ML20138L407
| ML20138L407 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/10/1985 |
| From: | Collette P, Geaney G PROTO-POWER MANAGEMENT CORP. |
| To: | |
| Shared Package | |
| ML20138B522 | List: |
| References | |
| EE-EQ-0019, EE-EQ-0019-RB, EE-EQ-19, EE-EQ-19-RB, TAC-59787, NUDOCS 8512190263 | |
| Download: ML20138L407 (74) | |
Text
{{#Wiki_filter:- - _ - . _ - - - 3 Public FORT ST. VRAIN NUCLEAR GENER ATING STATION Service'" PUBLIC SERVICE COMPAN # OF COLORADO d g ENGINEERING EVALUATION OF LINER COOLING WITH FIRE WATER FOLLOWING A HIGH ENERGY LINE BREAK
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ENGINEERING EVALUATION OF LINER COOLING WITH FIRE WATER FOLLOWING A BIGH ENERGY LINE BREAK FROM 35 PERCENT POWER OPERATION AT FORT ST. VRAIN EE-EQ-0019 REV B , o PREPARED BY PROTO-POWER CORPORATION 591 POQUONNOCK ROAD GROTON, CT 06340 DECEMBER 10, 1985 h._.
i. INDEX Page 1.0 Purpose 1 2.0 Summary 1 3.0 Scope 3 4.0 Approach 3 5.0 Evaluation 4 5.1 Background 4 5.2 Required Systems and Components 5 5.2.1 Primary Coolant Pressure Boundary 5 Isolation 5.2.2 PCRV Liner Cooling 5 5.2.3 Auxiliary Systems and Components 6 r 5.2.3.1 Circulating Water System 6 5.2.3.2 ACM Diesel Generator 7 5.3 Flow Adequacy and Control 7 5.3.1 Fire Water System Adequacy 7 5.3.2 Review of PCRV Liner Cooling Af ter 7 Flow Interruption 5.3.3 PCRV Liner Cooling 8 5.4 Operator Response Time 11 5.5 . Building Habitability 12 6.0 Conclusions 12 7.0 References 13 PCRV Liner Cooling Flowpath Figure 1 Brake and Seal Actuation Figure 2 , Electrical Items Located in Harsh Environment Table 1 Manual Actions Necessary to Initiate Liner Cooling Table 2 Summary of Manual Action Table 3 Highlighted P& ids for Liner Cooling Appendix A PCRV Depressurization Appendix B Reserve Shutdown System Actuation Appendix C Fuel Storage Facility Cooling Appendix D l
7 EE-EQ-0019 l REV B ENGINEERING EVALUATION OF LINER COOLING WITH FIRE WATER FOLLOWING A HIGH ENERGY LINE BREAK FROM 35 PERCENT POWER OPERATION AT FORT ST. VRAIN 1.0 PURPOSE This engineering evaluation was performed to document the acceptability of Prestressed Concrete Reactor Vessel (PCRV) liner cooling with fire water as an interim method of safely removing decay heat from the Fort St. Vrain reactor core following a high energy line break. This engineering evaluation also documents the method that will be utilized to actuate the helium circulator seal in order to maintain primary coolant boundary integrity. It is also the purpose - of this evaluation to identify the specific systems and components that would be utilized for liner cooling and for setting of the circulator seal and to confirm that no electrical components located in a harsh environment must function. This evaluation presumes that the Fort St. Vrain reactor has operated at power levels no greater than 35 percent of rated power since 1984. Liner cooling utilizing tire water is being considered as an interim option for safely removing core decay heat in order to eliminate reliance on electrical components in a harsh environment during the ongoing review of equipment qualification at Fort St. Vrain. Reactor power operation will be limited to 35 percent of rated in order to ensure that no fuel damage will result in the unlikely event that a total and permanent loss of forced circulation does occur following a high energy line break. 2.0
SUMMARY
The analysis of Reference 7.2 has shown that if the. Fort St. Vrain reactor has not been operated at power levels in excess of 35 percent of rated power, a permanent Loss of Forced Circulation (LOFC) can be tolerated without fuel damage resulting. This analysis also show.S that depressuri-zation of the PCRV is not necessary to limit heat transport from the active core and that liner cooling will still effectively'and safely remove core decay heat. Furthermore, this analysis concludes that actuation of the reserve shutdown system is not required to ensure reactivity shutdown margin. For the purposes of this evaluation, no credit is taken for any of the multiple and redundant methods of removing core decay heat utilizing forced circulation. (These. methods of Page 1
l EE-EO-0019 REV B torced circulation are identified in the Reference 7.1 FSAR) This evaluation also presumes that the reactor has not operated at greater than 35 percent rated power since 1984. This latter condition ensures that a total loss of forced circulation can be tolerated with no fuel damage to the Fort St. Vrain reactor core. Following a high energy line break, the fire water system, in lieu of the normally operating reactor plant cooling water system, would be utilized,for cooling the PCRV liner. Utilizing the fire water system eliminates reliance on electrical equipment in a harsh environment and also ensures that the intent of 10CFR 50.49 is satisfied during the interim. This method also provides full redundancy as only one of two fire water pumps, and PCRV liner cooling loops, are required for core cooling. Also, operating procedures are currently in place at Fort St. Vrain for PCRV liner cooling utilizing fire water (Section III, Option C of " Safe Shutdown Cooling Under Highly Degraded Conditions" and Section 5.3 of SOP 46-
" Reactor Plant Cooling Water System").
The simplified flow path for liner cooling is shown in Figure 1. These flow paths are established by positioning of manually operated valves or by utilizing manual overrides of air and motor operated valves. A Change Notice is currently being implemented to provide for setting the helium circulators brake and seal without reliance on electrical equipment items. A simplified diagram reflecting these expected system changes is shown in Figure 2. Systems that are required to be operable to perform <these two functions and to accomplish safe shutdown cooling are: SYSTEM OPERATION Brake and Seal System Provide Path for High Pressure Helium to Actuate Circulator Brake and Seal Fire Water SyLiem Cooling for PCRV Liner Reactor Plant Cooling Provide flow path for Water System fire water Circulating Water System Provide Inventory of Including Make-up Cooling Water Electric Power Power for Electric Motor Driven Fire Water Pump Power for Circulating Water Make-up Pumps Page 2
l EE-EO-0019 REV B Electrical equipment items that are required to function are not located in the harsh environment (e.g., circulating water make-up pumps, fire water pumps and associated instru-mentation). Although these two functions (liner cooling with fire water and setting of the helium circulator seals) do not rely on the operability of electrical equipment items in a harsh environment, some cables that are required to remain functional are located in a harsh environment. These i cables are identified in Table 1. This evaluation concludes that fire water is acceptable for liner cooling following a steam line break. 3.0 SCOPE This evaluation presumes that the reactor has not operated at a power level in excess of 35 percent of rated capa' city. Therefore, a total loss of forced circulation can be tolerated without fuel damage resulting. It is also presumed that the reactor has been scrammed, either auto-matica11y or manually, and that the steam flow into the building has terminated. This evaluation considers the specific flowpath, eystems and components that would be utilized for PCRV liner cooling with fire water and setting of the helium circulators seals following a high energy line break. Also considered in the evaluation are flow adequacy and control while utilizing the fire water pumps. Specific recommendations for implemen-tation of these method are provided under the conclusions section. In the event that depressurization of the PCRV, actuation of the reserve shutdown system, or Fuel Storage Facility coaling is desirable, these functions can be accomplished without reliance on electrical equipment items as distassed in Appendices B, C and D, respectively. 4.0 APPROACH The overall objective was to demonstrate the ability to safely recover from a high energy line break while elimi-nating, or at least minimizing, reliance on electrical equipment located in a harsh environment. It was concluded that the forced circulation method of removing core decay heat is not consistent with this objective. An analysis (Reference 7.2) was performed to determine the maximum power level at which Fort St. Vrain could be operated without fuel damage resulting following a total and permanent loss of forced circulation. This analysis Page 3
EE-EQ-0019 REV B concluded that at a power level of 35 percent of reted " capacity, up to 29.4 hours is available after LOFC to initiate liner cooling. This analysis concluded further that PCRV depressurization is not required for shutdown and cooling following a loss of forced circulation at 35 percent power operation. Additionally, this analysis concludes that the reserve shutdown system does not need to be actuated for these postulated conditions. The specific flowpath for liner cooling following a high energy line break was developed based on the following approach: Utilizing existing liner cooling inethods, systems and procedures to the maximum extent possible Eliminating reliance on electrical components in a harsh environment With the flow paths identified, all electrical components that are required to function were tabulated. Cables for these components and subtiers were obtained from the ongoing plant audit and plant electrical schematics. The listing of these cables which are also located in a harsh environment, is provided in Table 1. This evaluation does not require the reactor building louvers or HVAC to function either during or after the steam line rupture. Since no fuel damage will result, release of primary coolant via the reactor building louvers or via a rupture in the reactor building will not exceed the offsite doses already analyzed by Design Basis Accident No. 2, Rapid Depressurization. 5.0 EVALUATION S.1 Background Section 14.10 and Appendix D of the FSAR address the liner cooling method for heat removal from the reactor core in the event of a complete loss of forced circu-lation following full power operation (Design Basis Accident No. 1). Primarily, these sections evaluate the use of the normally operating reactor plant cooling water system to accomplish liner cooling. The original design of Fort St. Vrain considered the location east of 4A wall to be a mild environment, i.e. not subject to a high energy line break. This is the area in which the reactor plant cooling water pumps and much of the reactor plant cooling water system instru-mentation is located. As such, 't h e environmental Page 4
EE-EO-0019 REV B qualification requirements for this equipment and instrumentation were much less stringent than for electrical components located in a harsh environment. PSC's current re-evaluation of high energy line breaks for Fort St. Vrain has concluded that the area east of the 4A wall is potentially a harsh environment. Therefore, the alternate method of utilizing fire water for PCRV liner cooling was pursued for the interim period since it would not be feasible to qualify.the reactor plant cooling water. system in the time avail- 1 able. The major advantage of fire water for PCRV liner I cooling is the elimination of electrical components in a harsh environment that are required to function. Although fire water backup to the normal reactor plant cooling water system means of accomplishing PCRV liner cooling is mentioned in the FSAR, it*is not addressed in detail. In Reference 7.3, the plant designer has provided some supplemental information on liner cooling with fire water. The following paragraphs are intended to demonstrate further that liner cooling with fire water can be used as an alternate to the normal reactor plant cooling system for removing core decay heat. 5.2 Required Systems and Components 5.2.1 Primary Coolant Pressure Bcundary Isolation Presently, two methods are provided to isolate the primary coolant boundary when the circu-lators are shut down. Normally, the primary coolant boundary is established by either actuating the circulator break and seal system or by flooding the circulators with bearing water to establish a water seal. However, each of these existing systems- is dependent on operation of electrical equipment. In order to minimize reliance on electrical components during interim operation, a design change has been initiated to allow the brake and seal system to be manually set from outside the reactor building (i.e. from a mild environment). 5.2.2 PCRV Liner Cooling Either the diesel driven or electric motor driven fire water pump will be utilized for cooling of the PCRV liner for decay heat removal. The fire water supply will be from the circulating water storage ponds. This water is delivered to the fire water pump nuction via one of the two safety related circulating water Page 5
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EE-EQ-0019 , { , REV B l make-up pumps. Fire water flows will be admitted to one of the two reactor plant cooling < L water system (System 46) loops downstream of System 46 pumps and heat exchangers. Fire water from the-PCRV liner will-be returned to
. the circulating water system return lines where i' it will be discharged via the blowdown line to
!- the yard drain. During onc'e-through cooling, j the blowdown line flow should be aligned to t discharge only return flow. A simplified flow path for liner cooling is shown in Figure 1. Marked up P& ids highlighting the intended flow l- path are provided as Attachment A. In the l Reference 7.2 analysis, GA Technologies con-l cludes that liner cooling must be initiated
- within 29.4 hours following loss of forced circulation. However, liner cooling should be
+ initiated as soon as practical after Loss of Forced Circulation (LOFC). L ! The 29.4 hour maximum interruption in PCRV-cooling is the time available to restore cooling without jeopardizing the line. cooling system or i PCRV integrity, and is based on ' power operat' ion at' or be-low 35% of rated capacity. The Reference 7.2 analysis also concludes that
; depressurization of the PCRV and~ actuation.of i the reserve shutdown system are not required i following a LOFC accident during interim l- operation at or below 35% power.
