ML20234C717
| ML20234C717 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 06/29/1987 |
| From: | Capone D UNION ELECTRIC CO. |
| To: | Forney W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| REF-PT21-87 ULNRC-1514, NUDOCS 8707060628 | |
| Download: ML20234C717 (6) | |
Text
OM b N enliy Ec0TisG g
Etsc'}1%c V=W n
n': g 1901 Gratiot Street. St. Louis
.f L
i FILE June 29, 1987 Mr. W.
L.
Forney Chief, Reactor Project Branch 1 U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Forney:
ULNRC-1541 DOCKET NUMBER 50-483 CALLAWAY PLANT ENVIRONMENTAL QUALIFICATION INSPECTION CLOSEOUT Provided herewith are reports which address questions raised in a telephone conversation with Mr.
A.
Gautam of Region III on June 25, 1987.
The two issues discuss <ed involve the ongoing review of the main steamline break superheat issue and a recent Part 21 report filed by Sorrento Electronics on high range radiation monitors.
These were categorized as open items from the Environmental Qualification Inspection at Callaway from May 4-8, 1987.
We believe the information provided is responsive to the concerns raised and will allow completion of the inspection report.
Very truly yours, A
s D.
W.
Capc"ne Manager, Nuclear Engineering DS/plh l
p74We NJ gCo I
i QUM 301987 Mailing Address' P O. Box 149, St. Louis. MO 63166 a
i s
l cc:
Tom Alexion Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 316 7920 Norfolk Avenue Bethesda, MD 20014 Bruce Little Callaway Resident Office U.S. Nuclear Regulatory Commission RR#1 Steedman, Missouri 65077 k
j
bec:
3456-8575 3456-0021.6 Nuclear Date DFS/ Chrono D.
F.
Schnell J.
E. Birk 3
J.
F. McLaughlin A.
P. Neuhalfen R.
J.
Schukai M.
A.
Stiller G.
L. Randolph D. E.
Shain H. Wuertenbaecher D. W. Capone A.
C.
Passwater R. P. Wendling T. H. McFarland R. D. Affolter D. E.
Shafer D.
J. Walker 0,.
Maynard (WCNOC)
N. P. Goel (Bechtel)
G56.37 (CA-460)
Compliance (J. E. Davis)
NSRB (Sandra Auston)
CFA G.
Charnoff
ULNRC-1541 MAIN STEAMLINE SUPERHEAT/ PRESSURE TRANSMITTERS As a result of the NRC's review of WCAP-8822, superheating of steam during a main steam line break was identified as a concern for Westinghouse reactors.
After meetings and discussions with the NRC, Westinghouse determined in early 1984 that environmental qualification temperature envelops may be adversely impacted by the superheated steam for breaks outside containment.
Westinghouse identified this issue to the utilities in June 1984.
The NRC expressed a similar concern regarding equipment qualification and met with the WOG Regulatory Response Group.
Based on this meeting, the NRC agreed that an extended schedule was acceptable for addressing the superheat issue.
The NRC also requested that Westinghouse develop a generic JIO for use by affected plants.
Using the Westinghouse generic JIO and additional plant specific data, SNUPPS submitted a JIO for the superheat issue by SLNRC 84-0118 dated October 2, 1984.
This type of JIO is in accordance with 10CFR50.49(1).
To further address technical aspects of the superheat issue the SNUPPS plants joined a WOG subgroup.
This is discussed in the introduction to the main steam line break superheat report SLNRC 86-06 dated April 4, 1986 (attached).
By memorandum from Hugh L. Thompson, Jr. to Edward L.
Jordan dated July 15, 1985, the NRC accepted the schedule established by the WOG subgroup for resolution of the superheat issue which anticipated this issue being resolved later than November 30, 1985.
Westinghouse issued revised mass and energy release data considering superheat via WCAP-10961-P on October 21, 1985.
In SLNRC 86-06 the SNUPPS plants submitted the completed main steam line break superheat report to close out the previously submitted JIO and demons rate qualification of equipment per 10CFR50.49.
This report is under active review by the NRC.
As discussed in the attached documents, the steam line pressure transmitters perform their safety function of providing a Steam Line Isolation Signal well before their qualified temperature of 420 F is exceeded.
For long-term post-accident monitoring these transmitters may be unreliable; however, steam line pressure transmitters AB-PT-01, 02, 03, 04 can be used to monitor this RG 1.97 variable.
These are Class IE devices located outside the steam tunnel and are unaffected by the MSLB environment.
ULNRC-1541 QUESTION:
Why were the affects (e.g.; degradation of High Range Radiation Monitor Accuracy) of Cable IR Losses not addressed in the UE review of EQWP J-3617 ANSWER:
Union Electric reviewed'the qualification package for the High Range Radiation Monitors and found that the only test acceptance criteria for the cable was the successful passing of a voltage withstand test (See,
Excerpt 1 and 2).
This test consisted of passing 2000 VAC through the cable.
IR values that were measured were provided for engineering information only.
When G.A. Technologies Inc., supplier of the High Range Radiation monitors, was specifically asked to verify the acceptance i
criteria, they responded that the only acceptance criteria was the voltage withstand test (See Attachment 1, Excerpt 3).
