ML20054J293

Forwards Response to Questions CS 430.89 & 490.24 Re MSIV Steam or Feed Line Pipe Break Event,Per 820316 & 25 Requests.Info Will Be Incorporated Into PSAR Amend 69, Scheduled for Jul 1982

ML20054J293
Person / Time
Site:	Clinch River
Issue date:	06/18/1982
From:	Longenecker J ENERGY, DEPT. OF
To:	Check P Office of Nuclear Reactor Regulation
References
HQ:S:82:050, HQ:S:82:50, NUDOCS 8206280439
Download: ML20054J293 (17)

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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:050 JUN 181932 Mr. Paul S. Check, Director CRBR Program Office Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Check:

RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION

Reference:

Letter, P. S. Check to J. R. Longenecker, "CRBRP Request for Additional Infonnation," dated March 16 and 25,1982 This letter formally responds to your request for additional information contained in the reference letter.

Enclosed are responses to Questions CS 430.89 and 490.24 which will also be incorporated into the PSAR Amendment 69; scheduled for submittal later in July.

Sincerely, J n R. Longen ker Acting Director, Office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy Enclosures cc: Service List Standard Distribution Licensing Distribution t

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page 1 W82-0362 (8,22) 81 Question CSf3Q 89 (10.3) As expl ained in issue No. I of MUREG-0133, credit It taken f or all valves downstream of the Main Steam Isolation Yalve (MSIV) to limit blowdown of a second steam generator in the event of a steam line break upstream of the MSIV. In order to confinn satisf actory perf ormance following such a steam line break provide a tabulation and descriptive text (as appropriate) In the PS AR of al l fl ow paths that branch of f the main steam lines between the MSIV's and the turbine stop val ves. For each flow path originating at the main steam lines, provide the following Inf onnation:

a) System identification b) Maximum steam flow in pounds per hour c) Type of shut-of f valve (s) ui Size of valve (s) e) Quality of the valve (s) f) Design code of the valve (s) g) Closure time of the valve (s) h) Actuation mechani sm of the val ve(s) (i.e. , Solenoi d operated motor operated, ai r operated diagram val ve, etc. )

1) Motive or power source for the valve actuating mechanism In the event of the postulated accident, termination of steam flow from all systems identified above, except those that can be used f or mitigation of the accident, is required to bring the reactor to a saf e cold shutdown. For these systems describe what design features have been incorporated to assure closure of the steam shut-of f valve (s).

Describe what operator actions (if any) are required.

If the systems that can be used for mitigation of the accident are not available or decision is made to use other means to shut oown the reactor describe how these systems are secured to assure positive steam shut-off. Describe what operator actions (if any) are required.

If any of the requested information is presently included in the PSAR text, provide only the ref erences where the inf ormation may be f ound.

BespDRS9 Section 15.3.3 of the PSAR addresses steam or feed line pipe break event (updated and attached). Section 5.5 of the PSAR describes the design of the steam generator system. An updated list of steam generator system valves data in provided in the revised Section 5.5.3.4 and f igure 5.1-4.

QCS430.89-1 Amend. 69 July 1982

.- - _ - ____--___ - _ _ _ _ _ -_-___ - __________ _ ___ ____J

. Pcgo - 1 [8,223 #47 l .

l Power operators shall be sized to operate successf ully under the maximum j dif ferential pressure determined in the design specification.

l The main steam isolation valves (superheater outlet isolation valves) are capable of being closed to stop the venting of steam into the steam generator or turbine buildings in case of a steam line pipe break downstream of the isolation valves. The maximum steam flow rate is expected from a steam line break immediately downstream of the Isolation valve. The disc and stem will be designed to withstand the forces produced when closing the valve under choke flow conditions.

Figure 5.5-2A shows a main steam isolation valve. It is a conventional gate valve to provide a minimum resistance flow path when the valve is wide open.

