ML20138A464
| ML20138A464 | |
| Person / Time | |
|---|---|
| Issue date: | 10/04/1985 |
| From: | Taylor J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Stello V Committee To Review Generic Requirements |
| Shared Package | |
| ML20137Z209 | List: |
| References | |
| NUDOCS 8510080568 | |
| Download: ML20138A464 (33) | |
Text
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ge ucg'o UNITED STATES 8'
- 1 NUCLEAR REGULATORY COMMISSION
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WASHING TON, D. C. 20555 t
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OCT 84 985 s
MEMORANDUM FOR:
Victor Stello, Jr., Chairman Committee to Review Generic Requirements FROM:
James M. Taylor, Director Office of Inspection and Enforcement
SUBJECT:
REQUEST FOR CRGR REVIEW 0F PROPOSED IE BULLETIN 85-XX:
MOTOR-OPERATED VALVE COMMON MODE FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS Please schedule a CRGR review of the proposed subject bulletin (Enclosure 1).
This bulletin is considered to be a Priority B item as specified in the CRGR charter. A draf t copy of this bulletin was sent to all reactor vendor owner's groups, the AIF and INP0.
Coments from these groups were incorporated into the bulletin, where appropriate.
The proposed bulletin requests all holders of operating licenses or construction permits to determine the proper switch settings for certain safety-related valves, to implement these settings and to then test these valves under the highest differential pressure expected during normal or abnormal events, within the design basis, unless otherwise justified.
Implementation of this bulletin would help prevent the type of common mode valve failure that occurred at Davis Besse Nuclear Plant with valves AF599 and AF608.
s Evaluation of design basis in-situ performance testing of safety-related valves is addressed in Generic Issue II.E.6.1, In-Situ Testing of Valves.
This issue is scheduled to be resolved by May 1988.
IE proposes to issue the enclosed bulletin prior to the resolution of Generic Issue II.E.6.1 because comon mode -
failure of safety-related valves is a major concern in the bulletin and this failure mode does not appear to have been evaluated when Generic Issue II.E.6.1 was prioritized as Medium.
The Interim Reliability Evaluation Program (IREP) studies used in the prioritization of Generic Issue II.E.6.1 also apparently did not explicitly evaluate common mode valve failures.
In addition, there is nothing required in the proposed bulletin which potentially conflicts with the resolution of Generic Issue II.E.6.1.
IE anticipates that actions required by this bulletin can be used in the resolution of Generic issue II.E.6.1.
Based on the actions requested in the proposed bulletin, we estimate the industry cost over the life of the plants to be 1.9 million (1985) dollars and the NRC to incur a one-time cost of 268 thousand dollars.
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Victor Stello, Jr. The implementation of the actions required by the proposed bulletin would result in a significant reduction in core melt frequency and radioactive release.
The cost-benefit analysis is provided in Enclosure 3.
Original Signed ',i James f.1. Tayla, James M. Taylor, Director Office of Inspection and Enforcement-
Enclosures:
1.
Proposed IE Bulletin 2.
Responses To Section IV B of CRGR Charter 3.
Cost-Benefit Analysis 4.
Temporary Instructions (TI) for NRC Resident Inspectors cc:
H. Denton, NRR J. Heltemes, AEOD G. Cunninghan, ELD Distribution:
DCS WMinners, NRR EAB R/F OParr, NRR DEPER RF BShelton, AR4 JTaylor, IE SScott, ADM RVollmer, IE ELJordan, IE SASchwartz, IE CERossi, IE RBaer, IE HBAiley, IE RKiessel, IE BKGrimes, IE JGPartlow, IE DEisenhut, NRR HThompson, NRR RBernero, NRR d
FMiraglia, NRR TSpeis, NRR WRussell, NRR GHolahan, NRR
- SEE PREVIOUS CONCURRENCES
- DEPER:!E
- DEPER:!E DEPER:!E g
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ENCLOSURE 1 SSINS No.:
OMB No.:
3150-0011 Expiration Date: 08/01/88 IEB 85-UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555
, 1985 IE BULLETIN N0. 85-XX: MOTOR-0PERATED VALVE COMMON MODE FAILURES DURING PLANT TRANSIENTS DUE TO IMPROPER SWITCH SETTINGS Addressees:
All holders of nuclear power reactor operating licenses (0Ls) or construction permits (cps) for action.
Purpose:
The purpose of this bulletin is to request licensees to develop and implement a program to ensure that switch settings on certain safety-related motor-operated valves are selected, set, maintained correctly, and tested where' appropriate to accommodate the highest differential pressures expected on these valves during both normal and abnormal events within the design basis.
Description of Circumstances:
There have been two recent events, and a number of earlier events, during which motor-operated valves failed, on demand, in a common mode due to improper switch settings.
Event 1, Davis-Besse Plant - On June 9, 1985, the Davis-Besse Plant experienced a complete loss of main and auxiliary feedwater.
This event was described previously in IE Information Notice No. 85-50, " Complete loss of Main and Auxil-iary Feedwater at a PWR Designed by Babcock & Wilcox," and in NUREG-1154, " Loss.
of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9,1985."
