Text

Power-Operated Relief Valve Failure Reduction Methods, Final Rept
	ML19341B767
Person / Time
Site:	Calvert Cliffs
Issue date:	12/31/1980
From:	; ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, EMVC-EPS
To:	;
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ML19341B765	List: ML19341B764; ML19341B767;
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3, TASK-2.K.3.02, TASK-2.K.3.16, TASK-TM CEN-145, NUDOCS 8102270446
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CEN.145 0

PORV FAILURE REDUCTION METHODS Final Report l

Prepared for the C-E OW1ERS GROUP l

l NUCLEAR POWER SYSTEMS DIVISION December 1980 i

POWER a

SYSTEMS COMBUSDON ENGINEERING INC 8102270 i

t 4

h LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COM3USTION ENGINEERING, INC. NEITHER COMSUSTION ENGINEERING NOR ANY PERSCN A71NG ON ITS BEHALF:

f A.

MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTASILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS. OR USEFULNESS OF THE INFORMATICN CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATICN, APPARATUS, METHOD, OR PROCESS O!SCLCSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR '

B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF, OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS O!SCLOSED IN THIS REPORT.

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1 PORV FAILURE REDUCTION METH005 t

FINAL REPORT i

PREPARED FOR THE C-E OWNERS GROUP NUCLEAR POWER SYSTEMS DIVISION DECEMBER 1980 C-E POWER SYSTEMS COMBUSTION ENGINEERING, INC.

WINDSOR, CONNECTICUT

TABLE OF CONTENTS Section Title Page 1

BACKGROUND

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1 2

PURPOSE 1-3 DESCRIPTION OF PORV SYSTEM 1

4 PORY OPERATING EXPERIENCE 5

5 PRIMARY SAFETY VALVES 7

6 METHODS FOR REDUCING PORV SYSTEM FAILURE d

7 IMPLEMENTATION OF FAILURE REDUCTION. PROGRAM 12 8

ANALYSIS AND RESULTS OF FAILURE REDUCTION PROGRAM 13 9

SUMMARY

AND CONCLUSIONS 15 10 REFERENCES 17 F

TABLE 1 C-E PRIMARY SAFETY VALVE AND PORY DATR 18 TABLE 2

SUMMARY

OF EVENTS INVOLVING PORV OPERATION 19 TABLE 3

SUMMARY

OF EVENTS RESULTING IN POTENTIAL CHALLENGE TO PORV 20 FIGURE 1 TYPICAL PRIMARY SYSTEM OVERPRESSURE PROTECTION 21 FIGURE 2 TYPICAL ELECTROMATIC RELIEF VALVE 22 i

Appendix A

C-E ANALYSIS OF REFERENCE PLANT (SL-2) FAULT TREE FOR POWER OPERATED RELIEF VALVE LOSS OF COOLANT ACCIDENT l

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1.

BACKGROUND The failurs of a power-operated relief valve (PORV) to close subsequent to its actuation during an overpressure conditicn was a key. factor in the Three Mile Island-2 (TMI-2) accident.

As a result, the operating history of PORVs on all operating light water reactors (LWRs) was investigated by the Nuclear Regulatory Commission (NRC).

On an overall basis, the results of the invest tigation indicated that the probability of a small break loss of coolant accident (LOCA) due to the failure of a PORV to close appeared to be a major contributor to the total probability of a small break LOCA from all causes.II)

Consequently, the NRC has recuested(2) that methods for PORY failure reduction be evaluated by C-E for possible implementation to increase plant safety.

2.

PURPOSE The purpose of this study is to review PORY failures, to evaluate methods for failure reduction, to describe the plant ch'a'nges made or recommended to reduce PORV failures, and to evaluate the effectiveness of these changes for C-E operating plants.

3.

DESCRIPTION OF PORY_ SYSTEM i

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'3.1 Introduction l

A brief descripdon of the provisions for overpressure protection of the typical C-E Nuclear Steam Supply System (NSSS) primary coolant system and i

clarification of the supporting role of the PORVs is provided below.

Overpress'ure protection for the primary coolant system is based on the combined action of the primary safety valves, secondary safety valves, and the reactor protection system.

