ML19242A731

Forwards Re Improvements in LWR Safety Features & Rept Passive Containment Sys,New Concept to Solve Safety Concerns. Requests Timely Attention to Matter

ML19242A731
Person / Time
Site:	Crane
Issue date:	05/15/1979
From:	Falls O NUCLEDYNE ENGINEERING CORP.
To:	Stello V Office of Nuclear Reactor Regulation
Shared Package
ML19242A718	List: ML19242A722 ML19242A731
References
RTR-NUREG-0560, RTR-NUREG-560 NUDOCS 7908030592
Download: ML19242A731 (3)

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{{#Wiki_filter:- UCLEdrYNE ENGINEERLYG CORPOR.4 TION 728 West Michigan Asenue Jackson, Michigan 49201 May 15, 1979 Represented by O. B. Falls, Jr. Consultant Mr. Victor Stello, Director Reactor Operations Division Nuclear Regulatory Commission Washington, D. C. 20555

Dear Mr. Stello:

The energy supply situation in our country needs im-mediate attention to prevent electrical blackouts in the near future which could lead to extreme hardships to homelife and business. Majority public opinicn supports nuclear poser as a necessary energy source. The country needs a firm, definitive statement of support and encouragement by President Carter, Energy Secretary Schlesinger and the several Congressional Committees, having an interect in our energy policy and supply, as a basis for rejuvenation of the nuclear power industry. Also, there is a current need for sub-stantial safety improvements for light-water reactor (LWR) power plants. Consequently, we sent a Mailgram to President Carter with copies to Secretary Schlesingtv and NRC Chairman Joseph Hendrie. A copy of this Mailgram is enc _osed. We would call your attention to the references to the Three Mile Island incident. Your known interest in the energy situation in our country has prompted us to write to you. To support the claims made in the Mailgram a NucleDyne document is enclosed. This is a copy of a paper presented at the American Power Conference in Chicago on April 24, 1979; " Passive Containment System - A New Concept to Solve Safety Concerns". This paper responds specifically to the five safety research projects recommended to the Congress by the NRC in Report NUREC-0438, dated April 12, 1978. Also, some of the benefits that are derived frca a licensed nuclear power plant with the new safety features are enumer-ated in the enclosed " Application of the PCS produces the fallowing results". Extra copies of these publications are available or request. ,o- '] 7008030592 4'-

Mr. Victor Stello May 15, 1979 The FCS incorporates the substantial improvements needed for the LWR power plants. We request your urgent attention to our claims as stated in the Mailgram and discussed in the enclosed document. Furtherniore, we requc3t an opportunity to visit you to validate our claims. Your support in the application of these safety improvements will enable the LWR power plants to become a viable basic source of energy. This, in turn, may encourage President Carter and others to a firm statement in support of nuclear power. We await your response to our request to meet with you. Sinc,cyehy I O. B. Fallsf, Jr. Consultant / OBF/mr [ Enclosures I. 1 A/ 'l 0

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PASSIVE CONTAINMENT SYSTE'1 A NEW CONCEPT TO SOLVE SAFETY CONCERNS AMERICAN POWER CONFERENCE Chicago, Illinois April 24, 1979 Authors: O. B. Falls, Jr. F. W. Kleimola } UCLEDyns ENGINEERING CL)RIN1R.iTION T28 West Michi an Awnue Jackson, Michigan 49201 ,nn 'O [11,, l' ' b, a, )( _? j w

I TABLE OF CCNTE:TrS t INTRODUCTICN PROJECT A - Alternate Containment Concepts PROJECT B - Alternate Emergency Core Cooling Concepts PROJECT C - Alternate Decay Heat Removal Concepts j PROJECT D - Improved In-Plant Accident Response PROJECT E - Advanced Seismic Designs ? s LIST OF FIGURES t 1. Cell Arrangement 12. Steam Carryover Into Celuge Tanks 2. Cell Elevation 13. Containment Pressure Response 3. Reactor Vessel Cell 14. Steam Jet Injector Economy i 4. Reactor Coolant Pump Cell 15. Safety Injection Mass Flow Rates 5. Steam Generator Cell 16. Safety Injection Fluid Energy Capacity 6. Deluge Tank Cell 17. Safety Injection Refill Capability 7. Pressurizer Cell 1S. Secondary System Pressure Response i 8. Quench Tank Cell 19. Refill Tank Depletion 9. Refill Tank Cell 20. Alternate Decay Heat Removal System 10. Engineered Safety Systems for LOCA 21. Response for Alternate Decay Heat 11. Steam Flow Into Deluge Tanks Removal System f REFERENCES 1. NUREG - 0438, Plan for Research to I.tprove the Safety of Light-Water Nuclear t Pcwer Plants - A Report to the Cong:.ess of the United States of America, Nuclear Regulatory Commissicn - April, 1979. 2. NucleDyne E.gineering Corpcration, Safety Improvements With The Passive Ccntainment System - Prepared in Response to NUREG - 0438, June 12, 1978. J.