A closed system hydraulic-pneumatic operator, shown in Figure 5.5-28, Is

,~~W for opening and closing the valve during normal operation or during val ve exercising. Upon loss of electrical power, the pneumatic and hydraulic solenoids are opened by springs, which causes pneumatic pressure to shuttle the valve operating cylinder. The oil below the valve operating cylinder

( returns to ihe reservoir through the pilot check valves, which are p!:oted open by pressure acting through the hydraulic solenold valves. The gate valve is accelerated during the initial period of the blowdown and is decelerated at the end of the closing stroke by a hydraulic damper which enables sof t seating of the gate while providing f ast closing of the valve. A pressure compensated flow regulator ensures uniform closing times over variations in loed.

Position switches are provided to Indicate gate position remotely. The valve is repositioned by energizing the motor and solenoids.

A superheater bypass valve is Installed in parallel with the main superheater outiet Isolation valve and check valye for use durIng plant startup for preheating the BOP steam lines and following plant shutdown to maintain the BOP pressurized. This is an active valve, designed to f all closed.

Each valve used in the SGS will be evaluated as to its performance relative to plant safety and mode of operation in the event of f ailure (f all open, fall closed, etc.). As part of these evaluations, the need for a pneumatic l

accumulator adjacent to a valve and solenoid requirements for emergency operation wili be determined.

Tests and insoections Line valves will be shop tested by the manuf acturer for performance according to the design specifications for leakage past seating surf aces and for integrity of the pressure retalning parts. Selected line valves will be manually operated during loop shutdown periods to assure operability.

5.5.2.3.2 Recirculation Pumns ThekocirculationpumpwilIbeasinglestage,centrifugaltype,drivenbya

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constant speed, 4.0 KY,1000 HP motor. It wilI take suction from the steam drum, and provide 2.22 x 10 6 pounds of water per hour to the evaporators.

The pump and its support will be designed and f abricated per ASE Section 111, Class 3 as shown in Table 5.5-6.

5.5-7 Amend. 69 July 1982

Pcg] - 3 [8,22] #47 Saf ety/ power relief valves are installed on the outlet line of the evaporator units, on the steam drum and on the outlet line f rom the superheater. These valves all meet the requirements of Section 111 of the ASME Boller and Pressure Vessel Code f or protection against overpressure. Tabl e 5.5-8 Indicated design pressures and valve settings f or the steam generator saf ety/rel ief val ves. Additional valve data is provided in Table 5.5-8a.

5.5.3.5 Steam Generator Module Characteristics Each evaporator module will produce 1.11 x 106 lb/hr of 50% quality steam from subcooled water. Each superheater module will produce 1.11 x 106 lb/hr of superheated steam from saturated steam. The thermal hydraulic normal design operating conditions are given in Table 5.5-9.

The steam generator modules will supply the turbine with steam at design conditions over a 40% to 100% thermal power operating range f or both clean and fouled conditions. The steam generator modules are else capable of removing reactor decay heat with the natural convection in both the intermediate sodlum loop and the recirculaton water loop.

This hockey stick unit is of the same basic design as that of the Atomics International-Modular Steam Generator (Al-MSG) unit which was tested In a test program ca'rried out at the Sodium Canponent Test Installation. The Al->EG employed a 158-tube module with an overall length of 66 feet, as compared to the 757-tube CRBRP Steam Generator which has an overall length of 65 feet.

The Al-NSG heat exchanger was operated f or a total of 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Including operation both as an evaporator (slightly superheated steam out) and as a once through evaporator-superheater (from sub-cooled liquid to completely superheated steam).

The Al-F5G served as a preof test of the Al prototype hockey-stick steam generator design. The unit was operated for 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> under steaming conditions; all of these 4,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, the unit was at the same tanperature level at which the prototype wilI operate, with a steam pressure equal to or greater than prototype conditions. Table 5.5-9A compares various design operating conditions f or the CRBRP Units to the Al-MSG, and lists the number of hours which the Al-MSG operated under respective conditions. The Al-MSG operated at steam pressures equal to or greater than the CRBRP Units f or essentially the whole 4,000 hrs., and at CRBRP superheater inlet temperature for 750 hrs.