At operating conditions the PORVs are not formally part of the overpressure protection system; although the presence of PORVs increases the primary coolant system relieving l

capacity.

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3.2 Function of the PORY To reduce the number of challenges to the primary safety valves, and thus reduce the probability of gross safety valve leakage or weeping, pressurizers on all C-E operating plants (except for ANO-2) are provided with two PORVs having actuation set points below that of the primary safety valves.

Figure 1 shows a typical installation arrangement for primary system over-pressure protection.

Isolation valves are provided upstream of each PORV.

Throughout this report, the term "PORY System" is used whenever the PORY and its isolation valve is being considered in combination.

Design and operating parameters for the primary safety valves and PORVs at C-E operating plants are given in Table 1.(1)

Additional functions, not considered in the initial NSSS design, have since been assigned to th,e FORVs.

These functions include low temperature over-pressure protection, venting, and long term cooling subsequent to a LOCA.

These auxiliary PORY functions have been documented elsewhere and are not included in the scope of this report.

3.3 PORV Desien Basis The PORVs are designed to have an opening setpoint pressure below that of the primary safety valves and to provide sufficient relieving capacity to ensure that the primary safety valves do not lift or weep during over-pressurization transient conditions such as uncontrolled rod withdrawal, loss of load, or loss of all non-emergency AC power.

The PORV opening setpoint pressure is sufficiently high to ensure that the PORVs do not open in response to norral maneuvering transients.

3.a PORY Descriotion All PORVs in operating C-E NSSSs are Dresser electromatic relief valves wnich are pilot actuated, reverse-seated, and which use pressurizer pressure to operate the valve (Figure 2). When pressurizer pressure exceeds the valve setpoint pressure, the solenoid on the pilot valve is energized; this causes its plunger to actuate a lever to open the pilot valve.

The main valve's pressure chamber above the valve disc is vented 7~

5 A

t through the open pilot valve and the resulting pressure difference across the main valve disc causes the main valve to open and discharge pressurizer fl ui d.

When pressurizer pressure decreases below the setpoint value, the solenoid is deenergized, the pilot valve closes., and steam pressure builds up in main valve pressure chamber and force's the valve disc closed.

3. 5 PORV Ooeration The PORVs are designed for automatic or manual operation.

In automatic I

operation, the PORVs are opened by the high pressurizer pressure trip signal in the reactor protective system, which is actuated by a two out of four channel logic system.

The PORVs, which are actuated by the same bistable trip units which actuate the reactor trip, open whenever the pressurizer pressure exceeds the high pr. essure reactor trip setpoint and l

.they remain open until pressurizer pressure fails below the valve reset i

i pressure.

In the manual mode the PORVs can be operated independent of system temperature and pressurizer pressure.

The PORV actuation setpoints vary somewhat from plant to plant, at a nominal value of approximately 2400 psia, about 100 psi below the primary safety valves setpoint and 150 psi above normal operating pressure (Table 1).

I 3.6 PORV Isolation Valves To permit isolation of a PORV in case of excessive seat leakage or failure to close, motor-operated block valves are provided upstream of each PORV.

During power operation the block valves are normally open.

However, one or both PORVs may be isolated (block valves closed) because of excessive leakage.

Also, operation with one PORV isolated may be considered to j

avoid excessive reactor coolant discharge due to both PORVs lifting.

3.7 PORV Leakaoe Detection Several methods were used prior to the TMI accident for the detection of excessive PORV leakage or failure to close.

These methods include moni-toring PORV discharge piping temperature, PORV pilot valve position indica-tion, and quench tank pressure, tenperature, and level.

Readouts from each

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I of these measurements are generally available in the plant main control room.

Subsequent to the TMI-2 accident, the NRC required a reliable, direct means for PORV position indication.

Action to respond to this requirement is described in Sections 6 and 7.

3.3 Electric Power Sucolies In performing their function to reduce the frequency of primary safety valve challenges, the PORVs provide equipment protection and as a con-sequence, are not considered as part of the plant safety system. There-fore, the valves as installed in the field were not provided with safety grade power sources and no credit was taken for their operation in safety analyses.

