ML20072F565
| ML20072F565 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 03/17/1983 |
| From: | Wuller G ILLINOIS POWER CO. |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| U-0615, U-615, NUDOCS 8303240346 | |
| Download: ML20072F565 (46) | |
Text
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-500 SOUTH 277H STREET. P. o. BOX 511. DECATUR, ILLINOIS 62525-1805 K~~
3D6cket No. '50-461 March 17, 1983 7
',x director of Nuclear Reactor Regulation Attention:
Mr. A. Schwencer, Chief Licensing Branch No. 2
'D_(ivision of Licensing U S. Nuclear Regulatory Commission
,fWefhington, D.C.
20555
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DearMr. Schwencer:
Subj ect :
Clinton Power Stat ion Unit 1 Humphrey Concerns
Reference:
IP letter.U-0596 dated 2/4/83, G. E. Wuller to A. Schwencer, NRC, subj ect : Submittal addressing some John Humphrey concerns.
The referenced letter addressed some of the John Humphray concerns as applicable to the Clinton Power Station (CPS).
Enclosed are CPS responses on some additional Humphrey issues for NRC Staff review.
Included are Action Plans #15, 18, 19, 22, 25, 28, 31 and 32.
We believe that these responses will resolve the
=particular concern involved.
If there are any questions regarding this material, please contact me or J. H. Shepard (217) 424-6785.
Sincerely, G. E. Wuller Supervisor-Licensing Nuclear Station Engineering GEW/jmm enclosure cc:
D. H. Abelson, NRC Clinton Project Manager Mr. M.
B. Fields, NRC CSB Mr. H. H. Livermore, NRC Resident Inspector Illinois Department of Nuclear Safety R. W. Evans, Quadrex Corporation C303240346 830317 PDR ADOCK 05000461 A
~
Action Plan 15 4.6 The initial suppression pool temperature is assumed to be 95 F while the maximum expected service water temperature is 90 F for all GGNS accident analyses as noted in FSAR table 6.2-50.
If the service water temperature is consistently higher than expected, as occurred at Kuosheng, the RHR system may be required to operate nearly continuously in order to maintain suppression pool temperature at or below the maximum permissible value.
Response
The program for resolution to close this issue for Clinton is as follows:
15.1 A discussion of peak service water temperature which is expected under nonaccident conditions will be provided.
Also the expected peak suppression pool temperature under normal operating conditions will be discussed.
15.2 The conservatisms in the existing analyses defining peak service water temperature will be quantified to the extent possible.
Item 15.1 The maximum worst-case initial temperature of the lake where shutdown service water takes suction it approximately 90 F for the hiehest one (1) day average water temperature.
This temperature is based upon the once-in-fifty-years worst-case situation (drought, temperature, plant at 100% power, etc.).
The peak suppression pool temperature under normal plant conditions if 95 F as limited by the technical specifications.
The high temperature problem which prevailed at Kuosheng is not a concern for the Clinton Power Station because of the unlikelihood of the shutdown service water temperature being 900F.
Concluding that the continuous operation of the CPS RHR System is very remote, no reduction in the service life of the equipment is expected.
Item 15.2 Regarding the conservatisms in the present design, the ability of the suppression pool to accommodate accidents at the maximum (design) temperature is addressed in the response for Action Plan 28.
Thus, from the information provided above this issue is closed for Clinton.
15-1
Action Plan 18 4.9 The effect on the long term containment response and the operability of the spray system due to cycling the containment sprays on and off to maximize pool cooling needs to be addressed.
Also provide and justify the criteria used by the operator for switching from the containment spray mode to pool cooling mode, and back again.
5.3 Leakage from the drywell to containment will increase the temperature and pressure in the containment.
The operators will have to use the containments spray in order to maintain containment temperature and pressure control.
Given the decreased effectiveness of the RHR system in accomplishing this obj ective in the containment spray mode, the bypass leakage may increase the cyclical duty of the containment sprays.
Response
A criunion for transferring the RHR System from containment spray mode to suppression pool cooling mode will be developed to close this issue for Clinton.
The results of Action Plans 10 and 11, which evaluated conservatisms in calculating pool temperature and containment pressure, showed that, during a DBA without drywell bypass, there will be no need for containment spray activation to control containment pressure or temperature.
The. realistic analysis performed by GE demonstrated that the containment pressure will not exceed 8.1 psig - well below the design pressure of 15 psig, and also below the automatic spray actuation setpoint of 9 psig.
The peak suppression pool temperature was 1650F, with the airspace temperature always lagging behind the pool temperature, by as much as 36 F.
Sincerthe peak pool temperature is quite insensitive to the size of the break, these results would apply across the entire spectrum of LOCAs.
Only when a drywell break is combined with a large drywell bypass leakage, will containment spray be needed to control containment pressure.
For this case, the results of Action Plan 19 show that the decreased RHR heat exchanger effectiveness due to operation in the spray mode will nog result in exceeding the suppression pool design temperature of 185 F, even with containment spray in continuous operation.
This result was obtained by conservatively accounting for structural heat sinks in both drywell and containment airspaces.
Hence, it is concluded that the current operator procedures are adequate and need no modifications for the operation of containment sprays.
And thus this issue is closed for Clinton.
18-1
Action Plan 19 5.1 The worst case of drywell to containment bypass, leakage has been established as a small break accident.
An intermediate break accident will actually produce the most significant drywell to containment leakage prior to initiation of containment sprays.
5.6 The test pressure of 3 psig specified for the periodic operational drywell leakage rate tests does not reflect additional pressurization in the drywell which will result from upper pool dump.
This pressure also does not reflect additional drywell pressurization resulting from throttling of the ECCS to maintain vessel level which is required by the current EPG's.
9.2 The continuous steaming produced by throttling the ECCS flow will cause increased direct leakage from the drywell to the containment.
This could result in increased containment pressures.
Response
The following program for resolution was used to close this issue for Clinton:
19.1 A complete spectrum of analyses for varying break sizes will be completed neglecting depressurization of the drywell prior to initiation of containment sprays, but including the effects of containment heat sinks.
19.2 Analyses will be completed to show that the allowable leakage of A/fR equal to 1.0 is valid for Clinton.
19.3 Evaluate the'need of reducing the allowable tech. spec.
limiting conditions for the drywell leakage.
In response to 19.1 and 19.2 a spectrum of breaks rangin from 2
SBA to DBA with a bypass leakage equivalent area of A/ K = 0.9 ft were analyzed to determine the degree of containment pressurization prior to the initiation of containment spray.
The analyses were i
perfe Tued using GE proprietary computer program VACBR04.
Structural heat sinks and heat and mass transfer between suppression pool and containment air space were considered in the model.
