ML20054K642

Forwards Response to Questions CS 760.7,760.9,760.38,760.57, 760.58,760.86,760.112 & 760.136 Transmitted in NRC 820430 Request for Addl Info.Encl Responses Will Be Incorporated Into PSAR Amend 69 to Be Submitted in July l982

ML20054K642
Person / Time
Site:	Clinch River
Issue date:	06/29/1982
From:	Longenecker J ENERGY, DEPT. OF
To:	Check P Office of Nuclear Reactor Regulation
References
HQ:S:82:059, HQ:S:82:59, NUDOCS 8207060006
Download: ML20054K642 (18)

Text

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Department of Energy Washington, D.C. 20545 Docket No. 50-537 HQ:S:82:059 JUN 2 91982 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 Information," dated April 30, 1982 This letter fonnally responds to your request for additional information contained in the reference letter.

Enclosed are responses to Questions CS 760.7, 9, 38, 57, 58, 86, 112, and 136 which will also be incorporated into the PSAR Amendment 69; scheduled for submittal in July.

Sincerely, VL .

J n R. Longe. ker Acting Director, Office of the Clinch River Breeder Reactor Plant Project Office of Nuclear Energy .

Enclosures cc: Service List Standard Distribution Licensing Distribution 1

8207060006 820629 l PDR ADOCK 05000537 l A PDR

, . Pcge 1 (82-0374) [8,22] #99 Ouestion CS760.7 Are there any items that should be addressed or have been Included in the CRBR design resulting from the lessons learned from the TMI accident?

Resnonse The lessons learned f rom TMI have been caref ully reviewed and appropriate actions have been initiated. Lessons learned from the TMI accident bcIng addressed for the CP, and their inclusion in the CRBRP design, are discussed in Appendix H to the PSAR.

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l QCS760.7-1 Amend. 69 Julp _1982

Pcg3 3 (82-0374) [8,22] #99 Ouestion CS760.9 Tabl e 15.1.3-1, page 15.1-94 in the PS AR, lists PPS trip levels and/or equati ons. Several of the terms are not defined, and no Indication of units is given f or those that are defined. Please supply definitions for all terms including units. Please also indicate the level for each that leads to the latest trip. Provide the impact of the new settings on the frequency of occurrence of eve 7ts. Do these changes impact the duty cycles used f or plant design?

Resoonse The variables of Table 15.1.3-1, page 15.1-94, are def ined on page 15.1-95.

The term "S" represents the Laplace Operator. A variable expressed as a f unction of S is written in the f requency domain.

The units on single trip level subsystems are defined in the Table. For multiple trip level subsystems whose performance is given by a trip equation, the variables are normalized (i.e., 1.0 = 1005).

For single trip level subsystems, the Table Indicates the latest trip value.

For multiple trip level subsystems, the latest trip varies as defined by the trip equations. Each accident analyzed in the Section 15 Saf ety Analysis identifles the latest trip level assumed in the analysis.

The various trip f unctions with their respective trip settings are provided primarily for the PPS diversity and redundancy such that dif ferent types of f ault events are adequately protected against. They are not considered to have any impact on the f requency of occurrence of the events specified.

The Plant Duty Cycle is a set of conservative transients which provide an envelope f or plant operating conditions. Trip settings assumed in the Duty Cycle accident scenarios account for worst case PPS performance.

In general, trip settings are detennined such that changes to the duty cycle are not required. In a f ew Instances, redefinition of duty cycle events may j be necessary to accommodate limitations in the PPS trip subsystem such as j sensor accuracy or response time, l l

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l QCS760.9-1 Amend. 69 July 1982

Pcge 1 (82-0382) [8,22] #101 TABLE 15.1.3-1 PPS SUBSYSTEM TRIP LEVELS OR TRIP EQUATIONS Primary Shutdown System High Flux Trip at 115% power

. Flux-Rate Positive:

g -1 141.

1+28S

-0.99 (t) + 0.1706Np + 0.036410 L

Negative:

1.01p(t)-g-1 6 (s) i Idi- 0.1969Np 1, +0.041610 1+28 s .