Nonnally open, Limitorque motor-operated auxiliary feedwater (AFW) gate valves failed to reopen on either an automatic or manual signal from the main control room after they were inadvertently closed during the event. While other failures also occurred in the AFW system, the failure of these two valves was itself enough to prevent AFW from reaching either steam generator. During the recovery from this event, the valves were opened with the handwheels.
The results of licensee troubleshooting activities after the event led to the conclusion that the setting for the torque switch bypass limit (torque bypass) switch in each valve's control circuit had not been set to remain closed long enough to provide the necessary bypass function on valve opening with differen-I tial pressure conditions across the valve. During the event, the valves experienced a high dif ferential pressure af ter closing.
The torque bypass switch on both valves was improperly set.
Therefore, the torque switches were I
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IEB 85-
, 1985 Page 2 of still open when the torque bypass switches dropped out. This condition stopped valve travel before the valve discs were fully off their seats.
The torque bypass switches were set to drop out af ter the valves opened to 5%
full-stroke.
The 5% full-stroke setting was based on a number of handwheel turns.
In a 10 CFR 21 report, submitted subsequent to the event, Toledo Edison Company identified two reasons why the torque bypass switch settings were not adequate:
(1) the 5% full-stroke settings were not adequate for unseating the AFW system discharge valves with large differential pressures across the valves and (2) the procedure for setting this switch was defective in that the 5% full-stroke was not specified to be in addition to the handwheel turns required to take up the motor-operator coast and backlash. The torque bypass switch setting errors were revealed only when high differential pressure conditions across the valves caused higher loadings. The valve failures were reproduced during tests performed by the licensee with differential pressures applied across the valves. During the tests, the valves operated properly when low differential pressures were applied across them, but failed to open when high differential pressures (loadings) were applied.
The valves were instrumented during these tests to obtain signature traces of critical parameters.
Event 2, Sequoyah Plant Unit 2 - An event involving partial loss of main feedwater occurred on May 2, 1985, at Sequoyah Nuclear Plant Unit 2 while in Mode 2 and returning to power af ter a reactor trip.
Feedwater was being sup-plied through the main feedwater (MFW) system isolation valve bypass lines.
Operators attempted to open the MFW system isolation valves to supply water to the steam generators; however, two of the four MFW isolation valves would not open.
The startup was discontinued and the unit was returned to hot shutdown.
During examination to determine the reason for the valve failures, the licensee discovered that both valve stems had sheared from their discs.
The discs were found in the closed positions within the valve seats.
The stems had suffered fracture failures through approximately three-quarters of the diameter of the shaft:, in addition to stress failures of the remaining quarter.
The Limitorque motor-operators on the valves use limit switches to control valve motion in the open direction.
These MFW system isolation valves are large (18 inch diameter),
fast acting (154 inches per minute travel speed) valves. Because of the high speed of these valves and the large mass of the discs, the selection of the limit switch setpoint needs to account for the large momentum of the disc and its continued motion af ter the limit switch deenergizes the valve motor-operator.
The setpoint was not correctly established and the disc impacted the backseat during opening.
The failure mechanism of these valves was identified by the licensee to be impact loading of the stem on the opening stroke as a result of the disc impacting the backseat, combined with a stress failure of the remaining portion of the stem on the opening stroke.
NRC Field Evaluation - As a part of the resolution of Generic issue II.L.6, "In-Situ Testing of Valves," the fiRC contracted with the Oak Ridge National Laboratory in 1984 to perform a limited study to determine the ef f ectiveness of signature tracing techniques in determining the operational readiness of
IEB 85-
, 1985 Page 3 of safety-related motor-operated valves.
It was hoped also that this study could provide some insight as to current conditions of valve switch settings at nuclear power plants. Signature traces of motor current, torque and limit switch actuations and axial motion of the worm gear (an indication of operator torque) were obtained from 36 motor-operated valves at 4 nuclear plant sites.
Although the formal technical letter report [0RNL/NRC/LTR-85-13. " Evaluation of the Motor-0perated Valve Analysis and Test System (MOVATS) to Detect Degradation.
Incorrect Adjustments, and Other Abnormalities in Motor-0perated Valves"] has not been issued, the current draf t of the letter report indicates that (1) this inspection method can be used to improve current ASME methods and (2) there were abnormalities with nearly every valve tested.
Table I contains a summary of the study's findings with respect to switch setting abnormalities. Of particular interest with respect to the events described above is the finding that 75% of the valves had irproperly set torque bypass switches (50% of the valves had the close-to-open torque bypass switch set so that it was opening before the valve fully unseated) and 8% of the valves were unintentionally backseating. The abnormalities in Table I have not been fully evaluated at this time, and they should not be interpreted to mean that any abnormality resulted in an inoperable valve.
Background:
The NRC has previously identified common mode failures, on demand, of valves.
IE Circular No. 77-01, " Malfunction of Limitorque Valve Operators," reported that on October 28, 1976, two motor-operated (Limitorque) valves located between the refueling water storage tank and the charging pump suction at the Trojan Nuclear Plant failed to open in response to a spurious safety injection (SI) signal.
The malfunction in both valves resulted from the torque switch in the opening circuit becoming activated before the valves were fully off their seats.