Free convection heat transfer coefficients were conservatively calculated to minimize the hea; sink effectiveness.
The duration of drywell pressurization is maximized by assuming that the operator controls the RPV water level, in a manner such that the break will not be flooded by ECCS system operation.
Thus, there is no drywell steam condensation caused by a relatively cool break liquid flowing into the drywell to promote condensation.
19-1
The analyses show that the ligiting drywell to containment bypass leakage results from a 2.5 ft steam-line break accident.
The peak containment pressure in this case, at the dne of spray initation, was 14.5 psig - which is below the containment design pressure of 15 psic.
Theref re, the previously calculated bypass leakage capability of A/(R'= 0.9 ft for GGNS is still valid.
This analysis envelgoes CPS even thoug h the bypass leakage A/ dlC for Clinton is 1.0 ft The bypass leakage variable is not as significant in the analysis as the structural heat sinks and the heat and mass transfer between the suppression pool and containment air space.
These heat sinks and heat and mass transfer coefficients used in the analysis are conservative for Clinton.
F4 pre U9.l shows the peak containment pressure as a f"" ti "
f 2
break size, with bypass leakage of A/ dlC = 0.9 f t In response to 19.3 a worst-case scenario for drywell bypass leakage, obtained as a result of the study under item 19.1, was evaluated for long-term containment pressurization.
At 13 minutes after the break, the containment pressure is reduced due to automatic spray actuation.
This worst-case scenario is predicated on operator action (per EPGs) to control the RPV water level, thus preventing ECCS water overflow into the drywell and resulting in continuous drywell pressurization.
At 30 minutes post-LOCA, the upper pool dumps and the drywell-to-containment AP increases due to increased vent submergence.
This, in turn, increases the steam bypass leakage, and results in additional containment pressurization.
At approximately 50 minutes into the accident, the containment design pressure of 15 psig would be reached.
However, the operator is directed, again by EPGs, to terminate the drywell (and thus containment) pressurization by flooding the drywell with ECCS water before the containment design pressure is reached.
It is therefore concluded that there is no need for reducing the allowable bypass leakage technical specification in consideration of the additional pressure produced by upper pool dump.
Thus, this issue is closed for Clinton.
19-2
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FIGURE 19d CONTAINMENT BYPASS LEAKAGE CAPABILITY
Action Plan 22 5.8 The possibility of high temperatures in the drywell without reaching the 2 psig high pressure scram level because of bypass leakage through the drywell wall should be addressed.
Response
The following program for resolution was used to close this issue for Clinton:
22.1 A new analysis will be performed using the capability of bypass leakage.
This analysis will show that a temperature of 3300F is not reached in the drywell until after ten minutes.
In this interval, the operator will have received sufficient information to manually scram the reactor.
22.2 Develop a list of alarms & displays which inform the operator of conditions in the drywell.
In response to 22.1, there are two reasons why the drywell temperature may increase from its equilibrium value during normal operation:
(a) increased heat addition (steam leak in the drywell), or, (b) decreased heat removal (degradation or total loss of drywell fan coolers.)
The worst possible case is the combination of both (a) and (b).
The scenario analyzed, the worst possible case, was a steam break in the drywell with a non-isolated containment, no drywell b
leakage fan cooling, and the drywell capb)ility of bypass (effective GGNS area of 0.9 feet taken into account.
The analysis was performed using the General Electric proprietary computer program VACBR04.
The purpose of the analysis was to determine the peak drywell temperature that could occur during the first 10 minutes after a break with the drywell pressure maintained below 2 psig.
After 10 minutes it is assumed the operator will take some action to control drywell temperature e.g., scram and depressurize the reactor.
The heat sources were obtained from data for Perry, a BWR/6 238 Mark III plant.
It is estimated the total drywell heat load for GGMS is within 30% of the total drywell heat load for Perry.
An increase of 30% in the drywell heat load used in the analysis would result in less than a 30% increase in the drywell temperature rise.
A spectrum of break sizes was considered to determine the limiting case.
The peak drywell temperature of 246 F occurred at 10 minutes for a break area of approximately.007 ft2, at ll10F temperature rise from the initial drywell temperature of 135 F.
This peak drywell temperature is well below the design temperature of 3300F, even if the calculated temperature rise is increaged by 30%.
Even though the bypass leakage for Clinton is 1.0 ft' the drywell heat load is a more significant variable than the bypass leakage value.
This analysis is still conservative for CPS because the drywell heat load for Clinton is much smaller than for GGNS.
22-1
In response to 22.2, a Clinton specific list of alarms and parameter displays which are available to the operator to inform him of conditions in the drywell are developed in Table 22.1.
This would include drywell-cooling performance, temperature, airflows, leak detection, etc.
Thus, from the information provided above, this issue is closed for CPS.
22-2 t
4 Table 22.1 Alarm & Displays For the Drywell CONTAINMENT MONITORING 1EI-CM(later)
H /0 Concentration-2 2
LEI-CM(later)
H /0 Concentration 2
2 lLT-CM91 ster)
Drywell Sump Level 1ME-CM019 Drywell Air Humidity Lower Drain Cooler Inlet 1ME-CM020 Drywell Air Humidity Lower Drain Cooler Inlet 1ME-CM021 Drywell Air Humidity Upper Drain Cooler Inlet IME-CM022 Drywell Air Humidity Upper Drain Cooler Inlet 1ME-CM023 Drywell Air Humidity CRD Area 1 MIS-CM025 Drywell Air Humidity Analyzer 1PT-CM063 Drywell Pressure 1PT-CM064 Drywell Pressure 1RE-CM059 Drywell High Range Gamma Radiation 1RE-CM060 Drywell High Range Gamma Radiation 1RIX-CM059 Drywell High Range Gamma Radiation 1RIX-CM060 Drywell High Range Gamma Radiation 1RO-CM059 Drywell High Range Gamma Radiation 1RO-CM060 Drywell High Range Gamma Radiation 1TE-CM065 Drywell Atmosphere Temperature 1TE-CM066 Drywell Atmosphere Temperature 1TE-CM067 Drywell Atmosphere Temperature 1TE-CM068 Drywell Atmosphere Temperature 1TE-CM069 Drywell Atmosphere Temperature 1TE-CM070 Drywell Atmosphere Temperature 1TE-CM071 Drywell Atmosphere Temperature 1TE-CM072 Drywell Atmosphere Temperature 1XX-CM125 H /0 Gas Chromatograph 2
2 1XX-CM126 H /0 Gas Chromatograph 2
2 1XX-CM139 H /0 Teleprinter 2
2 1XX-CM140 H /0 Teleprinter 2
2 LEAK DETECTION lE31-N017A Drywell Ambient Temperature lE31-N017B Drywell Ambient Temperature 22-3
LEAK DETECTION - (cont'd) lE31-N017C Dryuell Ambient Temperature i
lE31-N017D Drywell Ambient Temperature I
lE31-N021 Drywell Air Cooler IVP01CC Drain 1E31-N033A Drywell Cooler IVP01CD Drain Flow.