Subsequent to the TMI-2 accident, consideration was given to providing the PORVs and their isolation valves with emergency power sources.

Further actions on PORY system power supplies are discussed in Sections 6 and 7.

3.9 Comoarison with Other PWRs The PORV systems provided in pressurized water reactors (PWRs) supplied by BabcockandWilcox(B$W)(3), Westinghouse (W) and C-E differ in details such as the type, number, capacity, setpoint, valve vendors j

and control circuitry.

Certain important differences among the PWR j

vendors' systems are described in the following sections.

On C-E plants, the initial design function of the PORVs was solely to reduce the challenges to the primary safety-valves during power operation.

The PORVs on B&W and W plants had an additional function, namely, to reduce ~the frequency.of reactor trips due to high pressure.

The PORY actuation set point on C-E plants coincides with the high pressure reactor trip satpoint, whereas, the other PWR vendors required that the PORV actuation pressure be below the high pressure reactor trip setpoint in order to reduce the number of high pressure trips. The C-E design allows the specification of a higher PORV actuation pressure, and therefore a greater margin above the norm &l plant operating pressure than do the other PWR designs.

Typically, the margin between normal operating pressure and a-x'

the PORV actuation setpoint was about 150 psi for C-E plants,100 psi for 11, plants, and 70 psi for B&W plants.

This difference provided an incremental margin to PORV challenges in C-E plants compared with those of the other PWR vendors.

The B&W plants are equipped with the same type of PORVs as those of C-E, namely, the Dresser electromatic solenoid pilot-operated valve described in Section 3.4.

The majority of W plants use Copes-Vulcan spring-loaded, air-operated valves. Air pressure on the control diaphragm overcomes the spring force to open the valve.

Venting the air pressure from the control diaphragm allows spring force to close the valve.

A few W plants use PORVs manufactured by Masoneilan (3 plants), Dresser (1 plant), ACF Industries (1 plant), and Control Compon'ents (1 plant).

4 PORV OPERATING EXPERIENCE 4.1 Combustion Engineering Plants The operating experience of PORVs in C-E plants has been compiled in j

Table 2 based on information supplied by the various plant operators I

durinc a survey conducted in early 1980.

The PORV actuations noted in l

Table 2 do not necessarily represent the total number which have occurred, l

since PORV actuations were not reportable events and were not routinely recorded.

Therefore, some actuations may have been overlooked.

Also, since the available means for the detection of PORV actuation was not direct, but generall'y dependent upon an integrating effect, such as increasing quench tank level, for example, some actuations may have gone undetected.

. Table 3 is a tabulation of high pressurizer pressure reactor trips l

occurring in C-E operating plants for which PORV actuations were not reported.

The data was obtained from a review of published data, mainly from the NRC.

Since, by design, a high pressurizer pressure reactor trip should be accompanied by PORV actuation, it is inferred that the actuation did occur, though it was not reported. _

Table 2 indicates a total of seven confirmed PORV. actuation events.

Four events occurred during PORV testing or system maintenance.

In two of these events the PORVs failed to close satisfactorily.

The remaining three actuation events occurred during power operation, witn the PORVs operating.

satisfactorily in each case.

Table 3 indicates a total of sixteen high pressurizer pressure reactor trips, eleven of which resulted from turbine runbacks.

Tables 2 and 3 extend the PORY actuation data presented in I1)

NUREG 0635 It was inferred that the high pressurizer oressure trips listed in Table 3 were accompanied by PORV actuations.

Comoining the confirmed PORV actuation events during power operation listed in Table 2 with the inferred actuation events from Table 3, a total of nineteen events or thirty-eight PORY challenges is obtained, with no failures being reported.

A total of about 29 reactor-years of operation is covered by this data.

The two PORY failures-to-close on C-E plants listed in Table 2 occurred.

during maintenance or testing.

The Palisades incident occurred when the Reactor Protection System (RPS) was deenergized for maintenance, which caused the PORVs to open.

Due to an ambiguity in the pertinent wiring diagrars the technician failed to perceive that his action would cause PORV actuation.