The valves also were equipped with a torque bypass switch. Each of the valves that malfunctioned was found to have its torque bypass switch adjusted such that it allowed the torque switch to be operable in the circuit before the valve was moved from its seat.
The Licensee's investigation revealed that in each case the valve had been manually closed hard on its seat following a maintenance operation.
Examination by the licensee revealed similar improper adjustments of the torque bypass switches on several other motor-operated valves in safety-related systems.
IE Infonnation Notice No. 81-31. " Failure of Safety injection valves To Operate Against Dif ferential Pressure," reported on September 3,1981, that both trains of the San Onof re Unit I saf ety injection system were found to be inoperable when challenged to operate against dif ferential pressure, improperly set switches were the principal cause of these failures.
There were no adverse consequences in this particular event because there was no loss-of-coolant accident.
The reactor pressure remained above the saf ety injection pump's shutoff head; therefore, no actual injection of water would have occurred if the valves had opened, llowever, had reactor pressure decreased and actual injection been required, injection flow would nut have been automatically
lEB 85-
, 1985 Page 4 of available as designed. These valves had been regularly tested at each refuel-ing outage, but the tests were not required to be performed with differential pressure across these valves.
Florida Power Corporation reported an event at Crystal River Unit 3 in LER 77-9.
During plant cooldown with the unit in hot shutdown, decay heat removal valves in the decay heat removal pump suction would not open with remote actuation. These failures were caused by pressure acting on the gate valve i
discs. The valves were opened manually with the handwheels.
The torque I
switches were reset.
In addition to common mode valve failures on demand, there have been numerous common mode failures discovered during testing or as a result of investigating a single failure. NUREG/CR-2270, "Comon Cause Fault Rates f or Valves "
February 1983, contains reports of 99 common cause valve fault events f rom 1976 through 1980.
The NRC has previously identified other problems with motor-operated valve switches in Bulletin No. 72-3, "Limitorque Valve Operator Failures"; IE Infomation Notice No. 79-03, "Limitorque Valve Geared Limit Switch Lubricant;"
l Circular No. 81-13. " Torque Switch Electrical Bypass Circuit for Safeguard Service Valve Motors;" Information Notice No. 82-10. "Following Up Symptomatic Repairs To Assure Resolution of the Problem;" and Information Notice No.84-10
" Motor-operated Valve Torque Switches Set Below the Manufacturer's Recomended l
Value."
The failure and potential failure of Westinghouse Electro-Mechanical Division motor-operated gate valves to close are discussed in IE Bulletin No. 81-02 and l
IE Bulletin No. 81-02, Supplement 1, " failure of Gate Type Valves To Close Against Differential pressure."
Copies of the above referenced NRC Bulletins, Circulars and Information Notices can be obtained from your local public document room.
i Actions for All lloiders of Operating Licenses or Construction Pemits l
For motor-operated valves in the high pressure coolant injection / core spray and errergency feedwater systems (RCIC for BWRs) that are required to be tested l
for operational readiness in accordance with 10 CFR 50.55a(g), develop and l
implement a program to ensure that valve operator switches are selected, set l
and maintained properly.
This should include the following components:
1 a.
Review and revise, as necessary, the design basis for the operation of each valve. Unless otherwise justified, the design basis shall include both opening and closing the valve with the largest differential pressure expected during both normal and abnormal events within the design basis.*
% Fe~rtist valve movement from the safety position should be considered in the design basis unless the motor-operator is disabled.
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i IEB 85-1985 page 5 of 1
b.
Using the results from item a above, review and revise as necessary the methods for selecting and setting all switches (i.e., torque, torque bypass, position limit, overload) for each valve operation (opening and closing).
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c.
The settings determined in f tam b above shall be denonstrated to be correct by testing the valves at the largest dif ferential pressure determined in item a above. Justification should be provided for any cases where testing with this differential pressure cannot be performed to demonstrate correct switch settings.
This justification should include the alternative to testing which will be used to ensure correct settings.
However, each valve shall de stroke tested, to the extent practical, to verify that the settings defined in item b above have been properly implemented even if testing with differential pressure cannot be performed.
d.
Prepare or revise procedures to ensure that correct switch settings are utilized tnroughout the life of the plant." Ensure that applicable industry reconnendations are considered in the preparation of these i
procedures.
l e.
Within 120 days of the date of this bulletin, submit a written report to I
the tiRC that describes the program to accornplish items a throuqh d above and contains a schedule for conpletion of the items. A copy of the 1
program shall be submitted to the flRC Resident inspector.
l.
For plants with an OL, the schedule shall ensure that these items are completed as soon as practical and within two years from the date of i
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this bulletin.
I 2.
For plants with a Cp, this schedule shall ensure that these items are f
j completed before the scheduled date for OL issuance or within two years from the date of this bulletin, whichever is later.
I f.
Provide a written report on completion of the above program, ihls report should provide (1) a verification of completton of the requested program,-
t (2) a suntnary of the findings as to the valve operability.ind setting of 1
the valve switches in the pre-test rondition, and (3) a su mary of the valve switch settings in th! post-test condition.
The f4RC staf f intends to use this data in the resolution of Generic Issue ll.E.6 This report shall be submitted to the flRC within 60 days of ctnnpletion of tho proqram.
