ML19345F777
| ML19345F777 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 01/13/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19345F771 | List: |
| References | |
| NUDOCS 8102190270 | |
| Download: ML19345F777 (5) | |
Text
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4 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 37 TO FACILITY OPERATING LICENSE NO. OPR-6 CONSUMERS POWER COMPANY BIG ROCK POINT PLANT DOCKET NO. 50-155 1.0 INTROCUCTION By-letter dated December 5,1980 and supplement dated December 19, 1980, Consumers Power Company (the licensee) requested an amendment to Facility Operating License No. DPR-6 for the Big Rock Point Plant.
This amendment would modify the Technical Specifications to reflect plant modifications designed to mitigate the effects of postulated steam line breaks. The plant modifications will incorporate (1) a prompt automatic containment spray system, (2) a remote manual backup containment spray system, and (3) the addition of spray nozzles in the steam drum cavity.
2.0 DISCUSSION Big Rock Point is a 71 MWe BWR in a cry containment.
It received its operating li ense in 1962 and was originally built as a demonstration pl ant.
While performing steam line brm (SLB) analyses for our Systematic Evaluation Program's (SEP) equipment qualification review, the licensee determined that the contr.inment design temperature could be exceeded.
This was reported in Ref. 1.
The original Big Rock Point safety analysis only considered the large, two-phase LOCA. When the smaller, single-phase SLB's ( < 75 gpm) were considered with the addition of superheat, the 0
analyses ~ predicted that the cgntainment design temperature of 235 F could be exceeded by as much as 100 F.
Big Rock Point is unique in two aspects that significantly affect the SLB analyses. Two fire pumps are u;ed to deliver water to both the core and containment spray systems. However, if one of the fire pumps was postulated to fail, insufficient water would be available to the two spray systems.
In response to this concern, the licensee previously took two steps to alleviate this matter.
First, one of the redundant containment spray trains was blocked out by removing power to the containment isolation valve (Operator action could restore power to this valve if needed). Second, a 15 minute delay time was placed on the containment isolation valve of the other containment spray train. The licensee
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calculated that after the initial 15 minutes the core would not need the additio.1al spra) cooling.
810278n 7. D
. The second unique aspect of Big Rock Point is their continuous purge system. During normal operation a slightly negative containment pressure
(-1/2" w.g.) is maintained by a pressure controller in the purge system's supply line. The pressure controller automatically adjusts the supply line flow rate while a constant exhaust flow rata maintains a negative pressure in containment.
If a small SLB is postulated, the pressure controller could close the supply side completely, and with the exhaust line open, the containment high pressure (isolation) setpoint may not be reached. Thus, a small SLB without containment isolation could enable the containment atmosphere to heat up without any automatic system to mitigate it.
The licensee concluded that some form of containment spray is needed immediately following an accident and that the present design which delays the sprays for 15 minutes is no longer acceptable. Modifications to the containment spray system have been proposed (Ref. 2). The modifica-tions include replacing the current containment spray no::les with icwer capacity no::les. One 1 the spray trains will now begin spraying a minimum of 50 gpm 75 secon s after the containment pressure reaches the spiay actuation setpoint of
- 2. psig.
The other spray train; which has had the power returned,,
to be manually actuated 15 minutes fo' lowing a break in order to remove ai borne iodine.
In addition, containment spray no::les will be insta'.itd in the steam drum cavity to protect safety related equipment from an overtemperature condition in the event of a break within this cavity.
The effect of these modifications is twofold.
For the large SLBs (> 75 lb/sec),
containment design temperature will only be exceeded by approximateTy 20 F for durations of one to two minates. However, for breaks with blow-downs of 75 lb/sec and less, the pressure controller in the purge system will prevent the containment pressure from reaching its automatic isolation signal setpoint of 1.7 psig. The licensee relies on ocerator action as the primary means to isolate containment and tric the reactor for a small SLB.
Calculations by the licensee show that for a break of 75 lb/sec, it would take approximately 20 minutes to reach the containment design temperature of 235*F.
3.0 EVA: lTION The purpose of this Safety Evaluation is to evaluate the proposed Technical Specification change to return power to both containment spray headers. By diverting ECCS water from the core spray system to the containment spray, there must be assurance that adequate core spray flow exists.
Finally, there must be some assurance that the operators will have adequate means available in the control room to alert them of a small SLB.
The core spray system consists of two subsystems, a core spray ring sparger and a core spray nozzle. Both of these spray systems have under-gone full scale testing with a varying flow rates and reactor vessel pressures to determine the necessary spray distribution on the core.
At reactor vessel pressures following actuation of the reactor depressuriza-tion system (RDS), minimum flow rates from the ring sparger and nozzle core spray have been detennined to be 292 and 296 gpm respectively.
These flow rates have been determined to meet core cooling requirements and assumed the decay heat generated to be equal to 1.2 times the ANS values as required by Appendix K to 10 CFR 50. Therefore, the licensee needed to provide reasonable assurance that the modified containment spray system will result in minimum ring sparger or nozzle core spray flow rates of 292 and 296 gpm respectively being available.