' Union Electric does not possess the necessary information to challenge the manufacturer on the capabilities of his product to overcome inaccuracies caused by cable IR losses.
Because the manufacturer did not specify an acceptance criteria based on cable IR losses, it was assumed that these losses did not affect the operability of the monitors.
In February of 1987, Sorrento Electronics (formerly G.
A. Technologies Inc.)
transmitted a 10CFR21 report to the NRC which stated that cable IR losses did affect instrument accuracies.
Because of the 10CFR21 report Union Electric performed a Callaway specific evaluation (preliminary) based on Sorrento supplied methodology, which shows that cable IR will not drop low enough to affect instrument accuracy.
Because cable IR losses will not affect the high range monitor accuracy, the factor of 2 accuracy requirement contained in the Union Electric response to Reg. Guide 1.97 was never violated.
In summary, Union Electric believes that the High Range Radiation Monitors are qualified to 10CFR 50.49 and that cable IR loss affecting instrument accuracy is a new I
criterion which is currently being addressed.
l
6:
ULNRC-1541 2
Excerpt 1 From"J-361-0212-CO3 (Pg 14 of 373).
.The acceptance criteria were that there could be no more than a 3.75 V fluctuation in the voltage readings at TP-1 and TP-3_during a LOCA, and that there should be no visible damage to the equipment which could affect its operation (Para. 6.5.1 and 6.5.2 of the test procedure, Appendix 1).
The allowable voltage fluctuation corresponds to 3 decades, which should be minimal in the high radiation fields encountered in a LOCA.
Excerpt 2 From J-361-0212-CO3 (Pg 38 of 373) 6.5 Acceptance / Failure Criteria 6.5.1 Voltage measured at the RP-2C TP-1 and TP-3 must remain within i 3.75 V of the Voltage measured in Step 6.3.2.
6.5.2 There can'be no visible damage to the equipment which could affect its operation.
6.5.3 Test article deviations shall be analyzed by General Atomic Engineering and appropriate action determined and initiated as required.
Excerpt 3 From EQWP J-361 (Check Sheet 2, Reference E)'
T. A. Moshenrose-(Project Manager,.G. A.
Technologies, Inc.)' letter to D. R. Quattrociocchi (Project Engineer, Bechtel) dated June 6, 1983.
Comment 2:
Please verify that the acceptance criteria used in Report No. 2806 were:
a.
Insulation Resistance greater than 10,000 megohms.
b.
Passage of a voltage withstand test of 2000 VAC, 60 HZ for 5 minutes.
RESPONSE: The acceptance criteria for the tests reported on in QR2806 was the parsing of a 2000 VAC dielectric withstand test which calculates to 80 volts per mil MAX stress at the surface of the inner conductor.
IR data was supplied for engineering information and since time was limited to one minute, values reported are lower than actual.
i
SNUPPS y
Standardized Nuclear Unit Power Plant Syrtem 7
RMid Cherry FfqM Nicholas A. Petnck
,y5.
gyg8[*girn$
f, Executive Director
- O N'
g, m st i i n rm
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( E 11'7
! l!l I i n.i April 4,1986 M. CbH f
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- {,,
N SLNRC 86-06 FILE: 0278 SUBJ:
Ma1n Steam Line Break Super-gg
- j heat Effects on Equipment c
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Oualification i it i I/
I i
.jgM RETURN CCMf,lENTS TO o. E. sHAFER - CooE 47o RECEIVED JE Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation M
0M SR8 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 NUCLEAR ENGR.
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 0. 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 Gutside Containment (HELB/SB0C) Subgroup of the Westinghouse Owners Group.
The Subgroup was formed to implement the program discussed in Reference 2.
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
% febLk \\ O 289k
m a
SLNRC 86- 06 Page Two that the equipment which must function to mitigate a postulated High Enargy
)
Line Break with superheat effects and to bring the SNUPPS plants to a safe i
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
- yours,
,f h
- 2. M g
(
- 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. Cummi ns USNRC/WC W. L. Forney USNRC/RIII J. E. Gagliardo USNRC/RIV 4
\\
1
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Evaluation of Environmental Qualification of Equipment Considering Superheat Effects of High Energy Line Breaks for Callaway Plant and Wolf Creek Generating Station ao 9%
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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
. estinghouse analyses had system piping ruptures outside. containment.
W 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 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 (WOG) ' 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 rel~ ease information for different classes of plants for use in
'I plant-specific evaluations. The SNUPPS Utilities joined the High Energy Line Break /Superheated 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/SBOC 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 repoct 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 SNUPPS 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. ---
4 e
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 ef fects as discussed in FSAR Section 36.4.2.
An MSLB size equivalent to a 3.41 ft2 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 3B.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 break areas 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 limt eing 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 Eneroy 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 (04), 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 co' pared Model F, Model D4, Model 05 and Model 51 steam m
generators using actual steam generator operating characteristics. This study demonstrated that use of the Model 04 steam generator was conserva-tive for the SNUPPS plants in that peak break enthalpy is 13 BTV/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 BTU /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 approximate 1.y 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
- g. Single failure of one safety injection train The plant response sequence of events is provided in Table III.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 SNUPPS plants 2
~~ /
with the exception of the time of steamline isolation for the 0.7ft,
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.