WThis Ttem is intended to be completely consistent with at t ton item 3.2, " post-
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1 Maintenance Testing ( All Other Saf ety-Related Components)," of 6eneric letter 83-28 " Required Actions Based on Generic Implications of Salem ATWS ivents "
These procedures should include provisions to monitor valvo performance to en*,ure the switt.h settings are correct, ibis is porticularly irrportant if the torque or torquo bypass switch ',etting has boen slynif f t.antly raised i
i above that required.
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IEB 85-
, 1985 Page 6 of.
The written reports shall be submitted to the appropriate Regional Administrator under oath or affirmation under provisions of Section 182a, Atomic Energy Act of 1954, as amended. Also, the original copy of the cover letters and a copy of the reports shall be transmitted to the U.S. Nuclear Regulatory Commission.
Document Control Desk, Washington, DC 20555 for reproduction and distribution.
This request for information was approved by the Of fice of Management and Budget under a blanket clearance number 3150-0011. Comments on burden and duplication may be directed to the Office of Management and Budget Reports Management, Room 3208, New Executive Office Building, Washington, DC 20503.
Although no specific request'or requirement is intended, the time required to complete each action item above would be helpful to the NRC in evaluating the cost of this bulletin.
If you have any questions regarding this matter, please contact the Regional Administrator of the appropriate hkC Regional Office or the technical contact listed below.
James M. Taylor Directnr Office of Inspection and Enforcement Technical
Contact:
H. A. Bailey, IE (301) 49-29006 R. J. Klessel, IE (301) 492-F119 At tachire n t s :
1.
Table 1 2.
List of Recently issued IE Bulletins
IEB 85-
, 1985 Page 6 of The written reports shall be submitted to the appropriate Regional Administrator l
under oath or affirmation under provisions of Section lH2a, Atomic Energy Act of 1954, as amended. Also, the original copy of the cover letters and a copy of the reports shall be transmitted to the U.S. Nuclear Regulatory Conmission, Document Control Desk, Washington, DC 20555 for reproduction and distribution.
This request for information was approved by the Of fice of Management and l
Budget under a blanket clearance number 3150-0011. Comments on burden and
~
duplication may be directed to the Office of Hanagement and Budget. Reports Management, Room 3208, New Executive Of fice Building, Washington, DC 20503.
l Although no specific request or requirement is intended, the time required to complete each action item above would be helpful to the NRC in evaluating the cost of this bulletin, i
If you have any questions regarding this matter, please contact the Regional Administrator of the appropriate NRC Regional Of fice nr the technical contact i
listed below.
Janes M. Taylor. Ofrectnr i
Of fice of Inspection and Enforcement r
t Technical
Contact:
H. A. Dailey, IE (301) 49-29006 R. J. Klessel, IE (301) 492-8119 l
Attachments:
1.
Table 1 2.
List of Recently issued IE 11ulletins i
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IE:
DGable 9/ /85 DEPIR:lE DIPf.R:lE OfPIR:ll OLPER:ll 00:DIPIR:ll U: DLPl.H : ll HAlialley:mj Cf.kogs t khaer Hf1.her SA',c hwa r t t iLJordan
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l-IEB 85-1985 TABLE 1 Summary of Significant MOV Abnormalities identified by M0 VATS Bypass switch improperly set 75*
Incorrect thrust 50 Unbalanced torque switch 33 i
Valve backseating 8
High motor current 3
Torque switch abnormalities 2
Hiscellaneous abnormalttles 33 t
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- Percent of valves experiencing abnormality.
The total does not equal 100 percent as trost valves had more than one abnormality.
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EtlCLOSURE 2 CRGR Charter Section IV !! Item
- Relevant Information (i) provided in proposed bulletin.
(11) provided in the package.
(iii)
The timing foe the requirements is stated in the proposed bulletin.
The proposed bulletin may be supplemented by additional requirements when Generic issue ll.E.6, "In-Situ Testing of Valves" is resolved.
The schedule for resolution of !!.E.6 is May 19M.
Implementation of this bulletin would require the design basis for the operation of certain valves to be reviewed and revised as necessary.
Switch settings are to be revtewed and revised as necessary.
Most valves will require the plant to be shutdown.
Several valves may require the plant to be in cold
- shutdown, i
plant shutdown will be recessary, but testing of valves requiring plant shutdown is not espected to l
be the critical path during a typical refueling outage.
fio specific design changes or valve switch setpoints require f4RC pre-implermntation approval.
fio new equipemnt is required by the bulletin.
fio new equipment is required by the bulletin.
(iv)
Providnd in proposed bulletin:
(v)
Provided in Enclosure 3, Co*.t.llenefit Analysi%.
(
I provided in fnclosure 1. Co'.t llenefit Analysis.
Cost to f4RC is 7(Ji thou*,and dollar'..
Co'.t to
- l. i c en',res ( *, 1.9 million dollar.
The e.chedule of two year".
I', long enough to ensure i
a refueling nutage durinq the linplementation period.
The *.chedule l'.
Ilmi t ed t o two year, due to the
'..t f e t y r.orn e rn.
'See, t ion ! V H.1 t, t ir he.l.