Croft, T.,

Duffin, D.

J., Steam Power Plant Auxiliaries and Accesscries, Nes York; McGraw - Hill, 1946. .<c r, i - 1

BIBLIOGRAPHY 1. Passive Containment System for Nuclear Power Plants, NEC-1, NucleDyne Engineering Corporation, 1976. 2. Passive Containment System for Nuclear Power Plants, NEC-2, NucleDyne Engineering Corporation, 1976. 3. Technical Presentation on Passive Containment System to Nuclear Regulatory Commission, NEC-3, NucleDyne Engineering Corporation, July 21, 1976. 4. Passive Containment System for Nuclear Pcwer Plants, NEC-4, NucleDyne Engineering Corporation, 1976. 5. F. W. Kleimola, N. A. Rautiola and O. B. Falls, Jr., The Passive Containment System, International Conference on World Nuclear Pcwer, Washington, D. C. (Nov. 14-19, 1976). 6. F. W. Kleimola, Resolution for ACRS Generic Items by The Passive Containment System, NucleDyne Engineering Corporation (June 15, 1977). 7. F. W. Kleimola and O. B. Falls, Jr., The Passive Containment System in High Earthquake Motion, American Nuclear Society Topical Meeting on Thermal Reactor Safety, Sun Valley, Idaho (July 30 - August 4, 1977). 8. Passive Containment System for Nuclear Power Plants, NEC-5, NucleDyne Engineering Corporation, 1977. 9. F. W. Kleimola and O. B. Falls, Jr., Passive Containment System for Boiling Water Reactors, /merican Nuclear Society Winter Meeting, San Francisco, California (Nov. 27 - Cec. 2, 1977). 10. F. W. Kleimola and O. B. Falls, Jr., Accessibility into the Passive Containment System, NucleDyne Engineering Corporation (January,1978). 11. F. W. Kleimola and O. B. Falls, Jr., Inservice Inspection and Plant Surveillance with the Passive Containment System, NucleDyne Engineering Corporation, (January 23, 1978). 12 C. B. Falls, Jr., Plant Surveillence with the Passive Containment System, Ameican Power Conference, Chicago, Illinois (April 25, 1978). 13. Saf ety Improvements with the Passive Containment System, Prepared in 3espense to NUREG-0438, NucleDyne Engineering Corporation (June 12, 1973), 14. F. W. Klei=ola and O. B. Falls, Jr., Recommissioning an Alternate to Decem-mis s ioning, American Nuclear Society Winter Meeting, Washington D.C., (Nov. 12-16, 1978). ([ (, ]h0

I I f t t i l 1 PASSIVE CONTAINMENT SYSTEM i A NEW CONCEPT TO SOLVE SAFETY CONCEPflS i O. B. Falls, Jr. & F. W. Kleimola NUCIIDYNE ENGINEERING COPPORATION AMERICAN POWER CCNFERENCE Chicago, Illinois April 24, 1979 INTRCDUCTION In the Passive Containment Sys:em (PCS-2) the innovative design features incorporate alternate and advanced engineering safety features for light-water reactors. These features interface on the five research projev - suggested by the Nuclear Regul.atory Commission (NRC) in NUREG-0438 to improve the safety of light-water nuclear plants (Ref. 1). A preliminary response to NUREG-0438 on the safety improvements offered by the PCS was prepared anu submitted to the NPC (Ref. 2). This presentation is specifically directed toward the improvements of fered by PCS-2 to the five research projects for a four-loop pressurized water reactor (PWR). The improved safety features are equally applicable to other PWRs and to the boiling water reactors (BWR). PROJECT A - ALTERNATE CONTAINMENT CCNCEPTS The Primary Reactor Containment System in PCS-2 is a thick-walled steel structure composed of interconnected cells. Individual cells house the components of the reactor coolant system (RCS) and the engineered safety systems. These compenents include: reactor vessel, steam generaters, reactor coolant pumps, reactor coolant pipes, and the pressurizer; also the deluge, refill, and quench tanks. A typical cell arrangement for two loops in a four-loop PWR is shown in Figs 1 to 9 inclusive. For accident purposes the air within the free volume of the centainment is evacuated to less than 2 psia. Electrical equipment requiring cooling bs' housed in separate ccmpartments; namely, the reactor coolant pump motors, the centrol rod drives, and the greater portion of the presuuriner. The containment is designed with a free volume of 250,000 cu ft. The deluge and refill tanks along with the steam generator secondaries contain suf-ficient liquid to flood the free volume to an elevation above any postulated RCS pipe break. ]m') 1 Copyright () NucleDyne Engineering Corporation 1979 - ]r, I