Steamline isolation time includes an appropriate delay from the time the signal is generated 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 0.5f t2 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 af ter 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 f t2 break case, since the worst-case room temper-ature did not exceed the peak room temperature used in the FSAR analysis of pipe breaks in the steem 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 concluded 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.,
3.2 Environmental Conditions The mass / energy release data developed as discussed in 3.1 above _was used i
to calculate the environmental conditions in the steam tunnel of the SNUPPS plants for four break sizes: 0.5ft2, 0.7ft2, 1.0ft2 and 4.6ft. The 2
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 compartments (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 feedwater 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 beyond 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 38-2 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 and pressures are:
Break Case Peak Pressure / Time Peak Temperature / Time 4.6ft2 17.5 psia /0.29 sec 469.9"F/200.0 sec 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/758.0 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, the calcu-lated temperature values exceed the qualification requirements previously -
)
used for equipment in the tunnel.
Therefore, the thermal response of the equipment was determined as discussed in the following section.
3.3 Eouipment 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 postulated 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 6iscussed 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 may be required following an MSLB in the steam tunnel. The representative equipment for which surface temperature calculations were performed enveloped the equipment listed on Table 3.3 which must function following an MSLB. The most severe room conditions (those for the break compartraent) and the flow characteristics from the FLUD calculation results were used in the calculation of the time dependent equipment surface temperatures. Currently, there is no ' formal
_ )
NRC guidance for convective heat transfer correlations for equipment i
f outside containment. Therefore, based on existing NUREG-0588 guidelines, the equipment surface temperatures were evaluated through the use of con-servative, yet reasonable, heat transfer coefficients.
At any given time, the greater of four times the Uchida condensing heat transfer rate (based on the compartment air to steam mass ratio) or the convective heat transfer rate was used to evaluate the transient surface temperature response of the selected equipment.
The Hilpert correlation, for flow past an object in a fluid stream, with consideratton for system turbulence, was used to calcu-late the convective heat transfer coefficient.
In the evaluation of the heat transfer coefficient for a component, the characteristic velocity was taken as the time dependent average velocity of the flow between the east and west rooms of the steam tunnel.
The flow between these two compartments represents approximately half of the blowdown.
This results in fluid velocities well in excess of those expected to occur in The vicinity of the equipment modelled.
The film properties used in the evaluation of the Hilpert equation were based on the state of the air and steam in the stream. As only the outside casing of equipment was modelled, the Lumped-Capacity method was used to calculate the surface temperature response of the equipment.
This ap'proach is justified by the thinness and the high conductivity of the modelled external casings. For the Main Steam Isolation Valve and Main Feedwater Isolation Valve terminal blocks located in terminal boxes on the valve actuators, more detailed, two-dimensional thermal anal-yses were performed to determine equipment temperatures.
The calculated room environment conditions and the same condensing and conventive heat transfer coefficients, as discussed above, were applied.
The heat transfer to the terminal blocks was modelled via conduction through the back of the.
I,_a,-JM e***
+
terminal box and via convection and radiation across the air gap from the inside surface of the front of the box.
Thus, using conservative and reasonable assumptions and techniques previously approved by the NRC, the 1
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 equipment performance.
j Equipment in the steam tunnel which is required to mitigate an MSLB is actuated by the following control signals:
- a. Steam Line Isolation Signal (SLIS), or I
b.lFeedwater Isolation Signal (FWIS), or
- c. Steam Generator Blowdown System Isolation Signal (SGBSIS), or i
- d. ' Auxiliary Feedwater Actuation Signal for. Turbine Driven Auxiliary Feedwater Pump (AFAS-TD)
In addition, the equipment can be manually actuated by the operators in the Control Room.
Following an MSLB, the response time of the Control Room operators from the time of receiving adequate warning until taking mitigating action is assumed to be 10 minutes.
Adequate warning is considered to be a reactor trip which would cause the operators to initiate event diagnosis under the plant operating procedures.
A. Main Steam Pressure Transmitters The mass / energy release analysis of reference 6 relied on the Low Steamline Pressure (LSP) signal only to initiate a Steam Line Isolation Signal (SLIS).
j-The High Steamline Pressure Rate signal was not used.
Steamline pressure and pressure rate signals are generated by twelve pressure transmitters 1
(3 per each steam line) located in the steam tunnel. The transmitters have been qualified under Westinghouse program ESE-1 to inside containment con-ditions, including 420*F and 57 psig. As shown on Figures 3.2-I through 3.2-4, the room temperature may exceed 420*F in the vicinity of the break location for some postulated break sizes; however, when the temperature response of the transmitters in a superheated steam environment is consi-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, the qualified temperature of the instrument cable (340*F_). is exceeded by the conduit surface temre ature prior to an SLIS for breaks in the 0.7ft -1.0ft2 2
range. Also, the qualified temperature is eventually exceeded for all break sizes which precludes use of the transmitters for long-term post-accident monitoring. A f ailure modes and effects analysis concluded that the cable failure modes would either have no effect on transmitter performance or result in a loss of sicjnal which would cause an SLIS and key the plant oper-ators to use alternate pressure transmitters for post-accident monitoring.