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i If licensees find that there is a problem which cannot be corrected in two years, such as valve and/or valve operator replacements, then an extended irrplementation schedule would seem in order.
The schedule for staff actions based on a hypo-thetical approval date of November 1,19fi5 for issuance of the bulletin is Receive program descriptions by March 1, l
l
- 1986, t
Program descriptions for plants will be reviewed within 6 months, with priority on plants goino into a refueling outage. Some of these plants inay require review in less than 2 n:enths. The bulletin, however, does not require program pre-implementation t
approval by the NRC.
The Generic issue !!.E.6 "In-$ltu Testing of Valves" has been given a priority of MEDIUM. Due l
l to the comon trode valve f ailures at Davis-Besse, j
IE believes this priority should be raised to HIGH l
for ausillary feedwater and high pressure injec.
r tion valves, r
I The bulletin has only specifled that procedures be prepared or revised to ensure correct switch f
settings are utilized throutjhout the life of the plant.
The degree of detail is not specified.
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l This safety concern is applicable for all plants.-
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Genoric issue II.E.6 rated this issue a MLDIUM priority for both HWRs and PWRs.
)
(vi)
The proposed bulletin requests motor operated valvt testing that is within the requirements of 10 CIR $0, Appendix H Section XI, fest Control, in addition, l
these requirements are specifically delineated for i
l emergency core cooling systerr, in 10 CIR 60 Appendia A i
45 Criterion 3/ - Testing of errergency core cooling i
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system.
i llowever, valvo testing under difforential preuute l
conditions ha*. not generally been conducted before L
although Criterion il rv<psire'. functional te', ting
... under conditions a+. rinso to doslyn a'.
practical"...
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(vii)
The package h.is been sent to OELD for comment.
i (viii)
The proposed bulletin is covered under OMB h1anket
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cicarance number 3150-0011.
The Cost-Benefit Analysis is provided in Enclosure 3.
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ENCI.05URE 3: Cost-Benefit Analysis
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1.
Benefit Analysis l
The benefit from implementing the proposed bulletin is a reduction in i
l frequency of core melt per reactor-year and in the associated risk of offsite radioactivity release caused by accident and transient sequences that require motor-operated valve operation for mitigation.
The safety
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l concern in the proposed bulletin is an increase in core melt frequency l
I caused by both single failure and connon mode faillures of switches in the l
I motor operated valves. Conrnon mode failures are not considered in the i
design basis for nuclear plants.
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The NRC's concern for valve testing requirements has been previously i
i identified in Reference 1 as Generic issue ll.E.6.1. "!r-Situ Testing of i
Valves."
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Core Mc__1 t Redu_c t iory i
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Reference 2 is a scoping study on the safety significance of connon modo i
t f ailures of saf ety-relatro valves due to misad.justinents of valve operator i
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$ witches.
This study concluded thiet correcting the connon mode failure component in all saf ety related valves would reduce the average core rnelt t
j frequency by at least 6.51 6 per reactrir-year and would reduce public
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risk by at least 15 person rem per reactor year.
The proposed bulletin h
t is limited to high pressure inlection druf aunillary f eedwater (RCIC for I!WRS) valves.
Ih h reduced +.cnpo h more than POI of the +ropo con +.idered l
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in Reference 1 and intuitively seems to include the more important valves.
l Therefore, the reduction in core melt as a result of implementation of I
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this bulletin will be assumed to be at least 1.0E-6 per reactor year.
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A study reported in Reference 3 shows that a core melt reduction of 1.0E 6 j
per reactor year can be achieved through a 2% reduction in the single, i
randon failure rate of all safety related valves. Since the bulletin includes approximately 20% of the more important safety-related valves, a l
10% reduction in the single, random failure rate of these valves corres-j ponds to the 2% overall reduction rate, The above reduction in core snelt as a function of reduction in valve failure rate would be expected to be j
even more f avorable if the contron mode f ailure reduction had been I
modeled.
Reference 4 shows that 422 of a total of 600 valve failure reports between 1
191(2 and 1980 were motor-operated valve faults.
This report also shows i
j that torque switch /llmit witch failures and misadjustrnents have directly j
resulted in over 30% of the 422 motor operated valve failure reports.
in addition a strong relationship is reported in Reference 5 to esist between torque switch adjustrnents and the failures attributed to mechanical or i
i motor failures.
Therefore a 101 reduction in the failure rate of the j
valves addrened in this bulletin throurlh a program to enwre correct i
switch settings is t.onsidered reasonable and conservative.
(
the above discunion *.upports the antFption that a t. ore einlt reduction of 1.01-(, per reat. tor year as a re w it of thi+, bulletin is rea wnable and l
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l 3-l conservative. The cost-benefit analysis below is based on the guidelines l
of Reference 6.
l Public Dose Benefit - PWRs l
l Based on an average public risk consequence of 2.3E6 person-rem / event, the public risk reduction is 2.3 person-rem / reactor-year.
The public benefit in person-rem from implementing the bulletin is estimated as j
follows:
Public Benefit Reduction in Public Risk x Number of Reactors x
=
Average Life of Reactors 2.3 rerson ren/rcactor-year a 90 reactors x 28.8 l
years 0.(d4 person-ren Public Dosn_Henefit
_BWRs liased on a average public rish contequerice of S ?t6 pere,nn-rern/ event, the public r19 reduction i+, $ 2 per'on rem / reactor year.