Since the Al-NSG unit was operated in the once-through mod, simultaneous j simulation of both inlet and outlet CRBRP conditions for the separate CRBRP evaporator and superheater units was not achieved, but operation over the CRBRP temperature and pressure range was achieved on both the sodium and steam conditions for significant portions of the test.

5 .5-23 Amend. 69 July 1982

Page 1 (82-0362) [8,223 #78 TABLE 5.5-8A VALVE DATA SL994ARY (a) (b) (c) (d) (e.fi (g) (h) (I)

VALVE IDENTIFICATION MAX '6 E STEAM GENERATm FLOW SIZE SECTION 111 CLOSURE TIE ACTUATOR POWER SYSTEM lb/hr TYPE INCHES DIV ISION SEC ECHMilSM SQJR Superheater Gate 16 Class 3 3 max. Electro-Hydraulle IE Electric

• Outlet (53SGV012) 1.11x106

, Superheater Flow 4 Class 3 3 max. Electro-Hydraulic 1E Electric

• Bypass (53SGV016) 3.41x104 Control Superheater Gate 12 Class 3 3 max. Electro-Hydraulic 1E Electric

• Inlet (535GV011) 1.11x106 Evaporator Gate 10 Class 3 3 max. Electro-Hydraulle IE Electric Inlet (53SGV006) 1.11x106 1

Steam Generator Bldg. Gate 10 Class 3 3 Electro-Hydraulic IE Electric

• Feedwater inlet isolation (53SGV001) 1.22x106 Maln Feed Water Flow to Class 3 5 Air Diaphran Instrument Air Inlet (535GV002) 1.22x106 Control Starbup Feedwater Flow 4 Class 3 5 Air Diephran instrument Air inlet (535GV003) 2.44x105 Con trol Steam Drum Drain Gate 6 Class 3 3 Electro-Hydraulic 1E Electric

• Valves (53SGV014,15) 1.1x105

1. Active Function (Safe Position) is 1E Electric 5.5-50a Amend. 69 July 1982

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Figure 5.5-2A. Main Steamline Isolation Yalve (Superheater Isolation Valve Outlet)

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SEAT 5.5-56 Amend. 69 July 1982

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4 Page - 1 L8,22J 87 af fecting the three steam supply systems and is provided if needed on a per loop basis. By def inition, this zone of protection will include the high pressure steam supply system downstream frce the Individual loop check val ves. j 7.4.2.1.2 Eaulsment Dssiso A high steam flow-to-feedwater flow ratio is indicative of a main steam supply leak down stream f rom the flow meter or insuf ficient f eedwater flow. The l superheater steam outlet valves shall be closed with the appropriate signal l supplied by the heat transport instrumentation system (Section 7.5). This action will assure the isolation of any steam system leak common to all three )

loops and also provide protection against a major steam condenser leak during a steam bypass heat removal operation.

7.4.2.1.3 JnJfjating CJrruJfs The OSIS is initiated by the SGMRS Initiation signal described in 7.4.1.1.3.

This initiation signal closes the superheater outiet Isolation valves in alI 3 loops when a High Steam-to-Feedwater Flow Ratio or a Low Steam Drum Water level occurs in any loop. In each Steam Generator System loop, the trip signal s for High Steam-to-Feedwater Flow Ratio and the Low Steam Drum Water level are input to a two of three logic network. If two of three tri p si gnal s occur in any of the 3 loops, SGMRS is initiated, and all 3 loops are isolated f rom the main superheated steam system by closure of the superheater outlet isol ation val ves.

7.4.2.1.4 By.passss and Interlocks Control Interlocks and operator overrides associated with the operation of the superheater outlet isolation valves have not been completely def ined.