An evaluation of the engineered safety systems (Fig. 10) in the loss of coolant accident (LOCA) is made for the worst case, a derble-ended, guillotine-type, pump-suction pipe break. Steam carryover into the deluge tanks (Fig. 6) initiates at less than 5 psia. Each of the four deluge tanks is designed with twelve (12) or more 12-inch vent pipes. Each vent pipe penetrates the top head of the deluge tank and extends almost the length of the tank. These vent an immediate pipes are perforated by thousands of small orifices to facilitatr quench of the steam carryover by the borated deluge water. Each vent pipe is encircled by a shroud pipe to promote thermal circulation past the orifices and within the tank. Sufficient freeboard space is prceided for the steam mass carryover and for thermal expansion in the LOCA. Additional heat sink capacity for the LOCA is provided by the quench tank fluid in the post-accident time period (Fig. 8). One cr more vent pipes (patterned af ter the vent pipes in the deluge tanks) are installed at each quench tank. Thus, the quench tanks provide for a vented containment with a heat capa-city equal to that of the deluge tanks. RCS blowdown in the LOCA initiates steam flow thrcugh the deluge tank vents and the quenching of the steam at the orifices. Steam flow into tne deluge tanks for the postulated pump suction break is traced in Fig.11. Maximum ateam carryover occurs at a little over four seconds into the accident with about 13,300 lb/sec of steam, representing 15.7 million Stu/sec of energy, quenched by the deluge water. The steam carryover increases the liquid volume within the four deluge tanks as traced in Fig. 12. The total volume increases from 53,400 cu ft to about 58,000 cu ft and a corresponding increase in tenperature from 50 F to about 128 F. With initial vacuum conditions, both in the containment free volume and at the freeboard, the liquid volume increase as a result of steam carryover does not impose an added pressure on the containment for the post-accident period. Approximately 500 cu f t of post-accident freeboard is allotted for each deluge tank. The post-accident freeboard is at a vapor pressure of 2.1 psia (the satu-ration pressure of the 128 F water) toward the end of the containment pressure transient. The evacuated containment and steam carryover into the deluge tanks have a decided beneficial effect in the LCCA. A curve cf the containment pressure response to the pump suction break is shown in Fig. 13. The centainment pressure peaks at about 75 psia. At this point the amount of energy in the cLeam ficwing into the deluge tanks, plus the energy retained in the saturated water in the contain-ment, starts to exceed the RCS blowdcwn energy and the ccntainment pressure reducers. By the end of the RCS blowdown, approximately 27 seconds into the accident, the containment atmospnere has reduced to sub-atmospheric pressure.

n the post-accident pericd of a LOCA, any radiolytic hydrogen released from the bcrated water flecding the containment is safely handled by a vacuum pumping system.

