ML20199K938

Forwards Evaluation of Environ Qualification of Equipment Considering Superheat Effects of High Energy Line Breaks for Callaway Plant & Wolf Creek Generating Station. Rept Concludes That Equipment Will Perform Safety Functions

ML20199K938
Person / Time
Site:	Wolf Creek, Callaway, 05000000
Issue date:	04/04/1986
From:	Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SLNRC-86-06, SLNRC-86-6, NUDOCS 8604100258
Download: ML20199K938 (23)

Text

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SNUPPS Standardized Nucteer Unit Power Mant System i 5 Choke Cherry Road Nicholes A. Petrick I Rockvino, Maryland 20060 Executive Director (301) 8694010 I

April 4, 1986 SLNRC 86-06 FILE: 0278 SUBJ: Main Steam Line Break Super-heat Effects on Equipment Qualification Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Docket Nos.: STN 50-482 and STN 50-483

References:

1. SLNRC 84-0118, dated October 2, 1984: Equipment Qualifica-tion Justifications for Interim Operation.

2. Westinghouse Owners Group Letter 0G-136, G. Goering (Northern States Power) to H. Denton' (NRC), dated October 24, 1984.

3. Westinghouse Owners Group Letter 0G-162, J. Cermak (SNUPPS) to D. Wigginton (NRC), dated October 28, 1985.

Dear Mr. Denton:

In June 1984, Westinghouse Electric Company informed the SNUPPS Utilities of a potential safety concern related to the analysis of equipment qualifi-cation following postulated steam system piping ruptures with superheated steam releases outside containment. A Justification for Interim Operation (JIO) for this concern was submitted to the NRC in Reference 1 to support the licensing review of the SNUPPS plants - Callaway Plant and Wolf Creek Generating Station.

To develop plant-specific information for an evaluation of this issue, the SNUPPS Utilities joined the High Energy Line Break /Superheated Blowdowns l Outside Contaimnent (HELB/SBOC) Subgroup of the Westinghouse Owners Group.  !

The Subgroup was fomed to implement the program discussed in Reference 2. i Reference 3 reported the completion of the program and outlined a tentative '

schedule for submittal of reports to the NRC.

Enclosed is the report entitled: " Evaluation of Environmental Qualification of Equipment Considering Superheat Effects of High Energy Line Breaks for  ;

Callaway Plant and Wolf Creek Generating Station." The report concludes ,

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m e se s e PLR ADOCKOSOOg2 N*y P

i SLNRC 86- 06 Page Two that the equipment which must function to mitigate a postulated High Energy Line Break with superheat effects and to bring the SNUPPS plants to a safe shutdown condition will perform their safety functions following such a postulated event.

The enclosed report provides the basis for terminating the JIO submitted by Reference 1.

Very rul ours, i \ 6. M W

' Nicholas A. Petrick MHF/dck/5a27 Enclosure cc: G. L. Koester KGE J. M. Evans KCPL D. F. Schnell UE B. Little USNRC/ CAL J. E. Cummins USNRC/WC W. L. Forney USNRC/RIII J. E. Gagliardo USNRC/RIV i

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1.0 Background

By letter dated June 13, 1984 (reference 1), Westinghouse Electric Corpor-ation notified the SNUPPS Staff and Utilities of a potential safety issue concerning the effects of steam superheating following postulated steam

! system piping ruptures outside containment. Westinghouse analyses had shown that steam generator tube bundle uncovery may occur during a High Energy Line Break (HELB) resulting in superheating of the steam exiting from the steam generator. This could result in an increase in the energy 1 of the steam and may impact the environmental qualification temperatures of safety-related equipment outside containment which may be required to function during or after an HELB.

Meetings were held with Westinghouse on July 24 and August 3,1984, to assess the impact of this postulated safety issue on the SNUPPS plants.

At these meetings, the Westinghouse modeling of the steam line break mass /

energy release rate was discussed. Several questions were raised, and it was determined that the Westinghouse modeling did not include the effects of froth, entrainment and compressibility, all of which should tend to decrease superheat. Although SNUPPS-specific mass / energy release data were not available, Westinghouse agreed to provide typical mass / energy release information for the purpose of performing scoping studies. Scoping studies were performed, and the results were provided to the NRC in a Justification for Interim Operation (reference 2).

In references 3 and 4, the Westinghouse Owners Group (W0G) advised the NRC of actions being taken to evaluate the potential safety issue. These letters indicated that a program was being defined which would result in mass / energy release information for different classes of plants for use in plant-specific evaluations. The SNUPPS Utilities joined the High Energy Line Break /Surerheated Blowdowns Outside Containment (HELB/SB0C) Subgroup of the WOG which was formed to implement the program defined in references 3 and 4. In reference 5, the HELB/SB0C Subgroup proposed a schedule for completion of the program.

A report of mass and energy release information, including superheat effects, appropriate for the SNUPPS plants was issued on October 21, 1985 as WCAP-10961-P. (The report was provided to the NRC via reference 6.)

This information was used in the development of SNUPPS-specific environ-mental conditions resulting from the postulated mass and energy releases and the evaluation of the performance of safety-related equipment at the SUUPPS plants. A non-proprietary version of WCAP-10961-P is being prepared by Westinghouse. The results of this evaluation are provided below.

