ML20084B182
| ML20084B182 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 08/15/1972 |
| From: | Hunnicutt D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | James Keppler US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20084B161 | List: |
| References | |
| NUDOCS 8304060164 | |
| Download: ML20084B182 (11) | |
Text
{{#Wiki_filter:r 9 U+ af o .~ peu re, ]f(~;1# f i, UNITED STATES '. i ATOMIC ENERGY COMMISSION DIVISION OF COMPLIANCE REGION lli 799 ROOSEVELT ROAD TELEPHONE GLEN ELLYN, ILLINols 60137 (312) e58-2660 August 15, 1972 J. G. Keppler, Chief, Reactor Testing and operations Branch Directorate of Regulatory Operations, Headquarters COMMONWEALTH EDISON COMPANY (QUAD-CITIES UNIT 1) DOCKET NO. 50-254 EVALUATION OF POSSIBLE CONSEQUENCES FOLLOWING A LONG TERM OPERATING HISTORY AT FULL POWER DURING A POSTULATED FLOODING SIMILAR TO THE OCCURRENCE IN JUNE 1972 AT THIS FACILITY Attached is an evaluation of the consequences of a postulated flood of the condensate booster pump room subsequent to a long operating period at full power. The evaluation indicates that, loss of cooling capacity would not be experienced, the unit could be cooled adequately to prevent fuel melt-down or primary system overpressure and that no serious safety hazards to the plant, plant personnel, or the general public would occur. One of the " side effects" of this evaluation was discovering that the licensee is not required by Technical Specifications or other inspectable commitments to assure that the various water storage tanks are kept filled or that a minimum inventory of emergency cooling water be maintained. RO:III is of the opinion that further study of the safety significance of this item be undertaken. RO:III recommends that licensee's Technical Specifications be amended to require an adequate inventory of emergency cooling water to assure that sufficient cooling water is readily avail-able for evaluated abnormal and/or emergency situations. b ff! h m w W D. M. Hunnicutt, Chief Reactor Testing and Startup Branch
Attachment:
As Above >f i; / ^( b>'n<l' t 131.. ,{ \\ V? 3 C [, Cid liC 8304060164 720901 03A13rj PDR ADOCK 05000254 S PDR
\\ O o QUAD-CITIES UNIT 1 COMMONWEALTil EDISON COMPANY l REPORT AND
SUMMARY
l An evaluation of the possible consequences of a flooding occurrence in the condensate booster pump room while the reactor was operating at full power during along period of sustained operation is forwarded for information. The data for this evaluation was taken from the flooding occurrence that occurred at the Quad-Cities Unit i facility on June 9, 1972. The purpose of this evaluation was to estimate the possible safety signifi-cance of a postulated flooding condition of the same area during operation at full power after the Unit has a significant operating history. This analysis assumes that other than the loss of the feedwater system, none of the remaining systems designed to supply water to the core were affected. Previous analyses, which are included in the FSAR indicate that sufficlent water would be avaliable over all pressure ranges to assure that the core would never become uncovered. Sufficient water capacity exists in the various condensate storage tanks and in the suppression chamber to meet the initial core cooling and heat dissipation demands, even though the main heat dissipation routes (the main condenser and hot-well, and the RllR heat exchangers) were assumed to be operable. This initial capability is continously supplemented by the heat removal capacity of the reactor water cleanup system, such that (depending on the initial water temperature in the suppression chamber and water storage tanks) by the time the suppression chamber is no longer available as a heat sink, the reactor water cleanup system could dissipate the heat load without assistance. If the initial water temperature is assumed to be 85 F or less (normally the water in the suppression chamber and condensate storage tanks is about 80 F) the calculations indicate that the combined heat dissipation is sufficient without the use of other heat dissipation capabilities. Under the conditions assumed, the suppression chamber with no other cooling system in service could function as the heat sink for approximately fifteen hours (900 minutes). No assumptions were made concerning the use of portable equipment, re-routing or installing cooling water lines, or other changes to improve the cooling capacity of the stated systems. (
