing. The commenters licensee IST programs indicates that manual suggested that a licensee might incorrectly valves are exercised every 3 months except interpret 50.55a(b)(3)(ii) as requiring that in instances where it is impractical to operate only MOV stroke-time testing be performed valves during unit operation. Valves are then in accordance with the OM Code. The exercised when the unit is in a cold shutdown NRC believes the current regulation in condition, and the exercise frequency cannot 10 CFR 50.55a(b)(3)(ii) clearly states that exceed 2 years. Therefore, a 2-year interval licensees must meet all of the ASME Code for exercising manual valves is justified provisions for testing MOVs. The NRC is because the available manual valve exercise not aware of any misunderstanding among data supports the 2-year interval. The NRC NUREG/CP-0 152, Vol. 4 4-28

NRC/ASME Symposium on Valve and Pump Testing to specific portions of the standard if those has approved longer test intervals for other provisions are deemed to be "inconsistent types of valves in the ASME OM Code but with applicable law or otherwise impractical."

the longer test intervals include additional Furthermore, taking specific exceptions means to determine component degradation. furthers the Congressional intent of Federal For example, although the ASME OM Code reliance on voluntary consensus standards test strategy for Class 2 and 3 relief valves because it allows the adoption of substantial has a testing interval of 10 years, Class 2 portions of consensus standards without the and 3 relief valves are subject to grouping need to reject the standards in their entirety and sample expansion if there is a test because of limited provisions which are failure. Manual valves that are required to not acceptable to the agency. Moreover, be exercised are not subject grouping and there is no legislative history suggesting sample expansion. Furthermore, obstruction that Congress intended agencies to take an from silting or blockage, or corrosion of "all or nothing" approach to endorsement valve internals are possible failure modes of voluntary consensus standards under the for safety-related manual valves that are not Act, and the OMB guidance implementing applicable to other types of valves with longer Pub. L. 104-113 does not address the matter.

test intervals. Exercising manual valves Finally, there is legislative history on Pub.

minimizes both of these failure modes and L. 104-113 indicating that Congress did not also allows for more immediate detection if intend each agency to prepare lengthy reports an obstruction or corrosion induced failure justifying the agency's decision not to adopt occurs.

a voluntary consensus standard, much less an in-depth report detailing the reasons for Comments on Use of Consensus Standards each modification or limitation that an agency The National Technology Transfer and imposes on the use of a consensus standard.

Advancement Act of 1995, Public Law Several commenters stated that the large (Pub. L.) 104-113, requires agencies to use number of modifications and limitations in the technical standards that are developed or proposed rule is an indication that the NRC adopted by voluntary consensus standards participation in the development of the ASME bodies unless the use of such a standard is Code is not promoting the endorsement of the inconsistent with applicable law or is ASME Code in 10 CFR 50.55a as approved otherwise impractical. A number of by the consensus process. The commenters commenters stated that the NRC approval emphasized that the NRC representatives of the ASME Code with exceptions (i.e., participating in the ASME consensus process modifications and limitations) does not meet should voice concerns or propose alternative the spirit of Pub. L. 104-113. Although Pub. options, and cast negative votes when there L. 104-113 requires Federal agencies to use are technical and regulatory concerns.

industry consensus standards to the extent This would allow other members on the practical, it does not require Federal agencies committees to evaluate the NRC technical and to endorse a standard in its entirety, nor does regulatory concerns during the development it forbid Federal agencies from endorsing of the Code, and thereby, reduce the number industry consensus standards with limitations of modifications and limitations needed when or modifications. The law does not prohibit incorporating the ASME Code by reference in an agency from generally adopting a voluntary 10 CFR 50.55a. The commenters also stated consensus standard while taking exception 4-29 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve and Pump Testing that Code changes are based on more than and limitations in the final rule, limitations 30 years of plant operations and experience, or modifications were imposed on a small years of research into better ways to inspect fraction of the ASME Code non-editorial components or evaluate the results of changes published in 1997 through 2000.

inspection results, or the use of risk insights. Approximately 165 of the 448 non-editorial changes reviewed were considered reductions The NRC imposes limitations or modifications of Code requirements, and the NRC approved on the use of the consensus standards that all but a small fraction of these non-editorial are used in its regulatory process when the changes. In conclusion, the NRC finds consensus standard does not adequately the concern that the NRC participation in address a specific regulatory issue, the the development of the ASME Code is not standard is technically incorrect, or it is promoting the endorsement of the ASME inconsistent with current regulations. In Code as approved by the consensus process, is accordance with NRC internal procedures, not justified.