A cold trap positioned at the intake to the vacuum punp removes in9 k, I, I L water vapor carryover. The hydrogen is pumped into the holding tanks of the gaseous radwaste system for processing through recombiners. Plant recovery from a LOCA is immediate. With the reactor refueling enclosure removed from the reactor containment (Figs. 3& 4), the fuel is removed from the core before containment decontamination proceeds. Af ter the fuel is retrieved, the borated water flooding the containment is processed through the radwaste demineralizers and stored for reuse. Containment decontamination is more readily acccmplished in the thick-walled steel structure; there is no contaminated thermal insulation to be removed, and there are not concrete surfaces requiring decontam-ation. Equipment within the containment requiring decontamination is at a minimum. All moisture is readily removed with the vacuum pumps present. Any faulted RCS component can be replaced through access openings (Fig. 4), or through roof closures (Fig. 5). Steps, to recommission the plant, can be undertaken during recovery operations. In summary, PCS-2 offers an alternate containment concept that has an inherent venting feature. As stated by the NRC (Rcf.1), "the objective of al-ternate containment designs is to reduce the probability of contaimment failure and subsequent release of airborne radioactivity." The heat sink capacity of the deluge, refill, and quench tanks is over 1300 million Btu. This heat capacity is more than 3.5 times that required for the pump suction pipe break (367 million Stu), There is no danger of subsequent release of airborne radioactivity in that the venting is contained and the containment does not become pressurized above atmos-pheric even in consideration of the 1300 million Btu of energy. PROJECT B - ALTERNATE EMERGENCY CORE COCLING CONCEPTS FCS-2 employs steam from a stored energy scurce for initiating emergency core cooling in the LOCA. In tne PWR the enormous amount of stored energy in the steam generator secondary is employed as the motive fluid for steam jet injectors positioned within refill tanks (Fig. 9). With the reactor in operation the re-fill tanks - completely filled with chilled, borated water - are maintained at secondary system pressure. Cepressurization of the RCS in the postulated LOCA passively initiates ECCS. The cacck valves (Fig. 10) positioned in series at the piping interconnecting the refill tanks to the RCS automatically cpen as socn as the RCS is 'epressurized below secondary system pressure. Safety injecticn piping is rcuted f rem the bottom of each refill tank through an adjcining deluce tank cell. From the deluge tanks the piping is routed to a " cold leg" within a reactor coolant pump cell or to a " hot leg" within the steam generator cell. Cepressurization of the refill tanks autcmatically initiates steam flow to the injectors within the refill tanks. A steamline is routed from each steam header through the tcp end of an adioining quench tank cell. Frem the quench tank cell each steamline is directed into a refill tank where the line branches to a number of injectors in a parallel array. Steam ficw through the iniectors, . a :) / - at i entrains the borated water, providing rapid safety inlection at high pressure. Each refill tank is paired to a steam generator. Each ' cold leg" and eac' ho t leg" in the reactor system has an interconnecting pipe from a refill tank for safety injection purposes. A more detailed analysis of the spectrum of pipt breaks may prove that safety injection to each " cold leg" only is more effective than injection into both cold and hot legs. Detailed analyses of ECCS requirements for the spectrum of pipe breaks in the intermediate range may show further advantages for a separate stored energy source at a temperature and pressure higher than the secondary system for passive safety injection utilizing injectors. This stored energy source would entrain borated water from a separate refill tank also maintained at a higher pressure. but at a lower temperature than the secondary system. e The economy of the steam jet injectors utilized in safety injection is defined in terms of the pounds of water entrained by each pound of steam flow i (Ref. 3). In Fig. 14 the econcmy shown ranges from about one and one fourth at 1000 psia back pressure to about seven at 10 psia back pressure; this economy is based on 1000 psia steam and 50 F intake water. Suppliers of iniectors anticipate f better econcmy than is shown in Fig. 14. Performance tests on injectors are [ required to establish the actual eccnomy at the higher steam pressures and high j back pressures. F J For the four-loop PWR in the postulated double-ended, guillotine-type pump E suction break, the tafe'y injection mass flow rate is traced for an 80 second f time period after the LOCA (Fig. 15). The back pressure curve plotted in respect ( to " time after LOCA" shows RCS depressurization. The safety injection mass flow rate corresponds to the back pressure at any point in time. The ECCS design is based on a core reflood rate equivalent to six inches per second at 100 psia RCS back pressure. As may be noted in Fig. 15, the mass injection rate almost doubles as the back pressure decreases from 100 psia to 14.7 psia. t l A rapid reduction in the RCS back pressure as shewn in Fig. 15 is a point of interest. This stems from a refill system that overwhelms the *CCA. The heat sink capacity of *he injection fluid as displayed in Fig. 16 provides the basis for the statemenu that the re fill system everwhelms the LCCA thic results in the rapid reduction of the RCS system pressure. The heat sink capacity shcwn does not take credit for the injection fluid lost (spillage) through the pipe break. I Curing RCS depressuriza tion the safety injection fluid entering the RCS i through the intact legs is at the liquid saturation temperature corresponding to the back pressure. This injected fluid is subject to rapid heating (boiling) cn contact with the coolant remaining, and by the stored energy in the RCS system ccmpenents (i.e., the reactor vessel internals and the core elements) and by the !o?I9 a '} l, ~4-

steam generator secondary fluid remaining. Blowdown of the injected fluid requires increasing its internal energy bv the amount of latent heat required f ar its evapcration. The safety injected mass provides significant heat sink capacity (Fig. 16). As shown, the mass injected within 30 seconds af ter the LOCA has a capacity of 54 million Btu, which is equivalent to the stored energy in the core elements and the reactor ve.el interals. The capacity increases to 246 r..illion Dtu at 65 seconds and to 328 million Stu at 80 seconds af ter the LOCA. Another way to depict the refill system's capability in overwhelming the LOCA is 'ay looking at the time required to refill given volumes (Fig. 17). The mass injected into the reactor system during 30 seconds after the LOCA - the blewdown time for the most part - refills the reacter vessel to the bottcm of the core. The entire core is flooded within 43 seconds and an cverflow through the postulated pipe ' break starts within 50 seconds af ter the LCCA. In consideration of the safety injection fluid energy capacity (Fig. 16), and t::e refill capability (Fig. 17), emergency cool. 3 of the core fuel is effective. The high turbulence resulting from reactor coolant blowdown enhances energy t: Tnsfer from the fuel; this continues with the rapid injection of emergency cooling water. With high1, borated water starting to reflood the core within 30 seconds a f ter the LOCA the fuel is rapidly quenched preventing an excessive cem-perature r.se. The steam generators contain an adequate amount of stored energy (steam) for safety injection and continued post-accident decay heat remcval (Fig. 18). With the core reflooded 43 seconds into the accident, the secondary system pres-sure is at 523 psia. At 50 seconds with an overflow out the pipe break the steam pressure is at 486 psia. As shown, at 80 seconds after the LOCA, the seccndary pressure is still above 300 psia. The refill tanks have an adequate supply of borated water for emergency core cooling and ccre reflood (Fig. 19). Curing the 80 second time period shown, the stored volume of water has reduced by only 30 percent, from 24,000 to 16,667 cubic feet.

n summary, the refill system in FCS-2 provides emergency core ccoling water passively injected into the reacter system to flocd the core in a timely and effective way.