2.0 Scope of Review Because the superheat effect is caused by steam generator tube uncovery following a postulated HELB, high energy systems connected to the steam generators were reviewed to determine if a postulated HELB could lead to tube uncovery and superheated steam. In reference 1, Westinghouse stated that the impact of superheated steam on temperature response inside containment had been addressed and determined to be negligible for both dry and ice containments. The SNUPPS plants have the large, dry contain-ment design. Based on this review, it was determined that only a postu-lated Main Steam Line Break (MSLB) in the main steam tunnel could result in a safety concern.

1-T

High energy line breaks in the main steam tunnel were evaluated in the SNUPPS plant FSARs. Section 3.6.2.1.1 of the FSAR identifies the break locations for SNUPPS high energy piping systems. The main steam piping in the main steam tunnel is a "no break zone" as shown on FSAR Figure 3.6-1.

Nevertheless, in accordance with the NRC position discussed in reference 7, an MSLB in this area was analyzed for environmental effects as discussed in FSAR Section 3B.4.2. An MSLB size equivalent to a 3.41 ft 2 single-ended rupture was required. This is referred to as the 1A case. The calculated maximum pressure and temperature were 21.4 psia and 324 F for the analyzed MSLB based on saturated blowdown data provided to SNUPPS by Westinghouse during preparation of the FSAR.

Therefore, the licensing basis steam line break outside containment for the SNUPPS plants, required by previous NRC position, is the 1A break which is discussed in Section 38.4.2 of the plant FSARs.

In reference 1, Westinghouse indicated that the time of steam generator tube bundle uncovery and consequent superheating of the steam varies depending on several factors, one of which is the break area of the postu-lated pipe rupture. In addition, the program established by the WOG Subgroup included consideration of a spectrum of break sizes. Evaluation of a spectrum of break sizes resulted from analytical work performed by Westinghouse and regulatory positions taken by the NRC for another reactor plant. The need to analyze various brear 7.reas for postulated steam line breaks outside containment goes beyond the licensing basis of the SNUPPS pl ants. Nevertheless, a spectrum of break sizes was analyzed to assure that the most limiting break was considered when evaluating time margins for equipment function as discussed in the Conclusion below.

3.0 Evaluation The evaluation of MSLB superheat effects on equipment is performed in three phases. Phase 1 is the determination of mass and energy releases which include steam superheat. Phase 2 is the calculation of room environ-ments which result from the mass and energy released to the room. Phase 3 is the evaluation of equipment performance in the calculated room environ-ments. Each of these phases is discussed separately below.

3.1 Mass and Energy Releases For the mass / energy release calculations in WCAP-10961-P, the SNUPPS plants were included in a plant category (Category 1) based on plant size and power level (4 loop, 3425MWt), steam generator type (D4), and steamline break protection system design (described on Table 3.1). The SNUPPS plants conform to these parameters with the exception of steam generator type since the SNUPPS plants have Model F steam generators. A sensitivity study was performed which compared Model F, Model D4, Model DS and Model 51 steam generators using actual steam generator operating characteristics. This study demonstrated that use of the Model D4 steam generator was conserva-tive for the SNUPPS plants in that peak break enthalpy is 13 BTU /lb lower for the Model F steam generator for the steamline break mass / energy release results.

i - -

An application for reload license amendment. for Callaway Plant has been submitted to the NRC (reference 8). This report analyzes the Callaway Plant at a power level of 3579 MWt. Sensitivity studies on core power have indicated that there is a small increase in break flow, break energy and superheat enthalpy as a result of increased power level, e.g.,1 BTV/lb in peak enthalpy between 3425 MWt and 3579 MWt. The use of the Model D4 steam generator in the analysis ensures a conservative mass / energy release even for a power level of 3579 MWt, because the peak break enthalpy is conser-vative by approximately 12 BTU /lb and the integrated break flow is lower by approximately 5% for the SNUPPS Model F steam generators.

Sections II.C and III.B of reference 6 provide analysis inputs and assump-tions used in the mass / energy release calculations. A modified version of the LOFTRAN Code (reference 9) was used by Westinghouse. Table 3.2 provides comments on the analysis inputs and assumptions and discusses additional assumptions which are applicable to the SNUPPS plants. Conservative assump-tions and inputs used in the analysis include:

a. Decay heat

b. Core reactivity coefficients

c. Availability of offsite power

d. Main feedwater control system response

e. Steam generator level

f. Protection system response times

9. Single failure of one safety injection train The plant response sequence of events is provided in Table 111.B-4 of ,

reference 6. The applicable case numbers on Table III.B-4 are cases 59 through 63. The data for these cases is applicable to the Sh0PPS plants with the exception of the time of steamline isolation for the 0.7ft2 ,

0.5ft2 and 0.2ft2 breaks. For these break sizes, a steamline isolation signal is assumed to occur ten minutes after reactor trip in accordance with assumption 17 on Table 3.2. Steanline isolation time includes an appropriate delay from the time the signal is gen 6cated until the valves are closed in accordance with reference 6, The mass and energy release results are presented in Tables A-1.59 through A-1.63 of reference 6.