O o 1 Introduction On June 9, 1972, the condensate booster pump room which is located directly below the main turbine condenser was flooded to a depth of 14.5 feet when a reinforced rubber expansion joint in a main circulating flow line broke. 1/2/3/ The flooding rendered a number of components and systems inoperable. Ilowever, since the reactor had been in a shutdown l status for nine days, the consequences of this occurrence were minimal. A question, and the purpose of this analysis, is, what would have been the consequences if the reactor had been operating at full power at the time of the occurrence? First, to define the problem, it is necessary to identify what components and system capability were lost. These, as they pertain to maintaining core water level and the dissipation of core heat are as follows: a. Four Reactor Heat Removal (RllR) service water pumps were submerged. Their loss removed the capability of the RIIR system heat exchangers to remove heat from either the primary system, its normal function during a shutdown period, or to remove heat from the suppression chamber, as it would during and following the operation of the emergency core cooling systems or containment suppression systems. b. Four condensate booster pumps were submerged. Their loss resulted in the loss of all feedwater flow to the reactor. Thus, normal reactor water level control was not available. The failed reinforced rubber expansion joint was physically c. located such that its failure completely drained the condenser hot well on the tube side. The failed expansion joint was not insolable (i.e., located directly below the hot well and above the first valve.) Thus, all circulating water pumps had to be shut down. This major heat dissipation system was not available. d. The circulating cooling water pumps for diesel generators No. 1 and No. 1/2 were submerged. Thus, neither of these emergency power diesels could have supplied power for any useful period before over heating and becoming inoperable. 1/ RO Inquiry Report No. 050-254/72-10. (6-12-72) 2_/ Directorate of Regulatory Operations Notification of an Incident or Occurrence. (6-12-72) 3/ RO Inspection Report No. 050-254/72-06. (7-12-72)
In summary, the most critical loss of capability had the reactor been operating at power at the time of the occurrence were: i a. The loss of normal reactor water level control - - - the feed-water system. b. The loss of the normal, major heat dissipation system - the condenser and hot well. c. The loss of the normal shut down heat removal system - the RIIR heat exchanger system. (Including the loss of the head cooling system) d. The loss of on-site emergency power. An examination of what alternate capabilities are available in each of these creas is as follows: a. Reactor Water Level Several means for maintaining the reactor water level above the core were available. They were: (1) The Reactor Core Isolation Cooling System (RCIC) The design purpose of this steam turbine driven systen is to provide cooling water to the reactor core in the event that the reactor becomes isolated from the main condenser simultaneously with a loss of the reactor feedwater system. The system is designed to supply 400 gpm of makeup water to the core over a reactor pressure range of 1135 psia to 165 psia. The RCIC system operates automatically to maintain the reactor water level between the low-low level (83 inches above the top of the fuel) and the high water level (193 inches above the top of the fuel). It may also be operated manually. Its normal water supply is from the condensate storage tank. The temperature of the water in this 100,000 gallon tank is normally maintained at about 800F. (2) The liigh Pressure Coolant Injection System (IIPCI) The design purpose for this steam turbine driven system is to provide water into the reactor vessel under loss of coolant conditions which do not result in rapid depressurization of the pressure vessel. The system is o o designed to deliver 5600 gpm over a range of 1135 psia to 165 psia and will automatically maintain the reactor water level between the low-low level and high level. It may also be operated manually. The normal source of water is the condensate storage tank. (100,000 gallons 0 /V 80 F) (3) The Low Pressure Coolant Inj ection (LPCI) Mode of the RHR System The purpose for this system is to provide water to the reactor vessel under loss of coolant conditions where rapid depressurization has occurred. Four 33% capacity electrical motor driven pumps provide 8000 gpm at 200 psia and up to 14,500 gpm at zero pressure. This system was available only on normal power (in this occurrence the diesel generators were inoperable). The normal water supply for this system is the suppression chamber. (4) The Core Spray System The purpose for this system is to provide a deluge spray over the core to provide cooling until the core could be reflooded during a postulated core uncovering accident. The system consists of two 100% subsystems which contain motor driven pumps each capable of supplying 4500 gpm at 90 psid. This system was available on normal power only. (The diesels were inoperable during this occurrence). Normal water supply is from the suppression