Flux to Pressure 1.318 8 - 9 + 0.0425 10 Primary to Np (0.14710.0022) + 0.059510.0007 - AbsVai intennedi ate [Np (1 1 0.015) - Ni (1 0.015) + 0.007510.01] SO Speed Ratio HTS Pump Frequency Trip at 57 Hertz Reactor Vessel Level Trip when level drops 18" from normal operating level Steam to Feedwater Trip at 30% mismatch FIow Ratio IHX Primary Outlet Trip at 8300F Temperature Secondary Shutdown System Flux to Total Flow 1.2fFp - 0.99 9 +0.087 .10 Startup Flux Trip before 10% power Primary to Fp (0.147 1 0.0022) + 0.050 1 0.0007 - abs Val intermediate [Fp (1 1 0.015) - Fl (1 10.015) + 0.007510.01310 Flow Ratio Steen Drum Level Trl,p at 8" drop f rom f ull power steady state level 15.1-94 Amend. 69 June 1982

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Page 2 (82-0382) [8,22] #101 TABLE 15.1.3-1 (Continued)

High Evaporator Outlet Temperature Trip at 750 F Sodium Water Reaction Trip initiated within 3.0 seconds HTS Pump Voltage Trip at 75% of rated voltage Definition of Variables j[I =I Laplace Operator P = Reactor inlet Plenum Pressure 0 = Reactor Flux F = Total Primary Pump Flow Np = Average Primary Pump Speed FP = Primary Pump Flow Np = Primary Pump Speed Fj P = Intermediate Pump Flow Nj = Intermediate Pump Speed The above variables are normalized such that their value at 1005 conditions =

1.0.

W 15.1-95 Amend. 69 July 1982

P:ge - 1 [82,0357] 8,22 #92 l

Ouestf on CS760.38 Provide the rationale for referring to the DHRS as a saf ety grade system when it also states that it must be " adjusted manually."

Response

DNRS is a saf ety grado system providing a fourth decay heat removal loop for CRBRP. DHRS is specified as a Saf ety Class System (see PSAR Section 3.2) with electrical equipment classified as 1E. Manually Initiating the DHRS does not degrade its saf ety grade status.

Manual Initiation of the Direct Heat Removal Servloe (DHRS) is appropriate based on the time period available and the number of operator actions requi red.

In the worst case transient analysis perf ormed f or the DHRS (extremely unlikely event of DHRS Initiation Following Reactor Shutdown f rom 100% Power with Loss of All Heat Transfer through the IHXs at time of reactor trip), it is assumed that no heat is transf erred to the DHRS for one-hal f hour af ter sh utdown. The heat capacity of the primary system is used as the principle heat sink during this period with no operator action required to assure this heat sink.. As shown in Section 5.6.2 of the PSAR, the temperatures associated with this event are acceptable.

Furthermore, the manual Initiation ref erred to actually consists of turning six switches in the Control Room on the DHRS panel from the normal to the DHRS position. This is an on-of f control rather than an adjustment. These control transfer switches and automatic sequences have been provided to automatically position DHRS valves and control the pumps and ABHXs In order to reduce the operator actions required. A conservative estimate of the time period required f or the control transfer switches and the automatic sequences to operate is twelve minutes; theref ore, the operator has an adequate time period to detennine the need for DHRS use and initiate DHRS.

Based on the time periods noted above and the number of operator actions required, it is judged acceptable f or DHRS Initiation to be perf ormed remote-manually by the operator in the Control Room.