Based on an evaluation of the reference 6 data, it was determined that break sizes below 0.5ft2 need not be considered in the calculation of environmental conditions in the steam tunnel. This follows because the 0.2ft2 case does not result in steam generator tube uncovery until well after the operator response time. If tube uncovery does not occur, then superheating of the steam also does not occur. In addition. the 0.5ft2 case does not result in significant superheating of the steam until after the operator response time. This was confirmed in the calculation of room temperature for the 0.5 ft2 break case, since the worst-case rcom temper-ature did not exceed the peak room temperature used in the FSAR analysis of  ;

pipe breaks in the steam tunnel until 10-1/2 minutes after a reactor trip '

occurred for this case. Therefore break sizes less than 0.5ft2 were not considered in the calculations of environmental conditions.

Finally, it was concieded that the full power (102%) cases are more limiting than the 70% power cases evaluated in reference 6. Therefore, only the full power cases were further evaluated for equipment qualifica-tion effects.

• 0 0

3.2 Environmental Conditions The mass / energy release data developed as discussed in 3.1 above was used to calculate the environmental conditions in the steam tunnel of the SNUPPS ,

plants for four break sizes: 0.5ft2, 0.7ft 2, 1.0ft2 and 4.6ft2. The temperatures and pressures in the steam tunnel as a function of time fol-lowing the postulated pipe breaks were calculated using a modified version of the Bechtel computer program FLUD (NE017). FLUD is a program to calcu-late pressure / temperature transients caused by steam and/or water blowdown into a system of interconnected compartments. The Westinghouse-supplied blowdown and the physical characteristics of the steam tunnel compartuents (i.e. volumes, vent areas, flow coefficients and heat sink dimensions and materials) were input to the program and the time dependent room environ-ments were calculated. The modified version of FLUD permits the revapor-ization of condensate which forms on the heat sinks. Per the guidelines given in NUREG 0588, Rev. 1, when the room environment was superheated, the rate of condensation was reduced by a maximum of 8%. This was done to approximate the revaporization of the condensate.

In the Westinghouse-supplied mass and energy release data (reference 6), it was assumed that auxiliary fe.edwater addition to the faulted steam generator continued for 30 minutes. As discussed in Table 3.2 and in Section 3.3 below, the expected SNUPPS operator response time is 10 minutes following adequate alertment to the accident situation. Thus, for break sizes for which the steam mass release rate is greater than the assumed auxiliary feedwater flow rate after 10 minutes, it is not possible to subtract the auxiliary feedwater flow rate from the reference 6 mass release rate with any degree of confidence regarding the resultant enthalpy of the blowdown.

For the 0.5 ft2 and 0.7 ft2 break sizes, the steam mass release rate is greater than the auxiliary feedwater flow rate well beycnd 10 minutes.

Therefore, the assumed termination of auxiliary feedwater to the faulted generator at approximately 10 minutes would result in earlier steam gener-ator tube uncovery and earlier superheating of the steam than assumed in reference 6 for the 0.5 ft2 and 0.7 ft2 break cases. However, the increased superheat would develop after equipment has been actuated to its safe, post-accident condition as discussed in the following Section and would not have an adverse impact on the conclusions of this report.

The results of the calculation are presented in Figures 3.2-1 through 3.2-4.

The SNUPPS steam tunnel consists of two volumes connected by a vent opening (ref. Figure 3B-2 of the FSARs). The temperature of the non-break volume (East compartment on the figures) lags considerably behind the volume where the break occurs (West compartment). The higher temperature in the West compartment was applied throughout the equipment evaluation for conservatism.

The peak temperatures ard pressures are:  ;

Break Case Peak Pressure / Time Peak Temperature / Time 4.6ft2 17.5 psia /0.29 sec 469.9*F/200.0 sec f 1.0ft2 16.0 psia /200.7 sec 441.7 F/436.0 sec 0.7ft2 15.8 psia /200.7 sec 431.6 F/758.D sec 0.5ft2 16.0 psia /1001.0 sec 367.0'F/1800.0 sec The calculated pressure values are well below the qualification requirements used for safety-related equipment in the steam tunnel. However, tha calcu-lated temperature values exceed the qualification requirement.s previously

used for equipment in the tunnel. Therefore, the thermal respon?. of the equipment was determined as discussed in the following section.

3.3 Equipment Performance Table 3.3 identifies the safety-related equipment located in the steam tunnel at the SNUPPS plants. This Table also discusses the equipment function following a po;tJlated MSLB and lists the temperature to which the equipment has been qualified. The performance of the equipment when subjected to the environmental conditions calculated in 3.2 above was evaluated and discussed below. In the evaluation of equipment performance, use was made of equipment thermal response (i.e., surface temperature or thermal lag analysis) to demonstrate the proper operation of equipment before it was calculated to be heated above its qualified temperature by the superheated steam. Failure modes and effects analyses were also employed, when required, to evaluate certain electrical circuits and determine equipment performance. In these analyses, conductor-to conductor and conductor-to-ground short circuits were evaluated. Cable failures of the " hot short" type were not considered to be credible failure modes.