The scring-return-to-Auto feature of the PORV selector switch contributed to the incident since the selector switch could not be retained in the " Manual" mode and

" Shut" position unless" held there by the operator.

Corrective action was taken to clarify the pertinent wiring drawings and eliminate the spring-return-to-Auto feature of the PORV selector switch.

The PORV failure-to-close in this instance was not due to the failure of the valve.

The second PORV failure-to-close occurred in Calvert Cliffs #1 during valve operational testing following valve maintenance.

The valve failed to shut completely.

Modified replacement parts had been installed in the.,.

valve because original replacement parts were unavailable due to vendor upgrading of the valve design.

Following adjustment of the pilet valve stroke, satsifactory valve closure was obtained.

4.2 Exoerience at Other PWRs b4)

Westinghouse PWRs in the U.S. have not reported any PORV failures but since they are equipped with a different type of PCnv their reliability experience is not relevant to C-E PORVs.

It has been estimated that in B&W plants there have been approximately 150 actuations of PORVs(3) with six cases of failure-to-close properly.

One failure occurred during low power testing upon loss of a vital bus, another during startup testing due to improper venting, and a third was a leaky valve.

Three failures occurred during power operation, giving approximately 3/150 =.02 failures per demand.

5.

PRIMARY SAFETY VALVES 5.1 Oceratina Exoerience No primary safety valve lif ts have been reported for C-E operating plants during approximately 30 reactor-years of operation. Westinghouse plants also have not reported any primary safety valve lifts. One primary safety l

valve lift has been noted(#) in a B&W plant, but no details were given.

l In view of the lack of challenges to the primary safety valves, a direct f

quantitative estimate of their reliability based on experience cannot be j

made.

I 5.2 Probabilistic Analysis The main ste'am safety valves (MSSV) are much more subject to challenges l

than are the primary safety valves, so that data regarding their reliability has been developed.

This data does not have direct applicability to the l

primary safety valves since, even though the MSSV bears some similarity to

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the primary safeties, there are distinct differences with rescect to service conditions, materials, and other design features.

Lacking data on the primary' safety valves, the MSSV data may provide some indication of primary safety valve reliability.

A study of PWR MSSV operating experience up to May, 1978 was performed by C-E.

The Jata sources used were NPRDS Failure Report Summaries, License Event Report Summaries, and Operating Units Status Reports.

The period reviewed included 137 reactor-years of operation at 38 PWR plants with an estimated population of 570 MSSVs.

During this period there were an estimated 2070 MSSV test deminds (pre-operational and annual).

Assuming one demand on MSSVs for every ten scrams or turbine trips, about 2580 operational MSSV demands were estimated. The total number of MSSV demands in the study period were estimated to be 5650.

During this period two events were reported (none from C-E operating plants) in which MSSVs failed to close following a demand.

The first event occurred at Turkey Ppint Unit 4 in 1974 when a missing cotter pin caused one MSSV to fail open.

The second event occurred at Three Mile Island Unit 2 in April, 1978.

A common mode failure of six MSSVs to close occurred due to cocked sleeves in the bellows assembly.

Thus, the total number of MSSV failures to reseat reported during the study period was seven.

Based on the seven reported MSSV failures and the 5650 estimated MSSV demands, a failure rate of 1.24 x 10-3 per demand is estimated, This failure rate is lower than the value of 2 x 10-2 estimated for power operated reliaf valves in NUREG 0560.(3) Assuming that the MSSV reliability data are tc, some degree applicable to the primary safety valves, the data suggests that the primary safety valves may be more reliable than the PORVs.

More definite conclusions must await development of ope:ational and/or test data on primary safety valves.

6.

METHODS FOR REDUCING PORV SYSTEM FAILURE 6.1 Reduction of PORV Challenaes The frequency of PORY system failures can be reduced by de:reasing the fre-quency of challenges to the PORVs.

These reductions must be made without adversely impacting safety or incurring unacceptable economic or performance p

penalties.

Methods for potentially decreasing the frequency of PORV challenges on C-E plants and a brief summary of their impacts on the plant are provided below.

7 6.1.1 Raise PORY Setooint High pressurizer pressure trips the reactor when the pressure exceeds the trip setpoint pressure and the output from the same bistable comparator also actuates the PORY.