L lE31-N530 Drywell Cooler IVP02SA Inlet!
lE31-N531 Drywell Cooler IVP02SA Outlet 1E31-N532 Drywell Cooler IVP02SB Inlet lE31-N533 Drywell Cooler IVP02SB Outlet l!
lE31-N534 Drywell Cooler IVP02SC Inlet f
lE31-N535 Drywell Cooler IVP02SC Outlet lE31-N536 Drywell Cooler IVP02SD Inlet 1E31-N537 Drywell Cooler IVP02SD Outlet lE31-N540 Drywell Cooler IVP02SA Diff. Temp. Amplifier lE31-N541 Drywell Cooler IVP02SA Diff. Temp. Trip Unit-lE31-N542 Drywell Cooler IVP02SB Diff. Temp. Amplifier j
lE31-N543 Drywell Cooler IVP02SB Diff. Temp Trip Unit-1E31-N544 Drywell Cooler IVP02SC Diff. Temp. Amplifier lE31-N545 Drywell Cooler IVP02SC Diff. Temp. Trip Unit lE31-N546 Drywell Cooler IVP02SD Diff. Temp. Amplifier lE31-N547 '
Drywell Cooler IVP02SD Diff. Temp. Trip Unit lE31-N572 Drywell Floor & Equipment Drains Inlet Flow Total lE31-N57h Drywell Floor Drain Inlet Flow
-lE31-N576 Drywell Equipment Drain Inlet Flow lE31-N578 Drywell Equipment Drain Weir Box Level for Flow Meas.
lE31-N5781 Drywell Equipment Drain Weir Box Level for Flow Meas.
lE31-N5782 Drywell Equipment Drain Weir Level-Flow Convert lE31-N5783 Drywell Equipment Drain Weir Flow Increase Alarm
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lE31-N5784 Drywell Equipment Drain Weir Flow Increase Timer lE31-N580 Drywell Floor Drain Weir Box Level for Flow Meas.
lE31-N5801 Drywell Floor Drain Weir Box Level for Flow Meas.
lE31-N5802 Drywell Floor Drain Weir Level-Flow Convert lE31-N5803 Drywell Floor Drain Weir Flow Increase Alarm 1E31-N5804' Drywell Floor Drain Weir Flow Increase Timer lE31-N590 Drywell Cooler IVP02SA Water Inlet i
22-4
.-.-.-.,_---._.a-----_.__.-,
r 4
LEAK DETECTION (cont'd) lE31-N591 Drywell Cooler IVP02SA Water Outlet lE31-N592 Drywell Cooler IVP02SB Water Inlet lE31-N593 Drywell Cooler IVP02SB Water Outlet lE31-N594 Drywell Cooler IVP02SC Water Inlet lE31-N595 Drywell Cooler IVP02SC Water Outlet lE31-N596 Drywell Cooler IVP02SD Water Inlet lE31-N597 Drywell Cooler IVP02SD Water Outlet lE31-R520 Drywell Cooler lVP02SA Diff. Temp. Indicator lE31-R521 Drywell Cooler IVP02SB Diff. Temp. Indicator lE31-R522 Drywell Cooler lVP02SC Diff. Temp. Indicator lE31-R523 Drywell Cooler lVP02SD Diff. Temp. Indicator lE31-R600A Drywell Cooler IVP02CD Drain Flow lE31-R603 Drywell Floor Drain Sump Pump Out Timer lE31-R604 Drywell Floor Drain Sump Fill Up Timer 1E31-R605 Drywell Equipment Drain Sump Pump Out Timer lE31-R606 Drywell Equipment Drain Sump Fill Up Timer lE31-R608 Ambient Monitoring Multipoint Recorder lE31-R609 Drywell Cooler IVP02CC Drain Flow RADIATION MONITORING lur-RM021A Line Isolator 1UT-RM021B Line Isolator 1UX-RM021 Communications Plug 1UT-RM022A Line Isolator 1UT-RM022B Line Isolator 1UX-RM0-2B Line' Isolator 1UX-RM022 Communications Plug 1UT-RM023A Line Isolator 1UT-RM-23B Line Isolator 1UX-RM023 Communications Plug 1UT-RM024A Line Isolator 1UT-RM024B Line Isolator 1UX-RM024B Line Isolator 1UX-RM024 Communications Plug 22-5
RADIATION MONITORING (cont'd) 1UT-RM027A Line Isolator 1UT-RM027B Line Isolator IUX-RM027 Communications Plug 1UT-RM028A Line Isolator 1UT-RM028B Line Isolator 1UX-RM028 Communications Plug 1UT-RM029A Line Isolator 1UT-RM029B Line Isolator IUX-RM029 Communications Plug 1UT-RM030A Line Isolator 1UT-RM030B Line Isolator 1UX-RM030 Communications Plug HVAC 1TE-VP033A Lower Drywell Supply Fan 01CA 1TE-VP033G Upper Drywell Supply _ Fan OlCC 1TE-VP033C Lower Drywell 1TE-VP033D Upper Drywell 1TE-VP033E Drywell Head Cover 1TE-VP033F CRD Area 1TE-VP033G RPV Annulus 1TE-VP034A Lower Drywell Supply Fan 01CB 1TE-VP034B Upper Drywell Supply Fan OlCD 1TE-VP034C Lower Drywell 1TE-VP034D Upper Drywell ITE0VP034E Drywell Head Cover 1TE-VP034F CRD Area 1TE-VP034G RPV Annulus l
l 22-6 1
MISCELLANEOUS 1B21-N094A Drywell Pressure LPCS RHR-A ADS-ARI 1B21-N094B Drywell Pressure RHR-B-C ADS-B RCIC 1B21-N094E Drywell Pressure LPCS RHR-A ADS-A 1B21-N094F Drywell Pressure RHR-B-C ADS-B f
1C61-K501 Drywell Temp-Lower 1C61-K502 Drywell Temp-Upper 1C61-N501 Drywell Temp-Lower 1C61-N502 Drywell Temp-Lower 1C61-R501 Drywell Temp-Lower 1C51-R502 Drywell Temp-Upper i
4 22-7
Action Plan 25 8.1 This issue is based on consideration that some technical specifications allow operation at parameter values that differ from the values used in assumptions for FSAR transient analyses.
Normally analyses are done assuming a nominal containment pressure equal to ambient (0 psig) a temperature near maximum operating (900F) and do not limit the drywell pressure equal to the containment pressure.