The surface temperature response was calculated for various representative pieces of equipment and components which nay be required following an MSLB in the steain tunnel. The representative equipment for which surface temper ature calculations were performed enveloped the equipn.ent listed on Table 3.3 which must function following an MSLB, The most severe room conditioris (those for the break ccmairtment) and the flow characteristics from the FLUD calculation results were used in the calculation of the time dependent equipment surface temoeratures, Currently, there is no formal NRC guidance for convcetive heat transfer correlations for equipment outside containment. Therefore, based on existing NUREG-0530 guidelines, the equipment surface temperatures were evaluated thiough the use of con-servative, yet reasonable, heat transfer c'aefficfeats. At any Oiven time, the greater of four times tire bchida condensing neat transfer rate (based on the compartment air to steam mas: ratio) or' the convective heat transfer rate was used to evaluate the transient sortac#e te.mperature esponse of the selected equipment. The Hilpert correlatior., for flow past an object in a fluid stream, With consideration for syStent turbulence, was used to calcu-late the convective heat transfer coefficient. In the cvoluation of ths neat transfer coefficiera for a corycnant, (.he chractepistic velocity was taken as the time depciident iverhce velocity of the flee between the east and west rooms oY the' steam tenne). The flow h tween these two compartments ,

represt;nts apprcximdtely Aalf of the bloudown. This results in fluid vclocities well in excess of those expected to occur in the vicinity of the eyof pment inodelled. The film oropertie3 used in the evaluation of the Hilpert equatioa were based on the state of the air and stean in tile strean:. AC Oniv the outpide casing of cavipment was modelled, the Lumppd-CapacL+y taEhori was ssed to calcolate f.no surface toQerature response of the equipment. This approach is justified by thc thinness ar:d the high j conductivity of the i.ndalled exte(dal casings, For the Main Steam Isolation Valo W Mair 'feehater Isolation Valve terminal blocks l'ocated in terminal boxes on the valve actuators, more netailed, two-dinensional thermal anal-ytes wtre prformed tc detennine equipment tersperatures. 'fhe calculated roca environment Conditions and tne same condensing and tonvertive heat transfer coefficients, as discussed above, were applied. The heat transfer to the tensinal blocks uas modelled via conduct $on through the back of the

-5s

f terminal box an,d via convection and radiation across the air. gap from the t

inside surface of the front of the box. Thus, using conservative and l

reasonable assumptions and techniques previously approved by the NRC, the

temperature responses of several limiting, representative equipment types '!,

were evaluated. The results of the evaluation are provided in Table 3.4 ,

which will be referenced frequently during the following discussion of -

i. equipment performance.

, Equipment in the steam tunnel which is required to mitigate an MSLB is actuated by the following control signals:

I j a. Steam Line Isolation Signal (SLIS), or j b. Feedwater Isolation Signal (FWIS), or j c. Steam Generator Blowdown System Isolation Signal (SGBSIS), or

! d. Auxiliary Feedwater . Actuation Signal for Turbine Driven Auxiliary l Feedwater Pump (AFAS-TD) 4 j In addition, the equipment can be manually actuated by the operators in the Control Room. Following Ln MSLB, the response time of the Control Room operators from the time of receiving adequate warning until taking mitigating i action is assumed to be 10 minutes. Adequate warning is considered to be a j reactor trip which would cause the operators to initiate event diagnosis l under the plant operating procedures.

4 j A. Main Steam Pressure Transmitters L

! t j The mass / energy release analysis of reference 6 relied on the Low Steamline 4 Pressure (LSP) signal only to initiate a Steam Line Isolation Signal (SLIS).

j The High Steamline Pressure Rate signal was not used. Steamline pressure j and pressure rate signals are generated by twelve pressure tra..smitters i (3 per each steam line) located in the steam tunnel. The transmitters have

been au811fied under Westinghouse program-ESE-1 to inside containment con-

ditions, including 420*F and 57 psig. As shown on Figures 3.2-1 through

' 3.2-4, the room temperature may exceed 420*F in the vicinity of the break l location for some postulated break sizes; however, when the temperature response cf the transmitters in a superheated steam environment is consi-i dered, as shown on Table 3.4, the transmitters are shown to provide an SLIS well before their qualified temperature is exceeded.

! The transmitters provide their signals via instrument cable which is run

' in solid and flexible conduit in the' steam tunnel. As shown on Table 3.4, i the qualified temperature of the instrument cable (340*F) is exceeded by the conduit surface temperature prior to an SLIS for breaks in the 0.7ft2 -1.0ft2 range. Also, the qualified temperature is eventually exceeded for all break j sizes which precludes use of the transmitters for long-term post-accident i monitoring. A failure modes and effects analysis concluded that-the cable j failure modes would either have no effect on' transmitter performance or

result in a loss of signal which would cause an SLIS and key the plant oper-l ators to use alternate pressure transmitters.for post-accident monitoring.

As discussed in FSAR Appendix 7A, Data Sheet 4.2, the steamline pressure '

j transmitters which are used for control of the atmospheric relief valves can )

be used to monitor this variable. These Class 1E transmitters are not l

! located in the steam tunnel and can be used for long-term post-accident ,

l monitoring of steamline pressure. '

[ .

An SLIS initiates closure of the Main Steam Isolation Valves (MSIV) (Spec-ification M-628), the HSIV Bypass Valves (J-601A), Steam Line Drain Valves (J-601A) and the Turbine Driven Auxiliary Feedwater Pump (TDAFP) Keep-Warm Valves (J-601A).