Rapid quenching prevents fuel temperatures from rising much above the temperatures existing during reacter operation. As stated ( Re f. 1), the NRC " concern about EC2 effectiveness has been related mainly to the difficult and complex calculations needed for analyzing the per-formance cf ECC systems in large pipe-break accidents in pressurized water re-acters.' The response of the ECC systems presented herein were analyzed with straightforward calculations performed by " hand" A computer program was not utilized. g- -f T l (, k 0 _3_

PROJECT C - ALTERNATE DECAY HEAT RE"O'.'AL CCNCEPTS PCS-2 encompasses alternate decay heat removal systems that are bunkered (Figs. 6, 8, and 9). These systems transfer decay heat for an extended time period after the nuclear chain reaction has been stopped. These systems are effective in a postulated pipe break accident condition in the RCS or in the secondary system; also, whenever the normal feedwater sources are unavailable as on the loss of electric power or some other ;nalfunction. LCSS OF COOLANT ACCICENT As described under Project B in the postulated LOCA, the reactor vessel refill system provides emergency core cooling (Fig. 9). After core reflood, the refill system continues post-accident heat removal for a number of minutes. As soon as the secondary system pressure is expended by the refill system, residual heat removal automatic. ally continues with gravity flow of borated water from the deluge tanks (Fig. 10). Each deluge tank is interconnected to a high pressure safety injection pipe leading to a " cold leg" in the RCS. Each pipe leading from the bottom of a deluge tank branches into a safety injection pipe at a point between the re-fill tank and the first valve. Each pipe from a deluge tank has an isolation valve and two check valves in series. Residual heat removal initiates as the refill system injection pressure decreases below the static head at the deluge tanks. This deluge tank water, heated from 50 F to 130 F by steam carryover from the containment, has over 50 feet static head. In that the containment-free volume and the deluge tank freeboard are at abcut the same pressure, the drivi e force continuing emergency cooling for the core is over 20 psi. The stored " ass in the deluge tanks con-tinues passive decay heat removal for about four hours into the accident. During this time, the containment is flooded with berated water to an elevation arove any postulated pipe break in the RCS. Heat exchange units (not shown) located at the bottom end of the deluge tank cella can be interconnected to heat exchange units in an outdoor ultimate heat sink at a higher elevation. This could provide residual heat transfer for the balance of the accident period. The fcur hour deluge water flow period prcvides time for this passive heat exchange system to "take over" A vented containment is provided for the post-accident time period (Fig. 8) Pipe vents leading from the centainment-free volume into the chilled 'luid within the quench tanks prevent the vapor pressure in the free volume from rising to atmospheric pressure. _.3 _

The research project on alternate decay heat concepts recommended by the NRC (Ref. 1) places emphasis on passive systems for high reliability. The PCS-2 post-accident decay heat removal system is passive and bunkered thus providing high reliability. LOSS OF NORMAL TEEDWATER A second alternate decay heat removal system is presented (Fig. 20). This system is ef fective in core decay heat rezaval for an extended tbme period whenever the normal feedwater sources are unavailable. This alternate system also enables RC5 cooldown at 50 F per hour. Emergency feedwater is autcmatically injatted into the secondary system along with steam blowdown to the contained heat sinks. Decay heat is transferred by conduction and natural convection from the core elements to t a secondary system for rejection from the RCS. On a loss of normal feedwater flow, power-operated relief valves on the steam header for each steam generator automatically open. Cne set of relief valves initiate steam blewdown to both the deluge and quench tanks (Figs. 6 and 8). A second set of relief valves initiate steam flow through steam jet injectors which entrain chilled water frcm the quench tanks for injection into the feed-water hcaders. The steam heats the entrained water to the saturation temperature: thus, the initial injection is about 545 F corresponding to the 1900 paia sacorAary system pressure. Steam flow to the deluge and quench tanks rejects the energy resulting from decay heat generation, sensible energy flow frcm (50 F per hour) RCS cooldown, and secondary system temperature and pressure reducticn. The latter enables continued thermal conduction and natural convection of energy from the RCS to the secondary system. The initial mass flow of steam into the deluge and quench tanks is in the range of 80 pounds per second rejecting 95,000 Btu /sec. The steam is dissipated in the deluge tanks through small orifices with an en-circling shroud promoting circulation of water past the orifices. Steam flow to the injectors positioned within the quench tanks is used to replenisk the mass lost through secondary system steam bicwdown: also the added amount

quired for the change in tne specific volume during steam generator ccol-down.