Bypass of OSIS may be required to allow use of the main steam bypass and condenser for reactor heat removal. In case the OSIS is initiated by a leak

! In the f eedwater supply system, the operator may decide to override the closure of certain superheater outlet isolation valves.

7.4.2.1.5 Redundanev and Diversity Redundancy is provided within the initiating circuits of OSIS. The primary trip f unction takes place when a high steam-to-feedwater flow ratio is sensed by two of three redundant subsystems on any one level sensed by two of three

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l 7.4-7 Amend. 69 July 1982

Page - 2 [8,22] 87 redundant channels in any one loop provides a backup trip f unction.

Additional redundance is provided by three independnt SGS steam supply loops serving one common turbine header. Any major break in the high pressure steam system external from the Individual loop check valves will be sensed as a steam feedwater flow ratio trip signal in all three loops.

7.4.2.1.6 Actuated Device The superheater outlet isolation and superheater bypass valves utilize a high rel iabil ity el ectro-hydraul ic actuator. These valves are designed to f all closed upon Ioss of electrical supply to the control solenold.

7.4.2.1.7 .SapACAtlDD The OSIS Instrumentation and Control System, as part of the Decay Heat Removal e-l

• Is designed to maintain required Isolation and separation between redundant channels (see Section 7.1.2).

7.4.2.1.8 .Qperator Informatlos Indication of the superhocter outlet isolation valve position is supplied to the control room. Indicator lamps are used for open-close position Indication to the plant operator.

7.4.2.2 Design Analysis To provide a high degree of assurance that the OSIS will operate when necessary, and in time to provide adequate isolation, the power for the system is taken from energy sources of high reliability which are readily available.

As a saf ety related system, the Instrumentation and controls critical to OSIS operation are subject to the safety criterla identified in Section 7.1.2. t Redundant monitering and control equipment wIII be provided to ensure that a single f ailure will not impair the capability of the OSIS instrumentation and Control System to perf orm its intended safety function. The system will be designed for f all saf e operation and control equipment, where practical, will assure a f ailed position consistent with its intended saf ety function.

7.4.3 Remote Shutdown Svstem A Remote Shutdown System is provided, it consists of the following provisions:

7.4-8 Amend. 69 July 1982 m%25m _ - _ _ _ _ _ _ _ _ _ _ _ .

Page 5 (82-0321) [8,22] #61 i

10.3 MAIN STEAM SUPPLY SYSTEM The Main Steam Supply System is shown in Figure 10.3.1. ,

10.3.1 DasIgn Rates The Main Steam Supply System includes steam piping and components downstream of itie steam piping anchor at the steam generator building penetration and conveys superheated steam fra each of the three steam generator loops to the high pressure turbine. Each steam generator loop is desiped to f urnish approximately 1,110,000 pounds per hour of 1535 psig, 906 F steam at the l superheater outlei nozzle.

The portion of the Main Steam Supply System downstream of the piping anchor at the steam generator building penetration up to the turbine stop valves and including the turbine by-pass piping is designed to MSI B31.1. The piping and component upstream of that point are safety related and designed in accordance with ASE Code Section lli as discussed in Section 5.5. Piping downstream of the turbine by-pass valves and the isolation valve for the steam seal regul ator are designed in accordance with ANSI B31.1, or manuf acturer's standard. This portion of the system has no safety function, accordingly, no special precautions have been taken for protection f ran environmental ef f ects.

A turbine by-pass system bypasses up to 80 percent of the rated steam flow (975 Mit) directly fran the main steam header to the condenser and the deaerator.

No saf ety-related equipment is located in the turbine building. Therefore, a maln steam iIne broak cannot jeopardize any safety-related equipment. The ventilation system for the turbine generator building is not saf ety-related and ef fluent resulting f ran a main steam line break will not af fect the HVAC system for any vital area.

10.3.2 Deserlotion Three separate 1Ines convey the superheated steam f rom the three steam genera 1or loops to the main steam header. Following temperature and pressure equalization in the main steam header, the steam is carried to the turbine by four parallel pipes. Each of these pipes contains a stop valve and a turbine governor controf val ve.