The technical specifications permit operation under conditions such as a positive containment pressure (1.5 psig), temperatures less than the maximum (60 or 700F) and drywell pressure can be negative with respect to the containment (-0.5 psid).
All of these differences would result in transient response different than the FSAR descriptions.
Response
The following program for resolution was used to close this issue for Clinton:
25.1 A detailed summary of all conservatisms which currently exist in the containment response analyses which are part of the FSAR will be provided.
Conservatisms in the suppression pool temperature analysis will be identified in Action Plan 12, 25.2 Complete an end point analysis to demonstrate that with all initial containment parameters at worst case values, the containment design pressure is still not significantly exceeded.
25.3 Perform an analysis with worst case values taking credit for realistic temperature differences between containment and suppression pool and the containment heat sinks.
25.4 A complete review of the technical specifications for contain-ment conditions versus accident analysis assumptions will be made.
A comparison of technical specification values and values used as initial assumptions in the accident analysis will be submitted.
Item 25.1 All conservatisms in the containment, pressure temperature and suppression pool temperature response including those referred to in Itan 25.'1 of the Program for Resolution are quantified and discussed in the information submitted under Action Plan 12.
25-1 l
Item 25.2 MP&L submitted a sensitivity study (see Reference 1) involving _
drywell and containment initial conditions similar to those at CPS, which affect Design Beeis Accidene (DBA) long-term contain-ment response.
That study basically drew on end-point calculations to establish sensitivity trends governing DBA peak containment pressure.
The study concluded that even under conservative (adverse) drywell and containmnet initial conditions, peak containment pressures would not exceed design (15 psig).
The response provided by MP&L in Reference 1 also discussed at length the non-realistic nature of end-point analyses.
As two examples:
1)
Such end-point analyses neglect the DBA pressure-reducing action of the safety-grade redundant containment spray trains.
2)
They also neglect the inherent energy-absorbing (pressure-reducing) action of the containment and drywell heat sinks --
energy sinks that become significant over the (typically) 4.0-5.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> post-LOCA when peak DBA pressure is reached.
The Containment Issues Owners Group (CIOG) has continued to evaluate varying combinations of conservative initial containment and drywell conditions.
The CIOG has expanded the range of initial drywell pressures evaluated up to an initial drywell pressure that initiates reactor scram and generates a LOCA signal.
These studies computed the peak containment pressure under hypothetical conditions where containment design temper-atures of 1850F, and 100% RH, are attained in the containment airspace.
The entire drywell air mass is assumed to be trans-ferred to the containment with no redistribution to the drywell.
The resulting sensitivity trend to varying initial drywell pressure, under " worst-case" initial conditions for all other parameters, ig given in Figure 25-1 for initial drywell temper-atures of 105 F and 135 F.
These results are excessively conservative with respect to GGNS & CPS.
As noted in reference 1, the actual calculated peak long term post accident containment temperature is 180 F assuming that thermal equilibrium exists between the suppression pool and the containment air space.
This is lower than the end point cemperature used in the CIOG sensitivity study.
The CIOG analyses also include the vapor pressure of water at 185 F which is also higher than the vapor pressure which would be predicted at GGNS & CPS using the conservative licensing basis assumptions.
In additicn, the air.velume ratio of containment to drywell at CPS is 21% higher than the volume ratio considered in the CIOG analysis, i
(
25-2
These results show that under excessively conservative, non-realistic assumptions and a methodology which neglects operator mitigating actions (EPG procedures), it is possible to compute end-point states for the containment airspace which exceed the containment design pressure.
The CIOG does not believe that such end-point calculation results are appropriate for assessing the adequacy of containment design.
The CIOG feels that no purpose is served in pursuing further contrived end-point computations of this nature and, accordingly, no further analysis on this issue is planned.
Item 25.3 A more realistic analysis, evaluated the conservatisms collectively associated with such end-point calculations.
To recap that response, FSAR licensing basis assumptions were used in GE's latest proprietary long-term containment response code, SHEX, to establish a reference DBA containment response transient.
- Then, a re-run was made with a conservative (adverse) initial conditions mentioned above in the first paragraph, and with realistic accounting for containment and drywell heat sinks and (non-equilibrium) contain-ment airspace temperatures that results from the counter-effects of pool surface evaporation and heat transfer, and heat transfer from airspace to heat sink.
This comparison showed that the resulting "more realistic" peak containment airspace pressure, relative to the FSAR reference" case, is lower by 4.3 psi.
Item 25.4 IP has made a complete review of technical specification requirements and compared these with accident analysis assumptions used in the generic end point analysis response'to item 25.2.
The following list is a comparison of the major assumptions used in this analysis.
1)
CPS technical specification 3.6.1.6 requires containment pressure to be no higher than +1.5 psig during operation.
Generic end point analysis used a pressure of +1.0 psig.
The generic end point analysis reported in item 25.2 utilized the standard BWR 6-238 technical specification parameters and performed a sensitivity study to determine the effect of variation in these parameters.
Based on the analysis results and the consideration
.of 1) larger containment volume 2) lower _ power level and
- 3) the compressibility of the containment atmosphere, i.t. is judged that the additional 0.5 psig allowed in the Clinton technical specification will have an insignificant effect on the final pressure in the containment following an accident.
2)
CPS technical specification 3.6.2.5 requires drywell to containment differential pressure to be maintained between
.10 psid and +1.5 psid.
This establishes maximum drywell initial pressure between 1.4 psig and 2.0 psig (SCRAM).
Generic end point analysis considered a wider range of between.85 psig and 2.0 psig (SCRAM).
25-3
The wider range of initial drywell pressure considered in the generic end point. analysis is a direct result of the higher initial pressure considered at CPS.
This results in a narrower range of initial drywell pressure at CPS (i.e. 1.4 psig to 2.0 psig) which is a conservative assumption.
The end results are a narrower range of final pressures but no change in the maximum pressure.
3)
CPS technical specification 3.6.3.1 requires the suppression pool temperature not to exceed 950F.
A final temperature of 185 F, used in the generic end point analysis is obtained by calculation of a 900F temperature swing.
Also initial containment air temperature used in the analysis is 95 F.
The initial (950F) and final (185 F) temperatures used in the analysis are based on technical specificationg and FSAR design values.
However the final temperature of 185 F is conservatively calculated to be 1800F which, if used, would lower the final calculated pressure considerably as shown by MP&Ls response to item 2 (Ref. 1) 4)
CPS technical specifications do not limit low temperatures in the drywell, however the containment issues owners group have determined a minimum temperature of 1050F.
The nominal temperature expected initially is 1350F, compared with a maximum operation temperature of 150 F per CPS technical specification 3.6.2.6.