NRC representatives on ASME committees coordinate with other NRC to ensure that Comments on Backfit Requirements for the views of NRC representatives on ASME committees are consistent with the views Modifications and Limitations of the NRC. This coordination minimizes The NRC is not imposing or mandating any the need for modifications and limitations new requirements in the limitations and and, thus, reduces unnecessary regulatory modifications to Code provisions. It most burden. The NRC strives to develop technical instances, where limitations and modifications positions in a timely manner for use in the are imposed, the NRC requires the use of standards development process. However, provisions of the ASME Code that have been in instances when it is not practical for NRC previously approved. This is the case when to develop a position on an issue prior to those provisions have been unacceptably casting its vote, NRC representatives on changed in later ASME Code editions and ASME committees are authorized to use their addenda. Several modifications restrict the best judgement based on their experience, use of a new Code provision while allowing technical expertise, and discussion with other a relaxation in the use of an earlier Code NRC staff. The goal that the NRC develop a provision.

final technical position on every Code change prior to voting on the change on the Main A number of commenters stated that the NRC Committee level is not always achievable imposition of exceptions (i.e., modifications because of higher priority activities and and limitations) to the ASME Code are current NRC staffing levels. backfits and should be analyzed in accordance with the regulations in 10 CFR 50.109. To The NRC reviewed approximately 448 non the contrary, the NRC finds that many of the editorial Code changes during the rulemaking modifications and limitations imposed during process to incorporate by reference the 1997 previous routine updates of 10 CFR 50.55a Addenda, 1998 Edition, 1999 Addenda, and have not been considered backfits. The final 2000 Addenda of Section III and Section XI rule dated August 6, 1992 (57 FR 34666),

of the ASME BPV Code and the ASME OM incorporated by reference in 10 CFR 50.55a Code. Although it may appear that there the 1986 Addenda through the 1989 Edition of are a significant number of modifications Section III and Section XI of the ASME BPV NUREG/CP-0 152, Vol. 4 4-30

NRC/ASME Symposium on Valve andPump Testing not considered by the NRC to be backfits. In Code. The backfit analysis section of the final conclusion, modifications and limitations have rule (57 FR 34672) stated that a modification historically not been considered to be backfits that simply retains an existing Section X1 unless they expand the scope of the Code to requirement is not a backfit. The final rule include components that were not considered also added a requirement to expedite the to be within the scope of ISI, or expedite the implementation of the revised reactor vessel implementation of new Code provisions.

shell weld examinations in the 1989 Edition of Section XI. Imposing these examinations was Limitations are also used to restrict the use of considered a backfit because licensees were a new Code provision while expanding the use required to.implement the examinations prior of an earlier Code provision. For example, to the next 120-month ISI program inspection 10 CFR 50.55a(b)(3)(vi) in the proposed rule interval update. prohibits the extension of the exercise interval for manual valves from 3 months (existing The final rule dated August 8, 1996 Code provision) to 5 years (new Code (61 FR 41303), incorporated by reference provision). 10 CFR 50.55a(b)(3)(vi) requires in 10 CFR 50.55a the 1992 Edition with that manual valves be exercised every 2 years.

the 1992 Addenda of IWE and IWL of In resolving this issue, the NRC could have Section XI to require that containments be retained the existing Code requirement to routinely inspected to detect defects that exercise manual valves every 3 months.

could compromise a containment's structural However, the intent of the ASME consensus integrity. This action was considered a process was to extend the exercise interval backfit because the Commission endorsed for manual valves, and in this case, the NRC new subsections of the Code that expanded is accommodating the ASME consensus the scope of 10 CFR 50.55a to include process to the extent that the NRC believes components that were not considered by the the extended exercise interval to 2 years is existing regulations to be within the scope of justified.

ISI. The final rule dated September 22, 1999 (64 FR 51370), incorporated by reference in In conclusion, modifications and limitations 10 CFR 50.55a the 1989 Addenda through the are not considered backfits because they either 1996 Addenda of Section III and Section XI retain existing Code provisions that have of the ASME BPV Code, and the 1995 been previously approved by the NRC, or are Edition with the 1996 Addenda of the ASME a compromise between new and old Code OM Code. The final rule expedited the provisions. Furthermore, the final rules dated implementation of the 1995 Edition with the September 22, 1999 (64 FR 51370), August 1996 Addenda of Appendix VIII of Section XI 8, 1996 (61 FR 41303), and August 6, 1992 for qualification of personnel and procedures (57 FR 34666), were reviewed by the NRC's for performing UT examinations. The Committee to Review Generic Requirements expedited implementation of Appendix VIII prior to publication to ensure that backfits are was considered a backfit because licensees identified and dispositioned in accordance were required to implement the new with the requirements in 10 CFR 50.109.