As discussed in FSAR. Appendix 7A, Data Sheet 4.2, the steamline pressure transmitters which are used for control of the atmospheric relief valves can be used to monitor this variable.
These Class 1E transmitters are not located in the steam tunnel and can be used for long-term post-accident monitoring of steamline pressure.
An SLIS initiates closure of the Main Steam Isolation Valves (MSIV) (Spec-ification M-628), the MSIV Bypass Valves (J-601A), 5 team Line Drain Valves (J-601A) and the Turbine Driven Auxiliary Feedwater Pump (TDAFP) Keep-Warm j
)
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; however, the actuator components and appurtenances have various qualification temperatures:
- 1. Hydraulic Components - 450 F
- 2. Pneumatic Components (including solenoid valves) - 450 F
- 3. Wi ring - 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 after 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 fol-lowing.steamline breaks in the tunnel.
Table 3.4 shows that, at the time SLIS i ; :nitiated for each break size, all the components, appurtenances aM terminal boxes are below their qualified temperatures with the ex.eption of the electrical contr'ol cable for one MSIV at Wolf Creek Genera.ing Station and instrumentation cable for MSIV limit switches.
The qualified temperature for the affected MSIV con-trol cable is exceeded by 12*F, w tile the instrumentation cable qualified temperature is exceeded by 42 F.
Valve position indication is the only post-accident function of the MSIV 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.
9 The control cable insulation system for the one affected MSIV consists of cross-linked polyethylene (XLPE) with a neoprene jacket.
Test resul ts for similar control cable used at the SNUPPS plants (XLPE, with hypalon jacket) have demonstrated a qualified temperature of 385 F.
Also, the qualified temperature of the MSIV control cable is exceeded for a short 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 quali fication temperature is exceeded.
Nevertheless, should the control cable for this one MSIV fail in a made which woul.d prevent that MSIV 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-Operated Control Valves The J-601A valves which receive a closure signal upon SLIS are air-operated 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 function acceptably until an SLIS occurs; nevertheless, a failure modes and effects analysis for this cable concluded
)
that environmentally induced failures would either cause the valve to fail 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-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.
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 in all cases potential failure modes result in a loss of indication or a failure of the valve to its safe position or both.
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 lo occur prior to the valve receiving its actuation signal.
Valve position indication for the MSIV Bypass Valves and the Steam Line Drain Valves is a post-accident monitoring function.
As discussed above for the MSIV position indication, the plant operators can determine steam gen-i erator isolation by use of alternate indications such as steam generator i
pressure and level and steam flow. So loss of position indication for these J-601A valves is not a safety concern.
6 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. Main 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 af ter 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 harsh environment.
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. Air-Ocerated 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 tiSLB.
After the valves are closed, the postulated failure of limit switches or limit switch instrument cable would result in a loss of ~fndication of valve position; however, the valves would not reposition.
F. Turbine Driven Auxiliary Feedwater Pumo Steam Supply Valves The steam supply valve's 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 steam 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
,i
+
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-6018 to MSLB conditions including a temperature of 335*F.
For all MSLBs with superheated steam can'ditions in the tunnel, the quali-fied temperatures of ARV-related equipment will be exceeded. When the qualified temperature of the ARV 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 -
i fication 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.
,' s 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 uncontralled 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 T0AFP 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 SNUbPS 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 tiFIVs (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
.c mechanical equipment exceed, with greater than 15'F margin, the calculated
- ~,
MSLB superheat temperatures for all break sizes.
)
I. Structures Ouring the evaluation of steam line breaks in the steam tunnel, the effects of.superheated steam temperatures on the tunnel structures was considered.
l It was concluded that the higher temperatures resulting from superheated 2 break) will have no detrimental blowdown (maximum 469.9 F for a 4.6ft 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 mitigation.
Most of the electrical equipment which must function following an MSLB in the tunnel can be characterized 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 l
regulations. The issue of adequate qualification time margin for equipment which performs its function within a short time period into the event is I
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 detrimental 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 s
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.
. 1
i References 1.
SNP(S)-1005, dated June 13, 1984: Notification of Unreviewed Safety Questions.
j i
- 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), OG-128, dated July 26, 1984 4.
Westinghouse Owner'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" MHF/dck/Sa29 1
Table 3.1 Design of Steamline Break Protection System A.
Safety Injection Signals
- 1. Low steamline pressure (LSP)
- 2. Low pressurizer pressure (LPP) 1
- 3. High containment pressure (for breaks inside containment only)*
B.
Steamline Isolation Signals I
- 1. Low steamline pressure (LSP)
- 2. High steamline pressure rate
- 1
- 3. High-High containment pressure (for breaks
]
)
inside containment only)*
l
- These parameters were not used in the analysis.
e
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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.
i 1 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 Of fsite 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' i
-Driven Auxiliary Feedwater Pump is assumed to start automatically
-)'
following a reactor trip,.with a conservative time delay included.
sw
- 11. Steam Generator Fluid Mass: -No comment.
i
-12. Break Sizes: No comment.