The minimum initial drywell temperature of 105 F is a very conservative temperature used in the generic end point analysis and is substantially lower than the normal operating temperature at CPS.
Reference 1)
MP&L's AECM-83/574, Item 2 of Action Plan 25.
25-4
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L-25-5
Action Plan 28 9.3 It appears that some confusion exists as to whether SBA's and stuck open 3RV accidents are treated as transients or design basis accidents.
Clarify how they are treated and indicate whether the initial conditions were set at nominal or licensing values.
9.4 The design basis calculations for stuck open relief valves should not rely on a manually initiated scram to occur within a short time interval of the SRV occurrence.
The analysis should assume that scram is initiated when the pool temperature reaches 110 F or 10 minutes after occurrence of the SORV.
Response
The program for resolution to close this issue at Clinton is as follows:
28.1 IPC will submit a letter confirming that the small break accident and the stuck open relief valve transient were treated as design basis accidents.
The analyses for these transients are completed using licensing basis values for the initial conditions.
28.2 IPC will verify that the Chapter 15 analysis of a SORV assumes reactor scram occurs when suppression pool temperature reaches 1100F.
Item 28.1 IPC has verified that small break accidents, as discussed in FSAR Section 6.2.1, have been treated as design basis accidents with initial conditions consistent with other licensing analyses.
The stuck open relief valve transients and small break accident transients in the suppression pool have been evaluated (see attachment I).
The results of this evalu ation shows that the maximum 0
suppression pool temperature gxpected is 171 F.
This is less than the design temperature of 185 F.
Item 28.2 The analyses of SORV events discussed in Chapter 15 of the CPS-FSAR 1
have been reviewed.
In all cases, scram is not required to return the system to a stable operating mode or scram is initiated by the event.
1 28-1 I
)
Attachment I I
i Plant operational transients, such as a turbine trip or loss of normal feedwater flow will actuate the SRV's to maintain the vessel pressure within desirable limits.
Once SRV's open, the steam released from the reactor is discharged through SRV. lines into the suppression pool.
Steam is then condensed in the pool in a stable condition.
Extended steam blowdown into the pool at an elevated pool temperature could yield an unstable steam condensation which might cause severe dynamic loads on the containment structure.
The current practice to deal with this phenomenon is to restrict the allowable operating temperature envelope of the pool in the Technical Specification such that this instability will not occur.
Early in 1981, the NRC issued the'NUREG-0783 to establish the operational temperature limit for suppression pool and requested that analysis be performed to assess Mark I, II and III containment designs.
The events requiring analysis by the NRC were carefully evaluated and a revised set of six transient cases which bounded all these events was identified.
The six identified transients are: (1) Stuck open relief valve (SORV) at full power with a loss -of 1 residual heat removal (RHR) heat exchanger and with the' condenser available to an RPV pressure of-150 psia; (2) SORV at full power with spurious isolation; (3)
Isolation scram with 1 RHR heat exchanger available; (4). Isolation j
scram with-SORV and loss of shutdown cooling; (5) Small break accident (SBA) with a loss of 1 RHR heat exchanger; and, (6) Small break accident (SBA) with loss of shutdown cooling.
A detailed-description of assumptions relevant to each of the cases is presented in Table 28.1 an'd'28.2.
Table'28.3 presents the.Clinton specific data that were used in these evaluations. ' Pertinent parameters related to the SRV's are presented in Table 28.4.
The assumed system delay times are summarized in Table 28.5.
The event sequence for each analysis is given in Table 28.6.
Final results are shown in Table 28.7.
I 4
J f
't i
28-2
TADLE 28.1 CLINTON POWER STATION COUTROL LOGIC Case Description Shutdowp Cooling Mod 5'
' Suppression Pool Cooling Mode HPCS FitR 1 I-RHR 2 RilR l -
RIIR 2 1.
.SORV0 POWER
+16 minutes after Not used TSi + -10 minu,tes Not used Not used Loss of 1 RIIR operator action (PR <135 psia) 2.
SOttV0 POWER Not used Not used TS1 +i10 minutes TS1 + 10 Not used
-SPURIOUS minutes ISOLATION 3.
ISOLATION
+16 minutes after N6t used TS1 +.10 minutes Not used Not used SCRAf1 operator action Loss of 1 Ri!R (PR <135 psia)
?4 ISOLATION /
Not used Not used TS1 + 10 minutes TS1 + 10 Not used ks
~
minutes W.
SCRAll SORV 5
+16 minutes dfter.
Not'used TS1 + 10 minutes Not used Ifigh drywel pressure Loss' of.1 RHR operator action signal with (PR <135 psia) 27 second delay cycle
~
on RPV wate t
level.
6.
SBA, Loss
. Not used Not'used TS1 + 10 minutes TS1 + 10 Same as 5 minutes of shutdown cooling.
I i
6 e
e e
TABLE 28 2 CLINTON POWER STATION CONTROL LOGIC Case Description Main Condensate MSIV Depressurization Rate Reactor -
Feedwater B'ooster Pump Closure ~
rGovernor Power-1.
SORV@ POWER
, Fully Not Used None
- 1),35% of rated bypass SCRAM @
Loss of 1 RHR available
- 2) 100 DEG F/HR until TS3 RPV pressure = 135 psia.
- 3) SORV 2.
SORV@ POWER Available Not Used
@T=0 seconds
- 1) 100 DEG F/HR SCRAM @
-SPURIOUS for fully closed
- 2) SORV T=0 ISOLATION
, makedp
@,3.5 seconds seconds 3.
ISOLATION /
Available Not Used
@T=0 seconds 100 DEG F/HR until RPV SCRAM @
9m SCRAM for fully closed pressure = 135 psia T=0 e
Loss of 1 RHR makeup
@ 3.5 seconds seconds 4.
ISOLATION /
Available Not Used
@T=0 seconds
- 1) 100 DEG F/HR SCRAM @*
SCRAM for fully cl6 sed
- 2) (SORV)
T=0 SORV makeup
@ 3.5 seconds seconds 5.
SBA' Flow-in =
On when
@T=0 seconds
- 1) -
SCRAM @
UPCS Loss,of 1 RHR Flow out PR <500 psia fully closed
- 2) 100 DEG F/HR T=0
""ttl
@ 3.5 seconds ntil RP pressure seconds e
e -off for rest of
- 3) Break flow transient 6.