QCS760.38-1 Amend. 69 July 1982

h$g$-8[82,0357]8,22#92 Ouest ion CS760.57 in Sections 15.7.3.1, Leak in a Core Component Pot (CCP), the T across the wall is shown to be about 15000F, Simple hand calculations show that the stainless steel wall can f all T's 15000F.

a. Demonstrate the wall structural Integrity for all cases considered.

b. Provided this f allure occurs, how does this alter the design and operationa! procedure?

c. With at least I hour per assembly for refueling, how long does it take to ref uel (in light of the new coro design)?

d. Demonstrate numerical and model accuracles of TAP-4F and DEAP computer Codes,

e. Why do clad temperatures keep Increasing af ter reaching the melting temperature? Does a phase change occur and how is it considered?

f. If the clad temperature stays at the melting point during the change of phase .(in the case of stacked f uel pellets), there should be significantly less CCP temperature rise as compared to what is calculated. A similar situation applies to the packed bed case, except in this case there is significantly more thermal Interaction as compared to the stacked pellet case, and thus, due to contact conduction, CCP temperature must be higher than the previous one. Demonstrate as in item d above f or the conditions considered herein.

Resoonse The 15000F temperature dif ference to which the question is addressed is the dif f erence between the CCP and EVTM cold wal l temperatures. It is not the temperature gradient across the CCP wall. The question is apparently the result of a lack of clarity in PSAR Figure 15.7.3.1-5, in which the identification of the dif ferent regions in the thermal calculation is such that " cold wall" was interpreted as the cold side of the CCP wall rather than as th e EVTM col d wal l . The approximately 15000F temperature dif ference would be across the approximately 7-inch argon gas space between the CCP outer wall and the EVTM cold wall. It is the driving f orce f or radioactive heat transfer of core assembly decay heat to the cold wall limiting the f uel temperature ri se. The temperature dif ference across the CCP wall itself would be only about 100F. .

QCS760.57-1 Amend. 69 July 1982

'P;ga 7 WB2-0320 [8,22] 59 The following sections of the response discuss the corresponding parts of the --

question.

a. There is no question of structural Integrity with the gradient of 100F.

tne maximum radial thermal gradients across the CCP wall occur during its insertion into the EVST sodium pool at the maximum rate of 24 ft/ min af ter transf er f rom the reactor vessel. This case was analyzed with the CCP assumed to contain a maximum powered f uel assembly and to have been In the EVTM long enough to reach steady-state temperatures. The transient thermal gradients are high, with a peak of approximately 5000F, a f ew tenths of a second af ter Insertion; however, the duration of the gradient is short. The ASE Code structural analysis accounted for thermal stresses resulting from the design basis number of these immersion cycles and showed that the allowable number of these thermal cycles is greater than the design basis number.

b. No response required.

c. There are between 3 and 379 core assemblies exchanged per ref ueling, with an average of 159. At 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per transfer (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per exchange) between the EVST and the reactor vessel as assumed in the event analysis in PSAR Section 15.7.3.1, the time f or en average ref ueling would be 13-1/4 days plus time for setting down and picking up assemblies. The transit time used f or the event analysis is a conservative estimate which is longer than the expected ref ueling times. (These times cover only core assembly transfers; preparation of the reactor for ref ueling and termination from refueling are not included.)

d. The TAP-4F and DEAP computer codes are described in Appendix A of the l PSAR. Verification of the codes is covered in these descriptions.

e. Cladding temperatures keep increasing af ter reaching the melting point because the thermal analysis does not model the stainless steel phase ch ange. This is conservative since it results in reaching steady-state i temperatures more quickly. It does not af f ect prediction of steady-state l temperatures.

The ef fect of cladding melting was considered by running two additional cases which assume no cladding at all is present. The accident analysis in PSAR Section 15.7.3.1 covers the CCP loss of sodium event to its expected conclusion and two extensions of this event (first, to collection of f uel fragments in the f uel assembly duct, then to a hypothesized relocation of the f uel fragments to the bottom of the CCP). The temperature distribution aha f uel assembly configurations f or these cases are shown in PSAR Figures 15.7.3.1-4, 15.7.3.1-6, and 15.7.3.1-6, respectively.

f. The question of cladding melting does not arise in the second case of a packed bed since this case assumes no cladding is present and f uel pellets have collapsed into a packed bed f ollowing loss of cladding. For conservati sm, all cladding of the entire 3-ft high f uel region was assumed to be absent.