- 13. Safety Injection System Flowrate: No comment.
l
- 14. Boron Injection Tank: No comment.
)
1
- 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, j
a single active failure of safety equipment was not assumed concurrent 4
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 i
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 coment" for the items below means that the assumption used in the Westinghouse analysis is applicable without modification or clarification.
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1901 Gratiot Street. St. Louis Donald F. Schnell March 24, 1987 vice President Nuclear Reculatory Ccmmission U.S.
ATTN: Document Centrol Desk Washington, D.C.
20555 ULNRC-1473 Gentlemen:
COCKE"' NUMBER 50-4G 3 CALLA'#AY PLANT PSLB SUPERHEAT EFFECTS ON EO SLNRC 86-06 dated 4-4-SA, 1.
References:
same subject 2.
NRC Request for Additional Information, P.W. O'Connor to D.F. Schnell, dated 1-6-87 Reference 1 submitted the results of a generic SUUPPS review qualification of performed to assess the effects on equipmentwith superheated main steam line breaks outside containmentReference 2 transmitted sev additional information needed for completion of the"'he encicsure prov blowdowns.
our submittal.
us.
If you have any further questiens, please contact Very truly yours,
/
l
/ l s:V Conald F. Schnell i
GGY/ dis Enclosure f
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I Mailing Address: P.O. Dot 149. St. Louis. MO 63166 j
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cc:
Gerald Charnoff, Esq.
Shaw, Pittman, Potts & Trowbrid.ge 2300 N. Street, N.W.
Washington, D.C.
20037 J. O. Cermak CFA, Inc.
3356 Tanterrn Circle Brookville, MD 20833 W. L. Forney Division of Projects and Resident Programs, Chief, Section lA U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Office U.S. Nuc) edr Regulatory Commission RRf1 Steedman, Missouri 65077 Paul O'Connor (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Comnission
-)
Mail Stop 316 s.
7920 Norfolk Avenue Bethesda, MD 20014 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 l
1
1 ULNRC-1473 March 24, 1987 RESPONSE TO NRC OUESTIONS
_ I MAIN STEAM LINE BREAK SUPERHEAT ANALYSIS OUESTION 1 With respect to the nodalization schemes, provide:
1.A.
Number of nodes (compartments) analyzed; 1.B.
For each node (compartment).
1.
Initial temperature 1
II. Initial pressure I
III. Initial humidity IV. Compartment free volume V.
Number of vents and vent area (square feet) for each vent; and VI, Compartment height (feet);
1.C.
Simple nodalization diagrams; 1.D.
Identification of break compartment, and relative locations of safety related equipment; 1.E.
Number, volume, surface area, and material properties of thermal sinks within compartments;
)
1.F.
Descriptions of equipment and the'rmal models of the equipment, including geometry (surface area and volume), material properties (density, thermal conductivity, specific heat) of all materials included in equipment models and sketches of equipment models.
RESPONSE
1.A.
Three nodes (compartments) were used in the analysis. These were the Main Steam Tunnel (MST) - West, the MST - East, and the atmosphere.
1.B.
Initial conditions for each node are provided in the following table. Callaway FSAR (Wolf Creek USAR) Figures 1.2-15 and 1.2-16 or Figure 3B-2 provide the MST compartment height (approximately 62 feet).
Compartment Initial Initial Relative Volume Junction 3
Floy) Area Pressure Tegp.
Humidity (ft)
(ft (psia)
( F)
MST - West (1)
, 14.7 120.0 0.70 59,098.92 1-2:
666.81 f
1-atm.: 203.14 MST - East (2) 14.7 120.0 0.70 59,239.92 2-atm.: 203.14 f
Atmos.
14.7 95.0 0.50 i
l i
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a 1.C.
For a nodalization diagram, refer to Sheet 2 of Figure 3B-4 of the Callaway FSAR (Wolf Creek USAR).
1.D.
The MST - West was identified in the MSLB superheat submittal of 4/4/86 as the break compartment. All safety-related equipment is located in the lower half of the MST compartments. Relative locations of safety-related equipment may be determined from Callaway FSAR (Wolf Creek USAR) Figures 1.2-12, 13, 15, 16; 38-2 and 3.6-1 (Sheets 1, 2, 3, 29, 30 and 49). However, since worst-case environmental conditions (MST-West) were used in calculating the surface tempera-tures of all equipment without consideration of spatially varying conditions, equipment locations are not a factor in the conclusions stated in the MSLB superheat submittal.
1.E Table 1 provides the requested information on heat sinks.
1.F In preparing Table 3.4 of the MSLB superheat submittal, nine equip-ment types were modelled. In addressing the request for additional information, the need to perform a more detailed analysis of one equipment type (Equipment 10) was identified. The ten equipment thermal models are discussed in the following sections. All models are one-dimensional except those for Equipment 9 and 10.