SBA Same as 5 Same as 5
' @T=0 sec~nds
- 1) HPCS
- SCRAM @
o Loss of fully closed
- 2) 100 DEG F/HR T=0 Shutdown 0 3.5 seconds 3). Break flow seconds
~~
cooling I
?
s o
9 84
=
1 i
TABLE 28.'3 Clinton Plant Sobcific Data' a
Reactor and Associated System Soecifications Reactor Core Power (102% rated) 2.797 x 10 Btu /sec 4
3
~ '
1.659 x 10 ft Reactor Volume.
t 5
Initial RPV' Liquid Mass 4.392 x 10 -lbg.
4 Initial RPV Vapor Mass 1.557 x 10 lb, 5
i Initial RPV Water Mass 4.5477 x 10 73m knitialSteamFlow (Appendix B) 3.459 x 10 lb /sec 3
m Initial Reactor Pressure 1040 psia i
lbm ft (Appendix B) 26.51 Turbine Bypass Flow Coefficient sec lb
- psia m
6I RPV Neat Structure Mass 2.534 x 10 lb RPV. Heat Structure Specific Heat
. 111 Btu /lbg
- F
~
RPV Heat Structure Area 6984.8 ft RPV Heat Structure H.T.C.
36/000. Btu /,f t
- F hr Feedwater liquid-level control efficient.
100 lb,'/sec-in.
RPV maximum pressure for shutdown cooling 135_ psia Feedwater Enthalpy Integrated FW FW Flow since Scram (lb )
Enthalpy (Btu /lbg) m 0.
398.0 124579 398.0 124580 358.0 605076 358.0 605077 310.0 r
642248 310.0 642249 271.0 1000000 271.0-28-5 l.
4 T
TABLE 28.3(cont'd)
Clinton Plant Snecific Data Flow Il m/sec)
HPCS (see note 2 Pres u e (osia) on page 28-19) 0 677.5 214.7 677.5 1161.7 193.6-1191.7 64.6 3
Suppression Pool and'Associ'ated System Soecifications 8.42 x 10 lb Initial Pool Mass m
Initial Pool Temperature 95*F
- Service Water Temperature -
95*F RHR-RX Effectiveness
.449 RHR Mass Flow Rate - Pool Cooling 698.3 lbg/sec lb /sec.
RHR Mass Flow Rate - Shutdown Cooling 698.
g Pool Cooling Technical Specification Temperature
~
(TS1) 95*F
- SCRAMSTechnical Specificatidn Temperature (TS3).
110*F.
Depressurization Technical Specification 120*F Temperature (TS4) f e
e 4
28-6 s
1
A l
TABLE 28.4 SRV Soecifications Number of SRV's 16 SRV Seat Area (In )
14.65 Wetwell, backpressure on SRV,(PSIA) 20.5 SRV Loss Coefficient (choked flow) 1~. 0 SRV Loss Coefficient (friction flow) 1.88 l
SRV Rescat Differential Pressure (PSIA) 50 SRV Setpoints (PSIA) 1103.0 8 x 1113.0 7:. x 1123.0 1
e i
t o
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9
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e l
l I
e 9
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=p y,p.
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. -.r p-
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44.-
-n..--
- ~
--m-----.
~
TABLE 28.5 System Delay Times Delay time to establish turbine bypass flow 1200 sec.
Switchover time for shutdown cooling 960 sec.
Delay time for start of pool cooling 600 sec.
Delay time for initial IIPCS operation 27 sec.
Time for complete MSIV. closure 3.5 sec.
Time for complete turbine stop valve closure 20 sec.
+
e es O
9 e
4 e
e e
S*
e e
g e
+
t, b
e aus e
9 e
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9 em.
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a'
+
7 j
i i t '~
TABLE 28.6 rEvent! Sequence.
I CASE 1 Time (sec)v SORV, pool temperature at 95B I
O TS3, SCRAM, start of turbine stop valve closure
- 406 1 RHR-HX in pool cooling 600 Turbine bypass flow established 1625 Turbine bypass flow stopped, start of manual depressurization (M.D.) @ 100*F/hr 1875 (PRPV = 150 pein)
Start of switch from pool cooling to shutdown cooling, end of M.D.
(P
= 135 psia) 3000 RPV Start of shutdown, cooling 3960c
'1
\\
CASE 2
. Tine (sec)
/
/
Oi SORV, pool temperature at TS1 c
Pool temperature at TS3, Reactor scramraed, start of MSIV Closure **
406 625-t Two RHR-HX's in pool.cocling
,\\
~ I 'j 0
Pool temperature at TE'4, start 65,0' of M.D.
@ 100* F/hr J
j
-i 1
I i
b) {
\\. >
l k.
/
j Turbine stop valve fully closed'20 'sec. following start 4 j
s l
of closure.
/
s
- MSIV fully closed in 3.5 sec. following' start of closure l
i I
28-9 l
- 1...
r TABLE 28.6(cont'd)
Eyent Sequence CASE 3 Time (sec)
SCRAM, TS1, start of MSIV closure
TS4, start of M.D. @ 100*F/hr 2000.
Start of switch from pool cooling to shutdown cooling, and of M.D.
(P
= 135.0 psia) 9955.
RPV Start of shutdown cooling
- 10209, CASE 4 Time (sec)
TS1, SCRAM, SORV, start of MSIV closure
- O
SRV Operation on high.RPV pressure 0-40 2 RHR-HXs in pool cooling 600 TS4, Start of M.D.
@ 100*F/hr 990
- A.~.._
CASE 5 O
TS1, SCRAM, Small break, start of MSIV closure *,
Initial HPCS Operation 27-235 10-2840 SRV Operation on high.RPV pressure 1 RER-HX in pool cooling 600 TS4, start of M.D. @ 100*F/hr 1700 Second HPCS Operation
.~G80-1290 Pressurizati'on of the RPV
~1290-1910 Third HPCS Operation
~2995-3150 TR4,. start of M.D. @ 100 F/hr 3305-
- MSIV fully closed in 3.5 sec. following start of closure.
28-10
N TABLE 28.6(Cont'd)?
Event Sequence Time (sec)
FW condensate booster pump flow 3615 Start of switch from pool cooling to shutdown cooling, end of M.D.
8485 (P
= 135.0 psial RPV Start of shutdown cooling 9445 4 CASE 6 TS1, SCRAM, Small break, start of MSIV closure
- O Initial HPCS Operation
~27-23 5 SRV Operation on high RPV' pressure
~ 10.e28 4 0 2 RHR-HX's in Pool Cooling 600 Second HPCS Operation
~ 9 8 0-12 9.0 Pressurisation of the RPV
~1290~1910 Third HPCS Operation
,L3150-3305 j
Pool temperature at TS4, start of M.D. at 100*F/hr 3460
'390 FW Condensate Booster Pump Flow 4
i t
- MSIV f ully closed 3. 5 sec. following start of closure l
l 28-11
~
~.