5.2.3 Auxiliary Systems and Components I 5.2.3.1 Circulating Water System Operation of one of the two safety ; related' circulating water make-up pumps and associated piping is required to ensure an adequate-source of make-up will be available'at fire water pump
- l. + for use during liner cooling. The "
L normal make-up flow of 5500 gpm to the ! circulating water tower basin is more than adequate for: cooling of the PCRV-liner. _The minimum circulating water 7 i storage pond capacity of 20 million gallons is sufficient to supply the
- required flowrate for eight days, which l
is considered adequate time to re-
- establish make-up flow from the river l water pumps to the storage pond.
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EE-EC-0019 REV B 5.2.3.2 ACM Diesel Generator The Alternate Cooling Method (ACM) provides an alternate source of power when the normal and emergency power systems are not available. The ACM will provide power for the circulating water make-up pump and, if necessary, for the motor driven fire water pump. 5.3 Flow Adequacy and Control 5.3.1 Fire Water System Adequacy The safety analysis for the Design Basis Accident No. 1 (DBA No. 1) presented in Appendix D of the PSAR, was based on PCRV Liner cooling using one loop of-the reactor plant cooling water system. Although the PSAR also discusses the use of fire water for cooling during DBA No. 1, specific details of fire water cooling are not provided. To verify the adequacy of the fire water system for liner cooling, a detailed flow analysis has been performed. The analysis (Reference 7.4) concludes that the fire water system flows and pressures will provide the same level of adequacy as does the reactor plant cooling water system. 5.3.2 Review of PCRV Liner Cooling after Flow Interruption As part of the Design Basis Accident No. 1 (DBA No. 1) Analysis, Appendix D.2.3 of the PSAR provides a detailed evaluation of the effect of a 30 hr interruption of PCRV liner cooling water flows after a LOFC and concludes that the resulting effects on the PCRV liner or support floor casing cooling do not affect the core or reflectors significantly, and will not result in severe thermal damage to the liner and cooling system. The effects of coolant flow restoration are also discussed in Reference 7.7. Since similar overall criteria applies to PCRV integrity following the LOFC at 35% power, the PSAR and the Reference 7.7 evaluations are considered applicable to this evaluation. At 30 hours after LOFC, the calculated maximum cooling tube-concrete interface temperature reported in the PSAR is 1500*F at the PCRV top head centerline above the core. The 30 hour cooling flow interruption is based on the time Page 7
EE-EQ-0019 REV B i required to reach a 1500* temperature, which, per paragraph D.2.3.2 of Reference 7.1, is a conservative estimate of the maximum liner p cooling tube temperature at which cooling water I could be re-established without . endangering the integrity of the cooling tubes by rapid steam flashing. FSAR Figure D.2-9 indicates that maximum top head and side wall liner temperatures at 30 hours of 1500*F and 750*F, respectively, rapidly decrease af ter initiation of cooling water flow. The peak liner temperature is quenched to 650*F and 300*F at the top head and side walls, respectively, as a result of restoration of cooling water flow. FSAR Figure ' D.2-10 show that the liner-concrete interface temperature drops from 1500*F to 250*F at two hours af ter restoration of cooling' water flow. During the rapid cooldown, fire water admitted to the open, once through cooling system will flash to steam in the cooling tubes, and be discharged via the cooling tower blowdown line. As the fire water flow continues the location at which the coolant flashes to steam would travel downstream as the tube temperatures decrease. Eventually, the entire tube will be cooled below the saturation temperature of the cooling water, and the steam flashing will terminate. The once-through, open system liner cooling using fire water which will be utilized during liner cooling will enhance steam quenching of the liner. Since the vent path is open to the atmosphere, steam pressure will not build up in the tubes, and the pump discharge head will continue to feed the tubes to provide cooling water. The inability of the open system to contain the steam prevents pressure buildup, which could otherwise cause vapor binding of the cooling system. 5.3.3 PCRV Liner Cooling The DBA No. I accident analysis presented in Reference 7.1 reports a heat load of 10.3 x 106 BTU /hr from the PCRV 2 hours af ter restoration of liner cooling. Page 8
EE-EQ-0019 REV B l The Reference 7.4 flow analysis has confirmed f that the fire water pumps can deliver the redistributed liner cooling flow rates listed in Reactor Plant Cooling Water System Operating Procedure (Reference 7.5). The flow through the I top head and upper barrel during LOFC cooling will be 850 gpm, with a 109 psia inlet pressure and a minimum tube pressure of 25 psia. With a worst case minimum top head tube outlet pressure of 25 psia, a temperature of 240*F can be tolerated without bulk boiling. This I corresponds to a maximum heat flux of approxi-mately 24,000 BTU /hr-FT2 at which boiling will occur. In Reference 7.2, GA reports a peak top head heat flux of 14,460 BTU /hr-FT2 occurring 36 hours after LOFC. Thus, conservatively assuming the maximum heat flux occurred over the entire tube, only 60% of the heat flux necessary to raise the cooling water to its saturation temperature would be transferred to the fire water. Additionally, GA reports an average side wall heat load of 486 BTU /hr-FT2 occurs after restoration of liner cooling. A conservative top head and upper barrel liner outlet tempera-ture may be calculated by applying the maximum top head heat flux and average side wall heat flux to their respective areas. To be con-sistent with GA, areas of 587 FT2 and 3370 FT2 for the top head and side walls, respectively, yield a maximum heat load of 10.1 X 106 BTU /hr. This yields a 25*F temperature rise across the PCRV top head and sidewalls, and an outlet temperature of 110*F, based on 85*F inlet temperature. It is noted that the 10.1 X 106 BTU /hr is close to the 10.3 X 106 BTU /hr value reported in Appendix D.2 as the heat load occurring 2 hours af ter a 30 hour interruption of liner cooling. In the DBA No. I analysis, the PCRV adequacy is based on heat removal from the above regions only, with reduced cooling water flow to the bottom head, bottom head penetrations and PCRV side walls below the core support floor. The PSAR reports that heat transfer from these regions will be negligible during the accident. Page 9
EE-EO-0019 REV B a During liner cooling, flows through the PCRV cooling tubes are " redistributed". The redis-tributed flows provide increased cooling water flow to the top head and upper barrel tubes where the heat flux is the greatest, and decreased flows to the lower barrel and core support floor tubes. Redistribution, when fire water is utilized for liner cooling, may be accomplished using manual overrides on six , System 46 valves (3 per loop). SOP 48 provides I a procedure for manually adjusting valve positions for flow redistribution. The temperature of fire water exiting the PCRV liner can be monitored at local temperature indicators at the outlet subheaders (TI-46187 typical). To preclude potentially incorrect temperatures during liner cooling, spare indicators should be kept on hand. Addi-tionally, a hand held surface pyrometer could be used as backup to monitor individual tube and/or subheader temperatures. Outlet pressure would be monitored locally at PI-46396 or PI-46397 at the System 46 outlet headers. By monitoring outlet pressures and tenperatures, flows could be manually adj usted, if necessary, to optimize cooling. However, since the pressure indicators are rated for a temperature of 150*F, their accuracy after exposure to a steam line rupture may be suspect. If necessary, temporary test gauges from the Results Department may be used to monitor the PCRV coling water outlet pressure. During liner cooling, the fire water system flows must be throttled to maintain adequate pressures in top head tubes to prevent boiling. A pressure drop by throttling of approximately 35 psi is required to obtain the pressures and flows calculated in Reference 7.4. Reference 7.6 specifies that valves V-46121 and V-46122 be used for throttling. However, the pressure at the valves, which are located at elevation 4884, will be near atmospheric pressure. Hence, a valve closer to grade level, such as HV-4221-1, which has a manual overide, will be used. Valves in the blowdown line are less desirable for throttling because they are located downstream of a section of 84" diameter circulator water piping. This section of large Page 10