QCS760.57-2 Amend. 69 July twS2

. Pcge 8 WB2-0320 [8,22] 59 CCP temperature is determined, not by fuel temperature, but by the heat transfer resistance between the CCP welI and the heat sink, which is the EVTM col d wal l. This resistance does not change among the three cases except for a minor reduction In heat transfer area as the fuel region gets shorter as the event progresses f rom one case to the next. This reduction is reflected by the smalI increase in CCP temperature ( 1000F) from one case to the next.

Thermal Interaction within the CCP is modeled to be less (i.e., thermal resistance is greater) In the packed bed case because of high contact resistance between particles and higher resistance to radiant heat transfer than when the f uel is in wel1-ordered stacks. This was done purposely in order to obtain conservative estimates of fuel temperature to assure that f uel melting would not occur. As noted above, CT temperature is not af f acted by increases in heat transfer resistance within the CCP; these changes af feet ony fuel temperatures. This can be seen by noting the substantial fuel temperature increases f rom one case to the next.

This conservative sequence of cases demonstrates that fuel melting wilI not occur.

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QCS760.57-3 Amend. 69 July 1982

l Pago 9 W82-0320 [8,22] 59 Qunstion CS760.58 in 15.7.2.5, page 15.7-16a, please elaborate on the portion of the statement j that roads, " Loss of fluid due to solidification of the concentrated waste." -

This statement ref ers to what process.

Response

As discussed in Sections 11-2 and 11-5, liquid radioactive waste is concentrated in an evaporator system to an approximate concentration of 24% by weight. The evaporator bottoms (concentrated liquid radioactive waste) are then transf erred to the Radioactive Waste Solidification System for immobilization In coment. Since the evaporator bottoms retain a portion of the influent water (and therefore some tritium), the tritium activity in the decontaminated water is somewhat reduced. Tne subject statement in PSAR Section 15.7.2.5.2 has been reworded for clarification.

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QCS760.58-1 Amend. 69

! July 1982

p;go 1 WB2-0358 (8,15) 52 The activity levels In the liquid are given in Table 11.2-4 of Section 11.2 of the PS AR. There are no gaseous radioactive lodine species which can be released because the fluids used to renove contaminated sodlum from components f orm salts which are stable. Any radioactive inert gas which may have been trapped in the sodlum that is eventually reacted with water and processed by the Radweste System is negligible. This is true because the quantity of these gases dissolved in sodlum Is small. The spilled fluid contains fission and corrosion products which are not evaporated. Thus, only water vapor containing tritlated water (HTO) can be released in the event that a f ailure occurs.

15.7.2.5.2 Analysis of Effects and Consecuences Gaseous Release The highest activity resulting from a radweste system f ailure involves collection tank leakage or rupture. 100% of the average annual collection tank Inventory of 20,000 gallons of water contains 1.44 x 105 Cl of tritium as HTO. The bulld-up of tritium in the recycle liquid over the 30 year life of the plant is a function of: (1) Input f rom the primary sodium removal system, (2) radioactive decay, (3) retention of a portion of the Influent in the evaporation bottoms which are transferred to the solid waste system for Immobilization, and (4) the release of a fraction of the storage tank Inventory to the cooling tower water blowdown. The value of 1.44 x 105 was conservatively estimated by using a loss of only 4700 gallons per year out of the 40,000 gallons of storage capacity.

A conservative analysis was made to calculate the of f-site doses if 10% of the tritium contained in the spilled liquid radwaste was released to the atmosphere in two hours f ollowing the spill. This highly conservative assumption resulted in a Beta Skin Dose of 4.47 x 10-5 REM and a whole body inhalation dose of 3.7 x 10-6 REM, at the site boundary. The potential beta skin and whole body doses at the LPZ are 0.68 x 10-9 REM and 3.05 x 10-7 REM, respectively.

Llauld Release For conservatism, the event has been analyzed assuming no credit f or the fIoor, drelns or operator actions.