B. Main Steam Isolation Valves The MSIVs are fast acting and close in 5 seconds or less upon receipt of either a protection channel 1 or 4 SLIS. Each MSIV has a dual actuator with redundant active electropneumatic/ hydraulic components (ref. FSAR Section 10.3.2.2). The MSIV actuators have been qualified for MSLB con-ditions by test and analysis to 450*F; nowever, the actuator components and appurtenances have various qualification temperatures:

1. Hydraulic Components - 450*F

2. Pneumatic Components (including solenoid valves) - 450*F

3. Wiring - 346 F*

4. Terminal Lugs - 352 F*

5. Terminal Blocks - 300 F (Wolf Creek), 312*F (Callaway)*

6. Limit Switches - 342 F

7. Conax Seals - 420*F

• Inside terminal box on actuator.

In addition, the electrical signal to close the MSIV's is provided via electrical control cable routed in conduit and junction boxes in the steam tunnel. And the valve limit switches provide their signal via instrumentation cable routed in conduit and junction boxes inside the steam tunnel.

The MSIVs are not required to be used again af ter they initially close, and a failure modes and effects analysis has verified that the MSIVs will not reopen as a result of environmentally induced failures of MSIV actuators or control cable. After the valves are closed, the subsequent failure of

the actuators or appurtenances would not mislead plant operators into per-forming actions adverse to plant safety.

The mechanical portions of the MSIVs (valve bodies, packing, etc.) are qualified to much greater temperatures than are postulated to occur foi-lowing steamline breaks in the tunnel.

Table 3.4 shows that, at the time SLIS is initiated for each break size, all the components, appurtenances aad terminal boxes are below their qualified temperatures with the exception of the electrical control cable for one MSIV at Wolf Creek Generating Station and instrumentation cable for MSIV limit switches. The qualified temperature for the affected MSIV con-trol cable is exceeded by 12 F, while the instrumentation cable qualified temperature is exceeded by 42*F.

Valve position indication is the only post-accident function of the HSIV limit switches. Failure of the limit switch instrument cable would result in loss of MSIV position indication. However, indication of steam generator isolation can be determined by use of alternate equipment such as steam generator level transmitters, steam generator pressure transmitters, main steam flow transmitters, auxiliary feedwater flow transmitters and reactor coolant system temperature detectors, all of which are not affected by the tunnel environment.

. - _ _ - - - _ - - . - - _ . . - - . .. . = - _ _ - --- _ _ -

The control cable insulation system for the one affected MSIV consists of cross-linked polyethylene (XLPE) with a neoprene jacket. Test results

for similar control cable used at the SNUPPS plants (XLPE, with hypalon j jacket) have demonstrated a qualified temperature of 385'F. Also, the

! qualified temperature of the MSIV control cable is exceeded for a short i period of time (approximately 1 minute) prior to SLIS. Based on these

! considerations, the control cable is expected to perform its function even

though its qualification temperature is exceeded. Nevertheless, should the control cable for this one MSIV fail in a mode which would prevent that MSIV t

from closing, the results would be identical to the assumed rupture of a steamline in the pipe-break-exclusion-area upstream of an MSIV. Because of the SNUPPS steam system design, closure of any three MSIVs allows only one steam generator to blow down following an MSLB as analyzed in the plant FSARs.

C. Air-0perated Control Valves i

The J-601A valves which receive a closure signal upon SLIS are air-operated l globe valves which fail closed on loss of air or electrical control power.

The valves would also fail closed if the actuator diaphragm were perforated

! as a result of a high temperature environment. The actuator and appurte-nances for J-601A valves were qualified by test to 335'F.

]

{ The control cable for the J-601A solenoid valves is routed in conduit and junction boxes in the steam tunnel. As discussed above for MSIV control cable, this cable is expected to Tunction acceptably until an SLIS occurs; 3

nevertheless, a failure modes and effects analysis for this cable concluded j that environmentally induced failures would either cause the valve to fail j to its closed position or not prevent an SLIS from tripping the valve to the closed position. Therefore, control cable qualification for the J-601A, SLIS applications is not required. Similarly, a failure of the solenoid would cause the solenoid valve to reposition such that air pressure would

) be vented from the actuator diaphragm and the air operated valve would fail

closed. As shown on Table 3.4, the solenoid valves would respond to an SLIS

well before their qualified temperature is exceeded- for all break sizes.

]

These valves do not need to be repositioned from their safe (closed) posi-j tion following an MSLB, and failures of these valves in the closed position

{ would not mislead the plant operators to perform actions adverse to safety.

j Each J-601A valve is equipped with one or more limit switches which provide valve position indication and, for some valves, a control function. Failure

! modes and effects analyses have been performed on the valve circuits and l in all cases potential failure modes result in 0 loss of indication or a i failure of the valve to its safe position or uoth. Limit switch circuit failures would not result in repositioning of the valve once it is in its safe position, nor would they prevent the valve from moving to its safe position if they were to occur prior to the valve receiving its actuation

signal.