L [ EE-EQ-0019 L REV A diameter piping will dampen pressure fluc-tuations caused by throttling, and could increase dif ficulty in maintaining the desired outlet pressure. Pressure switches (PSH-46225 and PSH-46226) on the cooling tube outlet headers are normally used to automatically isolate cooling water flow on high (140 psig) pressure in the tubes, which is indicative of a cooling tube crack by closing t subheader inlet and outlet isolation valves. Since the PCRV will be pressurized during liner cooling, a high pressure condition could initiate closing the System 46 valves. These subheader inlet and outlet isolation valves are also designed to close on lov surge tank level or on a low tube flow rate when the liner cooling system is redistributed remote-manually from the control room. However, since the System 46 valves will be manually redistributed rather than using the handswitch in the control room, liner cooling will not be interrupted by a low tank level or low tube flowrate. In any event, all of the subheader valves are air-to-close and may be manually opened by isolating the air supply, as necessary. The total fire water flow to the PCRV following LOFC will be approximately 1500 gpm, yielding a bulk outlet temperature of only 100*F. The low bulk temperature is low enough to preclude flashing of fire water return flow at either the blowdown line discharge or at the valve used to throttle the fire water flow. 5.4 operator Response Time Various manual operations must be performed to set the circulators brake and seal and to establish PCRV liner cooling. A Change Notice is currently being prepared to allow the helium circulator brake and seals to be set by performing several manual valve operations. It is anticipated that these operations will be performed from the Helium Storage Facility or from some other mild environment, and 16 manual operations will be required. It is conservatively estimated that all operations associated with manually setting the brake and seal can be accomplished within one hour following the high energy line break. Table 2 lists the manual actions required to accomplish liner cooling. The Reference 7.2 analysis states that y 29.4 hours is available following the high energy line Page 11
l EE-EQ-0019 L REV A break to perform these operations. The manual actions were classitied as either Class (1) o'r Class (2). Class
- (1) actions must be performed. Class (2) actions are desirable but not necessary to accomplish liner cooling. Class (2) actions result in isolation of fire water flow to non essential users thereby maximizing the time to deplete the inventory from the storage ponds. Table 3 summarizes the sequence of the l functions associated with liner cooling, the total number of manual actions needed to perform each tunction, and the amount of time available to perform each of the functions. Fourteen Class I actions are required to establish liner cooling. Of these, one action is performed in the turbine building, eleven are in the reactor building, and two are performed from the control room.
S '. 5 Building Habitability To ensure that the steam environment following a steam line break does not preclude building accessibility for prolonged periods, cool suits are being procured. These suits will allow building access in ambient tempera-tures of up to 180*F for periods of 45 minutes. The worst case steam line ruptures at 35% power yield temperatures'of 180*F and 140*F within 4 hours and 12
, hours after the rupture, respectively, in both the reactor and turbine buildings, with no credit taken for operation of the respective ventilation systems. ,
If bearing water is lost prior to setting the circu-lator seals, primary coolant can leak through the helium circulator labyrinth seals and be discharged into the reactor building, via the low pressure separator relief valves. It is conservatively estimated that the manual actions required to set the circulator
~ seals can be accomplished within one hour following the high energy line break. Per Reference 7.2, it is estimated that 1250 pounds of primary coolant (approxi-mately 17 percent of total PCRV inventory) can escape into the reactor building. This would result in an estimated dose to personnel entering the building of
.55 rem /hr 12 hours after LOFC, or .13 rem /hr 24 hours after LOFC.
6.0 Conclusion on the basis of this evaluation, it is concluded that decay heat removal from the Fort St. Vrain reactor core can be accomplished using liner cooling, with fire water, following a high energy line break in either the reactor or turbine building while at 35 percent power. Page 12
EE-EQ-0019 REV B As a result of this evaluation, the following recommen-dations are intended to identify specific actions which should be accomplished in order to implement the liner cooling method. Prepare detailed procedures to be used by plant operators to initiate and maintain liner cooling using the fire water system.
- Perform a plant walkdown to ensure that all manual valves required for liner cooling are accessible and to verify all manual operators on valves normally remotely operated are functional. .
- Perform a Mock Run to record actual operator response time required to perform all functions necessary to initiate liner cooling using fire water to demonstrate viability of manual operations within the time required.
7.0 References 7.1 Fort St. Vrain Updated Final Safety Analysis Report (FSAR) 7.2 Attachment A to PSC letter No. P-85460, " Confirmatory Action in Support of 35 Percent Power Restriction During EQ Schedule Extension Period", dated December, 1985. 7.3 General Atomic Correspondence ME:WSB:215:85, dated September 24, 1985. 7.4 Proto-Power Corporation Calculation No. 63-01, "FSV Fire Water System Flow Analysis During Depressurization and PCRV Liner Cooling." October, 1985. 7.5 SOP 46: Reactor Plant Cooling Water System 7.6 Fort St. Vrain Procedure " Safe Shutdown Under Highly Degraded Conditions" 7.7 General Atomic Letter No. RIC:253:76/ SED: RIB:67:76, Dated June 25, 1976, "PSC Transient Coolant Flow and Thermal Response of Upper Barrel PCRV Liner to Coolant Flow Restart After a 48 Hour LOFC Accident." O
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EE-EQ-0019 REV. B TABLE 1 CABLES LOCATED IN A BARSB ENVIRONMENT SCHEM. ELECTRICAL E-1203 FUNCTION COMPONENTS POWER SUPPLY PAGE CABLES Cire. Water P4118 480V SWGR Bus No. 3 P.73 3480 Make-up Pump (Mild) 3481 Cire. Water P4118S 480V SWGR Bus No. 1 P.74 3410 Make-up Pump (Mild) 3412 3413 Fire Water P4501 480V SWGR Bus No. 1 P.78 7262 Pump Circuit (Mild) 7263 Interface 7264 Components 7265 in Harsh Area 7289 CR-4503,4504, 25083 4505, 4508, 26520 4509, 4514 26528
- Fire Water P4501S Diesel Driven Pg1664 7024 Pump P.1665 7070 P.1666 7071
, 7266 7268 7269 7281 7288 O
EE-EQ-0019 REV. B TABE 2 i MMKIAL ACTIONS NECESSARY TO INITIATE LINER 000 LING Class Incation li Function Manual Action Description 1 2 (Note 5)l Align Start P-4501 or Fire Water Pump 1 CR Fira Water Start P-4501S Dnergency Fire Water Pump 2 CR to Users Close V-4571 To Isolate T-4501 2 ACB Close V-4576 (4) Isolate Headers as Needed 2 OtTr Close V-45116 (4) to Limit Pressure Boundary 2 OlTP Close V-45109 (4) 2 is Close V-4542 (4) 2 IB Close V-4553 (4) 2 is Close V-4589 (4) 2 TB Close V-4541 (4) 2 7B Close V-4552 (4) 2 is Close V-45123 (4) 2 RB Close V-45124 (4) 2 RB Close V-45130 (4) 2 RB Start P-4118 or Circ. Water Make-up Pump 1 CR Start P-4118S or Cire. Water Make-up Punp 2 CR
, Start P-4118SX Circ. Water Make-up Pump 2 CR 7b Return Close V-46123 (46124) Normal Return to Surge Tanks 2 RB Firo Water Open V-46121 (46122) Return thru Service Water 1 RB to Cooling Close V-461633-P Tb E-2101/E-2106 1 RB Tbwer Close V-45895-P To Control Station 3,4,5,6 2 RB Open V-46166 Alt. Tie to Circ. Water Return 2 RB Clom V-46612 Serv Water Return frm S-4602,T-2303 2 RB Close V-42398 Serv Water Return fr m HVAC 2 TB Close V-42102 Serv Water Return frm E-4202 2 1B Close HV-4221-3 (3) Serv Water Ieturn 2 7B Open HV-4221-1 (3,6) Serv Water Tie to Cire Water Return 1 is Close HV-4138-1 (3) Cire. Water Return frm Condenser 2 TB Close HV-4138-2 (3) Cire. Water Return frm condenser 2 is O
Page 1
EE-ECH)019 REV. B TABLE 2 Class Incation! Function Manual Action Description 1 2 (Note 5)l PCW Open V-46119 (46120) Fire Water to KRV Top 1 RB Di"tributed Open V-4614-% (4614-94) Penetrations 1 RB Liner Crack Open V-46103 (46104) Fire Water to K RV Bott e Head 1 RB Cooling Open V-4614-98 (461500) and Penetrations 1 RB Open V-4691 (4692) Fire Water to PCRV Top Head, 1 RB Open V-4614-90 (4614-92) Barrel, Core Support Floor 1 RB l Open HV-46227 (46228) (3) Upper Barrel & Top Head 1 RB Set the following valves to red line on position indicator dial HV-46229 (46230) (3) Iower Barrel 1 RB HV-46231 (46232) (3) Oore Support Floor 1 RB Class 1: Action necessary to accomplish stated function. Class 2: Action desirable, but not necessary, to accmplish stated function. (i.e., isolation of non-essential users of fire water, isolation of lines fr a other systems, etc.) Not3 3: Valves referencing this note have manual overrides and may be operated manually if necessary. Not') 4: Valves referencing this note isolate automatic fire suppression systems.