Steam flow through the injectors entrains the chilled water, and develcps a velocity head with suf ficient resultant pressure to open the dcwnstream check valves for emergency feedwater injection into the adjacenc f eedwater headers. Initially, the high pressure steam entrains about 1.24 pcunds of water per pound o f s te am. The starting feedwater flow rate is in the range of 100 pcunds per seccnd. As the secondary sis: 2m pressure decreases, the eccncme of t:,e in'ector improves essentially as shown in Fig. 14. In this application the sceam pressure and sec0niary system back pressure decrease simultaneously. ,m, 'N' / I -7_

i A beneficial feature of the rejection of the energy to the heat sinks 8 provided within the deluge and quench tanks is that the release of potentially i radioactive steam from the secondary system is completely contained. After the first hour of cooldown, steam blowdown into the deluge tanks has increased the water temperature from 50 F to about 99F. This produces a vapor pressure slightly under 2 psia in the containment through evaporation at the deluge tank vents. As shown in Fi,g. 21, the heat sink temperature increases to about 177 F in four hours. The pressure in the containment is at 7 psia, over 7 psi below a tmo s phe ric. At this point in time the RCS pressure and temperature has reduced to 400 psia and 350 F. Decay heat removal can be switched to the normally provided residual heat removal system. As shown in Fig. 21, the heat sink temperature increases from 177 F to about 200 F from the fourth to the sixth hour; the containment pressure increases from 7 psia to about 12 psia. If there are reasons to continue operation of the alternate decay heat removal system it can effectively continue decay heat transfer as long as required by cooling the liquid in the deluge and quench tanks with the heat transfer systems provided for these tanks. PROJECT D - IMPRC'v*ED IN-PLANT ACCICENT RESPCNSE This research project as recommended by the NRC ( Re f. 1) Ddeals with what the plant operators can and shculd do in a developing accident situation". The NRC believes that human f actors have a major influence en tha availability of safety systems when needed under stress conditions. Safety system availability is also influenced by performance tests and maintenance operations along with ccmponents left in an unavailable state through an oversight. The engineered safety systems in PC5-2 inherently respond to the pro]ect proposed. 1. A respense frcm the plant cperatcrs in a developing accident situation is not required: 2. Cperator action under stress conditions during an accident is not required: 3. Test performance and maintenance operations, which may jecpardice saf ety system availability, are not required for passive systems. 4. Passive systems do not require test operations which may initiate accidents as is the case fcr active systems. FRCJECT E - ACVANCED 3EdMIC CESIT;3 This recommended ?!RC research project entails a study of varicus concepts for improved seismic resistance ( Pe f. 1). On '( O i _3

In PCS-2, embedment of the reactor building to a depth whereby the refueling floor is at grade elevation lowers the center of gravity sub-stantially. This places the major portions of the massive components in the nuclear steam supply system below grade. In addition, the massive reinforced - concrete required in the supporting and shielding structure, the large water tanks for the engineered safety systems, and the refueling structures; along with the large inventory of water therein, below grade, further lowers the center of gravity. The compact arrangement of the reactor building makes undergounding feasible. This embedment of the reactor building, lowering the center of gravity reduces amplification of the ground motion up through the structure, moderating the structural forces and the subgrade bearing pressures and reducing the amplified respcse spectra on the equipment. This deep embedment also provides stability against sliding and overturning and moderates the tce pressure at the soil-foundation interface. The heavy reinforced - concrete mat foundation for the reactor building is common to all engineered safety features. In this manner all radioactive and all safety class structures, systems and components are commonly based. Interstructure relative displacements are not a concern with the reactor building being the baric seismic Category I structure. Faulting displacements of safety related umbilicals are limited to the underground piping to the ultimate heat sink. These are not required for at least four or more hours into the LOCA. The reactor cuilding is not subject to the pressure and temperature transient in the LOCA therefore, the building does not require a structural acceptance test (SAT) at the LOCA conditions. In the LOCA the reactor building provides secondary containment and is not subject to the mass and energy release frcm RCS blowdown. This postulated accident is acccmmodated by the primary reactor containment's thick-walled, steel structure. $}fr }l _9_

REACTOR VESSEL CELL 4 ~~~~ lf'N, ,\\ (\\ \\ \\, \\' REACTOR COOLANT PUMP CELL STEAM GENERATOR I /* '] k k' CELL ,*- DELUGE TANK CELL i [ [ PRESSURIZER CELL s A REFILL TANK CELL j x QUENCH TANK CELL ra g; FIGURE 1 CELL ARRANGEMENT Copyright h thicleDyne Enqirteerire; Cornoration 1979

STEAM GENERATOR CELL QUENCH TANK CELL l DELUGE TANK CEL ,fm m,e, ,_,T ' ~ ~ A, REFUELING I ENCLOSURE ^ O C2 \\ /

==-- I 1 __ L_ 17 CELL ~ O'v x Nx is t~' REACTOR VESSEL CELL O REACTOR COOLANT PUMP CELL \\ rc O FIGURE 2 CELL ELEVATION c,vr w,< e rem ie,>v,m,:m, i,mm < i,, ce<,-<meiem 102,

i t t G.h. I t l-4: N REFUELING t l l: ENCLOSURE i ~ s STEAM {_' l'j GENERATOR CELL G i \\* N i w l H,ql.gf eba[ q Q: j; i REACTOR VESSEL t 1 j 16 .e. , i, n \\ / / I e \\ i.. % 3 N.. c s I r 9:?'- s FIGURE 3 REACTOR VESSEL CELL Copyright Q *;ucleOyne Ene:ineering Corporation 1979