The turbine bypass is connected to the main steam header located bef ore the turbine stop valves. Figure 10.3-1 shows a diagrammatic arrangement of the Main Steam Piping System.

10.3-1 Amend. 69 July 1982

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15.3.3 Extrempiv Unlikelv Events 15.3.3.1 .Sf. gam or 3F 3d Line_fjpe Break 15.3.3.1.1 JIlerfj.flsat.lon of Causgs and Acejdent De3rrJRfjDD The breakage of a steam or feed pipe in the steam generator system is considered an extremely unlikely event. If such a break should occur, the resulting accident might have one of several forms, depending on where the break is located in the system, its size and whether or not it is insolatable.

It should be noted that a reactor trip by the Plant Protection System will shut down the reactor bef ore any of the steam system temperature changes have been transported back to the reactor oore (at pony motor speed approximately 150 seconds) henco no problem results with immediate reactor saf ety. The event instead is considered in the plant design for its ef f ect on plant

- - ' service l if e through ther.T.;l tr;c.;i c..1- induced stress.

The plant has incorporated design features to protect against the steam line break. For instance the Superheater Outlet isolation Yalve and Superheater Bypass Valve in each loop are active valves and will close within 3 seconds f oll owing a steam line break. Closing of these valves in the f ailed loop will prevent blowdown of more than one loop through the postulated pipe break.

The valves in the f ailed l oop will close by either a Low Superheater Outlet Pressure (< 1100 psig) or a High Steam /Feedwater Flow Mismatch. When a high steam /feedwater flow ratio occurs, the Superheater Outlet isolation Val ves and Superheater Bypass Valves in the other two loops will close. A detai l ed

. description of the Outlet Steam isolation Subsystem (OSIS) is presented in Section 7.4.2. The superheater Outlet Check Valve provides additional back-up to prevent blowdown but is not relied upon in any analysis. The Superheater Bypass Valve is normally closed during operation, in the event of f ail ure of an active val ve to close, the Superheater Outlet and Bypass Valves in the other two loops preclude their blowdown.

Breaks at the following locations have been Investigated:

a. Mair. steam l'Ine rupture.

b. Steam line from a superheater to the main steam header

c. Saturated steam line between the steam drum and the superheater.

d. Feedl ine break

e. Recirculation lir.e break The saturated steam line break has been selected as the most severe thermal transients of the events presented above. Analysis results for this event are presented in Figure 15.3.3.1-1. All of the above cases are swanarized as f ol l ows:

15.3-38 Amend. 69 July 1982

Pege - 2 [82-0321] 8,22 #90 Main steam iIne rupture:

A steam break at the main steam header would, if not isolated, produce a severe cold leg temperature transient in all three loops consisting of a down transient due to initial excess cooling followed by an up-transient after dryout, it is not plausible, however, to assume that isolation woul d f all to occur in all three loops, hence for case (a) actanatic isolation was assumed at three seconds with isolation initiated by the Plant Protection System (PPS).

15.3-38a knend. 69 July 1982

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Once the superheater outlet isolation valves close, the plant achieves a l new operating point based on steam load through the safety valves and hence no other excessive plant temperatures are produced. As noted bel ow, l a reactor shutdown is Initiated by the PPS based on either the primary shutdown system (steam / feed flow mismatch) or secondary system- (Low Drum Level), terminating high power operation bef ore excessive loss.of eater inventory. Elther the high steam-to-feedwater flow ratio or the Lot Steam Drum Water Level Trip al so activates the steam generator auxil!ary heat removal system (SGAHRS) as noted below and discussed in Section 5.6. All three loops would provide heat removal from the core. With the superheat steam line isolated, pressure In the steam system will build up to the rel ief setpoint. The drum water level will drop due to steam venting and the low steam drum water- level trip will then activate SGAHRS If it has l not been activated earlier in the transient by the High Steam to Feedwater Fl ow Ratio.