'Ibese valves need be operated only if the stean environment results in inadvertant activation of the autmatic suppression system.
Nots 5: Incation are abbreviated as follows: RB Reactor Building
'IB Turbine Building ACB Access Bay O(7f Outside Not'a 6: V-41278 which is located outside may be used as an alternate for HV-4221-1.
v Page 2
EE-EQ-0019 REV. B i TABLE 3
SUMMARY
OF MANUAL ACTIONS BY FUNCTION Number of Manual Actions Frquence Function Time Allowed Necessary Desirable 1 Manually Set I hour 16 0 Circulator Brake and Seal 2a Align Fire Water to Within 29.4 hrs. 2 15 (Note 1) PCRV Liner 2b Align Fire Water Within 29.4 hrs. 3 9 Return to Main Cooling Tower 2c Redistribute Liner Within 29.4 hrs. 9 0 Cooling NOTES:
- 1) Eleven of these fifteen actions isolate autodatic fire suppression systems. These valves need be operated only if the steam environ-mental results in inadvertent actuation of one or more automatic suppression systems.
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, REV. B A_PPENDIX B INITIATION OF PCRV DEPRESSURIZATION Initiation of depressurization will limit free convection heat transport from the active core to the upper plenum.
One train of the helium purification system is required for depressurization of the PCRV. Components that are required to function are: 1) High Temperature Filter Adsorbers (HTFAs), 2) Helium Purification Coolers (HPCs), and 3) Low Temperature Adsorbers (LTAs). The liquid nitrogen system is required to provide cooling to the low temperature adsorbers during depressurization. Fire water would be utilized for cooling of the on-line high temperature filter adsorber (HTFA) and the on-line helium purification cooler (HPC) during depressurization of the PCRV. Spoolpieces are being permanently installed to reduce the time to complete the manual actions necessary to align fire water to the HTFAs. A design modification to remove the internals from check valves V-46171 and V-46172 is necessary in order to use fire water to cool the High Temperature Filter Adsorbers. The fire water supply will be from the' circulating water storage ponds. This water is delivered,to the fire water pump suction via a circulating water make-up pump. A simplified flow path for cooling of components necessary for depressurization is shown in Figure Bl. A simplified deprissurization flow path is shown in Figure B2. Marked up P& ids highlighting the intended flow path are provided. Fire water from the HTFA would be returned to the circulating water system return lines where it will be discharged via the blowdown line to the yard drain. During once-through cooling, the blowdown line flow should be aligned to discharge only return flow. This evaluation does not take credit for the reactor building exhaust fans. Since depressurization is through the purification system, this is comparable to normal PCRV pump down. In which case, the radiation levels found in the helium will not be harmful to personnel. As the exhaust fans would not be operable the purification system helium will discharge to the reactor building ventilation. No electrical equipment is needed to initiate depressurization. 5 Page 1
r b EE-EO-0019 j E REV. B h APPENDIX B I L Flow Adequacy and Control f HPC Cooling f The helium purification coolers are normally cooled via the H s closed loop purification cooling water system with a 75.5 gpm 3 flow. During depressurization, the HPC will be cooled by fire b j water. The flow analysis has confirmed that the fire water i l system can provide the design flow rate.
! It is important that the dual drain path from the on-line HPC, as P I shown on the attached P&I diagrams, be maintained during de- d } pressurization, as the parallel path drain is required to obtain 1 5 the required 75.5 gpm flow. The maximum flow through the 1" i ! drain lines, without throttling, is approximately 84 gpm.
y ( During the 8 hour depressurization period, the approximately k 40,000 gallons of 115'F fire water will drain into the reactor building sump, causing a level rise approximately 3 feet in the
$ 27 foot deep keyway.
y
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f During depressurization, HTFA Cooling will be accomplished using 1 fire water. (This flowpath requires that a Change Notice be y( prepared and issued to remove the internals of check valves i V-46171 and V-46172.) The fire water system flow analysis has
- demonstrated that the inlet pressure at the HTFA will exceed the
[ normal System 46 pressure, and will provide the 40 gpm flow e g required during depressurization. The fire water flow will h p return to the cooling tower blowdown line via the System 46 tie l-L_ to the circulating water return header. Throttling of the return flow is required to maintain high enough pressure to provide j L required flows to the HPC. As discussed in paragraph 5.3.3, the i flow must be throttled using a valve at or near grade level, to maintain a PCRV outlet pressure of approximately 12 psia as read , on PI-46396 or PI-46397. = l 1 E LTA Cooling During depressurizatior. of the PCRV, liquid nitrogen system E cooling of the low temperature adsorbers is required. L.C.O. " $ 4.2.12 requires that sufficient LN2 (650 gallons without the E [ nitrogen recondensers operating) be available during power g operation so that depressurization can be accomplished. Con-p tinued cooling of the LTA's for decay heat removal is not i in- required due to the slow heat up f rom this source (more than a ? week to reach a temperature level above design) . Therefore, an h i. adequate supply of liquid nitrogen is available in the normal C. storage tank and surge tank. L
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Page 2 .i w
EE-EQ-0019 REV. B APPENDIX B Since the LN2 system is predominantly a passive system, utilizing gravity teed and natural circulation, the reliance on active components is minimized. The nitrogen recondensers are not utilized in the current mode of operation. Also, to ensure nitrogen circulation, surge tank venting must be accomplished. This requires that either V-25203 (manually operated) or, V-25204 (manually operated), HV-2510 (pneumatically operated, F.C., manual override) and PV-2521 (pneumatically operated, F.O.) are open. The supply and return valves for the operating LTA (HV-2505, HV-2502-1 or HV-2506, HV-2504-1) must be open. These valves are pneumatically operated fail open, with manual over-rides. The preceeding system configuration is required by Reference 7.5. During depressurization the non-operating LTA and the moisture monitors are not required. To conserve liquid nitrogen it is desirable to close the nitrogen supply valves to this equipment. The LTA supply valves are HV-2505 or HV-2506. For this moisture monitors these are HV-2534 and HV-2535. Page 3
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EE-EQ-0019 REV. B TABLE B1 MANUAL ACTIONS NECESSARY TO INITIATE DEPRESSURIZATION Class location Function Manual Action Descri tion 1 2 (Note 5 HTFA Open V-45821-P (45822-P) Frm Fire Water 1 RB Cooling Open V-46173 (46174) Frm Fire Water 14 RB Close V-45823-P (45824-P) Ib Reactor Bldg. Drains 1 RB Close V-46167 (46168) Supply frm System 46 1 RB Close V-46185 To C-1301, 1302 1 RB Open V-46606 (46605) Inlet to A-2301 (2302) 1 PB Open V-46602 (46601) Outlet frm A-2301 (2302) 1 RB Open V-23104 (23106) Inlet to A-2301 (2302) 1 RB Open V-23105 (23107) Outlet frm A-2301 (2303) 1 RB Close V-23322 HTFA Spoolpiece Telltales 1 RX Close V-23323 HTFA Spoolpiece Telltales 1 RX Close V-23324 HTFA Spoolpiece Telltales 1 RX Close V-23325 HTFA Spoolpiece Telltales 1 RX HPC Open V-4525 Frm Fire Water 1 IB Cooling Open HV-31122 (3) To Emerg. Condensate Header 1 TB Close HV-31191 (3) Emerg. Condensate Bypass Header 1 TB Close V-4567 Tbil Tale 3/4" 1 is Close V-211567 Frm Discharge of P-2109, 2110 2 is Close V-211565 To Suction of P-2109, 2110 2 TB Close V-21867 To Iow Pressure Seperator 1 RB Close V-211658 ib Turbine Water Drain Tank 1 RB Open V-47101 Emerg. Condensate Header to Sys. 47 1 RB Open V-47102 Emerg. Condensate Header to Sys. 47 1 RB Open V-4776-P (4711-P) Emerg. Condensate Header to Sys. 47 1 RB Icop 1 (2) RB Open/close depending on train in use HV-4704-1 (3) (4704-3) Inlet Train A HPC (E-2301) 1 RB HV-4704-2 (3) (4704-4) Outlet Train A HPC (E-2301) 1 RB HV-4705-1 (3) (4705-3) Inlet Train B HPC (E-2302) HV-4705-2 (3) (4705-4) Outlet Train B HPC (E-2302) Open V-4745 Drain to RB Sump 1 RB Open V-4746 Drain to RB Sump 1 RB Close HV-4736-1 (3) (4736-3) Inlet E-2308 1 RB Close V-4704 (4705) Pump Suction P-4701 (4702) 2 RB LTA Close HV-2534 (3) Moisture mnitor Nitrogen Supply 2 RB Cooling Close HV-2535 (3) Tanks 2 RB Close HV-2556 (3) misture Monitor Outlets 2 RB Close HV-2557 (3) 2 RB Page 1
EE-EQ-0019 REV. B TABIE B1 Class Incation tu ion Manual Action Descri ion 1 2 Note 5 'Ib Depress- Verify Open HV-2301(3)(2302) Train A(B) Inlet to E-2301(2302) 1 RB urize PCIN Verify Open HV-2303(3)(2304) Train A(B) Inlet to F-2301(2302) 1 RB Close HV-2311-1(3) (2312-1) Train A(B) Inlet to C-2301(2301S) 1 RB Open HV-2311-2(3) (2312-2) 'Ib PLmp Down Line 1 RB open V-23279 To Ventilation first 1-1/2 hrs 1 RB Open V-23271 'Ib Ventilation after 1-1/2 hrs 1 RB Class 1: Action necessary to accmplish stated function. Class 2: Action desirable, but not necessary, to accomplish stated function. (i.e., isolation of non-essential users of fire water, isolation of lines from other systems, etc.) Note 3: Valves referencing this note have manual overrides ard may be operated manually if necessary. Note 5: Incation are abbreviated as follows: RB Reactor Building TB 'Ibrbine Building ACB Access Bay O(11' Outside O Page 2
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EE-EQ-0019 REV. B APPENDIX C RESERVE SHUTDOWN SYSTEM MANUAL ACTUATION The Reserve Shutdown System is operated following a total and permanent loss of forced circulation from full power operation to ensure reactivity shutdown margin. There are 12 Class (1) manual actions required to actuate the reserve shutdown system. These actions listed in Table C1 would all be performed in the Reactor Building. There are no electrical components required to function if the Reserve Shutdown System-is actuated as described in Table C1. The reserve shutdown system has been highlighted on the attached P&ID's (PI-11-1, PI-11-2, PI-82-4). w,