ACCESS OPENING s'Nd' t 'N REFUELING cr'.H: ENCLOSURE .. A. m. u: . / [-{ /) w ) ,) y { N 1 REACTOR COOLANT j PUMP ddN ~ I gg.t_ ___ __- __ n _. -- f k jjd = - - h / .b d,' l 7I N /[ REACTOR a i ,j VESSEL m s u 's: \\_ FIGURE 4 REACTOR COOLANT PUMP CELL

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c, (, [g [_. [O3 C' cricht C 'JucleDyne Encineering Corporation 1979

ROOF CLOSURE A " ~ ' - REFUELING ./ a..:.,/- ) i ENCLOSURE i .q l ef..wj sc i e + .c V. v"b._ s -- .y, i' 1, :: i . L~ Ji.LN s b lt-N, I I N l S TEAM i GENERATOR ./ 4 (, i n '~~- l N I s' "2 a); i l

c n N

1 REACTOR VESSEL rh d' I( 3.t '. Vl x

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.g, 7 I s N\\g FIGURE 5 STEAM GENERATOR CELL ( ii .t e Cc p.c ri:n t @ Nu:-leC ne Engineering ccrpc ation 1979

ROOF CLOSURE u 3/ ~ l y- ' '~ ..}O..., I ).. .n. !!], y,,. I i i DELUGE TANK 'h !l 0 f q) ?nn i i !.4,4f%!'., s 1:. i .r 'u

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l GENERATOR a i

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l? j s e~ 3 I %I f /) ~ ~ / y\\ / .. ~. ~ >, / FIGURE 6 DELUGE TANK CELL -) n I) [_ \\_ ] i} t. 23Firicht @ ';ucleCyne Engineering Corporation 197 9

w 4 ,N D7% ./ i N "e m s~ r3 W' p I b[. wlhi

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.i J '.: ./ 2T,hW.:.; I q:i c: -,.1 q 1 np; q l ? l STEAM PRESSURIZER ] I GENERATOR I l =. t ji t l l l i l i i t i 'l +jR fj 4...,- e h ,. h / tI '5 V N L r*, '~j OH re7, ~ -e s ^ f

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1 / .. - 4% t, A q' s r l -_v t I / N i y,. h FIGURE 7 PRESSURIZER CELL .,.,t_ 'r, /i d U

.fri ht @ ',acleDyne Engineering Corporation 1979

ROOF CLOSURE 7 i. STEAM GENERATOR ,v./ f l ',a p.,4 ~ir'ci l' ~ f# .f ~ $y[Dff?fx \\ x)! l S i 2 Ih;f \\'L! \\ a,diT! ;l ! U "i : ,l' e

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F.,.. -- N tj s / )[3" Nl. vs N [ v' l QUENCH TANK l -1 FIGURE 8 QUENCH TANK CELL ') O /., 'T C 0r'/ r ig ;t @ i.ucleDyne Engineering Corporation 1979

s ROOF CLOSURE STEAM GENERATOR 7-QUENCH TANK CELL -N~s. s I Q.fy,*f,l f m % s/ e q _s[o.d. \\ \\ e x l N, x i ~, A ,l ,- / P,r' i w* e-l,i /,- ![ g 4 l! 1

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, r' pf <4s / o.- l 7 I;. ', f(f } Lm) ~ N i FN. D' / c ,) ,/ / ,/ t.. - / / REFILL TANK CELL ~/. .v' ") .%4' I.' FIGURE 9 REFILL TANK CELL i '\\'s Copvricht @ NucleDvne Encinaarinq Corporation lo

e __ OUTSIDE g ISOLATION VALVE 3_.x INSIDE CONTAINMENT STEAM m GENERATOR (N* DELUGE QUENCH TANK i i i TANK h TA ( ( J S ~ \\ / I 8 I l gigi s TO RCS INJECTOR s ~4 COLD LEG e FIGURE 10 c-C.'* ENGINEERED SAFETY SYSTEMS FOR LOCA Cui.ynight O this i t i",..t 1:iDI ilu ' r IIMI l'"f l '"Fl 5 0" I 'J 7

(D O Mo 16 - 15 x ~ x O 12 - 12 $ m 3 \\ y =l 9 9 = tu b l-- 4 <C OC oc 6 6 3 g o o 2 .s 1 LL m 3 3 O = 'l s 0 0 i i i i i i 0 4 8 12 16 20 24 28 32 z', TIME AFTER LOCA (SECONDS) FIGURE 11 STEAM FLOW INTO DELUGE TANKS ce1v<s,st e necieny,,e em,i m_,,e<, om.ere u co -

