ML20151X037

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Forwards Request for Addl Info Re Site,Plant Layout, Containment Structural Design,Instrumentation,Control,Power Sys,Plant Sys,Esfs & Accident Analysis
ML20151X037
Person / Time
Site: Diablo Canyon, 05000000
Issue date: 06/30/1967
From: Morris P
US ATOMIC ENERGY COMMISSION (AEC)
To: Peterson R
PACIFIC GAS & ELECTRIC CO.
Shared Package
ML20151W779 List:
References
FOIA-88-156 NUDOCS 8808250204
Download: ML20151X037 (19)
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%n .i mmv ama 'o JUN 3 01967 Docket No. 50-275 Pacific Cas and Electric Company 245 Market Street San Francisco, California 94106 Attention: Mr. Richard H. Peterson Vice President and General Counsel Centlemen:

This refers to your application dated January 16, 1967, for a construction permit and facility license which would authorize construction and operation of a nuclear power reactor at the Diablo Canyon site located in San Luis Obispo County, California.

We have requested additional information pertaining to the site, plant layout, containment structural design, instrumentation, control, and power systems by letters dated May 5,1967 and May 18, 1967. We also indicated that additional questions related to other plant design features would be submitted as our review progressed. These questions have now been formulated. Accord-ingly, you are requested to provide the information pertaining to plant systems, engineered safety features, and accident analyses listed in Enclosure (1).

In addition to the information requested in Enclosure (1), we also refer you to the report, dated June 15, 1967, of the Advisory Comittee on Reactor Safeguards (ACRS) on the Vermont Yankee Nuclear Power Station. In this report, which is Enclosure (2) to this letter, the ACRS made reference to several matters which it "believes are of significance for all large water-cooled reactors." These items are identified by an asterisk (*) in the report. Consequently, you 8808250204 080721 PDR FOIA ,,.

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. 2 JUN 3 0 567 Pacific Gas and Electric Co.

are also requested to review these matter,- in conjunction with your design and provide appropriate inforr -ion, as necessary, on each of them including any considerations for special pro-grams to obtain further information during the plant final design phase.

Your reply to all of these questions and matters should be submitted as an amendment to your application. We will be avail.

able as may be required to discuss and amplify the meaning of the questions.

Sincerely yours, OkW st r: u

s. Le, ce Peter A. Morris, Director Division of Reactor Licensing

Enclosures:

(1) Request for Additional Information (2) ACRS Report dated June 15, 1967 AIRMAIL Distribution:

AECPub.

SAN Doc. Room Doc.Rm.kk 4-Formal Suppl.

DRL Reading RPB 2 Reading Orig: KWoodard R. S. Boyd C. Henderson

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H. Steele bee W. B. Cottrell, ORNL DRL/)tT DR(

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RF/)UEST FOR ADDITIONAL INFORMATION_

PACIFIC CAS AND ELECTRIC COMPANY DIABLO CANYON REACTO_R DOCKET NO. 50-275 V. Plant Systens A. Core Physics

1. State the number of incore flux monitor positions and evaluate the ability to adequately demonstrate the nuclear peaking factors used for the design. Discuss the expected margin between operating and design peaking factors.
2. Provide an analysis of the ef fect of a misplaced fuel assembly on steady state and transient conditions in the core. Discuss the provisions to preclude this occurrence. How would this condition be detected?
3. Evaluate the reactivity ef fects which could occur in the event of movement of control rods or fuel elements due to seismic loadings.

4 Discuss the length of time during core life that the moderator reactivity coef ficient will be positive.

5. Provide an analysis at 'beginning of life' and 'near end of life,'

of the equilibrium xenon reactivity at full power, the maximum xsnon reactivity buildup following shutdown from full power, and the xenon reactivity rate of burn out on return to full power from maximum shutdown xenon. Discuss the ability of the dissolved boron and control rod systems individually to compensate for this reactivity burn-out.

B. Thermal-Hydraulic and Fuel Rod Design

1. Provide a detailed discussion of the experimental prograne which are in progress or will be accomplished prior to operation to demonstrate that the proposed core can be safely operated at a linear power generation rate of 18.9 Kw/f t and withstand levels of 21.3 Kw/f t without fuel failure during operating transients at the maximum expected fuel exposure.*

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  • Response may be in conjunction with ACRS report on Vermont Yankee dated June 15, 1967.

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2. Identify clad damage limits in terms of Kw/f t, clad temperature,

% strain, internal gas tressure (state the amount expected including bases for prediction), burnup, fuel centerline temper-ature, and extent of center fuel melting.*

3. Provide a sensitivity analysis in terms of flux depression fac-tors, thermal conductivity, and gap conductance values used to calculate fuel temperatures.
4. With regard to thermal margins, provide the following:

(a) A distribution curve showing the fraction of the core (or number of rods) operating at the various power levels for design and overpower conditions.

(b) A determination of the corresponding DNB ratios and number of fuel rods and the area that could expe-rience DNB, using the statistical W-3 DN3 correlation and the distribution in (a) above.

(c) A sensitivity analysis arbitrarily assuming certain errors in major parameters used in calculating the number of rods experiencing DN3. For example, calculate the number of rods with DNB, as a function of possible percentage errors in the DN3 correlation, power distributions, flow rates, and power levels.