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EE-EQ-0019 REV. B TABLE C1 MANUAL ACTIONS NECESSARY TO INITIATE RESERVE SHUIDOWN Class Iocation Function Manual Action Descri ion 1 2 Note 5 Reserve Disconnect N2 supply at V-82887 1 RB Shutdown Connect to Solenoid Bypass Fitting (right hand fitting) 1 RB (Asstning Open N2 bottle valve 1 RB Solenoid h is will open Valves HV-1102-1,5,7,14,15 Inoperable) -28,29,30,31,34 Disconnect N2 supply at V-82888 1 RB Connect to Solenoid Bypass Fitting (right ham fitting) 1 RB Open N2 bottle valve 1 RB t is will open HV-1102-6,16,17,18,20
-32,33,35,36 Disconnect N2 supply at V-82889 1 RB Connect to Solenoid Bypass Fitting (right hand fitting) 1 RB Open N2 bottle valve 1 RB W is will open HV-1102-2,3,8,9,10
-19,21,22,37 Disconnect N2 supply at V-82890 1 RB Connect to Solenoid Bypass Fitting (right hard fitting) 1 RB Open N2 bottle valve 1 RB W is will open HV-1102-4,11,12,13,23
-24,25,26,27 Definitions:
Class 1: Action necessary to accmplish stated function. Class 2: Action desirable, but not necessary, to accomplish stated function. (i.e., isolation of non-essential users of fire water, isolation of lines frcm other systems, etc.) Note 5: Iocation are abbreviated as follows: RB Reactor Building TB Tbrbine Building ACB Access Bay OUP Outside Page 1
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EE-EQ-0019 REV B APPENDIX D FUEL STORAGE FACILITY COOLING The fuel storage facility is used to store irradiated fuel until it can be removed from the site. The wells can be cooled by one of the two System 46 reactor plant cooling water system loops, each of which is sized to remove the design heat load of 1.02 X 106 BTU /hr. Limiting Conditions of Operation L.C.O. 4.7.3 of the Technical Specifications requires that if cooling flows are interrupted, immediate actions must be taken to restore cooling, and, if cooling is not re-established within 24 hours, the fuel must be - transferred to a well or wells where cooling may be accomplished. A steam line rupture could render the fuel storage well cooling system components inoperable. Ref erence 7.4 demonstrates that the fire water system can provide the design flowrate of 102 gpm to the wells once the reactor building becomes accessible and the manual actions listed on Table D1 are performed. The flow path for fire water is shown on Figure D1. Since the reactor building environment will be accessible within 4 hours after LOFC (see paragraph 5.5), it is expected that Technical Specification compliance will be maintained. For information, the present heat load from stored fuel is approxi-mately 40,000 BTU /hr, or only 4% of the design value.
EE-EO-0019 REV. B TABLE D1 ACTIONS NECSSSARY TO INITIATE FUEL STORAGE FACILITY COOLING USING FIRE WATER _ MANUAL ACTION DESCRIPTION LOCATION Open V-45821-P (48222-P) Fire Water Isolation RB Open V-46173 (46173) Fire Water Isolation RB Close V-45823-P (45824-P) Drain Line RB Close V-46176 (46168) System 46 Supply RB
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l i I l ATTACHMENT 5 1 SIGNIFICANT HAZARDS CONSIDERATION FOR CPERATION AT 35 PERCENT POWER I-( l l l i
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Attachment 5 P-85460 r~'S ('j SIGNIFICANT HAZARDS CONSIDERATIONS FOR OPERATION AT 35 PERCENT POWER WITHOUT THE SAFE SHUTDOWN EQUIPMENT FULLY ENVIRONMENTALLY QUALIFIED EVALUATION PSC has proposed a 35 percent power level restriction on the Fort St. Vrain Nuclear Generating Station during the period of November 30, 1985 through May 31, 1986, since the environmental qualification of the safe shutdown equipment located in a potential harsh environment area resulting from a high energy line break has not yet been fully established. Procedures for responding to a high energy line break during the proposed schedule extension period will be based upon availability of only that electrical equipment which would not be exposed to a potential harsh environment. The purpose of this analysis is to demonstrate that operation of the Fort St. Vrain plant at a power level not to exceed 35 percent poses no significant hazard to the public. In the event of a high energy line break, reactor shutdown is accomplished by either an automatic or manual scram, depending on the magnitude of the break. The reactor would remain in a subcritical condition following reactor scram even without manual insertion of
.. ,the reserve shutdown system poison material. Procedures for responding to a high energy line break during this proposed schedule extension period will nevertheless require manual insertion of the reserve shutdown system poison material. Decay heat removal is to be accomplished by the use of one of the two liner cooling flow paths with fire water as the cooling medium. Information supporting the use of fire water as an acceptable liner cooling source is contained in Attachments 1 and 4 of this letter. The following summarizes important points from these Attachments:
- 1. Temperature profiles in the reactor and turbine buildings for the worst steam line break in the appropriate building have been evaluated. These profiles, plus the relevant assumptions used in this determination, are contained in response to Confirmatory Action 3 of Attachment 1. The most important parameter obtained from these curves is the time at which personnel access into the affected building is possible. Two different access times are possible, one at 180 degrees Fahrenheit with personnel using cool suits and the second at 135 degrees Fahrenheit without special (Ov;
, I 4
'd , b Attachment 5
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P-85460 ( N (\ / ,; .
, protective' clothing. These two time periods are 4 hours for cool suit entry and 15 hours for normal entry.
- 2. Two methods for reactor cooldown were analyzed in response to Confirmatory Action 1 of Attachment 1, both dependent upon liner cooling supplied with fire water. The first, depressurization of the PCRV, results in somewhat higher peak fuel temperatures, but t the temperatures remain below fuel particle failure temperatures.
The PCRV internals are not significantly damaged. The second method, maintaining the PCRV pressurized throughout the cooldown, results in much lower peak fuel temperatures but may result in some damage to the PCRV internals. The integrity of the PCRV ' liner and concrete are preserved in both of these methods. The' method using depressurization of the PCRV is the preferred method for cooldown. A significant result of these studies is that if the PCRV is to be depressurized, then depressurization must be initiated within 12 hours following the steam line break. The access times noted, in 1 above assure the viability of the , depressurization method with the use of cool suits. The depressurization process does not rely upon any electrical j, equipment located in a potentially harsh environment. 7-m3 . Adequacy of liner cooling d th fire water to remove decay heat t ') was evaluated in Engineering Evaluation EE-EQ-0019, Attachment 4 G/ of this submittal. The case analyzed in this report is the pressurized PCRV cooldown involving loss of the top ' head cover plates and both layers of the Kaowool insulation. This was chosen since it results in the' maximum heat flux to the liner and hence presents tho greatest ch'allenge to tre liner cooling system utilizing fire water. The conclusion of the,,evaluatirp : tyt liner cooling supplied by fire water is adequate to remo n the are decay heat with no bulk boiling, even when the liner insulation is postulated to completely fall off and initiation of liner cooling is delayed until the insulation falls off. 4.- Leakage of the primary coolant from the PCRV following the steam line break was conservatively estimated in response to Confirmatory Action 2 of Attachment 1. This leakage is in two parts: I & C\
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Attachment 5 Q. P-85460 N.] a) Short term leakage through the circulator brake and seal system. No credit was assumed for the automatic actuation of this system since the associated electrical equipment has not as yet been environmentally qualified. This system can be sealed manually from outside the reactor building. A very conservative estimate of one hour was assumed for personnel to isolate the system. The amount of helium leaking out in the one hour was calculated as 1250 lbs. To determine the 2 hour dose at the exclusion area boundary the maximum primary. coolant activity levels allowed by Technical . Specification LC0 4.2.8 were assumed (equivalent to an equilibrium core with 5 percent failed fuel) to be associated with the above helium leakage. The results are shown in Table 1. The two hour dose at the exclusion area
~ boundary is well below the guidelines of 10CFR100.
b) long term primary coolant leakage through the PCRV is primarily through the PCRV penetration closures. This was conservatively estimated as 400 lbs/ day, which is the maximum leakage permitted by Technical Specification LC0 4.2.9. Actual measured leakage to date has only been a small fraction of this amount. The maximum primary coolant activity level allowed by the Technical Specifications was assumed to determine the 30 day dose associated with this leakage rate. The dilution factors used to calculate doses for cooldown from 35 percent power with PCRV liner cooling are those specified in Table 14.12-1 of the FSAR, assuming no helium buoyancy. The results shown in Table 1 are well below the guidelines of 10CFR100 for the low population zone boundary. CONCLUSION The proposed interim operating procedure does not create a significant hazard because:
- 1) The consequences of a steam line rupture accident occurring during the period of proposed interim operation are no greater than those created by the Maximum Credible Accident, which has been previously evaluated in the FSAR. The dose consequences at the Exclusion Area Boundary and at the Lu Population Zone -
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Attachment 5 P-85460 V Boundary (16,000 meters) are a small fraction of the 10CFR100 guidelines.