The public l
benefit in person-rem f rom linplementing the hulletin is estirnated as follows:
i
a Reduction in Public Risk x Number of Reactors x Public Benefit a
Average Life of Reactors
$.2 person-rem / reactor-year x 44 reactors x 27.4
=
years 1
l 0.CF4 person-rem
=
i Occupational Dose Penefit - Core Melt - BWRs and PWR$
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j Based on an average occupational dose risk consequence of 2L+4 person -
l remicore melt event, the occupational dose reduction is 0.02 person-rem /
I reactor-year the total for 134 reactors for 28.3 years is.8E+2 person-i i
r ren.
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At 1000 dollars /porson-rem, the total of the above benefits is 12E*6 i
dollars i
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Hased on a +.urvey of +.overal plants' consnitments for valve testing in t
accordance with 10 CfR $0.%a(9), the averayo plant is espected to test 20 valve. In responw to this bulletin, the total offort for this bulletin i
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is expected to require 24 person-hrs / valve for the initial effort and 1
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person-hr/ year / valve for follow-on effort.*
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Occupational Dose Cost - Industry l
i The cost of the initial industry effort due to radiation dose is based on l
an average field of 30 mr/hr for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> / valve or 300 mr/ valve.
This radiation field is based on NRC resident inspector estimates.
The
[
l remaining initial 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> /balve and the I hr/ valve follow-on would be out of the radiation field. For the initial ef fort, only the operating plants are totaled in the radiation dose:
93 plants x 20 valves / plant x 300 mr/ valve
= 558 rem total initial dose.
f 588 rem x 1000 dollars / rem = 0.6E6 dollars l
I IndustrylaborCnst i
The labor cost is based on 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> / valve initially and I hour / valve / year for follow on.
I l
'These times are based on e*.timates by NRC resident inspectors and represent i
only the increa*.e in time required beyond what is presently required to meet l
the requirements of 10 CFR 50.55a(g).
The f ollow-nn cost repre+.ents additional i
fjA racord keepinq cf fort to ensure the correct switch etting. are maintained.
l 1here is no f ollow on to'. ting in addition to that required for 10 CIR $0.%d(9) included in thl. co't c', tim.ite.
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m
i Initial cost is:-
24 hr/ valve x 20 valves / plant x 134 plants x $50/ hour = 3.2E+6 I
j dollars.
L J
l Follow-on cost is:
i 1 hr/ valve-yr x 20 valves / plant x134 plants x $50/ hour x 28.3 years =
j 3.8E+6 dollars.
The present worth of this cost (assuming an average of i
1 i
14 yr and 5% discount rate) is 1.9E+6 dollars.
i The total industry labor cost is 5.lE+6 dollars.
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Onsite Property Damage Cost - HWR5 and FWRs
\\
The cost (negative) from the reduction in onsite property damage as I
4 a result of the reduction in core trelt frequency is estirnated at -?E+9 I
i
)
dollars per event.
The property damaqe cost is:
I
\\
1
{
-2E+9 dollar / event x 1.0E-6 events / reactor-year a 134 reactors x j
t 28.3 yr.
-1.6E*6 dollar *,,
j j
the present worth of thl'. cost (asstimine) act,ldeants at 14 yr ar'el o 67 discount rate) is -3.81.t6 dollars, i
i IOt.n l Influ% try, CO% t (Do*.<>,1 thor, arief Projiett;ty 11 tm.igr')
1 f'
l j
Ihtl (Ota l lfidtlitry CD'e t l$ l.9l 66 <joIIar%.
I
t l
7-l Occupational Dose Cost-NRC i
1he NRC cost due to dose is assumed not to be increased.
i NRC Labor Cost i
The NRC person. hrs is estimated to be initially 2 person. hrs / valve with no j
additional follow-on required.
1 l
NRC cost is:
i l
2 hrs / valve x 20 valves / plant x 134 plants x $50/hr = 0.21E6 dollars i
TotalCost(NRCandIndustryl l
l t
The total cost is 2.2E6 dollars.
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Cost Benefit Ratio l
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2.20E6 dollars of cost /12.0E+6 dollars nf benefit = 0.111 j
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l I
1 I
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~.
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References 1.
NUREG-0933, "A Prioritization of Generic Safety Issues", July 1985.
2.
Memorandum to Henry Bailey, IE, from Harold VanderMolen, NRR, of September 5, 1985, subject:
Common-Mode Valve Failures Due To improper Adjustment Of Torque, Limit and Bypass Switches.
3.
NUREG/CR-2800, Supplement 2, " Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development", U.S. Nuclear Re9ulatory Commission, December 1983, Pages 2.296 and 2.300.
4 NUREG/CR-2770, " Common Cause fault Rate for Valves " February 1983.
5.
Electric Power Research Institute, EPRI NP-241, " Assessment of industry valve problen.s", Palo Alto, California 197fi.
6.