Therefore, only one setpoint is available.

Raising this Reactor Protection System (RPS) high pressurizer pressure reactor trip setpoint would invalidate the safety analysis and increase the challenges to the primary safety valves.

6.1.2 Lower High Pressunzer Pressure Trio Setcoint This requires the concomitan lowering of the PORY actuation setpoint as described above.

Doing so would increase the number of challenges to the PORVs.

6.1.3 Raise the setooint for the existing PORY Ooening/Hich Pressurizer Pressure Trio and Add Another Hich Pressurizer' Pressure Reactor Trip at 2400 osi The setpoint fcr the existing PORV Opening /Hi h Pressurizer Pressure C

Reactor Trip would need to be raised approximately no higher than i

20-40 psi to prevent primary safety valve challenges during a full loss of turbine load without a simultaneous reactor trip while simultaneously precluding PORV openings during milder pressure increases.

The benefits of this alternative would be very small since only a very small fraction of the PORV openings would have been. avoided by this modification (i.e., full load rejection where PORV opening was desired to preclude primary safety valve opening and the inadvertent initiations would not have been affected).

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Further, there is no more room in the protective system cabinetry in some of the operating plants to accommo.date additional bistable trip units and other circuitry that would be required.

Adding additional trips would be expensive and would'take a considerable amount of time to. incorporate.

6.1.4 Block Gut and/or Deactivate PORV During Power Ooeration In the event of a full power incident which causes the turbina admission valves to close rapidly (e.g. full load rejection, electrical system over-frequency, turbine control failure), the reactor would trip on high pressurizer pressure in the absence of a turbine trip signal.

The pressurizer pressure would continue rising above the 2400 psi setpoint until the reactor trip quenched i

the power output of the core and caused the pressurizer pressure to decrease.

If is prudent to use the power operated relief valves to preclude challenging the primary safety valves during this transient.

Therr. are PORV block valves which can be closed in the unlik'ely event of a -PORV fdiling to close.

Such block valves are unavailable to mitigate the consequences in the unlikely event that I

a safety valve fails to reclose.

6.1.5 Reduce Operating Pressure A reduction in operating pressure would tend to reduce the number of PORV openings, but by only a small proportion.

Also, the lower the operating pressure, the higher the overshoot in pressure after a load rejection is terminated by the high pressurizer pressure trip.

The higher overshoot in pressure results from the delay in the reactor trip.

This increases the potential for challenging the primary safety valves.

More importantly, decreasing the prircary ope;ating pressure would dectease the operating DNB ratio thus causing the i

core to be operated closer to one of the safety ifmits.

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6.1.6 Elimination of Turbine Runback Table 3 indicates that a relatively large number (11] of high pressuretrips(andpresumably22PORVactuations)otEurred during turbine runback events.

A review of this plant feature indicated that its elimination would not adversely affect plant operation, while at the same time reducing PORY challenges to a significant degree.

6.2 Imoro.'ed capabil3ty for Countermeasures The frequency of PORV system failures can also be reduced by improving the capability for appropriate countermeasures (PORV isolation) sub-sequent to a PORY failure to close. Methods for pbtentially improving the capability to take appropriate actio'n and a brief sumary of their f

impacts on the plant are discussed.

6.2.1 Automatically Close Block Valve Whenever F07V Fails to Close on Command

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There are several ways this could be implemented. The block valve j

closing signal could L" >rmed by an initial PORV opening signal so l

that the block valve u-d remain open in normal operation but would be au%matically closed if the PORV failed to close on conmand.

l Another approach would use the concurrence of an open PORV valve and I

and PORV valve closure command to automatically close the block valve.

r Although automatic valve closure would remove the requirements for i

operator action upon PORV failure, the additional control circuitry would introduce additional complexity to the system and would itself be subject to its own failure mades.

These schema require further detailed evaluation to determine their positive and negative imoacts j

on overall plant safety.

A simpler approach is to assure that the operator is able to utilize existing inplant instrumentation to

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identify a stuck-open PORV and to close the block valve.