chamber. (5) The Control Rod Drive Cooling Water The control rod drive hydraulic system, among other functions, provides water at a constant pressure of about 20 psi above the reactor pressure to supply cooling water to each control rod drive mechanism. This system provides a coolant flow of 80 to 120 gpm of water through the CRD mechanisms into the core at all reactor pressures. The normal supply of water is from the demineralized water storage tank. This system was available only on normal power (the diesels were inoperable during this occurrence.) r ~ O o (6) Availability of Water for Maintaining Reactor Pressure Vessel Level The makeup water system provides a main well water storage tank of 200,000 gallons of untreated well water as a supply for the condensate storage tanks. Directly in series with this tank is the demineralizer system, the 100,000 gallon cleanup condensate storage tank, and available as required are two 350,000 gallon capacity, potentially contaminated condensate storage tanks. Thus, immediately available for core cooling were up to 800,000 gallons of demineralized water, plus up to 200,000 gallons of well water which is ready for demineralization. All of the above systems were operabic, on normal power, at the time of this occurrence. b. Reactor Heat Dissipation Had the reactor been operating at full power at the time of this occurrence, the loss of capability to remove core heat by the normal heat removal routes (the condenser and the RHR heat exchanger) appears, potentiallv, to be the most serious aspect of the occurrence. The heat removal mechanisms which remained functional are: (1) Pressure Relief to the Suppression Chamber The five (5) reactor relief valves are sized to rapidly remove the generated steam flow upon the closure of the turbine stop valves coincident with the failure of the turbine bypass system thus precluding operation of the safety valves or overpressurization of the reactor vessel. In this particular instance, the relief valves provide the immediate automatic means of dissipating the core heat to the suppression pool. The total capacity of the five (5) relief valves is 3,038,000 pounds per hour (607,700 pounds per hour each) of steam. Each of these electromagnetic relief valves can be manually operated for long term pressure control of the primary system, as required. (2) Reactor Water Cleanup System The basic design purpose for this system is to remove fission and activated corrosion products from the primary system recirculating water; however, in this particular circumstance it has the capability to con-tribute a significant heat dissipation capability for L.
the recirculating system. The system is designed to continously remove a portion (102,584 pounds per hour) of the recirculating water from the suction line of one of the reactor recirculating pumps. This water is passed through three regenerative *and two non-regener-ative heat exchangers in the cleanup process, which reduces the primary water temperature to approximately 6 120 F by transferring the heat (about 42.5 x 10 BTU / hour) to the reactor building closed cooling water system. All systems and components required for this system to be functional were availabic (on normal power only) during this occurrence. Thus, this system had the capability to continously remove 102,548 pounds per hour of hot re-circulating water from the primary system, cool it to about 120 F, and return the coolant to the primary systgm. This coolant stream continously removes up to 42.5 x 10 BTU /hr of thermal energy from the primary system.** Conclusions This analysis assumes that other than the loss of the feedwater system, none of the remaining systems designed to supply water to the core were affected. Previous analyses, which are included in the FSAR indicate that sufficient water would be available over all pressure ranges to assure that the core would never become uncovered. Sufficient water capacity exists in the various condensate storage tanks and in the suppression chamber to meet the initial core cooling and heat dissipation demands, even though the main heat dissipation routes (the main condenser and hot-well, and the RHR heat exchangers) were not operable. This initial capability is continously supplemented by the heat removal capacity of the reactor water cleanup system, such that (depending on the initial water temperature in the suppression chamber and water storage tanks) (See Notes 2 and 3) by the time the suppression chamber is no longer available as a heat sink, the reactor water cleanup system could dissipate the heat load without assistance.
- The regenerate heat exchangers were not functional during this occurrence (feedwater system was shut down due to loss of booster pumps) thus, the clean up water would not be re-heated prior to return to the primary re-circulation system.