Equipment 1: Solenoid housing, modelled as a 2"-diameter cylinder (d =
0.167').
The material is constructed of stainless steel. The following l
)
data was used in the model:
488 lbm/ft3 Density (0)
=
Q-l 9 Btu /hr-ft OF Thermal Conductivity (k)
=
0.11 Btu /lbm OF Specific Heat (cp)
=
0.01' (t = 1/8")
thickness
=
refer to Figure 11 sketch Equipment 2: Terminal box, modelled as an equivalent cylinder of hydraulic diameter 0.95' (dh = 11.4").
The boxes are constructed of 14 gauge steel and were modelled using the following data:
487 lbm/f t3 0
=
27 Btu /hr-ft OF k
=
0.113 Btu /1bm OF
=
cp thicknessx =
0.0747" (t =.006')
refer to Figure 12 sketch l
i 2-I
o o
Equipment 3: Pressure transmitter.- top section, modelled as a cylinder of diameter 0.31' (d ~ 3 3/4"). The material is steel, and the data used in the model was as follows:
487 l bm/f t3 0
=
27 Btu /hr-ft OF k
=
0.113 Btu /lbm OF
=
cp 0.0747" (t =.006')
thickness
=
refer to Figure 13 sketch Equipment 4: Conax connector, modelled as a 0.75" diameter cylinder (d = 0.063'). The material is stainless steel:
488 lbm/ft3 D
=
f 9 Stu/hr-ft OF k
=
0.11 Stu/lbm OF
=
cp 1/16" (t = 0.005')
thickness
=
refer to Figure 14 sketch Equipment 5/6/7: Flexible conduits, modelled as cylinders with diameters
~)
varying in accordance with actual size. For Able hoses the thickness varies between 0.1" to 0.12" and 0.11" was used for the models. For flex conduits, the thickness is 0.0625".
The material is galvanized carbon steel. The zinc coating was not modelled.
487 l bm/f t3 0
=
27 Btu /hr-ft OF k
=
0.113 Btu /lbm OF
=
cp not required sketch Equipment 8:
Solenoid valve body, modelled as a cylinder with a diameter of 1.5".
The material is brass. The properties of red brass were used:
532 lbm/ft3 0
=
35 Btu /hr-ft 0F k
=
cp 0.092 Stu/1bm OF
'=
0.25" (t = 0.021')
thickness
=
refer to Figure 11 sketch
n
.o Equipment 9:.Two-dimensional model of a terminal block inside a termical box. The materials, dimensions and configuration of the equipment are listed on Figure 15. The properties of steel are as previously described for other equipment. The properties of Alumicum and Polysulfone are:
3 168 (Al),
78 (Polys.)
D (1bm/ft )
=
k (BTV/hr-ft UF) 138 (Al),
0.15 (Polys.)
=
0.22 (AT),
0.30 (Polys.)
p (BTU /lbm-F)
=
c Equipment 10:
Two-dimensional Skinner solenoid valve. The materials, dimensions and configuration are provided in Figure 16. The properties of steel and stainless' steel are as previously described for other equipment. The properties of copper and epoxy insulation are:
3 558 (Cu),
11.2 (Epoxy)
D (1bm/ft )
=
k (BTU /hr-ft OF) 230 (Cu),
0.36 (Epoxy)
=
0.092 (Cu),
0.24(Epoxy) p (BTU /1bm-F)
=
c Correlation of the equipment listed on Table 3.4 of the MSLB superheat submittal to the above equipment types is discussed below.
Main Steam Pressure Transmitters The transmitters correlate directly with
' Equipment 3.
The model used in the analysis is conservative because the thickness of the transmitter housing is approximately 0.25" instead of the
~
modelled.0747".
Main Steam Pressure Transmitter Instrument Cable - This cable is run in solid and flexiole conduit in tne MST. Ine thermal lag response of flexible conduit is limiting. Therefore, the cable' correlates to Equipment 5 with a conduit diameter of 0.75".
When comparing the thermal response of conduit, the surface temperature of smaller corduit followed more closely the room temperature than larger conduit. Also, for a given size, flexible conduit followed scre closely the room temperature than solid conduit.
It has been noted that an error existed for this equipment on Table l
3.4 of the MSLB superheat submittal. 2For the feepwater. isolation signal temperatures, with break sizes 1.g ft and 0.7 ft, thg temperatgres in Table 3.4 should be 250 F and 245 F instead of the 235 F and 232 F values which were incorrectly picked off the thermal lag curves. (A l
)
~ -...
[
revised Table 3.4 is attached. The table has been revised to correct the above error and to incorporate additional changes identified in the follow-ing discussion).
MSIV/MF1V Solenoid Valve - The valve correlates to the curves for a Skinner solenoid vaive (Equipment 10). Revised qualified and thermal lag temperature values have been incorporated into Table 3.4 based on revised qualification data and a two-dimensional analysis for this solenoid valve.
MSIV/MFIV Wiring and Lugs - This equipment is located inside terminal boxes mounted on tne M51V and MFIV actuators. The one-dimensional terminal box response curve (Equipment 2) was conservatively applied to this equip-ment.