60 150 4 w 58,000 CU. FT. E 53,400 CU. FT. 48 128 F-120g o o-m m @o D 90 $ E " 36 w* xw 2 ~1- = o-3 - 60 3 I 24 o>D 4 TANKS: 15,000 CU. FT. EACH 1-50 F o a-- - 30 D 5 12 d 9 a a O 0 0 4 8 12 16 20 24 28 32 e l TIME AFTER LOCA (SECONDS) FIGURE 12 STEAM CARRYOVER INTO DELUGE TANKS n y e/ y-i vili t (I') f1t s l I r/i n l'r e g i si... r i t o r G ir-[iv ir.i t j o ri ]<47ap

g 100 DOUBLE-ENDED, ~mo PUMP SUCTION BREAK 80 75 PSIA y Dw 60 m O. I-- 2 40 w M 2 j h 20 r ATMOSPi-lERIC PRESSURE \\ Z N o o 0 0 4 8 12' 16 20 24 28 32 i TIME AFTER LOCA. (SECONDS) er) Q FIGURE 13 CONTAINMENT PRESSURE RESPONSE Copyrlyht. @ rhicleDjne 1:ng i n ee r i t uj Corporation 1979

4 10 E ECONOMY: POUNDS OF WATER @5 8 PER POUND OF STEAM 4 oW WITH: 1000 PSIA STEAM o i W w 50 F INTAKE WATER T co 6 ou HOT ww m o-4 Em >w W> 7< 2 3 E. <m w =! 0 O 200 400 600 800 1000 BACK PRESSURE (PSIA) s FIGURE 14 STEAM JET INJECTOR ECONOMY m m,w,, n o,.. i...... ,:,o,.........,.,< ,.>, mom, v o <>

M 10 1000 o BACK PRESSURE Q e x D5 6 8 800 o h MASS RATE m a a 6 - 600 g ~ w w H T o-m 4 CORE REFILL RATE AT 400 s d= a SIX INCHES PER SECOND m 2 200 g 100 PSIA BACK PRESSURE 0 J 0 0 10 20 30 40 50 60 70 80 TIME AFTER LOCA (SECONDS) 'L'l FIGURE 15 SAFETY INJECTION MASS FLOW RATES Copyright @ thiclet;yne Engineerituj Corpora tion 1979

S 200 o e x 3 160 H 80 SECONDS, S 328 x 106 120 U 65 SECONDS, f 246 x 106 4 80 O STORED ENERGY: 2 CORE 29.3 x 10 6 E 40 INTERNALS 25.0 x 106 TOTAL 54.3 x 106 D l I I I e Q 0 10 20 30 40 50 60 70 80 TIME AFTER LOCA (SECONDS) ~ FIGURE 16 SAFETY INJECTION FLUID ENERGY CAPACITY Co; >y r i yli t. C IJoclei,yne I:nq i nee r imj Corg or.-it ion 1979

20 e OVERFLOW THRU BREAK 3 _ CORE REFILL RATE - AT / 16 f SIX INCHES PER SECOND CORE REFLOODED ~ ~ 12 m m <t lE 8-Z REFILL TO BOTTOM OF CORE l 4 !? 0 ,,sq 0 10 20 30 40 50 60 70 80 ~ TIME AFTER LOCA (SECONDS) Ti FIGURE 17 SAFETY INJECTION REFILL CAPABILITY Copyright h flueleDyne 1:ng ineeri ng Cor[> ora tion 1979

1000 1000 E 800 - 800 ~ PRESSURE y 55 = Sh a w m 600 - 600 a E w E TEMPERATURE E g 400 - 400 m 2 2 w 200 - 200 m H H CD M O i i 0 O 10 20 30 40 50 60 70 80 TIME AFTER LOCA (SECONDS) FIGURE 18 N SECONDARY SYSTEM PRESSURE RESPONSE m,,,,,,,,, e,_,.,,..,<,..,,.....,1,.,c.,,-.....>.,

/ 25 - ~ 24,000 CU. FT. ~~ Q 20 16,667 CU. FT. X cF 15 I wg 10 Da O 4 TANKS: 6000 CU. FT. EACH 5 0 0 10 20 30 40 50 60 -70 80 TIME AFTER LOCA (SECONDS) FIGURE 19 5 REFILL TANK DEPLETION c,

ISOLATION VALVE f q+ e e >c 5 n 3_g (N STEAM GENERATOR D f ) { h \\ I QUENCH DELUGE TANIC TANK a ){ N nj ' INJECTOR V ( y OUTSIDE INSIDE CONTAINMENT FIGURE 20 ALTERNATE DECAY HEAT REMOVAL SYSTEM Copyright C " ' c','ir..- Engineering Corporation 1979

20 o 200 c 5 c;; 5 o o. ~- - 16 8 160 y i 2g 2- <w m 120 12 m m g-a T o I-U- 4 2 l-mE og 80 8 z w dg 2 aw 2 g l-- - 4 a 40 0 2 a O w 0 0 i i i o 0 0 1 2 3 4 5 6 TIME AFTER LOSS OF NORMAL FEEDWATER (HOURS) FIGURE 21 RESPONSE FOR ALTERNATE DECAY HEAT REMOVAL SYSTEM Copyright @ NucleDyne Engineering Corporation 1979 ........,}}