NUREG/CR-3568, "A Handbook for Value-impact Assessment". December 1983.
i TEMPORARY INSTRUCTION 2515/XX,
(
INITIAL INSPECTION REQUIREMENTS FOR IE BULLETIN 85-XX,
" MOTOR-0PERATED VALVE COMMON MODE FAILURES DURING PLANT TRANS!ENTS 00E TO IMPROPER SWITCH SETTINGS" 2515/XX-01 PURPOSE To provide guidance for perfonning the initial inspection followup of the licensee's activities taker.
in response to.
IE Bulletin 85-XX,
" Motor-Operated Valve Comon Mode Failures During Plant Transients Due to r
-r improper Switch Settings."
i i
i 2515/XX-02 OBJECTIVE To ensure a minimum level of consistency in the review of the licensee's initial response and corrective program developed in response to IE Bulle-tin 85-XX.
The review of the licensee's actual tests and corrective actions, as well as the final response to IE Bulletin 85 XX, will be the subject of supplements to this temporary instruction (TI),
l 2515/XX-03 BACKGROUND 4
1 On June 9, 1985, the Davis-Hesse Plant experienced a complete loss of main and auxiliary feedwater, ibis event has beeri previously described in IE Information Notice 85-50, " Complete loss of Main and Aueillary leedwater at 1
I
a PWR Designed by Babenck & Wilcox " dated July 8, 1985, and NUREG-ll54 0
" Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9,1985," dated July 1985.
As a result, it will not be described in detail in this T!.
IE Bulletin 85-XX was issued to ensure that the switch settings on certain safety-related motor-operated valves are set and maintained so as to accommodate the most severe loading expected during design basis events. This IE bulletin also briefly described (1) a recent valve failure at Sequoyah Nuclear Plant Unit 2 which was caused by an improperly set open limit switch (2) the results of a recent study on the effectiveness of signature tracing techniques in determining the operational readiness of motor-operated valves, and (3) a summary of the previously issued generic communications pertaining to problems associated with motor-operated valve switches. A copy of IE Bulletin 85-XX is enclosed as Appendix l.
2515/XX-04 BASIC REQUIREPENTS 1
for each f acility with an operating license or construction permit:
04.01 Review the licensee's initial response to verify that it:
a.
Was submitted within !?0 days of issuance of Bulletin 85-XX.
b.
Provides sufficient details pertaining to the program.
While it is not necessary for the licensee to *.ubmit his entire program, the response should not sittply paraphrase the criteria in IE Hulletin HS-XX.
c.
C l ea r l y demons t ra t e',
that the program will meet the crite-ria in action items a,
t hrough d.
of 11 Hul let in H'2-X X.
04.02 Review the licensee's program to cerify that it:
a.
Includes all of the motor-operated valves in the identified systems which are in the licensee's approved inservice testing plan for valves.
This can be best accomplished by selecting a sample of valves from the list and ensuring that they are clearly identified in the licensee's program, j
Particular attention should be placed on the larger valves which are subjected to high differential pressure.
b.
Provides for a systematic review of all design basis events described in Chapter 15 of the licensee's Final Safety Analysis Report to determine the greatest differential J
pressure to which the valve may be subjected.
Whether or not the valve, in its normal position, is properly posi-tioned for the particular design basis event is not rele-vant.
If system operation is required, it should be i
assumed that any valve in the system will require cycling j
during the event due to either spurious operation of the valve (i.e.,
single failure) or operator error.
Valve operation required in the course of recovery f rom an event should also be considered.
c.
Provides a methodology for determining the proper switch settings. Nonnally, the best approach for determining this information is in consultation with the valve and/or operator manufacturer.
Simply referring to the valve's instruction manual settings will normally not be sufft-cient, unless it can be demonstrated that the valve's initial de'.ign specification enveloped the conditions determined above.
1 P
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d.
Will result in properly setting all of the valve switches.
The procedures for setting of the switches should be available in the valve manufacturer's instruction manuals.
e.
Verifies the setting of each valve's switches through signature or other similar testing.
If possible, this testing should be conducted at the greatest dif ferential pressure detennined above.
Where this is not practical, stroke testing with the valve operator artificially loaded to simulate the specified differential pressure may be used provided it clearly demonstrates the proper setting of the limit, torque, and torque bypass switches.
An example of such an artificial load would be a load cell on the valve stem.
f.
Ensures that each valve / operator / switch setting combination is tested at the greatest differential pressure determined above.
The key to accepting prototypical testing as the l
basis for accepting a particular valve is that the tested valve / operator / switch setting combination must be the same.
j Interpolation to other valve and/or operator sizes should not be permitted.
lloweve r, the test could be performed by another utility or tha valve manufacturer provided suffi-cient documentation is available to clearly establish the i
similarity between the two valve assemblies.
9 Ensures that maintenance and test procedures are consistent with the requirements of action item 3.2
" Post-Maintenance Testing ( All Other Saf ety-Related Compo-nents)," of Generic l.etter 83-78, " Required Actions Based on Generic Implications of Salem ATW5 Lvents," dated July 28 I 'J 83.
(Note - Inspe' tion rsquirements ' relating to review of the licensee's actual c
I tests and corrective actions, as well as the final response, will be the subject of supplements to this TI.