! Valve position indication for the MSIV Bypass Valves and the Steam Line j Drain Valves is a post-accident monitoring function.- As discussed above for-i the MSIV position indication, the plant operators can determine steam gen-

erator isolation by use of alternate indications such as steam generator l pressure and level and steam flow. So loss of position indication for these j J-601A valves is not a safety concern.

j L _ _ _ _ _ _ _ _____ - _ ____ _

As shown on Table III.B-4 of reference 6, the Feedwater Isolation Signal (FWIS) and Safety Injection Signal (SIS) are generated well before steam generator tube uncovery. The SIS causes an FWIS and an SGBSIS. These signals actuate the following steam tunnel equipment: An FWIS closes the MFIVs (Specification M-630) and the Feedwater Chemical Addition Valves (J-601A); an SGBSIS closes the Steam Generator Blowdown Isolation Valves (J-601A).

D. Mair. Feedwater Isolation Valves Like the MSIVs, each MFIV is fast acting (closure in less than 5 seconds),

receives closure signals from protection channels 1 and 4 and has a dual actuator (ref. FSAR Section 10.4.7.2.2). The MFIV actuators, appurtenances and control circuits have been qualified to the same parameters as the MSIVs. Therefore, in accordance with Table 3.4, the MFIVs will be closed following receipt of an FWIS well before the qualified temperatures are exceeded. The MFIVs are not required to be used again after they are closed, and a failure modes and effects analysis has verified that the MFIVs will not reopen as a result of environmentally induced failures in the actuators or control or position indication circuits. Plant operators would also not be misled into performing actions adverse to safety by the subsequent failure of actuators or appurtenances in a harch environment.

l The mechanical portions (valve stems, packing, etc.) of the MFIVs are qual-ified to much greater temperatures than are postulated to occur following steam line breaks in the tunnel.

E, fir-0perated Control Valves Certain J-601A valves receive a closure signal from an FWIS/ SIS or SGBSIS/ SIS.

These valves have the same failure modes as discussed in Section C above; however, because they close on receipt of an SI signal, the valves will have closed prior to the actuator, appurtenances, control circuits and terminal boxes exceeding their qualified temperatures (refer to Table 3.4).

Postulated failures of control circuits in the steam tunnel would not cause the valves to reposition from their safe position.

As shown on Table 3.3, the mechanical portion of J-601A Control Valves is qualified to much greater temperatures than are postulated to occur in the tunnel following an MSLB.

Af ter the valves are closed, the postulated failure of limit switches or limit switch instrument cable would result in a loss of indication of valve position; however, the valves would not reposition.

F. Turbine Driven Auxiliary Feedwater Pump Steam Supply Valves The steam supply valves for the Turbine Driven Auxiliary Feedwater Pump (TDAFP) are also J-601A valves. These valves open early in any transient which results in a reactor trip. The stearn generator level response for the SNUPPS plants following a reactor trip causes low-low steam generator levels to occur. Low-low steam generator levels in any two steam generators results in an AFAS-TD. Since reactor trip occurs at the same time as or before an SIS for all break sizes, the discussion above for J-601A valves which respond to an SIS applies to the TDAFP steam supply valves. In the ,

l l

_9

case of the TDAFP steam supply valves, the valves fail open on loss of air pressure or electrical control power, and a rupture of the actuator diaphragm causes the valves to open. A Failure modes and effects analysis of the electrical control and indicating circuits, including junction boxes, for these valves concluded that failures due to high environment temperature either result in the opening of the valves or would not prevent the automatic or manual opening of the valves.

G. Steam Generator Atmospheric Relief Valves The Steam Generator Atmospheric Relief Valves (ARV) are located, together with their actuators, controllers and appurtenances, in the steam tunnel.

Electrical control circuits for the ARVs are routed via conduit and junc-tion boxes in the tunnel. The ARVs are relied on to perform a controlled plant cooldown following an MSLB. The ARV actuators were qualified under Specification J-601B to MSLB conditions including a temperature of 335'F.

For all MSLBs with superheated steam conditions in the tunnel, the quali-fied temperatures of ARV-related equipment will be exceeded. When the qualified temperature of the xRV controller is exceeded, the controller is expected to fail in a mode which would keep the valves closed. Al so, if the valve diaphragm were to fail in the harsh environment, the valves would remain closed. In these cases, the function of steam generator heat removal would be performed by the Main Steam Safety Valves (refer to Spec-ification M-140 on Table 3.3). Long term plant cooldown could be performed if required, after the faulted steam generator was secured, via local manual control of the ARVs or Main Steam Safety Valves.

In the event that the ARV controllers failed such that the valves open, then the three intact steam generators would blow down via the ARVs to the atmosphere. This type of blowdown has been analyzed by Westinghouse during the development of the Emergency Response Guidelines (ERGS) for the WOG.

Guideline ECA 2.1, " Uncontrolled Depressurization of All Steam Generators" in Revision 1 to the ERGS, considers the case of an MSLB with failure of all MSIVs to close. This would result in a more severe plant transient than an MSLB with uncontrolled opening of ARVs on three intact steam gener-ators. The analysis of this event concluded that a stabilized plant and a safe cooldown can be achieved with a flow equivalent to one Motor-Driven Auxiliary Feedwater Pump (MDAFP). Based on the discussion in Section F above, the SNUPPS plants, which have a complement of two MDAFPs and one TDAFP, would have more than the minimum required feedwater flow to remove reactor decay heat and cooldown the plant, even though the TDAFP would not be operable after sufficient steam generator pressure was lost. Therefore, in the unlikely event that the ARVs failed open, the SNUPPS plants can be brought to a safe shutdown condition using the methods of ECA 2.1.