As pointed out In Section 2.4.13, accidental IIquid spflis are not seen as posing a danger to present or f uture groundwater users in that the ultimate destination of contaninants in the groundwater would be the Clinch River.

Movement of groundwater is f rom groundwater ridges to adjacent groundwater l ow s. Review of Figures 2.4-68 and 2.4-69 lends support to the assumption made of the cooling toeer blowdown discharge point as a conservative assumption (In terms of l

15.7-16a Amend. 69 July 1982

Pcg3 21 WB2-0320 [8,22] 96 Question CS760.86 Core Structure Cooling - 6% of the total coolant flow is bypassed to cool control assemblies, radial shleid assemblies, the reactor thermal liner and to account f or bypass and leakage f low. In optimizing the reactor perf ormance, the bypass flow allocation has only a minimal Impact on core perf ormance. The bypass flow allocation can be Increased by 20% and the coolant flow to core and blanket assemblies is only reduced by 1%. However, af ter the design is "f rozen" these flow allocations must be such that they provide suf ficient cooling to the structures under f ull flow as well as natural convection conditions.

While the bypass flow can be f ully suf ficient under full power operating conditions, there is concern that there is also suf ficient flow under low power - low flow conditions, especially during natural circulation transients.

There is always a problem to control low flow rates. For exenple, under f ull power operating conditions, the radial shield receives a total of 1.35% of total flow, which enounts to approximately 0.5 lb/sec of sodium flow per assembly. Under natural convection conditions, any flow through these assemblies will be Invisible. Even at reduced power operation, it Is not clear how much flow these shield assemblies will see. The flow path to get to the shleid assemblies, is rather complicated. First, the flowing sodium enters the l ower i nl et f l ow modul e. The leakage flow goes up the hydraulic balance bleed hole and enters the lower piece of the bypass flow module. From there it goes up and goes through the socket f or the RRS finally into the RRS.

For this flow to proceed this f ar, the LIM has to fit into the core support plate. Inside the core support structure are holes which allow the sodium to flow from the LIM to the BPFM. In the BPFM are holes which have to line up with the sockets f or the RRS. The tolerances in alI of those complex fittings wil l af f act the f low which goes into the RRS.

This exanple Illustrates the problem in core structure cooling at both full power and convection flow conditions. The f ollowing questions need to be answered:

1. What is the accuracy in the prediction of the bypass flow distribution et f ull power and low power conditions?

2. What are the temperature limits f or the core structures which determine the required coolant flows?

3. Is there a possibility to control the flow to the various core structures?

4. What margins are built into the bypass coolant flows? '

5. What happens if there is no flow through the RRS?

6. Can the bypass flows be measured directly?

7. Since there is no gamma heating at BOL in the RRS, what are the assurances that they will receive suf ficient cooling?

8. Is it possible that there is a local overheating of the vessel because of lack of cooling of the RRS? .

QCS760.86-1 Amend. 69 l

- July 1982 l __

,Pago 22 C 2-0320 [8,22] 98 Resoonse Before addressing the specific questions, some discussion of general conditions appears necessary to clarify the Project approach to assuring adequate bypass flow over the entire operating range including natural ci rcul ati on, o The flow to the RRS, reactor vessel / liner and other peripheral components is controlled through orifices located either in the LIM's or In the Bypass Flow Modules. Scme test data have been obtained for the flow entrol orifices to characterize the bypass flow.

o To date, bypass flow elevations have been done at low steady-state forced flow conditions as part of the flow management calculations.

o Regarding transients, the verlfled C6 BRA-WC code has the capability to dynamically model RRS and other low flow paths simultaneously with the high flow fuel and blanket assemblies over the entire operating range as shown during simulation of the FFTF natural circulation tests. For example, at

-3% total reactor flow, COBRA correctly predicted the flow split between the Row 8 reflector and driver assemblies.

o From the design standpoint, at low flow conditions the components cooled by bypass flow have very Ilttle heat generation, and any increase in temperature following flow reduction enhances the natural circulation phenomenon.