6.2.2 PORV Position Indication Reliable and positive cont:-O rocm indication of PORV position would provide vital information to the operator in a clear and timely manne- -

to permit him to take the appropriate action necessary to prevent escalation of a minor incident into a LOCA.

An ultrasonic flow-meter, located at the discharge piping of the PORV, with flow indication and alarm in the control room, would provide direct, positive, rapid-response, and reliabla indication of PORV position.,

An advantage of this instrument is that it does not require any penetration of the-piping.

Alternatively, the PORV could be provided with a position indicator for the main valve disc position.

6.2.3 Electric Power Sucolies The PORVs and their associated block valves, which were designed for an equipment protective function rather than a safety function, were not initially provided with emergency power supplies.

The provision of emergency power to these valves would maintain the availability'of the relief system and also permit its isolation, if necessary, upon loss of all non-energency power sources.

6.2.4 Imorovement of Ocerator Cacability The evaluation of the TMI-2 incident indicated that a program to improve operator performance, particularly during emergency conditions, would significantly reduce the potential for serious nuclear incidents.

Upgrading operator capability to recognize and to respond appropriately to a PORY failure-to-close should significantly reduce the possibility of the subsequent occurrence of a small break LOCA.

7.

IMPLEMENTATION OF PORV SYSTEM FAILURE REDUCTION PROGRAM The following actions to reduce PORV system failures have been completed or are pending:

1.

The turbir.e runback feature has been eliminated from C-E operating plants.

2.

The motor operators for the PORY block valves and the pilot solenoids for the PORVs have been provided with emergency power supplies to permit them

.to function upon the loss of all non-emergency power.,.

3.

Ultrasonic flowmeters are being installed on the PORV discharge piping to provide a direct measurement of steam flow and therefore, of PORV position, with indication and alarm in the control room.

4.

Operator training programs have been initiated to provide the operator with a more comprehensive understanding of plant operation under emergency conditions.

Guidelines and detailed emergency operating procedures have been developed to aid the operator to cope with a spectrum of emergency conditions.

This includes the conditioning of the operator to recognite and respond promptly to PORY failure to prevent escalation of the failure to a small break LOCA.

8.

ANALYSIS AND RESULTS OF FAILURE REDUCTION PROGRAM An analysis was performed to provide an estimate of the reliability of the PORY l

system as well as an estimate of the improvemen't'in reliability excected as a result of the various actions taken or to be taken as noted in Section 7.

Appendix A presents a description of the reliability analysis and the results obtained.

This section provides a discussion of the analysis and results.

Table A-1 gives challenge frequencies for the PORVs and demand failure rates used in the analysis for various aspects of PORV and block valve operation.

The frequency of challenges to the PORVs is based on the C-E operating plants' experience presented in Section (4.1).

The PORV demand failure (failure-to-close) rate is based on the B&W operating experience described in Section 4.2.

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The reasons for using the B&W data as a basis are that:

1.

The C-E PORV system design basis and other NSSS features as discussed in Section 3.0 teMed to keep PORV actuations to a minimum, so that only a small statistical cata base for PORY actuations on the C-E ~NSSS was a/ailable.

2.

B&W operating plants had experienced a relatively large number of PORV actuations, and in addition, their operating plants are equipped, with one exception, with the same type of PORVs from the same supplier as are C-E operating plants.

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3.

Westinghouse operating plant experience was not included due to the fact that, in general, they used a different type of PORV from different vendors than did C-E and B&W.

The specific value of the B&W PORV demand failure rate used in the Appendix A analysis was 0.02 failures-to-close per opening.

If the C-E plant experience (38 challenges with zero failures) was statistically combined with the B&W data, the demand failure rate would be reduced by about 20% to 0.016.

A value of 0.155 was used for the probability of ' failure of the operator to I

isolate the failed-open PORV.

This value is based on data in UASH 1400(5) and is.taken as the mean between the operator's normal stress level and severe stress level failure probabilities.

Table A-2 provides the es'timated frequency of an unisolated failed-open PORV, (i.e. small break LOCA due to a failed-open PORV) for a C-E plant to which various features have been incorporated.