requirements in Appendix VIII prior to the next 120-month ISI program inspection interval update. The final rule also imposed modifications and limitations that retained existing ASME Code requirements that were 4-31 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve andPump Testing II. Incorporation By Reference of implement as alternatives to the requirements "Code Case" Regulatory Guides in the ASME BVP Code. The NRC has found some Code cases unacceptable and has noted The NRC is proposing to revise its approach their unacceptability in the RGs. The NRC for approving ASME Code cases in order to revises these RGs as new Code cases are fully satisfy the Administrative Procedures published. Additionally, the reference to RGs Act (APA) (5 USC 553) and 1 CFR Part 51, in Footnote 6 does not give revision numbers "Incorporation by Reference." The NRC of the RGs as also required by 1 CFR Part 51.

is proposing to amend NRC regulations in 10 CFR 50.55a to incorporate by reference the Furthermore, the NRC incorporates by NRC's regulatory guides (RGs) that address reference various portions of the ASME the use of Code cases prepared for the ASME BPV and OM Code requirements in 10 CFR BPV Code and OM Code. These "Code Case" 50.55a. Because these Code cases are usually regulatory guides currently are designated as alternatives to ASME Code requirements and RG 1.84, 1.85, and 1.147. not interpretations of how the requirements may be met, it is not permissible to use the RG To date the NRC practice has been to process to approve licensee implementation review ASME BPV Code cases, assess the of alternatives to these requirements. The acceptability of each, and issue regulatory approval to use these Code cases must be guides providing its conclusions on the granted on a plant-specific basis or through acceptability of the Code cases. The NRC has rulemaking. Although the RGs are issued referenced these RGs in Footnote 6 of 10 CFR for public comment, general reference to the 50.55a. Footnote 6 reads as follows: RGs addressing the ASME Code Cases in ASME Code cases that have been Footnote 6 of 10 CFR 50.55a could be viewed determined suitable for use by the as contrary to the requirements of the APA, Commission are listed in NRC Regulatory which requires that the public be given the Guide 1.84, "Design and Fabrication opportunity to review, comment, and receive Code Case Acceptability-ASME Section appropriate consideration of their comments III Division 1," NRC Regulatory Guide prior to the imposition of Federal regulations.

1.85, "Materials Code Case Acceptability ASME Section III Division 1," and The NRC held many internal discussions on 1.147, "Inservice Inspection Code this matter in order to reach a decision on Case Acceptability-ASME Section XI how to endorse ASME Code cases in the most Division 1." The use of other Code cases efficient and effective manner that met Federal may authorized by the Director of the procedural requirements. The NRC also held Office of Nuclear Reactor Regulation upon public meetings with external stakeholders request pursuant to §50.55a(a)(3). to discuss the issue and obtain feedback on various approaches. As a result of these many Recently, it has come to the NRC's attention discussions, the NRC concluded that the most that specific incorporation by reference by the effective and efficient approach for permitting Office of Federal Register (OFR) of these RGs licensees to use Code cases as alternatives has not previously been approved as required to ASME Code requirements would be to by 1 CFR Part 51. The NRC deemed many of incorporate by reference the RGs that list the Code cases listed in these RGs acceptable acceptable, conditionally acceptable, and (some with limitations) for licensees to unacceptable Code cases into 10 CFR 50.55a.

NUREG/CP-0152, Vol. 4 4-32

NRC/ASME Symposium on Valve andPump Testing officials from the OFR, this approach would This would give the Code cases the same meet OFR requirements for incorporation by legal status as the portions of the ASME Code reference of documents in the regulations.

that are currently incorporated by reference The change in the Code case approval process in 10 CFR 50.55a. The approach would be will be seamless to licensees and would retain accomplished through rulemaking by making a process with which licensees are already the following revisions to 10 CFR 50.55a: familiar.

1. A new paragraph, 50.55a(i), containing In addition, this approach would meet NRC's the language of incorporation by reference performance goal of maintaining safety would be added to 10 CFR 50.55a. This by continuing to provide NRC review and paragraph would identify each Code case approval of new ASME Code cases. It would RG by title and revision number.

reduce unnecessary regulatory burden by

2. Footnote 6 would be removed in its eliminating the need for licensees to submit entirety. Note that Footnote 6 also plant-specific relief requests for NRC review contains the statement that the use of other and approval. It would also increase public Code cases may be authorized by the confidence by allowing public participation Director of the Office of Nuclear Reactor in the process used to update the NRC's Regulation. However, this provision is regulatory guides that approve, condition, or also contained in 10 CFR 50.55a(a)(3). reject ASME Code cases as alternatives to the Thus, its deletion from Footnote 6 will provisions of the ASME Code requirements.

have no impact.