For preliminary guidance, Appendices II and III contain the current drafts.
These drafts will be revised to reflect comments received. based on ' knowledge gained from review 'of ~the.
licensee's_progtam.
2515/XX-05 REPORTING REQUIREMENTS Document the activities associated with this TI 'in the routine inspection reports as.they are completed.
As a minimum, the following milestones should.be reported:
05.01 Completion of review 'and acceptance of the licensee's. initial response to IE Bulletin 85-XX.
05.02 Completion of review and acceptance of the licensee's program
~
to meet-the criteria of action items a. through d. of IE Bulletin f
85-XX.
1 1
2515/XX-06 EXPIRATION i
I This TI shall remain in effect until September 30, 1988.
2515/XX-07 IE CONTACTS Questions concerning this T1 should be addressed to either lienry A. Bailey at(301)492-9006 or Richard J. Kiessel at (301) 492-8119.
J' rwe,,,ss
2515/XX-08 STATISTICAL DATA REPORTING For 766 input, record all time spent in inspecting and reporting for this TI against Module No. 255XX.
Inspection activities that meet the general programmatic requirements also may be reported under the appropriate modules (e.g., 92703, "IE Bulletin, Confirmatory Action Letter, and Generic Letter Followup").
END e
I
APPENDIX II TEMPORARY INSTRUCTION 2515/XX, SUPPLEMENT 1 CURRENT DRAFT INSPECTION REQUIREMENTS FOR THE TESTS AND CORRECTIVE ACTIONS ASSOCIATED WITH IE BULLETIN 85-XX A.
~ PURPOSE To provide preliminary guidance with respect to the' inspection requirements for the tests and corrective actions associated with IE Bulletin 85-XX, " Motor-Operated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings."
8.
BACKGROUND Temporary ' Instruction 2515/XX provided guidance for performing the initial inspection followup (i.e., that associated with the review and acceptance of.the licensee's initial response and corrective ' program) to IE Bulletin 85-XX.
C.
BASIC REQUIREMENTS l.
Monitor the licensee's tests and corrective actions to verify that:
'a.
They are completed on time (two years from the date of issuance of the bulletin for plants with an opera tilig license on the date of the bulletin).
b.
They are consistent with the previously approved program.
This can best be accomplished by selecting a sample of valves and observing the actions.
Particular attention should be paid to large size valves which are subjected to high differential pressure.
However, it is recommended that the same sample not be used to verify each of the licensee's individual actions.
Selection of different vals es for different aspects of the program is expected to result in a better overall evaluation of the program.
2.
It is recognized that the final acceptance of the switch set-tings will be the signature or other similar test. However, the number of problems found during the test can be reduced if the maintenance personnel are properly trained in the appropriate procedures. Thus, it is useful for the inspector to be familiar with the proper procedures for setting the switches.
This can also provide' insight into the effectiveness of the licensee's procedures and training prog rams.
Some aspects to be given attention are:
a.
Does the procedure provide clear instructions as to the specific values the switches are to be set at?
l b.
With respect to setting the limit and/or torque bypass switches, does the procedure clearly specify whether the measurements are based on stem or disc movement?
i c.
Does the procedure clearly specify that the torque switches are to be set with the valve in mid-stroke and the belleville spring in a relaxed condition?
3.
Thsre are a number of maintenance related parameters that will significantly affect the proper switch settings.
These parameters should be-controlled during the tests required by Bulletin 85-XX and later monitored to ensure these parameters remain within the assumed boundaries considered in the selection of switch settings.
Some of these parameters are:
(a) Valve stem and operator lubrication.
(b) Packing - type; installation method and age.
(c) Spring pack and switch lubrication.
'(d) Replacement gasket (thickness).
(e) Torque and limit switch cleanliness and alignment.
EtID
)
APPENDIX III TEMPORARY INSTRUCTION 2515/XX, SUPPLEMENT 2 CURRENT DRAFT INSPECTION REQUIREMENTS FOR THE FINAL RESPONSE ASSOCIATED WITH IE BULLETIN 85-XX A.
PURPOSE To provide preliminary guidance with respect to the inspection requirements for the final response associated with IE Bulletin 85-XX, " Motor-0perated Valve Common Mode Failures During Plant Transients Due to Improper Switch Settings."
8.
BACKGROUND Temporary Instruction 2515/XX provided guidance for performing the initial inspection followup (i.e., those associated with the review and acceptance of the licensee's initial response and corrective program) to IE Bulletin 85-XX.
Temporary Instruction 2515/XX, Supplement 1, provided guidance for performing the inspection requirements for the tests and corrective actions associated with IE Bulletin 85-XX.
C.
BASIC REQUIREMENTS Review the licensee's final response to verify that it:
1-1.
Was submitted within' 60 days of completion of _ the corrective actions.
2.
Provides a meaningful summary of the findings.
Of particular interest are:
a.
The number of valves in the program.
b.
The number of valves found with switch settings which were not in-agreement with the previous procedures.
c.
The number of valves which had to have their switch settings changed because of the evaluations required by IE Bulletin 85-XX.
t d.
The number of valves where the original switch settings were such as to indicate that the valve probably would not. have operated in the most severe design basis event.
EliD u
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