The licensee used its own computer code FLOWNET to determine the minimum flow to the core spray systems.
FLOWNET is a best estimate hydraulic code that sums all line and nozzle losses and determines the flow division between the containment and core spray systems. A spectrum of breaks along with various single failures were assumed to determine the worst core spray flow conditions. The licensee's results show that all postulated breaks would receive the minimum spray flow rates and, therefore, the proper spray distribution. Only one break received the minimum flow rate (a non-ECCS line break assuming failure of the diesel generator which disabled the core spray nozzle) while all the other postulated breaks received more than the minimun flow.
The staff requested further assurance from the licensee that adequate core spray flow would exist. Testing of the modified piping network was ruled out because it would require spraying firewater (which comes directly -
from Lake Michigan) into the reactor vessel and into containment.
Temporary piping modifications would be necessary which would include disassembling welded pipe supports for the containment spray piping.
Instead the licensee performed sensitivity studies to verify adequate core spray flow.
Testing performed by the nozzle manufacturer shows an uncertainty l
margin of !5% for the containment spray nozzle flow. The licensee's ans!ysis showed that by conservatively increasing the flow rate of each containment spray nozzle by five percent, the resulting core spray flow rate was decreased by less than 1 gpm.
Sensitivity studies, with varying line losses, led the licensee to conclude that for every 10 gpm increase in containment spray, core spray would be reduced by 1 gpm.
Knowing the pressure drop caused by the nozzles i
and the elevation change in the piping system, the only remaining uncertainty is the piping losses. Piping losses in the Big Rock facility
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. only account for approximately 15% of the containment spray system pressure drop.
Considering that piping losses can be calculated fairly accurately, any uncertainty in assumed piping losses for the containment spray system can be considered negligible with regard to core spray fl ow.
Therefore, the staff concludes that adequate core spray flow will exist.
As discussed' earlier, for small steam line breaks (1 75 gpm), operator action will be relied upon to scram the rcactor.
Scramming the reactor will also isolate the containment which in turn will permit a pressure rise in containment so that the spray system will be automatically actuated.
We nave discussed the need for prompt operator action with the licensee.
Plant procedures have been revised and operators have received special training in order to recognize and take correct action following a small SLB.
Considering the size of the Big Rock Plant, a small SLB would produce a step change in both steaa flow and electric output readouts in the control room.
(A 75 lb/sec treak is 27% of rated steam flow.)
Break flows large enough to produce rapid heating of the containment (50-75 gpm) may even be heard by the operator.
Tne primary means to alert the operators of a sna11 SLB wil1 be the high air temperature and dewpoint alarms. There are 12 different temoerature monitors that will actuate an alarm in the control room by the time air or dewpoint temperature reach 120*F. The control room alarm remains audible until it is manually reset. The alarm is on the control room front panel within 20 feet of all charts, the containment pressure indicator, control switches for manual scram, emergency condenser, and switches for manual actuation of the containment spray. Revised procedures require the operator to examine the steam flow, feedwater flow and electric output charts immediately following a high temperature alarm so that appropriate action can be taken.
The licensee's The Big Rock Point Plant is being reviewed as part of SEP.
In addition, pipe break analyses will also be reviewed as part of SEP.
the NRC is reviewing the ability of safety-related. electrical equipment to withstand the effects of a harsh environment as a part of the on-going environmental qualification reviews.
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. We conclude that the Big Rock Point proposed modification and Technical Specification changes provide reasonable assurance that the required safety functions of containment spray and emergency core cooling will function effectively in the event of a steamline break.
Therefore, we find the applicant's modifications to the containment spray system and the proposed Technical Specifications acceptable.
3.0 ENVIRONMENTAL CONSIDERATION
Ke have determined that the amendment does not authorize a change in effluent ty;es or total amounts nor an increase in pcwer level and will not result in any significant environmental imcact.
Having made this determination, we have further concluded that the amendment involves an action which is insjgnificant from the standpoint of environmental impact and pursuant to 10 CFR 551.5(d)(4) that an environmental impact statement or negative declaration and environmental im act acpraisal need not be precared in connection with the issuance of this E9encment.
4.0 CONCLUSION
S ke. ave concluded, based on the considerations discussed above, that:
(1)
- acause :ne amendment does not involve a significant increase in the pre-
- 2:ility or consequences of accidents previously considered and does not
'ns:lve a significant decrease in a safety margin, the amendment does not invtive a significant hazards consideration, (2) there is reasonable
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assurance that the health and safety of the public will not be endangered l
- / ::eration in the proposed manner, and (3} such activities will be con-l c;c:ed in ccmcliance witn the Ccmmission's regulations and the issuance of
- nis amendment will not be inimical to the common defense and security or to l
tne nealth and safety of the public.
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5.0 REFERENCES
1.
Big Rock Point LER 80-023, dated August 26, 1980.
l 2.
Letter from Consumers Power Company to NRC (G. C. Withrow to Director, l
NRR), dated December 5,1980.
i Dated: Janua ry 13, 1981 l
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