It shows the progressive reduction in the recurrence frequency of a small break LOCA due to a failed-open PORV as the various methods for PORV system failure reduction noted in Section 7 are implemented.

Case 1 is the reference case prior to elimination of the turbine runback feature.

This case takes no credit for operator action to isolate the failed-open PORV on the assumption that the available instrumen-l tation did not provide clear, positive valve position indication to the oper-ator.

Case 2 assumes elimination of the turbine runback feature, with no

- credit for operator action.

Case 3 is similar to Case 2, except credit is taken for. operator action on' the basis that appropriate instrumentation has

-been added to give the operator clear, positive indication of PORY position.

-Cases 4'and 5 assume that provision for automatic closure o'f the block valve

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~upon failure of.the PORV to reclose has been incorporated.

Case 4 assumes l

a control grade design which involves reliable components but has only a single isolation valve and hence is not single failure proof.

Case 5 assumes a safety-grade design with series isolation valves to provide single failure protection.

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for closure.

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1 The estimates in Table A-2 show that the elimination of the turbine runback feature and taking credit for operator action (based on positive valve position indication and alarms) serves to reduce the estimated recurrence frequency of a small break LOCA due to PORY failure by a factor of about 14.5 (or about 18 for a PW demand failure rate of.016).

The estimated recurrence frequency for a small break LOCA due to a PORY failure is 1.8 x 10-3 per reactor-year

-3 (or about 1.4 x 10 per reactor-year for a PORV failure demand rate of.016),

which is well within the 90% confidence range of a small break LOCA due to a pipe break, 10-2 to 10 per reactor-year, as estimated by WASH-1400.

Two

~4 factors which would further reduce the recurrence frequency of a small break LOCA due to PORV failure from the value before the TMI-2 accident have not been quantified.

One is the improvement in operator capability and reduction in the probability of operator error due to new intensive operator training programs, and the updating of plant emergency procedures based on guidelines l

which consider the realistic response of the plant to transients and accidents.

The second is the provision of emergency power to the PORV block valves to allow PORY isolation, if necessary, after loss of non-emergency power.

These factors p. ovide some additional confidence regarding the conservatism of the analytical results.

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Tabis A-2 also shows that provision of control grade automatic block valve l'

closure upon PORV failure to close would reduce the recurrence frequency of a small break LOCA due to PORV failure nearly to the lower limit of the range of 10 10-4 per reactor-year estimated for the small break LOCA due to i

pipe rupture by WASH-1400.

The provision of a safety-grade, single-failure-proof design for automatic block valve closure by the addition of redundant isolation valves reduces the recurrence frequency to a negligable value.

f 9.

SUMMAPV AND CONCLUSIONS li The C-E operating plants af ter approximately 29 reactor-years of operation have experienced no PORV failures during power operation.

The elimination of the turbine runback feature and the provision of a direct reliable means for indica-

[-

ting PORV position to the operator provided significant improvements in system relieoiliti.

The recurrence frequency.of a small break LOCA due to PORY failure-y.

I

has been reouced by an estimated factor of about 15 to a value of about

-3 1.8 x 10 per reactor-year.

T'his recurrence frequency is well within the 90% confidence range of the recurrence frequencies u.' 10-2 to 10'4 per reactor-year for a LOCA due to a small pipe rupture estimated in WASH-1400.

Imoroved operator training programs and emergency procedures, as well as the provision of emergency power to the PORVs and to their block valves, though not quanti-fied, has reduced the small break LOCA recurrence frequency even further.

The incorporation of the feature of automatic block valve closure upon PORY failure would further increase PORV system reliability.

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REFERENCES 1.

NUREG-0635 - Generic Evaluation of Feedwater Transients and Small-Break-Loss-of-Coolant Accidents in C-E Designated Operating Plants, January 1980.

2.

NUREG-0737 Clarification of TMI Action Plan Requirements, Nov.1980 i

3.

NbREG-0550 - Generic Assessment of Feedwater Transients in Pressurf:ed i

I Water Reactor Designed by the Babcock and Wilcox Company, May,1972.

l

-4.

NUREG-0611 - Generic Evaluation of Feedwater Transients and Small-l Break-Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants, January, 1980.

j 5.