- The cleanup system is in a Group 3 isolation group which is initiated by a low reactor water level. For the purposes of this analysis it is assumed that the isolation is reset after the initial low level trip is I
reset and that the water level is manually maintained above the isolation l trip point. i
r nU, o The calculations indicate that the supression chamber alone can dissipate the core decay heat for up to 900 minutes, but that beyond that time6 some supplemental heat removal capability, on the order of 1.1 x 10 BTU per minute, is necessary to continue the cooldown of the primary system. Calculations 1. Total BTU Generated The total BTU of decay heat generated in the first 7000 minutes following a scram which terminates a 1000 hour run at 2511 Mwt is 8 calculated to be 66.16 x 10 BTU. This figure is derived by plotting the decay heat curve from data provided in the Point Beach Technical Specifications, Section 15.3.3 and calculating the total area under the curve. (See Attachment "A") The end point of 7000 minutes was chosen since at this point in time the reactor water cleanup flow which removes 708,000 BTU per minute continuously, is capabic of assuming the entire decay heat load. 2. BTU Removed by the Reactor Water Cleanup System This system continously removes 102,548 pounds per hour of water from the primary and cools it to 1200F. Thus, in this application this system will continuously remove: 102,548 pounds / hour (544-120) x 414 F cooling f-60 minutes / hour = rs708,000 BTU / minute. 3. Suppression Chamber Heat Removal Capability 3 Initially the suppression chamber contains about 112,203 ft of water (7,021,467.0 pounds). This represents an initial capability to dissipate, by direct relief valve blowdown to the suppression pool, 705,657,433.0 BTU of heat (assumes initial temperature of suppression pool water is 950F and ending temperature is 200 F). Secondly, as identified earlier, there is normally up to 800,000 gallons of de-mineralized water, plus up to 200,000 gallons of well water, ready to supply the emergency core cooling system requirements. Thus, an additional heat dissipation capacity of approximately 722,324,567 BTU is available to the suppression chamber. (This assumes that a free volume of 10,000 ft3 for expanison remains,4/ and that the 4/ Suppression Chamber Capacities from FSAR Section 5.2.3 Water Volume 112,203 cubic feet Free Air Volume 117,245 cubic feet Total Volume 229,448 cubic feet /_ O o initial temperature of the water used is 95 F, with an ending temperature in the suppression chamber of 200 F). In total, these calculations indicate that the suppression chamber has the capability of absorbing approximately 1,438,000,000 BTU of heat from the reactor core. The suppression chamber alone can absorb the core decay heat for the initial 900 minutes of the occurrence as follows: Area under curve for 900 minutes = 24,438 Mw minutes 24,438 Mw-min x 3.413 x 103 BTU /Kw hour 60 min /hcur = 1,392,966,000 BTU If it is assumed that the suppression chamber and condensate storage tank water temperature is initially 85 F, rather than 95 F (maximum allowable by Technical Specifications). Then the suppression chamber alone can dissipate the following heat: 13,693,680 pounds of water x 115 F temperature increase = 1,574,733,200 BTU From the Calculations Table (Attachment B) it can be seen that this would permit the suppression chamber alone to dissipate the generated decay heat for 1000 minutes. Attachment "B" also shows, using the 85 F initial water temperature assumption, that the combined heat dissipation capability of the suppression chamber and the reactor water clean up system is sufficient to accommodate the entire decay heat load. (Af ter approximately 7000 minutes, the suppression chamber is exhausted as a heat sink; however, at this time (7000 minutes) the reactor water clean up system can dissipate the decay heat load). Note No. 1 This analysis sheet indicates that, in 900 minutes, the suppression chamber, if unaided, will be exhausted as a heat sink (water temperature will have reached 200 F with 13.6937 x 106 pounds of water in the suppression chamber -- an end point chosen for this analysis). At this time some other heat dissipation capability, of approximately 1.1 x 106 BTU per minute, must be provided to dissipate the decay heat being generated. At 900 minutes the total decay heat generated is 1,410,600,000 BTU, the suppression chamber is exhausted at 1,438,000,000 BTU..
/ O O ~ Note No._2 This analysis sheet indicates that in 4000 minutes the suppression chamber, if aided continuously to the full capacity of the reactor water clean up systemheatdissipationcapabilgty,becomesexhausted(waterwillhave reached 200 F with 13.6937 x 10 pounds of water in the suppression chamber). At this time some other heat dissipation capability is required to supple-ment the reactor water cleanup system. An additional heat removal capability of.1 x 106 BTU per minute is required to dissipate the decay heat being generated. Note No. 3 These calculations indicate that depending upon the assumptions made regarding the initial temperature in the suppression chamber and in the condensate storage tanks, the combined heat dissipation capabilities of the suppression chamber and the reactor water clean up system can provide sufficient heat removal such that when the suppression chamber is exhausted the reactor water clean up system can dissipate the decay heat load by itself. If it is assumed that this initial water temperature is 95 F (the maximum temperature allowable by Technical Specification) then the combined heat dissipation capability is deficient after 4000 minutes, as stated in Note No. 2. However, if it is assumed that the initial water temperature is 850F (normally this water temperature is 80 F) then these calculations show that the combined heat dissipation of these two systems is sufficient to remove all of the decay heat without any other heat dissipation capability being required.
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