MSIV/MFIV Terminal Blocks - The terminal blocks correspond to Equipment 9.
. terminal box, it was found that results from thg 1.0 ft{ysis for the Based on an examination of the lumped-heat-capacity ana MSLB enveloped those of the other breaks, including the 0.7 ft break, at all times 2
until the end of the analysis for the 1.0 ft MSLB (800 seconds). For this reason, the detailed two-dimensional analysis was performed only for this break. Moreover, since the temperature of the terminal block remained below the qualification temperature for the entire duration of the two-dimensional analysis, the results were considered bounding and conservative.
Thys the terminal block temperature response was determined for the 1.0 ft break using a two-dimensional analysis. The equipment tempergtures for the other size breaks were determined as follows. The 1.0 ft 2
]
temperature response for the terminal blocks was compared to the 1.0 ft response of the terminal box (Equipment 2).
It was noted gnat the terminal block response lagged the box response by approximately 85 F in the time period of interest.
Since the terminal block temperature response is drfven by the terminal bcx response, it was assumed that the approximately 85 F difference in temperature gould exigt between the tgrminal block and the terminal box for the 4.6 ft, 0.7 ft and the 0.5 ft* cases.
In this way, the terminal block temperature values for steam line isolation in Table 3.4 of the submittal report were cbtained from the Equipment 2 thermal lag curves.
For the feedwatgr isolation temperatures in Table 3.4, it was assumed that the 4.6 ft break temperature was tge same as for steam /line isolation (this assumption wag used for tge 4.6 ft break case for all equipment);
and, for the 0.7 ft and 0.5 ft cases,thetempegatureoftheterminal blgckwasassugedtoincgeaseataconstant0.65 F/second from the 1.0 value (135 F). 2 65 F/second is the maximum terminal block rise 0
ft rate for the 1.0 ft break..
L MSIV/MFIV Control Cable - This cable is run in conduit and flexible conduit in the MST.
The temperature correlates to the thermal lag values for 1.5" flexible conduit (Equipment 6). The submittal Table 3.4 values for this j
equipment have been revised to reflect actual thermal lag values for the l
i four break sizes. The values previously provided were based on an average of thermal lag values for 0.75" and 2" flexible conduit.
l MSIV/MFIV Limit Switch - A thermal lag curve was not specifically developed
{
"for a 11 nic switen.
It was assumed that a limit switch housing thermal response would be similar to the response of the solenoid valve solenoid housing (Equipment 1). This assumption is appropriate because the thickness of a limit switch body (approximately 1/8") is equal to the thickness of the modelled solenoid valve solenoid housing. For a sketch of a limit switch refer to Figure 17.
-MSIV/MFIV Canax Connector - These connectors correlate to Equipment 4.
MSIV/MFIV Limit Switch Instrument Cable - Portions of this cable are routed in 0.75" Able nose.
Therefore, the tnermal lag response curve for Equip-ment 7 is appropriate for this cable.
J-601A Solenoid Valve - This valve correlates to the curve for a solenoid i
valve booy (Equipment 8).
J-601A Control Cable - This cable is run in conduit and flexible conduit in tne steam tunnel and correlates to Equipment 5.
Question 2 All assumptions used, including but not limited to the following:
l 2.A.
Junction loss coefficients; I
2.8.
Heat transfer coef ficient for heat transfer through walls; l.
2.C.
Condensation model used.
Response
2.A.
Assumptions used in the MSLB superheat submittal concerning mass and energy release, environmental conditions and equipment performance are discussed in sections 3.1, 3.2 and 3.3 of the submittal. Additi-onal assumptions pertinent to environmental conditions and equipment performance are:
~
Cooling by the MST ventilation system is conservatively neglected.
a.
b.
Owing to their low failure pressure, the MST blowout panels open soon into 'the transient. Hence, they are modelled as simple open
- vents, Thermodynamic equilibrium exists at all times in the MST compart-c.
ment atmosphere.. _ - _. _ _ _ _ _ _ _ _ - _ _ - _
d.
The MST volume is filled with a homogeneous mixture of air, steam and water.
e.
Air. is treated as an ideal gas, and steam is treated to second order in the virial expansion of the equation of state.
f.
Flows between compartments are calculated using steady state compressible flow equations for an ideal gas; inertial effects and vapor dropout are not considered.
g.
The junction loss coefficients between nodes have been calculated using simple contraction and expansion losses.
2.B.
For convection, the heat transfer coefficient was calculated based on-a velocity, a characteristic length, and the compartment thermodynamic conditions.
Hilpert's Equation was used to evaluate the forced convection heat transfer coefficient based on a Reynolds number at film conditions.
If this heat transfer coefficient was less than its natural convection counterpart, then the natural convection heat transfer coefficient was used assuming turbulent conditions (Ref.
NUREG-0588, Rev. 1).
2.C.
For the condensing heat transfer coefficient, four times the Uchida heat transfer coefficient was used (Ref. NUREG-0588, Rev. 1).
Question 3 Analysis results:
,_')
3.A.
Compartment temperature versus time curve (peak temperature specified);
3.B.