- 2) There is no possibility of a new or different kind of accident from any accident previously evaluated. Loss of forced circulation accidents and primary coolant leaks are currently evaluated in the FSAR.
- 3) It has been determined that there is no reduction in any margin of safety defined in the basis for any Technical Specification.
Applicable Technical Specifications are as follows: LCO 3.1.3, LC0 4.2.6, LCO 4.2.7, LC0 4.2.9, LC0 4.2.12, LC0 4.2.13, LC0 4.2.14, LCO 4.2.15, LC0 4.2.18, LC0 4.3.5, LC0 4.3.6, LC0 4.6.1, LC0 4.7.3,-and SR 5.2.10. Therefore, the interim operating procedure will not significantly increase the risk to the health and safety of the public, nor involve any significant hazards considerations as defined in 10CFR50.92. 1 O-4
TABLE 1 Attachment 5 P-85460
SUMMARY
OF 0FF-SITE DOSES RESULTING FROM POSTULATED ACCIDENTS TOTAL DURATION DOSE (REM) LOCATION OF ACCIDENT MAXIMUM DOSE WHOLE BODY THYR 0ID Complete Loss of Low Population 3.7 x E-4 3.6 x E-2 Forced Circulation Zone Boundary Cooling - DBA No. 1 (180 day) Worst PCRV Penetra- Exclusion Area 2.5 17.4 tion Failure (both Boundary closures of a steam (2 Hours) 4 generator penetra-tion) - DBA No. 2
" Maximum Credible Exclusion Area 1.62 x E-1 8.8 x E-2 Accident"(largest Boundary Potential PCRV leak (2 Hours) rate) 30 minutes to set Low Population 4 x E-4 2.1 x E-2 circulator seals and Zone Boundary 400 lbs leakage via (30 day)
PCRV penetration
.-~ closures Exclusion Area 3.6 x E-3 1.5 x E-2 '
Boundary s_ (2 Hours) 60 minutes to set Low Population 8 x E-4 5.9 x E-2 circulator seals and Zone Boundary 400 lbs/ day leakage (30 day) until entire primary coolant inventory is Exclusion Area 5.4 x E-3 2.3 x E-2 released Boundary (2 Hours) 10CFR100 Guidelines Low Population 25 300 Zone Boundary (Duration of Accident) Exclusion Area 25 300 Boundary (2 Hours) , O V
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i r-1 4 ATTACHMENT 6 i1 f'
?
GA LETTER GP-256, " FIRE WATER LINER COOLING DURING DESIGN BASIS ACCIDENT NO. 1" DATED OCTOBER 18, 1985 4 t 1 4 1 l , .9 S 3 o 4 1 f ]
Attachment 6 to P-8548 G A Technologies (,) G A Technologies Inc. PO BOX 856C8 SAN Cl ego. CAufoANIA 92138
'6?94 455-3000 A
October 13, 1985 GP-2656 Mr. H. L. Brey, Manager Nuclear Licensing & Fuels Division Public Service Company of Colorado 2420 West 26th Avenue, Suite 100D Denver, CO 80211
Subject:
Fire Water Liner Cooling During Design Basis Accident No. 1 (
Dear Mr. Brey:
Ns' Attached for record purposes is GA internal memorandum ME:WSB:215:85,
Subject:
"The Adequacy of the Fire Water System During a Permanent Loss of Forced Circulation Accident" dated Revised: 10/15/85. This information was requested by Art Wong in support of current environ-mental qualification activities. To satisfy PSC schedular demands, the original 9/24/85 issue and the revised 10/15/85 issue were pre-viously transmitted to your offices by Federal Express.
The original PSC request was to provide confirmation supporting the FSAR statement that fire water is adequate for liner cooling during Design Basis Accident No.1. The 9/24/85 issue addressed the original request with the conclusion that fire water is adequate. The second request was to address a delay of up to 28 hours in start of the fire water liner cooling and to address variation in delta temperature rise to be expected between liner cooling tubes. The revised 10/15/85 issue addresses these issues. Again it is concluded that the fire water liner cooling mode is adequate with the capability to suppress local boiling. Ov
~
10955 JoMN JAY HoPKINS OR SAN OfEGo CAUFORNIA 92121
.. .__ _ -_. . . ..__. _ _ _. __. . . _ _ __ _ - _ _ _ _ _ _ . .. __ _. .- ._____ = _ _ . _ _ . ~ . _ _ . . _ _ . _ _
GP-2656
. page 2 of 2 i
i
- i. ,
f If you have any questions regarding the attachment, contact
; Jack Kennedy at (619) 455-4116.
1
- . Very truly yours, i ..
D. J. Kowal, Director Fort St. Vrain Services Attachment 4 I 4 i 1 i b. i 1 4 l J t; 4 I 1 2 4 i ? i I r I l l l-'1wM Wa*v mM4
a Attachinent to GP-2656 INTERNAL CORRESPONDENG GA 1076 t I Q' PROJ: 1900 IM REPLY g FROM: W. S. Betts REFER TO: ME:WSB:215:85 10: A. J. Kennedy DATE: 9/24/85 MBS: 2970.503
, iltu s um 10/15/89
SUBJECT:
The Adequacy of the Fire Water System During a Pemanent Loss of Forced Circulation Accident i Ref: 1. " Safe Shutdown and Cboling with Highly Degraded Plant Conditions Abnomal Procedures for Shutdown Cooling," Public Service Company of Colorado FSV Nuclear Generating Station, February 13, 1984. - Introduction Design Basis Accident No. 1, " Permanent Loss of Forced Circulation," is reported in Section 14.10, Appendix D.1 and Appendix D.2 of FSAR. Section D.2.3 describes an accident correspending to Design Basis Accident No. 1 plus a simultaneous loss of water cooling to the PCRV for a temporary period (30 h). It is concluded that adequate supplies of water and paping power could be obtained within 30 h by using the fire water system. This memo assesses the l adequacy of the paping power of this fire water system for this accident. Discussion Per Section D.2.1.1 of the FSAR, the heat load on the PCRV side wall liner and cooling system is about 10 million Btu /hr maximm, 7.5 times the normal l operation heat loads for this region. The maximm PCRV side wall heat flux into the PCRV liner is about 7800 Btu /hr-ft2, which would result in a conservatively calculated maximum tube face heat flux of about 600,000 Btu /hr-ft2, 30% of the nucleate boiling heat flux and only 10% of the critical 3 heat flux. The maxima side wall cooling tube water temperature rise would be l 700F, resulting in an outlet temperature of 1700F, which is only 50% of the maximm water saturation temperature permitted to prevent bulk boiling. The maximm heat load on the PCRV top head liner cooling system is about 3.3 million Stu/hr. This value is only about 1.2 times the normal heat load on the l top head and penetrations. The maximm heat flux incident on the PCRV top head liner and penetrations is about 12,000 Btu /hr-ft2, resulting from radiation from the 17000F top reflector surfaces. These maxima conditions would give an l average PCRV top head liner heat flux of about and a i conservatively calculated maximm hot tube face heat 14,200 fluxBtu of/hr-ft2, about 85,000 l Btu /hr-ft2 using the 6 iri. tube pitch for a 50% dry tube operation. These heat l loads would result in about a 100cF water temperature rise, conservatively A assming that the maximm heat flux were incident upon the entire length of a t cooling tube. These conservative heat fluxes and heat loads are only about 40% of the permissible values when boiling begins. l ,
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l o 1
A. J. Kennedy 10/15/85 The above assmes that one loop of the PCRV liner cooling water system (system
- 46) is operational. Table 1 summarizes the actual flow rate in the PCRV liner cooling tubes recorded at 1001 power. The total flow rate is slightly less than 1500 gpm per loop.
3 The fire water system (system 45) can be used to supply water to the individual PCRV cooling water tubes as shown schematically in Fig.1 (Fig.1-2C of Ref.1) by the opening and closing of selected valves (see p. 32 of Ref.1). The fire water pe ps shown in this figure (i.e., P4501 and P4501S) have the characteris-tics shown in Figs. 2 and 3 Either p e ps can produce 1500 gpm with a head of 3 x 98 = 294 ft (126 psi). Therefore, these peps (from system 45) are capable of approximating the flow rate of the system 46 pumps in either one of the two PCRV cooling water system loops. This implies that the water temperature rise of 1000F predicted for one loop operation using system 46 (per Section D.2.1.1 of FSAR) will also apply if system 45 (the fire water system) is used. Assuning that the maximum initial water temperature is 1000F then the maximm exit temperature is 2000F. To prevent boiling in the individual PCRV cooling water tubes a ~45 psig back pressure on the PCRV cooling water is established if fire water is being supplied directly to the PCRV cooling water system (page 9 of Ref. 1). This 45 psig is recorded at pressure indicators PSH-46225 and PSH-46226. These indicators are located at elevation 4870 ft as shown in Fig. 4. The elevation of the PCRV top head liner is 4904 ft, therefore, the elevation head between the pressure indicators and the PCRV top head is 34 ft (14.6 psi). This implies that the minimum pressure' of the water in the PCRV top head (Pmin is: Pmin = 45 - 14.6 + Patm - where Patm is the atmospheric pressure = 12.3 psi P min = 42.7 psia I at 42.7 psia the saturation temperature is 2710F. This is sufficiently high (i.e., >2000F) to prevent boiling in the cooling tubes. See Attachment A for additional details.