The approach described above was discussed

3. There are currently 12 references to in SECY-01-0110, "Initiation of NRR Footnote 6 in 10 CFR 50.55a. Because Sponsored Rulemaking: ASME BPV and each footnote reference would be deleted, OM Code Cases," dated June 21, 2001.

a cross-reference to the appropriate The Commission approved the NRC's portion of proposed paragraph (i) would recommended approach in a staff require be added with a statement that pursuant to ments memorandum dated July 6, 2001. The 10 CFR 50.55a(i), licensees may use the proposed rule was issued on March 19, 2002 Code cases that the NRC has found to be (67 FR 12488).

acceptable or conditionally acceptable as alternatives to the provisions in the ASME In summary, the NRC believes that this Codes. approach is a reasonable and legally sound approach that will eliminate the litigious risks Adopting this approach would establish a associated with the existing approach. This process of periodic rulemakings to incorporate option is responsive to the industry's desire by reference the latest regulatory guides which for generic approval of ASME Code cases list all acceptable, conditionally acceptable, and is consistent with NRC's performance and unacceptable ASME Code cases in goals in that it maintains safety, makes 10 CFR 50.55a. This approach would provide more efficient use of NRC's and licensee's a sound regulatory basis for NRC's approval resources by eliminating the need for plant of the generic use of Code cases by licensees specific reviews, and provides an opportunity as alternatives to the provisions of the ASME for public involvement.

Codes as incorporated by reference in NRC's regulations. Based on consultations with 4-33 NUREG/CP-0152, Vol. 4

NRC/ASME Symposium on Valve and Pump Testing Revisions to NRC's Code Case Regulatory RG 1.85 continue to be used by licensees. The Guides title of RG 1.84 will be changed to reflect the In conjunction with the Footnote 6 rulemaking scopes of both RGs ("Design, Fabrication, and described above, the NRC is preparing its next Materials Code Case Acceptability-ASME revisions to RGs 1.84, 1.85, and 1.147. There Section III, Division 1").

are several major changes to these RGs and approvals of significant, new Code cases that There are no major changes to RG 1.147 will appear in these next revisions and are (Section XI ISI) other than to update the list of worth mentioning. Code cases to include the latest ASME Code,Section XI ISI Code cases.

The first major change is the combining of RG 1.84 (Section III design and fabrication) The second major change to the Code Case with RG 1.85 (Section III materials). RGs is the introduction of a new (draft)

Beginning with Revision 32, all Section III regulatory guide addressing OM Code case nuclear component Code cases that have been acceptability. Draft Regulatory Guide DG approved for use by the NRC will be listed in 1089, "Operation and Maintenance Code Case one regulatory guide. For this revision (32), Acceptability-ASME OM Code," is the first the NRC reviewed Section III Code cases time that OM Code cases will be endorsed in listed in Supplement 4 to the 1992 Edition a regulatory guide. The need for an OM Code through those listed in Supplement 10 to the case RG became apparent to the NRC when 1998 Edition (except for those Code cases the NRC incorporated by reference for the related to elevated-temperature, gas-cooled first time the OM Code in a final rulemaking and liquid-metal reactors;Section III Division issued on September 22, 1999 (64 FR 51370).

2 components; and submerged spent fuel OM Code Cases OMN-1 through OMIN-13 waste casks). This will be accomplished by were reviewed for inclusion in this draft RG.

placing all Section III design, fabrication, The Code Case RGs were issued for and materials Code cases into RG 1.84. It public comment on December 28, 2001 should be noted that RG 1.85 will no longer (66 FR 67335). The major changes to be updated, but it will not be withdrawn at this the Code Case RGs discussed above are time because some Code cases contained in summarized in Table 1 below.