Rupture in a Steam Line Between 'a Superheater and the Main Steam Header:

This event results f rom a break occurring in the superheater exit steam l ine upstream of the Isolation val ve. A similar event follows from a break downstream of the Isolation va.lve (including a break in the main steam line) If the Isolation valve f all s to close. For these cases.

Isolation can still be ef fectively accomplished by the superheater inlet Isolation valve, either by manual Initiation or automatically when steam drum pressure f alls bel ow 500 psig. Consequently, a break in the superheater-to-header line has an ef f ect similar to the preceding main steam line break case, but its ef f ects are limited to a single loop.

Saturated Steam Line Break:

In the saturated steam line break, case (c) above, the break may be located such that loss of water in the af f ected steam drum cannot be prevented. Isolation valves on the modules could still be closed, but saf ety val ve outflow will still lead to module dryout. Consequently, no credit is taken for isolation in these cases.

As steam is remo'ved from the system by the break, increased flashing of water into steam within the steam generator occurs, renoving additional heat and causing the sodlum tanperature initially to decrease at the evaporator exit. A plant shutdown, when Initiated by low steam feed flow, l will cause coastdown of the intermediate sodium pump, and hence will amplify the initial decrease in evaporator exit tanperature.

Subsequently, when most of the moisture has been discharged f rom the steam generator, both evaporators and superheater will dry out, and the evaporator exit sodium temperature will increase to approach the intermediate hot leg temperature. The cold leg f anperature increase will eventually be transported back to the reactor inlet, af ter being conducted through the IHX of the af f ected l oop. Due to extended transport delays at pony motor flowrates, the temperature increase -

15.3-39 l

. Amend. 69

July 1982

Page - 4 [82-0321] 8,22 #90 l An al ternate 1ocation for this breek is at tha extt of one avaporator modu l e. Cl osure of the other isolation val ves, including the inlet valve on the af fected module, would lead to a dryout of the generator simil ar to previous cases. If the inlet isolation valve on the module does not close, the contents of the drum woul d be dumped through the af f ected module, producing a severe temperature down-transient on that module. The remaining module will dry out and its resulting increase in sodium exit temperatura wIII mlx wIth that f rom the f aulted module to attenuate the not intermediate cold leg temperature transient.

For the steam and f eed break cases, the following conditions have been applied to assure a conservative analysis:

a. The largest possible break size is assumed, corresponding to the f ull guillotine severance of the pipe involved.

b. The earliest PPS trip is used to predict the largest span for the sodium temperature transient for cases in which the intermediate cold

! leg temperature is considered.

c. The transients were run f rom a starting point at the 1121 Wt reactor power design condition (stretch power).

d. Credit has not been taken for heat storage in shell and structural l

metal in active or unheated parts of the modules in mitigating the

! thermal transients. Credit was taken only for 75% of the tube metal in the heated part of the modules.

e. No isolation was perf ormed on the af fected unit during the drum to superheater break, feed break and recirculation line break cases and the steam generator was allowed to go to f ull dryout.

The action of the Plant Protection System (PPS) in the above casos is the following:

Primary Shutdown System

a. Reactor and pl ant tri p - steam-feedwater fI ow ratio Secondary Shutdown System l

a. Reactor and plant trip - hlgh evaporator outiet temperature 15.3-41 Amend. 69 l

July 1982 L -

Page 7 (82-0321) [8,22] #63 Questien CSdS(L2d The W-2 test is a slow, overpowered test (about 5 cents /s ramp rate) conducted on f ull length FFTF gecrnetry fuel pins in the Sodium Loop Safety Facility (SLSF) by the Hanford Engineering Development Laboratory (HEDL). It has not been f ully examined or analyzed; nevertheless, the test has several' important l ImpIIcations f or the CRBR.

First, there is the puzzle of very early cladding breaches, possibly as early as ten seconds into the transient, and w!th a breach definitely confirmed at about 15 seconds into the transient. These early f ailures were unexpected because of the low fluence that had been accumulated by the cladding.