- The elevation of the pap is 4790 ft, therefore, the maximm pressure of the water entering the top head tubes is Pmax = 125 - (4904 - 4790) 62/144 +
12.3 - 88.2 psia. The actual pressure in the tubes also depends on the frictional pressure - drop. WSB:cc cc: D. Alberstein l D. Carosella J. Lopez R. Rosenberg
- f. Swanson E.0F
4 Table 1 O Loop 1 Flow The Flow Rate Measured at 1005 Power 11/7/81 Loop 2 Flow Location Subheader GPM , Subheader GPM Served
.f 1F 155.5 2F 160.6 Core support floor 3F 48.9 4F 59.0 SF 114.6 6F 125.3
! 7F 122.4 8F 121.7
- Subtotal 441.4 Subtotal 466.6 1VB 91.8 2VB 71.7 Lower barrel sidewall and 3VB 114.3 4VB 86.9 upper part of lower SVB 80.1 6VB 78.0 penetrations Subtotal 286.2 Subtotal 236.6 1VT 65.7 2VT 63.3 Top head and upper barrel 3VT 56.4 4VT 61.9 sidewall SVT 55.3 6VT 68.2 TVT 38.4 8VT 43.9 9VT 39.1 10VT 38.8 Subtotal 254.9 Subtotal 276.1
~
1T 20.2 2T 24.3 Top head penetrations 3T 74.7 4T 75.6 (mainly refueling ST 56.9 6T 72.4 penetrations) 7T 50.7 8T 53.4 Subtotal 202.5 Subtotal 225.7 IB 129.3 2B 108.1 Lower penetrations and
- 3B 176.6 48 156.1 bottan head Subtotal 305.9 Subtotal 264.2 Total 1490.9 Total 1469.2 l
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. .. 8o- 5 46226 in reactor building .
per drasting number M-254 0i.t.! w. b, ' II
,.1. h1
" ^" ' M . "
PSC C0t!TR0nED DOCUMENT '
' ;IF USED FOR DESIGN REQUEST
) lSSUE VERIRCAT10tl FROM I
ggb - VELLUM CENTER COPY yedfy issue Status 4th oocument Conglel p J
._ rar to Usd
~
AITACHENT A
$ BOILING IN THE COOLING TUBES During the time (up to 30 h) of loss of water cooling to the PCRV, the temperature of the PCRV liner will continue to rise as shown in Fig. A-1 (i.e.,
Fig. D.2-9 of FSAR). The highest temperatures occur in the top head liner.
" With the restart of the PCRV liner cooling water, the temperature of the liner rapidly decreases.
1 4
' Analysis (Ref.. A-18) has predicte<1. that . although the cooling water punp capacity is rated at 1500 gpm at P = 125 psig (see Figs. 2 and 3 of main text),
.-the-initial-ficw-rate. through-the-punp-will be-substantially smaller due to boiling and subsequent relatively high velocity vapor flow through the cooling tubes.
It is also, predicted that the minimum pressure in the liner cooling tubes will be even higher than a normal condition with water through the entire cooling water system. The analysis in Ref. A-1* estimated the coolant working pressure
- to be 60 psia with a corresponding saturation tanperature of 2930F.
! The Ref. A-1* analysis predicts that water flow can be established in a typical ._ PCRV cooling tube within 12 minutes after fire pump has been operated. l Therefore, it is concluded that the boiling in the cooling tubes will be of a relatively short duration (-12 minutes). The above conclusion assunes the initial liner temperature is 11000F. This corresponds to the liner temperature at approximately 28 hours during LOFC plus , . temporary loss of water cooling accident (see Fig. A-1). 1 As expected if the liner temperature is less than 11000F (i.e., the cooling 4 water punps are turned on in less than 28 h), then the duration of the boiling I will be less than 12 minutes. Conversely if the liner temperature reaches the I maximun value of 15000F shown on Fig. A-1, the duration of boiling would be
! longer. Additional analysis would be required to quantify the duration of l boiling.
Once cooling water is reestablished it is predicted that the cooling tubes have ! a sufficient cooling capacity to avoid dry out even with the high heat loads l predicted (in Section D.2.1.1 of FSAR) when the thennal barrier fails during an
- LOFC accident. This heat load is conservatively predicted to result in about a 1000F water temperature rise. It is estimated that the maximun inlet i temperature of the fire water will be 850F. This implies that the maximun exit temperature of the cooling water is 1850F which is substantially less than the ,
minimun saturation temperature of 2710F predicted in main text. ' I i
- This reference is included in Attachment B.
i (
. _ _ _ , _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ ~ _
.. - - _ _ _ . _ - _ _ . . - = - _ - . . - . _ _ _
1 ; During nonnal operation at 1001 power the average temperature rise (AT) is 100F
, (see Table 5.9-1 of ESAR). However, it is recognized that there are tubes with larger and smaller values. Table A-1 sunmarizes the aT measured for the 38 cooling tubes serving the top head liner. It is seen that the maximum AT is 150F (i.e. , for tube T28). If it is conservatively assuned that the 1000F temperature rise indicated in Section D.2.1.1 of FSAR refers to a typical tube, it follows that AT for tube T28 could reach 1500F. Even with the additional .. . 4 conservative ..assunption the bulk boiling does not occur since the exit temperature for tube T28 would still be less than 2710F.
References i j- - A-1-- Chin,--E. , "PSC.. _ Transient- Coolant- Flow' and- Thennal Response of Upper
. Barrel PCRV Liner to Coolant Flow Restart After a 48 Hour LOFC Accident," & . - . . - . ~: RIC:253:76, SED: RIB:67:76, June 25, 1976.
l t I 4 1 1 i 1 e l t
.in.g 3 4-
~ ._ _. _ _ ._ _ _ . . , _ ._., _ _ .
. _ _ _ _ , _ . _ . . . _ _ , _ . _ _ . . . . _ _ . . ~ . _ _ .
Y .. i. . . . 2000 j ii I(i it. '
~ LINER COOLING WATER FLOW 1600 -
REESTABLISHED AT 30 HR w a me g. i sr 1200 - ! 4 l Y
- a. l I
N l - l
$z 800 -
! TOP HEAD LINER AT PCRV CENTERLINE 3
- d E . % *
;; f f r 4 'lU. 4
} l 00 - l SIDEWALL LINER AT ! CORE HIDPLANE ' I I O I I I I I I l 0 Is0 80 120 160 i 200 216 0 280 n r J . TlHE (HRS) C m j Fig. A-1 higure D.2-9) Maximum Top Head and Side Wall PCRV I.iner Temperatures During LOFC g% 1 1 -+g. . Plus Temporary Loss of Water Cooling Accident 4,k ! um t
~
!.. n' . .
M Table A-1 7m The Measured Temperature Rise ( T) of Top Head Cooling Tubes ( ) at 100% Pcwer (11/7/81)
\v/
OT (_T
' Tube No. Subheader CF Tube No. Subheader CF
! T1 ~
~ 3VT 10 - T31 SVT 14 T2 6VT 13 T32 2VT 7
' ~
T3 3VT 14 - T33 1VT 11
- - -- ~
T4 6VT 12 T34 2VT 10 15 l' ? 3VT 8- T35 .. 3VT 13 T6 ~ 6VT 13 T36 2VT 11 17 1VT 11 T37 1VT 10 T8 4VT 9 T38 2VT 3 T9 4VT 8 T10 6VT 10 Average 10 T11 3VT 7 T12 4VT 11 T13 SVT 14 6VT 8 (v T15 C)'T14 SVT 11 T16 6VT 11 T17 SVT 8 T18 4VT 11 T19 SVT 8 T20 6VT 3 T21 SVT 7 T22 4VT 7 T23 1VT 7 T24 1VT 10 T25 IVT 10 T26 2VT 12 T27 3VT 12 T28 2VT 15 T29 IVT 10 T30 4VT 11 v) h f6 ,
~~n.- , - - .. . . . ,. _ , __-_, - --
l l
. "p=r;-
}}}bac/rryt enr B
,c..unumse.semes====s i i
= aan.v RIC:253:76 i . * "' E. Chin g "*"" " SED: RIB:67:76 l
\- " H. Jones "" June 25.1976 9 ~ ~ - ' -
' i i .- ammer PSC . _ . . . . _ i
, ; E. . . . ' Transient Coolant Flow and Thermal Response of Upper l
- Barrel PCRV Liner to Coolant Flow Restart After a ;
l . j. , 48 Hour LOFC Accident j *;i, ,
- =_ . . . !
,,- REFERENCES _ * ; l ! t I l'.. 1. . F. He/R. Yo11msn, "FSV - Structural Integrity of PCRV Liner Cooling . l ! '.' Tubes During Rapid Cool Down from 1500*F*, RIC:243:75,1:./:.7/75 1 1 .- * ' jJ -- 2. W. $. 8etts. "PSC. The Response of the Liner and the Concreta to I LOFC Accident". TBC:002:76, March 3. 1976. l ' i. .
- *- 3. Sargent & Lundy Engineers. " Gasoline Driven Fire Pump - P45015*
1 Foreign Print No. 42-P-1-14, Pro,1ect 90. Fort St. Vrain Unit 1
- l. d: .