Compartment pressure versus time curve (peak pressure specified);
3.C.
Equipment temperature versus time curve (peak temperature specified) for each piece of equipment; 3.0.
Equipment qualification temperature; 3.E.
Identification of locations for A.,
B. and C.,
if spatially varying within a compartment.
Response
3.A.
Compartment temperature versus time curves were provided in the MSLB superheat submittal, Figures 3.2-1 through 3.2-4 3.B.
The peak pressures achieved in the MST for the MSLB superheat analysis are provided 1.n section 3.2 (page 4) of the MSLB superheat submittal.
These pressure'. values are lower than those calculated by an analysis that maximizes pressure (rather than temperature) conditions.
Therefore, reference is made to Figure 38-5 of the Callaway FSAR (Wolf Creek USAR) for the worst case pressure versus time curve for the MST. !
A
~
3.C.
Worst-case equipment temperature versus time curves are provided, for each of the ten equipment types identified in the response to question l.F., in Figures.1 through 10.
.The worst-case temperature values provided in Table 3.4 of the submittal report are obtained by using the curves and the appropriate actuation time for the equipment. The actuation times, for steam line isolation and feedwater isolation, are listed'in Table III.8-4 (cases 59-63) of WCAP-10961-P except that manual steamline isolation is assumed to' occur 600 seconds after reactor trip and allowance is made for steamline isolation valve closure time, i.e. the time from receipt of a closure signal until the valve is fully closed.
3.D.
Equipment qualification temperatures are provided on Table 3.4 of the MSLB superheat submittal (last column on table).
3.E.
As discussed in the response to question 1.0. above, spatially varying conditions were not assumed in the analysis.
b l
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TABLE.1 HEAT SINX PROPERTIES Thermal
~~
Heat Heat Sink Surface Conductivity D XCp Sink Material / Code Area (ftr)
Thickness (ft)
(8tu/hr *F-ft)
(8tu/ft3 *F) k Struc. Steel /l 4287.13 0.042 25.0 53.9 2
Struc. Steel /l 4300.96 0.042 25.0 53.9 3
Concrete Floo-/2 945.00 2.0 0.8 30.0 4
Concrete Floor /2 945.00 2.0 5
Concrete -
3200.00 1.0 Column /2 6
Concrete -
3200.00 1.0 Column /2 7
Concrete -
3352.50 2.0 Column /2
]
8 Concrete -
3352.50 2.0 Column /2 9
Interior 1665.00 1.0 Concrete /2 I
10 Interior 1665.00 1.0 Concrete /2 11 Concrete 1639.00 2.0 Column /2 12 Concrete -
1639.00 2.0 Column /2 13 Concrete Wall /2 1865.00 4.0 14 Concrete Wall /2 1865.00 4.0 15 Concrete Roof /2 365.00 2.0 Y
Y 16 Concrete Roof /2 365.00 2.0 Note: Odd numbered heat sinks are in Compartment 1, and even numbered heat sinks are in Compartment 2.
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These boxes sto fabncated from.16 gauge and 14 gauge steel anc have esternal screw cfamps on three sides of the cover.The scre.
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optional panets and terminal kits wmch must be ordered separatory When ordering special"CHNF" bones, be sure to completety spec.t.
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OPER ATION AL D ATA
' SPECIFICATIONS
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STANDARD Heavy Dutv, Machine Tool Type, m y,,,,,.c.overin e mina pai cmovitto Double Pole. Double Throw, Quick gfagog*y gYagsg8;g',tn Make, Quick Break, But: Type.
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~.i Enclosure is Water, Oil and Ourttight.
I PM. r;,g,,, 4 0
O J<.*. --. Meets (NEMA) Type 1,4, & 13 E O t o iii. g
',] Requirements.
j:,2' S Torque Necessary for Operation of se i
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,d} Switch. 23 in. -th. (Without Retum g#,
f.
.,; Spring. Itam 23,10 in..Ih.)
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Extemal Lever is Adiustable by 7*30' M
-. "fn-g Increments Thru 180.
i Ambient Temperature:.20*C to +90*C.
.C m o. - >
Double pole, doisbie break, doubli Ampere Rating See Fage 14 for Contact Configuration duty 'fimit switch throw, hoevy Volts AC DC having mechenion! trvvel of 10* to #
trip and with two riormany open 125 20 5
A. Pre-Travel Trip Position..
10*
8*
and two normeHy domed orants. t 250 15 1.5 B. Roset Position.
Can be fumished with sesridard, 480 10 C. Total Travel 37' M
S D. Recommended Travel
. 13*
W' M'*
3 DIMENSIONS and MOUNTINGS r..
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STYLE 1 MOUNTING STYLE 2 MOUNTING STANDARD MOUNTING ORDERING NUMBERS 'D OSDERING INFORM ATION Opposite Witeowt Standard Mot.eteert CCW Socing Matum
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EA170 21100 EA170 22100 EA170-23100 STYLE N0.1 MOUNTING (D2400X.1)
(D240cX.1.S R)
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EA170 31100 EA170-32100 EA170 33100 STYLE NO.2 MOUNTING (D2400X.2)
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