H. Mechanical Equipment The results of the SNUPPS mechanical equipment environmental qualification review have been used to evaluate the capability of safety-related mechan-ical equipment to function following an MSLB with superheat. The equipment specifications are M-140, M-157, M-224A, M-224B, M-231C and M-771. (Refer to Table 3.3.). Also, the mechanical portions of the Control Valves, Atmo-spheric Relief Valves, MSIVs and MFIVs (specifications J-601A, J-601B, M-628 and M-630) were evaluated using the mechanical equipment qualification program results. In all cases, the qualification temperatures for the

=

l mechanical equipment exceed, with greater than 15'F margin, the calculated MSLB superheat temperatures for all break sizes. j I. Structures During the evaluation of steam line breaks in the steam tunnel, the effects of superheated steam temperatures on the tunnel structures was considered.

It was concluded that the higher temperatures resulting from superheated blowdown (maximum 469.9 F for a 4.6ft2 break) will have no detrimental effects on the steam tunnel structural steel and reinforced concrete due to the relatively short duration of these events.

4.0 Conclusion Based on the above information, the environmental qualification of equipment for superheat effects at the SNUPPS plants was reviewed for compliance with regulatory requirements. .

The equipment which is exposed to the environmental conditions of an MSLB with superheat effects in the steam tunnel and which is required to func-tion to mitigate this postulated event and/or to bring the plant to a safe shutdown condition has been discussed in Section 3.3 above. During this evaluation, the Steam Generator Atmospheric Relief Valves were found to be not required for MSLB mitigation or safe shutdown following an MSLB in the tunnel. The failure of these valves, in any mode following an MSLB in the tunnel, was determined to be not detrimental to plant safety or accident nitigation.

Most of the electrical equipment which must function following an MSLB in the tunnel can be charactesized as equipment which performs its safety function early in the event prior to exceeding its qualified temperature and whose subsequent failure will not result in a safety concern. The qualified temperature of this equipment is based on the NUREG-0588 review performed during licensing of the SNUPPS plants. The SNUPPS NUREG-0588 review addressed the qualification program requirements of current NRC regu'l ations. The issue of adequate qualification time margin for equipment

, which performs its function within a short time period into the event is adequately addressed because the evaluation considered a spectrum of MSLB sizes, the evaluation considered the potential need for the equipment later in the event, the evaluation considered whether equipment failure was detrinental to plant safety or could mislead plant operators and the eval-uation was based on conservatively developed mass and energy release values to provide margin between the time the equipment function is completed and

, the time that environmental qualification temperatures are exceeded.

Therefore, it has been concluded that the equipment in the steam tunnel

, which must function to mitigate a postulated MSLB in the tunnel and/or to bring the plant to a safe shutdown condition will perform its safety func-tion in the environmental conditions following an MSLB including super-heated steam effects, i

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References

1. SNP(S)-1005, dated June 13, 1984: Notification of Unreviewed Safety Questions.

2. SLNRC 84-0118, dated October 2,1984: Equipment Qualification Justifi-cation for Interim Operation.

3. Westinghouse Owner's Group letter (G. Goering, Northern States Power) to NRC (H. Denton), 0G-128, dated July 26, 1984.

4. Westinghouse Owne. 's Group letter (G. Goering, Northern States Power) to NRC (H. Denton), 0G-133, dated August 20, 1984.

5. Westinghouse Owner's Group, High Energy Line Break /Superheated Blowdown Outside Containment Subgroup letter (J. Cermak, SNUPPS) to NRC (D.

Wigginton), 0G-145, dated February 25, 1985.

6. Westinghouse letter (P. Rahe) to NRC (H. Thompson), dated January 17, 1986: Submittal of WCAP-10961-P.

7. NRC letter (O. Parr, NRC) to the SNUPPS Utilities, dated October 17, 1977: Design of Valve Room for Main Steam and Feedwater Line Valves in SNUPPS Plants.

8. Union Electric Company letter (D. Schnell) to NRC (H. Denton), ULNRC-1207, dated November 15, 1985: Application for Reload License Amendment Using Westinghouse Optimized Fuel Assemblies.

9. Westinghouse letter (P. Rahe) to NRC (C. Thomas), NS-NRC-85-3009, dated February 27, 1985: Topical Reports WCAP-8822-P-SI and WCAP-8860-SI, " Mass and Energy Releases Following a Steam Line Rupture" l

I MHF/dck/5a29

s t

Table 3.1 Design of Steamline Break Protection System A. Safety Injection Signals

1. Low steamline pressure (LSP)

2. Low pressurizer pressure (LPP)

3. High containment pressure (for breaks inside containment only)*

B. Steamline Isolation Signals

1. Low steamline pressure (LSP)

2. High steamline pressure rate *

3. High-High containment pressure (for breaks inside containment only)*

I

• These parameters were not used in the analysis.