WASH-1400 - Reactor Safety Study, October,1978 Appendix III, Table III 6-1.

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C-E PRIMARY SAFETY YALVE AND PORY D/\\TA A.

PRIMARY SAFETY hlVES DATA Valve Valve Number Setpoint

Rated Minimum

Maximum Actual Plant Vendor Type per plant psig capacity lb/hr capacity,1b/hr Ft. Calhoun Crosby llB-BP-86 2

2530 216,000 240,000 2485 212,000 236.000 Palisades Dresser 31739A 3

2565 230,000 256,000 2525 230,000 256,000 2485 230,000 256,000 St. Lucie 1 Crosby llB-BP-86 3

2485 212,000 236,000 Maine Yankee Dresser 31709KA 3

2535 218,000 243,000 2510 216,000 240,000 2485 214,000 238,000 Calvert Cliffs 1 and 2 Dresser 31739A 2

2550 304.000 334,000 2485 296,000 329,000 Millstone 2 Dresser 31739A 2

2485 296,000 329,000

Capacity indicated corresponds to 3% accumulation above set pressure B.

PORV DATA Valve Valve Number Setpdint

Relievin,q Capacity Plant Vendor Type per plant psig lb/hr Ft. Calhoun Dresser 31533VX 2

2385 111,000 Palisades Dresser 31533VX 2

2385 Id5,000 St. Lucie 1 Dresser 31533VX-30 2

2385 159,000 Maine Yankee Dresser 31533VX 2

2385 150,000 Calvert Cliffs 1 and 2 Dresser 31533VX-30 2

2385 159,000 Millstone 2 Dresser 31533VX-30 2

2400 148,000

Rated value at 0% accumulation, provided by vendor

g Summary of Events Involving PORV Operation PLANT INITIATING PLANT DATE C0flDITIONS EVEllT DESCRIPTION I

Consumers Power

Palisades Sept. 8, 1971 Mode 3 Technician deenergized PORV opened when RPS deenergized.

RPS for maintenance Baltimore Gas & Elec.

Calvert Cliffs-1 July 6, 1979 Mode 5 Test of PORV During operational test of 00dV valve failed to fully close.

Adjusted pilot valve stroke 2

August 20, 1980 100%

MSIV Closure PORVs cycled on high pressure Florida Power a Light Feb. 21, 1977 100%

100% load rejection PORV cycled during test when St. Lucie -l reactor tripped on high pressure.

Omaha Public Power Dist.

Fort Calhoun May 28, 1978 80%

Turbine control valve PORV's cycled when plant tripped closed on high pressure.

Fort Calhoun Dec. 20, 1978 Hode 5 Troubleshooting PORV's opened when technician pressure recorder pulled recorder fuses, flortheast Utilities Aug. 10, 1979'

")de 5 Troubleshooting PORV opened on loss of AC Millstone-2 to emergency bus.

Maine Yankee Atomic Power Company No PORV Operation E"ents Maine Yankee

Palisades has operated since 1972 with PORV block valve shut.

[

Tant.E 3 Summary of Events Resulting In Potential Challenge to PORV i

PLANT INITIATING PLNIT DATE CONDITI0flS l

EVENT DESCRIPTION Consumers Power Mar. 19, 1973 85%

Circuit Noise Spurious high pressure trip Palisades (Note 1)

Aug. 31, 1976 100%

HSIV shutting liigh pressure trip due to MSIV shutting.

Nov. 26, 1976 15%

Generator Synchronization Spurious high pressure trip while bringing generator on line.

May 22, 1978 100%

Closure of both MSIV liigh pressure reactor trip.

Baltimore Gas & Elec.

Calvert Cliffs -l July 8,-1975 100%

Turbine runback liigh pressure trip due to turbine

runback, tinable to verify PORV

,/o

?

operation due to loss of plant computer.

Jan. 26, 1975 20%

Power reduction with liigh pressure reactor trip.

manual pressurizer spray control Northeast Utilities Apr. 13, 1976 80%

Turbine runtiack liigh pr<rssure reactor trip.

Millstone -2 Apr. 23, 1976 100%