TABLE 28.7
- Maximum'Suppr'essi'on Pool Temp ^erature RPV Pressure' at Time Initial Pool Itaximum Pool Case Maximum Pool of Max.-
Water Mass Water Mass Temperature
. Pool Temp.
(lb )
(lb )
(*F)
~ (p s ia') '
6 6
1 151.6 20.7 8.42 x 10 8.878 x 10 6
6 2
167.9 22.8 8.42 x 10 9.710 x 10 6
6 4
3 171.0 121.7 8.42 x 10 9.143 x 10 6
6 4
161.6 22.6 8.42 x.10 9.610 x 10 6
6 5
164.5 14.a' 8.42 x 10 8.761 x 10 6
6 6
163.5 23,1, 8.42 x 10 9.252 x 10 9
6
+
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800. E R
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3
-i 80.
D.
.i..
iiiij
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iiiig i
i i i i i g i i ii 10 10 10
. 10 10 TIME SINCE START OF TRANSIENT (SECONDS) l l
FIG 28 l-CASE 1:SORV RT FULL POWER - LOSS OF 1 RHR - PSJRE = 1:30 PSIR r
(Ulsd)~3HOSS38d Ad3
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(d 030) 3801883dW31 100d NOISS38ddnS 28-15
3 4
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h-g i.
a i
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t 800. E 5
160.-
g ta ta m
to m
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.i.i.,iiig
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.i.. iiiig
.i.
iiii 4
3 10 10 102 3g 10 10 TIME SINCE REACTOR SCRBM (SECONOS)
FIG Z8.5-CRSE 5: SBR - LOSS OF 1 RHR -
INSTANTRNEOUS C00LDORM R8TE
120d.
200 y
- 1000.
u-180.-
o w
Q n
cc w
m 800. E g
?
160.-
g W
w Ct' z
=>
a.
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w w
hl w
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- o Ae C
q e
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a
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e Q
m 200.
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=
. i i. iiiig
.i.i. iiiig
.ij. i,i i i g
. isiig
.i.i. i gii 80.
10 10 10 10 10 10 TIME SINCE REACTOR SCRAM (SECONOS)
FIG Z8.4 -CASE 4:ISOLRTION SCRAM - SORV ON ISOLATION
200.
1200.'
~
l l
lib$h n
1000.
h-180.-
8 a
n c:
to s.e x
800. E 0
160.-
a-n:
to to D
n:
cu C
m to y
l H
600. y
- j 140.-
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5 2
o
}
G 400.
0 17.0. -
E
~
!w 5
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l D.
00*
'g'l'I'gI11ll
' g ' i 'I'g i 11ll
- g'i'I'gI'llll
' g
- I ' I'g i I ll)
' g
- I ' I 'g i I l l 10 10 10 10 10 10 TIME SINCE.RERCTOR SCRRM.CSECONOS)
FIG 28.6 -CRSE 6: SBR-LOSS OF SHUTDORN C00L'ING-INSTANTANEOUS C00LDOWN RATE
Notes:
1.
In the' Events 3 and 4, it is not assumed that the containment accepts the " hot" portion of the feedwater in the feedwater system.
This is because Events 3 and 4 are considered isolation events.
The feedwater pumps are tripped at the time the break is detected.
Following this, the pumps will coast down and the isolation valves will close in a short period of time.
Therefore, consideration of feedwater addition is not modeled.
2.
For Events 1 through 4, HPCS is considered available but it is not used in the analyses.
In Event 5 and 6 HPCS is in the Standby Mode and is not pressurized so it will not be available in a DBA and should not be treated in a Small Break Accident.
3.
In event 1.the turbine stop valve closes completely 20 seconds after the reactor is scrammed which effectively isolates the reactor from the main condenser.
At 1200 seconds after scram, the turbine bypass valve opens the main condenser becomes available as a heat sink for reactor steam and RPV pressure drops until the RPV pressure becomes 150 psia.
Then the turbine bypass flow is cut off and the depressurization rate of the RPV is reduced.
At 1800 seconds after the scram, the manual depressurization of 1000F/hr. is initiated until the RPV pressure drops to 135 psia.
This pressure is maintained until the switch over of the RHR system from pool cooling to shutdown cooling is accomplished.
The RPV pressure continues to decrease until atmospheric pressure is reached.
4.
A single SRV is modeled as open throughout the transient for Events 1 and 2.
For Event 4, a single relief valve cycles throughout the transient.
For Events 3, 5, and 6, nine SRV's open shortly after isolation and then reclose.
These represent the two lowest setpoints j
(1103 and 1113 psi).
Following reclosure, the low setpoint valve will cycle throughout the transient.
5.
The RCIC is available in all Events but is not considered in the analysis.
Use of the RCIC would reduce the thermal load on the suppression pool, dumping the energy to the condenser, thus producing a non-conservative evaluation of the suppression i
pool temperature.
6.
In Events 1 through 4 the condensate storage system is available but is not considered in the analysis.
Use of the condensate storage system would increase the thermal capacity of the suppression pool and thus. produce non-bounding transients for the events under consideration.
2 8 - 19
7.
The source of the decay heat curve used in the analyses is the CPS specific calculation performed by GE(22A3759AM, Rev. 3).
This decay heat curve includes the effect of delayed neutrons and void collapse.
8.
For Event 5 shutdown cooling is considered.t 9.
In Events 4 and 5 manual depressurization is considered.
9 1
28 20
Action Plan 31 11.0. Mark III load definitions are based upon:the levels in the suppression pool and the drywell weir annulus being the same.
The CPS technical specifications permit' elevation differences between these pools.
This may effect load definition for vent clearing.
Response
The~following program was used to resolve-this issue:
.- 13.1 The maximum possible differences between weir annulus level and suppression pool level will be defined.
This definition will include an evaluation of changing the vacuum breaker set point.per Action Plan 29.
31.2 A discussion will be given of how pressure differences between the wetwell and th'e drywell will be controlled.
31.3 A discussion of how these pressure differences affect
-load definition will be provided.
Item 31.1 The maximum negative differential pressure which can exist between the drywell and containment is -0.20 psid which corresponds to the set. point of the drywell vacuum. breakers.
This produces a maximum increase in weir annulus water level above suppression pool level or.5.54 ~ inches.
In actuality, it-will not be possible to ' achieve any negative pressure in the drywell with respect to containment during normal plant operation.
f.
The maximum positive differential pressure which can exist between L
the drywell -and containment is l'.50 psid.
This positive differential-pressure produces a maximum increase in suppression
_ pool level above weir annulus level of approximately 3.5 feet.
Item 32.2 The negative pressure condition in the drywell is controlled automatically by the vacuum relief valves (lHG010A D,
~
1HG0llA D) in the Combustible-Gas Control System.