Second, gross f uel expulsion occured about as predicted by all of the

! prediction methods (as to time) at about 22 seconds Into the transient.

However, the site of the expulsion was apparently at axial midplane, which was unexpected.

i Third, it is speculated that the site of expulsion may have been influenced by

, the early failure, which is presumed to have occured at midplane, l

The applicant is requested to comment on: 1) the implications of the early t

cl adding breaches with respect to the adequacy of perf ormance evaluation

! models in cladding f ailure criteria being used for the CRBR, and 2) the implications of the midplane site of the f uel expulsion and of the influence l the early failure may have had on the location of the site, for beyond-design energeti cs.

Response

1) Evaluation of the W-2 Sodium Loop Safety Facility (SLSF) test was not intended to be used by the CRBRP as a primary requisite to test the validity of the CRBRP methodology in predicting incipient f ailure threshold (time). Since the completion of the test, considerable ef fort has been expended by the safety community reviewing the test results, however, complete test examination and Interpretation of test instru-mentation has not been reported in the open iIterature, although a preliminary data report is available. Once the W-2 test has been f ully examined and it can be determined that the test will provide a usef ul benchmark relative to predicting cladding breach initiation, the incipient f ailure threshold time can be evaluated using the CRBRP methodology. A schedule for the release of the available testing information will be provided to NRC by July 31, 1982.

2) The TOP event with fuel expulsion at the core midplane has been enalyzed

, extensively, and the resul ts are documented in Ref erence QCS490.24-1. The l analyses have shown that the midplane f uel expulsion would not result in a sustained superprompt critical excursion, whether fission gas or f uel QCS490.24-1 Amend. 69 May 1982

F LPcgs LTTMR%2NTDGEDFUiM vapor pressure causes the fuel expulsion. Ref erence QCS490.24-1 al so contains the results of an alternative SAS/FCI analysis to provide insite into the margin available. This less rigorous analysis assumed the superprompt critical excursion based on SAS/FCI calculations at near prompt critical, despite the f act the SAS/FCI calculations at such conditions were considered unrealistically conservative (Reference QCS490.24-2). The resulting work energy was calculated to be 33 MJ at sodium impact with the reactor head, which is well below the SMBDB value at 101 MJ.

The NRC question speculated that the site of fuel expulsion in the W-2 test may have been influenced by the early cladding breach which was not predicted by current analytical models. This implies that the fuel expulsion site may not be determined accurately within the current models. To address the implication, PLUT02 calculations have been performed to confirm that the midplane fuel expulsion, which has been analyzed as mentioned above, is the most energetic case. The results of these PLUT02 calculations are plotted in Figure QCS490.24-1. Examination of Figure QCS490.24-1 shows that the midplane fuel expulsion yields essentially the highest peak positive reactivity feedback from fuel motion. Theref ore, it can be said that an early cladding breach may cause at worst f uel expulsion at the midplane, which has been analyzed from the standpoint of the whole core response (Reference QCS490.24-1).

References:

QCS490.24-1 S. K. Rhow, et al ., "An Assessment of HCDA Energetics In the CRBRP Heterogeneous Reactor Core, "CRBRP-GEFR-00523, _ December 1981.

QCS490.24-2 J. L. McElroy, et al, "An Analysis of Hypothetical Core Disruptive Events in the Clinch River Breeder Reactor Plant,"

CRBRP-GEFR-00103, General Electric Co., April 1978.

l QCS490.24-2 Amend. 69 May 1982 82-03355

Pcge 9 (82-0321) L8,22J #63 e

Figure QCS490.24-1 PLUT02 Prediction of Peak Fuel Motion Reactivity vs. Fallure Location 3 , , , ,

Reactivity 2 -

h -

per Assembly, g

. O -

O b i i , i A

-30 -10 0 10 20 Distance from Midplane, cm QCS490.24-3 Amend. 69 July 1982 m* __