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Table 3.2 Analysis Inputs and Assumptions Comments *

1. Initial Power Level: No comment.

2. RCS Temperature: No comment.

3. RCS Pressure: No comment.

4. RCS Loop Flow: No comment

5. Decay Heat: No comment.

6. Core Reactivity Coefficients: No comment.

7. Rod Control: No comment.

8. Availability of Offsite Power: No comment.

9. Main Feedwater System: No comment.

10. Auxiliary Feedwater System: In accordance with actual response char-acteristics of steam generator levels at the SNUPPS plants, the Turbine Driven Auxiliary Feedwater Pump is assumed to start automatically following a reactor trip, with a conservative time delay included.

11. Steam Generator Fluid Mass: No comment.

12. Break Sizes: No comment.

13. Safety Injection System Flowrate: No comment.

14. Boron Injection Tank: No comment.

15. Low Steamline Pressure Setpoint: The analysis assumed a 379 psia set-point for Low Steamline Pressure. The comparable SNUPPS setpoint is 394 psia, with all errors included.

16. Single Active Failure: In accordance with established NRC position, a single active failure of safety equipment was not assumed concurrent with a postulated pipe break in a pipe break exclusion area as defined in NRC Standard Review Plan 3.6.1. An exception to this statement is the treatment of Safety Injection as described in item 13 above.

17. Operator Response Time The time required for the plant operators to take mitigating action was assumed to be 10 minutes following receipt of adequate warning.

This is consistent with expected operator response as discussed in FSAR section 6.2.1.4.3.3. Adequate warning was considered to be a reactor-trip.

• "No comment" for the items below means that the assumption used in the Westinghouse analysis is applicable without modification or clarification.

i Table 3.3 Safety-Related Equipment in the Steam Tunnel Specification Description Qualified Temperature Function E-028 Terminal Boxes 346 F Provide terminations for air-operated valve circuits.

E-057 600V Control Cable 346 F Electrical cable for operation and E-057A 600V Control Cable 447*F performance monitoring of safety-related equipment.

E-057B 600V Control Cable 385*F E-062 600V Instrument Cable 340*F J-601A Control Valves 335*F (actuator) Post-MSLB functions: Main steam isolation 800 F (valve) bypass valves, SG blowdown isolation valves, steam line drain valves, hydrazine addi-tion valves and turbine driven auxiliary feedwater pump steam supply and keep-warm valves.

J-601B Atmospheric Relief 335 F (actuator) Used but not necessary for control of plant Valves 800 F (valve) cooldown.

M-140 Safety Valves All metal Required for main steam pressure relief (rated 600 F) post-MSLB.

M-157 Steam Strainers 1300 F Required post-MSLB for pressure boundary function.

M-224A Anchor Darling Valves 1112 F Required post-MSLB for pressure boundary function.

M-2248 Velan Valves 1112 F Required post-MSLB for pressure boundary function.

Table 3.3 Continued Specification Description Qualified Temperature Function ~

M-231C Borg-Warner Valves 1112 F Required post-MSLB for pressure boundary function.

M-628 Main Steam Isolation 324 F (actuator) Required post-MSLB for steam line Valves 800 F (valve) isolation.

i M-630 Main Feedwater Isolation 324 F (actuator) Required post-MSLB for feedwater line Valves 1200 F (valve) isolation

• M-771 Instrumentation Devices 420 F Required post-MSLB for pressure boundary (NSSS) integrity.

Ji(ESE-1) Pressure Transmitters 420 F Required post-MSLB to generate low steamline pressure signal.

W

-(HE-8) Electrical Connectors 420*F Required for environmental qualification of electrical circuits.

I

~

Table 3.4: Room / Equipment Temperature ( F) at Actuation Time SLIS FWIS Qualified BREAK SIZE 4.6ft2 1.0ft2 0.7ft2 0.5ft2 4.6ft2 1.0ft2 0.7ft2 0.5ft2 Temperature Peak Room Temperature 308 392 366 323 308 295 294 290 NA Main Steam Pressure Transmitters 237 345 345 312 237 228 233 228 420 Main Steam Pressure Transmitter 248 363 358 320 248 235 232 245 340 Instrument Cable MSIV/MFIV Solenoid Valve 230 316 320 290 230 220 219 218 450 MSIV/MFIV Wiring and Lugs 233 337 340 303 233 222 223 223 346 (wiring)/

352 (lugs)

MSIV/MFIV Terminal Blocks (l) 155 250 255 219 155 135 139 144 300(2)

MSIV/MFIV Control Cable 244 358 355 319 244 245 242 241 385(3)

MSIV/MFIV Limit Switch 230 336 340 306 230 223 224 224 342 MSIV/MFIV Conax Connector 248 363 358 320 248 240 243 242 420 MSIV/MFIV Limit Switch 299 '382 364 322 299 287 283 283 340 '

Instrument Cable ,

J-601A Solenoid Valve 230 316 320 290 230 220 219 218 335 J-601A Control Cable 248 363 358 320 248 250 245 245 346 NOTES: (1) The temperatures provided are based on a detailed, two-dimensional thermal lag analysis.

(2) Wolf Creek terminal blocks are qualified to 300 F; Callaway, to 312 F.

(3) One MSIV at Wolf Creek has cable qualified to 346 F.

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