7/17/68. . I a.~., . .# ! .? p l '.<.-
- StsecutY_ '
- .~ .'
- This study examines the difficulties in forcing water through cooling :
t:sbes which have been heated to nearly 1100*F. The assumed initial 7 temperatures of liner and concrete were based on the results of Reference T
' *: 2.after 48 hours. These are very conservative assumptions since it is GA's
$ position that cooling water will be restored in less than three hours. .)
~
This study predicts that the water vapor will exit the tube at a -'
'Al saturation tempersture of about 300'F within 12 minutes under espected ,
canditions. At shorter times superheated steam will exit the tubes. ; l n The significance of this result is that it establishes the einimum time, T. - - after the water pumps are operating, that the thersal barrier is required .
~
to remain functional. i
- This study also predicted the temperature distribution in the cooling tubes ,
i
& and the liner as a function of time. This information should be used to "
predict the thermal stresses in the cooling tubes. (Actually a much worse thermal gradient was used in a preliminary stress evaluation resulting in . c,;?. no failure of the cooling tube for a one time occurrence, Reference 1.)-
- h. . INTRODUCTION _
An accident condition has been postulated where the liner cooling system y loses all forced circulation for a period of two days. At the end of l i I b
;, l
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. .s . m..$ )
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K.'.:.1 -u , E.CMn RIC:253:76 N. Jones SED: RIB:67:76 this 48 hour period, the nonnal PCRV circulating water system and the tuo standby backup pumps are assumed to be inoperative. In addition, the outside (nearoy community) supplied electrically driven fire water pump is also assumed to be out, so that the gasoline driven fire. water system pap is the enly PCRV liner cooling source.
- The mejor purpose of this study is'to verify that the available coolant flow through the liner cooling tubes in the upper core barrel region is -
sufficient-to cool 'the-liner;=:In addition, the transient tenperature_ .
< *- response of the cooling' tube 7sveld.' and liner due to introdue:1on .of - '
" ~ coolant flow is' required for~iise'in thermal stress considerations;-To
make the coolant flow rate estimates,' the pumo characteristics curve
~
~
and some coolant flow systems data is used. The region of the PCRV
, liner considered herein is the PCRV upper barrel sidewall. The effect
, on the top head region is also discussed. -k1though the cooling water pop capacity is ~ rated at 1500 gpm at P = 125 psig, the initial flow rata
-- through the pump will be substantially' smaller due to boiling and sub-l - sequent relatively high velocity vapor flow through the cooling tube.
. DISCUSSION A thenal model shown in Figure 1 was made using the TAC 30 code to simulate ,
a typical section of the upper barrel of the PCRV liner region when only ona cooling water loop is operational. It is assumed that during the period n fire water is introduced to the liner, the cover plate temperature for , thermal barrier is maintained at a constant temperature of 1500*F. The te boundary temperature is fixed at 700*F. This concrete temperature
- s consistant with tte corresponding concrete
- temperatures reported in Kaference 2.
Tha coolant working pressure is estimated to be 60 psia with a c5'rresponding ' saturation temperature of 293*F. For large coolant / wall temperature dif- ~ I farentials, film boiling will occur. Estimates made in 2the calculation ' file recosamend a conservative value of h = 50 Btu /hr-ft *F for the film heat transfer coefficient. Fer the 48 hour LOFC accident condition, the initial inlet water will at - some point downstream turn into superheated steam resulting in a significantly higher flow velocity due to the lower gas density and applicability of jhe law of conservation of mass. Since pressure drop is proportional to ov , '- tha staan flow (for the same mess rata) causes a significantly greater pressure drop than the water flow. To simulata the boiling phenomenon in the TAC 30 transient thermal model, the water properties were altered such that liquid water flow is asstaned from the initial entry point of z = 0.0 inch until sufficient heat is added from the cooling tube to raise the water condition from saturated liquid to _ saturated vapor. The specific heat for the water is altered so that when e s, , .. , I
- * . , '[ *
,< dgg , ,
.m...
.g a - **%j.
l . I
)
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)
RIC:253:75 l
,. Chia SED:tI5:67:75
. Je 3 l 43 water temperature rises about 10'F this amount of heat was added, -
, he downstream nodes in the computer model use the actual steam properties. i .- l enc 3, as time progresses the point at which superheated vapor begins enerally is shifting further and-further. downstream until a steady state 1 I andition is reachad. .If sufficient flow is achieved, the steady state . l' l onditicn is one which does not contain superheated steam and the coolant i ampereture is maint/ tined at a ' temperature no more than about 10*F above i I hr inlot tespereturec_If_the flow rate is too small, superheated steam
~ - - ! cy always" exist in 1ihVi:coling~ tube (for the boundary conditions _ assumed - ";
tr this thermal model).
- ]g
=- - _ , . , '
t
,1t
?
i wo different pressure Erop condiItorN (a'nd hence flow conditions) ars
.? .
nastdered herein for the liner cooling tubes. First. the pump character- , stics curve is considered and increased vapor flow due to the larger pimp 4 ' I i med available at reduced flow ratas are acc5hnted for. Second, only the T ! wuction in upstream pressure losses due to a reduced flow rate is con- f
~
- :idered and the nonnel pump outlet pressure head is assumed available at the .
] :coling tube. This estimated condition is conservative because upstream Q
. lasses art shown to be negligible in the esiculation file and yet the ,
attre available head is not taken into account. 4 l t tEss.T5 < l tent solution option of the TAC 30 code was used to generate the ? __ . lThe t results shown in Figures 2 through 5. F 1m the Ifmited flow f
'aracteristics available to this author r.f w e estimate on the !lC Mture distribution in the liner vicinity as a f, tion of time is . shame in Figure 2 and 3. The differential tesoerature across a cooling l tube weld reaches a maxisess of about 200*F. The detailed tempereture i J distribution near the cooling tube is shown in Figure 3 for an upstream location at the time 12.3 minutes after the fire water cooling circuit to ._,,
the liner has been initiated. My best estimate of the initial cooling water ficer rate through a typical tube is 180 lbs/hr (or & = 90 lbs/hr *
.i ,
through the TAC model of 1/2-tube). This flow rate is based on making use l j
> ef the pump characteristics curve with its associated higher pressure head .
I at lower than rated flow races. For this flow rate, the coolant temperature I will be maintained at the saturation temperature (acproximately 300*F) throughout the length of the Ifner cooling tube after flow has been initiated
- for t > 12.3 sinutes.
i
! If the pump characteristics advantages are not taken into account, the '
!i ofinitial 1/2 tube) flow rate will resulting in about be the 72 same lbs/hr. (or stemperature vostream = 36 lbs/hr. for the TAC model distribution With this lower flow rate the time re- l; ; ! Figure 4 (compared to Figure 2). .:
! spr red to estabitsh a saturation temperature for the coolant throughout-Times :" shorter than those ! the length of the tube is t > 3.0 hours. .
i above would mean that superfGated steam (T > 300*F) would exist in the - - - . s,i l demnstream portion of the liner tubes. ' 1 i ...-.; .: I r m* l 4.: . . .s ., , {;
m
, y . ** .i * * - ,
E. C in RIC:253:75 N. Jones 4- SED:AIB:67:76 e i The initial coolant flow rate assumed through the cooling tube will : I determine the time required to flood the cooling tube with water / steam
- - dat the saturation tamperature of about 300*F. The expected trend is -
shown in Figure 5.. . . C
.y
+-~-*
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- 2
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' ' ~ 1t is expected that a coolant tescerature of about 300*F can be established
" - "in a typical PCRV upper barrel sidewall cooling tube within 12 minutes after F
-, the gasoline driven fire pumo has been operated. However, even under a f l conservative assumotion, the time to establish saturated water in the ..
I cooling tubes should not exceed three hours. Due to the reduced flow rate $ l
, - - caused by superheated steam, it might be advisable to try to distribute ?
the flow from this pump simultaneously to both coolant loops (if the ." I. ! volving arrangenent for this circuit.would par. nit this) senicing the upper barrel / top head regions. These results are based on the pwp characteristics 4._ l 'l given in Reference 3. Any reduction in the ptmp performance would require l re-analysis based on actual measured pump performance characteristics. },. i j , m cc: W. 5. Betts, Jr# 3b 'O l E. Eischoltz R. E. Vollaan .! T. E. McKelvey ,~ i 5. A. Lightner J-0. Wistrom - . , , 4
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CA i543Ut Ev.11/80) 4 CALCULATION REVIEW REPORT . [ i% M df N F/W LVEDI h Cly A APPROVAL LEVEL
's 4 Dff Tsu: / de-~/ e M 4/ QAL LEVEL DISCIPLINE SYSTEM OD C, TYPE PROJECT 00CUMENT NO. ISSUE N0sLTR.
(1 8 ! W J 6 : 2,1 5.* 4 6 INDEPENDENT REVIEWER: NAME 7 A d tt , A R O S F_ LLA ORGANIZATION 105 . , ,, REVIEWER SELECTION APPROVAL: BR MGR s DATE S-M~N REVIEN METH00: YES l NO ERROR DETECTED ARITHMETIC CHECK LOGIC CHECK
^
NO4 ALTERNATE METH00 USED SPOT CHECK PERFORMED COMPUTER PROGRAM USED s REMARKS: (ATTACH LIST OF DOCUMENTS USED IN REVIEW)
@% N :.ata ) M b 8 ch? - 8 gwk k w ud-4 ~(
wWr s 6 nt " *GA d -wd- [.k btsuby - n/ dss i l 1 I umoulATIONS FOUNO TO BE VAll0 AND CONCLUSIONS TO BE CORRECT: INDEPENDENT REVIEWER
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DATE 28!8 Y a}}