Following are responses to the specific questions asked:

1. The flow accuracy requirmnents have been established in the Individual component design specifications; these translate directly into the uncertainties required for the flow control devices. This approach can be taken because the general orifice characteristics over the range of flow conditions are well known, and the desired specific characteristics can be designed into each device. It is recognized that the pressure drop ,

uncertainties are greater at low flows, and this is f actored into the design. The flow control mechanisms will be tested to verify that the requirements over the flow operating range have been met. Final uncertainty levels will be assessed following evaluation of experimental data (as discussed in the response to Question CS760.77). The results of these evaluations, together with the entire CRBRP reactor flow dlMribution network as-finalized, will be reported in the FSAR. <

2. The steady-state and transient thermal constraints of RRSAs are based on the cross-duct temperature gradients at the axial load pads as well as their absolute temperatures which are welI below any structural limits of i the 316 stainless steel and the bolling temperature of sodium. These analyses took into account the worst combination of uncertainties at the

+2r/-2w conf idence I evel .

The vessel flow was based on the requirenent that the vessel temperature remains under 9000F for nominal conditions and meets structural require-monts during transients.

QCS760.86-2 Amend. 69 July 1982

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Page 23 22-0320 [8,22] 98

3. Flow to the various core structuret is controlled by flow tested orifices in the Lower inlet hodules and the Bypass Flow Modules. The leakage flow in the flow path, particularly the leakage across the piston rings, is 1 taken into account. The Bypass Flow Module is fed by orifices which are l flow tested to accurately control the flow, not by leakage flow up the hydraulic balance bleed holes.

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4. 3 s' uncertainties are considered in tne CATFISH representation of the reactor flow network.

. 5. The RRSAs are cooled by two dif forent methods: directiy by orificed coolant flowing through the assemblies and Indirectly by Interstitial flow through the core. Therefore,- the case of "no f low" is not expected.

6. The bypass f low, as woll as.the flow to any other core component, cannot be measured directly in the reactor. However, out-of-pi l e -f l ow testi ng' of orifices has been or wilI be conducted to characterize the flows. -

7. If there is no heating, there shou'Id be no concern about_suf ficient cooling. It is not true that there is no gamma heating at BOL in the RRS.

Neutron radiation capture and inelastic scatter is practically constant throughout life. The only dif ference between BOL and EOL conditions is in the additional heating due to' fission product garnma from the core and neutron activation decay power, which is minor av f ull power conditions.

B. Vessel temperatures are Insensitive to RRS tenperatures since the vessel is cooled via a separate, para!'lel, controlled fIcv path. Actually, If less flow than designed went through the R9S, thic extra flow would be redistributed to the other flow paths in -the flod netwcek, including that of the vessel liner.

Finally, it should be reitereted that a detailed' description of the finel evaluation of the core siructures thennal-hydraulic design will be reported in th e FS AR.

l QCS760.86-3 Amend. 69 July 1982

prgo 1 162-0298 (8,22) 43 Ouestion CS760.112 What are the various setpoints for the Turbine Bypass system?

Response

See PSAR Section 10.3.2 for the requested Turbine Bypass System setpoints.

See PSAR Section 10.4.4 and revised Section 7.7.1.8 for description of the steam dump and bypass control system.

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l QCS760.112-1 Amend. 69 1 Ju_lp 1982

a ge 3 (82-0362) Ld,22J #79 Control of the argon supply and vent values is accomplished by an "on-of f" type pressure controller which cycles the supply and vent values to maintain the cover gas pressure between the lower and upper limits. Sufficient dead band is provided between lower and upper limit operation to prevent undue cycling of the supply and vent valves.

Ocarational Considerations The pressure controllers f or the sodium dump tanks are located at the local control panel in the Steam Generator Building. However, tmuel overrides f or the supply and vent valves are provided in the main control room and may be utilized at the plant operator's discretion.

High and low pressure alarms alert the operator to of f-normal conditions which may result f rom a mal f unction of the pressure control system. Pressure data is provided to the Data Handling and Display System and is available for display upon cal l by the operator.