These valves f;
automatically open at -0.20 psid to equalize the pressure between l
the dryw e ll and the containment.
'The positive pressure l
condition in the drywell is controlled by the 24" drywell purge system valves (lVQ001A, IVQ002)_which are opened only when the operator must reduce the differential pressure under abnormal conditions.
s 31-1 i
Item 31.3 The differential pressure between the CPS drywell and
~
containment air space is constrained within the normal range:
-0.10 6 P
+ 1.50 DW-W with differential pressure in psid.
If the drywell pressure is greater than the containment airspace pressure the water. level in the weir annulus will be depressed and consequently, the liquid inertia above the top vent will be reduced.
This will cause the top vent to clear earlier folladng a postulated LOCA resulting in lower drywell pressure when the vents clear and a lower peak drywell pressure than has been calculated in the existing accident analysis.
The' lower driving pressures decrease the pool swell velocities, accelerations and loads.
If, on the other hand, the initial containment airspace pressure is greater than the initial drywell pressure, top vent clearing would be delayed which would increase the peak drywell pressure.
An analysis was performed to determine the upper limit of this effect for the Clinton Power.. Station when the AP w_wu d
is -0.1 psid.
This corresponds to the water in the weir annulus being elevated by almost 2.8 inches.
The results of this analysis show this small change produces a negligible affect on the pool swell transient and drywell bypass leakage.
31-2
i Action' Plan'32 14~.0 A failure in the check valve in the LPCI line to the reactor vessel could result in direct leakage from the pressure vessel to the containment atmosphere.
~
This leakage might occur as the'LPCI motor operated isolation valve is closing and the motor operated isolation valve in the containment spray line is opening.
This could produce unanticipated increases in the containment pressure.
Response
The following program for resolution was used to close this issue out for Clinton:
32.1 The potential effect of maximum backflow which can occur will be estimated.
This will include calculating i
maximum backflow which can occur, evaluating thermal l
interaction with the relatively cool RHR spray flow and estimating the limitations on flashing created by flow through the spray nozzles.
32.3 An evaluation of-the possibility of adding interlocks to prevent simulat..neous actuation of these valves will also be performed.
In response to 32.1, a schematic diagram depicting the arrangement of the LPCI system and the containment spray system is included as Figure 32-1.
The diagram shows only one LPCI.- spray system since the analysis considers only the single active component failure on one LPCI check j
valve following the postulated LOCA.
A bounding calculation was performed to evaluate the I
maximum containment. airspace pressurization which could occur due to the postulated backflow. - This calculation determined the maximum mass and energy addition to the airspace, then used a standard subroutine '(THERMO) to determine the resultant containment pressure and temperature.*
i THERMO, a standard component of GE containment analysis computer codes, assumes thermodynamic equilibrium of all components (air, steam, liquid) in the airspace.
Then,-
for a given airspace volume, air and water (steam and liquid) mass, and total internal energy, the airspace pressure and temperature are calculated.
1 l
32-1 1,.
i
..,s
- -. _ -. - ~ _ _,. _, _ _ _,..., _,..-_...,_,
.--.s
System Performance Key features of the system are:
1.
The RHR pump shut-off head is 750 ft, which is approximately 325 psi.
2.
Containment spray is actuated with a simultaneous signal to close the LPCI motor-operated. valve and open the containment spray motor-operated valve.
3.
Containment spray cannot be automatically actuated until 10 minutes after the LOCA.
Automatic actuation then occurs only if the containment pressure is greater than or aqual to 9 psig.
Valve closure and opening times, based on GGNS start-up data, are shown in Table 32.1.
Conservatively assuming that the valve flow area equals the pipe flow area, and taking the maximum LPCI closing time and the minimum containment spray opening time, simplified valve flow area vs. time curves as shown in Figsre 32-2 are assumed.
These closure / opening times and valve flow area are conservative for CPS because the valves for Clinton are much smaller than GGNS meaning faster opening / closure times and lower flow areas.
Containment' Response Calculation The maximum mass and energy addition to the containment airspace due to reactor backflow is calculated by assuming critical flow at reactor pressure.
Pipe flow greas of the LPCI/ containment spray system vary from 0.56 - 19.7 ft.4, so the critical flow limiting area will be at the LPCI or containment spray valves.
Thus, the curve bounding the shaded triangle in Figure 32.2 represents the limiting flow area vs. time.
An equivalent constant area over the "both-valves-open" window of 18.5 seconds (0.15 ft2), was calculated to simplify the final calculation of containment pressurization.
Assumptions for analysis of net containment pressurization are:
1.
Reactor pressure is 325 psia.
(The failure of the check valve cannot occur until there has been LPCI flow to the vessel, opening the check valve, so the vessel pressure cannot be greater than 325 psia).
2.
The containment air mass includes all of the air which was initially in the drywell.
3.
Containment pressure = 9 psig at start of backflow.
Results and Conclusions Results of this analysis are summarized in Table 32.2.
The leakage from the reactor through the containment spray headers
~
pressurized the containment 0.8 psi.
Thus, this postulated reactor backflow pressure analysis is very conservative for CPS and does not represent a safety concern.
32-2
In answering 32.3, since the backflow increases the containment pressure by less than 1 psi, IP has. determined that a detailed evaluation of providing interlocks to prevent this backflow is unwarranted.
32-3
Table 32.1 GGNS VALVE OPENING / CLOSURE TIMES Opening Closing Valve (sec)
(sec)
Containment Spray Valve #E12-F028A 65.0 N/A Valve #E12-F028B 65.5 N/A LPCI Valve #E12-F042A N/A 18.5 Valve #E12-F042B N/A 17.5 Table 32.2 CONT /,INMENT PRESSURIZATION DUE TO REACTOR BACKFLOW THROUGH THE CONTAINMENT SPRAY SYSTEM Containment pressure at start of backflow 9
psig Pressurization due to reactor backflow
+0.8 psig NET RESULTANT CONTAINMENT PRESSURE 9.8 psig 32-4
SPRAY HEADERS
~_
3 I K-O n0 g A
L e
F, I
RPV M0 hM0 V
RHR PUMP SUPPRESSION P00L' C
c Figure 32-1 CPS Containment Spray and LPCI Piping Schematic 32-5
7 1
1.5 Containment Spray 1.4 MO Valve 1.3 1.2 i
1.1 LPCI 1.0 MO Valve b
9 E
A
.8 5
.7 E
6 m
.5 4
Maximum "Both-Valves-0 pen" Window = 18-1/2 seconds.
wt.
.lb
.AP
- * ~ ' -
c 0
10 20 30 40 50 60 70 80 90 100 110 Time, Seconds Figure 32.3. GGNS Valve Open Area vs. Time 32-6