Table 1 - Summary of Changes to Code Case Regulatory Guides NRC's Approval Document ASME Code Cases Current Proposed Section III RG 1.84 (design and fabrication) RG 1.84 (design, fabrication, RG 1.85 (materials) and materials) Rev. 32 Section XI RG 1.147 (ISI) RG 1.147 (ISI) Rev.13 OM Code none new RG (draft DG-1089)

NUREG/CP-0152, Vol. 4 4-34

NRC/ASME Symposium on Valve and Pump Testing It should be noted that many of the OM Code III. Conclusion cases approved by the NRC in the draft RG The final rule to update 10 CFR 50.55a implement risk-informed alternatives to IST to incorporate by reference a more recent requirements for pumps and valves. These edition and addenda to the ASME OM Code Code cases may be used by licensees (when is scheduled to be issued in September 2002.

the RG is issued in final form) without a need The next update to 10 CFR 50.55a will to request NRC review and approval provided incorporate by reference the 2001 Edition, they are used with any conditions as identified 2002 Addenda, and 2003 Addenda of the in the final RG. With the incorporation by Section III, Division 1, and Section XI of reference of the OM Code Case RG (draft the ASME BPV Codes and the ASME OM DG-1089), if a licensee voluntarily elects to Code. The final rule will become effective use the Code Case, the conditions specified 60 days from date of publication in the in the RG are regulatory requirements, not FederalRegister The final rule to amend the guidance or recommendations. regulations in 10 CFR 50.55a to incorporate OM Code cases that have not yet been by reference the NRC's RGs that address the reviewed and approved by the NRC in the use of Code Cases prepared for the ASME draft RG may be implemented pursuant to BPV Code and OM Code is scheduled to be 10 CFR 50.55a(a)(3) which permits the use issued in March 2003. The next revision to of alternatives to the regulations in §50.55a Code Case RGs 1.84, 1.85, and 1.147 are provided that the proposed alternative can be scheduled to be issued at the same time as the final rule.

demonstrated to provide an acceptable level of quality and safety and its use is authorized by NRC's Director of the Office of Nuclear Reactor Regulation.

4-35 NUREG/CP-0152, Vol. 4

1. REPORT NUMBER U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT (Assigned byNUMBER NRC, Add Vol, Supp., Rev.,

NRC FORM 335 and Addendum Numbers, if any.)

(2-89)

NRCM 1102, BIBUOGRAPHIC DATA ;HEET 3201,3202 (See instructions on the reverse) NUREG/CP-0152, Vol. 4

2. TITLE AND SUBTITLE and Pump Testing 3. DATE REPORT PUBLISHED Proceedings of the Seventh NRC/ASME Symposium on Valve YEAR MONTH I July 2002

4. FIN OR GRANT NUMBER Ii 6. TYPE OF REPORT

6. TYPE OF REPORT

5. AUTHOR(S)

Conference Proceedings Editor: T. G. Scarbrough

7. PERIOD COVERED (Inclusive Dates)

Office or Region, U.S. NuclearRegulatory Commission, andmailing address; if contractor,

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (IfNRC, provide Division, provide name and mailing address-)

U.S. Nuclear Regulatory Commission cind Engi neers Board on Nuclear Codes and Standards of the American Society of Mechanical Commission,

'Same as above'; if contractor,provideNRC Division, Office or Region, U.S. NuclearRegulatory

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (IfNRC, type andmailing address.)

Divis ion of Engineering, Washington, DC 20555-0001 U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, York, NY 10016-5990 ASME Board on Nuclear Codes and Standards, Three Park Avenue, New

10. SUPPLEMENTARY NOTES

11. ABSTRACT (200 words or less)

Board on Nuclear Codes and Standards of the The 2002 Symposium on Valve and Pump Testing, jointly sponsored by the and by the U.S. Nuclear Regulatory Coin mission, provides a forum for exchanging American Society of Mechanical Engineers v alves and pumps used in nuclear power plants.

information on technical and regulatory issues associated with the testing of to discuss the need to improve that testi ng to help ensure the reliable performance of The symposium provides an opportunity pe rsonnel, and consultants ensures the discussion valves and pumps. The participation of industry representatives, regulatory and methods at nuclear power regarding the improvement of test ing programs of a broad spectrum of ideas and perspectives plants.

- I i3.

AvAILABILITY STATEMENT

13. AVAILABILITY STATEMENT

12. KEY WORDS/DESCRIPTORS (List words or phrasesthat will assist researchers in locating the report.) unlimited

14. SECURITY CLASSIFICATION Inservice Testing (This Page)

Valves unclassified Pumps Motor-Operated Valves (This Report)

Risk-Informed Inservice Testing unclassified Air-Operated Valves 15. NUMBER OF PAGES ASME Boiler & Pressure Vessel Code ASME Code for Operation and Maintenance of Nuclear Power Plants

16. PRICE NRC FORM 335 (2-89)

Federal Recycling Program UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, DC 20555-0001 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE, $300