7.7.1.8. Steam Dumo and Bvonss Control System The Steam Dump and Bypass Control System provides the necessary control and Instrumentation nerdware to operate the Turbine Bypass System as described in Section 10.4.4 and shown in Figure 10.3-1.

Redundant Interlocks are provided to prevent bypass operation in the event the condenser is unable to accept steam flow (e.g., high condenser back pressure or loss of circulating water flow).

At reactor poer levels above 40% rated power, the Turbine Bypass Control System is operated in a load error mode where bypass valves are opened proportional to the dif ference between turbine demanded load and generated power. A velve position signal Is provided to the Turbine Bypass Control System by the Supervisory Control System which makes the comparison. The circultry includes a dead band with a 10% load setpoint.

A pressure control channel is provided for the regulation of main steam pressure following reactor trip, during decay heat removal operation and during turbine standby, loading and unloading operations. The pressure control mode is manually selected for reactor power levels below 40%.

At reactor power levels above 40%, the Steam Dump and Bypass Control System automatically positions bypass valve (s) to regulate bypass steam flow approximately proportional to reactor power, however; the pressure. control mode may be manually selected by operating personnel at any power level. The pressure control mode is automatically selected following a reactor scram and j a turbine trip condiflon.

7.7-11

- Amend. 69 July 1982

I Pcge - 3 (8,22) #95 l

i Question CS760.136 The list of design transients (Table 5.7-1) includes two-loop operational events yet two-loop conditions are not included in the heat transport system design. We feel that the Project's present position with regard to redundant heat removal capability is not consistent with two-loop operation. Any future two-loop operation may entall considerable design changes.

Response

Revised PSAR Section 1.1.1 describes the Project's approach to two-loop operation.

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QCS760.136-1 Amend. 69 July 1982

P g3 1 (82-0374) [8,22] #100 The plant is designed with three main coolant loops and the Intended mode of operation is that all three loops should be continuously in service.

While the system design is Intended to be capable of allowing for operation at power on two loops, the applicant is not requesting NRC review of this opera-tional mode. If, at some time in the future, the applicant considers that all

. safety requirements can be met under two-loop operatioin without significant additional design features, the applicant may elect to apply to NRC for a two-loop operation capability. This should not constrain the NRC review of the construction permit application.

The construction completion data for the plant was originally scheduled for September, 1961. Until current Congressional and Administration actions are canpleted, the reactor criticailty data cannot be re-scheduled.

1.1.2 Overvlaw of Safety Design Anoroach The design of the CRBRP is based on the defense-in-depth safety philosophy, commonly known as the Three Levels of Saf ety design approach. A summary of the design saf ety approach for the CRBRP is provided in Tables 1.1-1 and -2.

Level 1 Design The first level of saf ety provides reliable plant operation and prevention of accidents during normal operating conditions through the intrinsic features of the design, such as quality assurance, redundancy, maintainability, test-ability, inspectibility, and f all-saf e characteristics. The plant is designed not only to accommodate steady-state power conditions, but also to have adequate tolerance for normal operating transients, such as start-up, shut-down, and load-following. As a basic part of the CRBRP development program, a number of large-scale engineering proof tests are being performed to verify the design concepts. This testing process provides predictability of per-formance and, hence, saf ety through assurance of the use of proven methods.

materials, and technology.

Extensive pre-operational test programs will be conducted in the plant to assure conf ormance of components and systems to the established performance l requi rements. Key parameters will be monitored continuously or routinely and well-define surveillance, in-service inspection, and preventive maintenance programs will be carried out by a trained operating and maintenance staf f to provide assurance that as-built quality is maintained through the lif e of the plant.

Level 2 Design The second level of saf ety provides protection against Anticipated and Un-likely Faults (such as partial loss of flow, reactivity insertions, f ailure of parts of the control system, or f uel handling errors - Faults are defined in Table 1.1-1 A) which might occur in spite of the care taken in design, con-struction, and operation of the plant. This level of saf ety 1.1-3 Amend. 69 July 1982

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