5. What design features will be provided to enable the operator to detect the occurrence of fuel failure?

C. Reactor pressure Vessel ,

1. Discuss how small cracklike defects at the start of vessel service life will be accounted for in the fatigue c.alysis of the reac tor vessel.
2. State which of the vessel design transients listed in Table 4-5 results in stress amplitudes exceeding the endurance limit of the material.
3. Provide stress amplitudes for those transients listed in Table 4-5.

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4. List those areas of the reactor vessel where the :umulative fatigue usage factor exceeds 0.2. Es timate the percentage contribution to these usage factors from each type of transient cycle listed in Table 4-5.
5. Recent conclusions, drawn from the results of tests on irradiated and non-irradiated specimens indicate that neutron irradiation tends to reduce the low-cycle fatigue life. How will this effect be considered in the f atigue analysis of the portion of the vessel located between the top of the core and the reactor nozzles?
6. State if the material irradiation surveillance program will i comply with all the recommendations of the recently revised ASTM specification E 185-66T.
7. Describe the method which will be employed to determine neutron flux at the irradiation samples and estimate the accuracy of the proposed method.
8. Supply a drawing showing a vertical cross section of the reactor vessel, ins station and the primary shield. Particularly show the arrangement of these materials around the reactor vessel ,

nozzles, and indicate which parts are readily removable. Will there be clearance and provisions for water flow between the vessel and the insulation?

9. Please provide the design provisions that are to be implemented to provide local leakage detection capability in various regions of the primary system. Your reply should include:

(a) design criteria for leak detection; (b) potential leakage sources and provisions for detection in those areas; (c) a discussion of minimum leakage rate detection capability; and (d) monitoring stations and alarms for leakage detection systems.

4 D. With regard to thermal shock on primary system components upon rapid safety injection of cold water, provide the following:

1. The details of an analysis which indicate that the reactor vessel can accommodate the rapid temperature change without failure at the end of fatigue life.

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2. An estimate of the margin available for initial vessel temperature and pressure during the transient. Discuss possible mechanisms whereby the vessel wall temperature could be increased prior to injec tion. (Consider smaller breaks where the core is uncovered longer and wall heatup could ensue.)
3. An evaluation of the capability of the piping, safety injection nozzles, and vessel nozzles to withstand the transient.
4. An evaluation of the ability of the vessel nozzles, which provide support for the vessel, to withstand the Class I load combinations in addition to distortions caused by this postu-lated thermal transient.
5. An evaluation of the ef fects of this transient on the core barrel and other internals with regard to assuring that distortion would not restrict the floir path of the enerfeucy core coolant.

E. With regard to the rod drive housings located on the vessel head, j proviets the fol'owing:

1. A vertical section drawing showing the head adapters, tha pres-sure housings , magnetic latch assemblies , rod travel Soesing, and all restraining spacers.
2. The code and class to which these pressure housings are designed.
3. A detailed evaluation of the damage (including stresses and deformation) which could occur in adjacent rod pressure housings in the event of a circumferential or longitudinal failure of a rod housing. Sta:e all assumptions including the ef fect of restraining devices, jet forces, and collision of the failed housing with parts of the adjacent housing. Assume that the micarta sleeve does not remain intact and consider the rod housings at the outer periphery, a

F. Steam Generators

1. Provide criteria for the stops or other restraining devices which will be used to prevent excessive movement of the steam generators during seismic loading. Analyse the forces on the shell, tubes, and tube sheet upon striking these stops for the maximum earthquake.

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2. To what extent is the "sloshing" of water in the generators considered in the seismic design?
3. Provide a thorough discussion substantiating your belief that the occurrence of multiple tube ruptures at any time during the lifetime of the plant is highly improbable. The ef fects of corrosion should be considered.
4. Provide the approximate volume and energy content of a steam generator operating at various power levels.

G. Net Load Reieccion

1. Describe the equipment provided in the s team and turbine systems to enable capability to safely maintain the reactor at critical af ter a net load rejection. State the amount of s team released to the atmosphere during the transient.
2. Describe what information from sensors in the primary and second-ary systems will be used for controlling the steam dump valves.
3. What is the design leakage rate (total) for all valves located outside containment upstream of the steam line isolation valves?

M. Containment Leak Rate Testing

1. Can an integrated leak rate test be performed on the containment at design pressure at any time in plant lifet
2. When the leak test is to be performed only on the liner seas chan-nels and penetration pressurisation system, describe in detail the methods to be used to determine that leakage paths do not exist through isolation valves and other barriers which must be included in the 0.1% per day leakage specification at design pressure.
3. How will you determine that the lines leading to the seas channels and penetrations have not become blocked by accumu-lation of condensed water, debris or by other meanst 9
4. State your criteria with regard to the integrity of each of the double barriers (other than the seam channels and pressurized penetrations); i.e. , if one barrier were to f ail, would the second provide the appropriate leakage protection. If this is the case, how will each barrier be tested independently of the other?
1. Containment Isolation System
1. Provide a diagram for each penetration showing all valves and barriers or. both sides of the liner. For each case locate (a) missile shielding, (b) parts of the system which are not Class I, and (c) test connections for demonstrating that the two barriers (valves or closed systems) are leak tight.
2. For the valves shown above, list (a) the type of valve including leakage characteristics at accident pressures, (b) the control system type, (c) the signal to operate the valve, (d) the power source required to actuate or operate the valve, (e) the time to close the valve, (f) the normal position of the valve (state if locked) and (g) the failure position.
3. State the criterion for radiation shielding which will be provided for manual isolation valves in lines which could contain fission products or are located in the auxiliary building or other poten-tially contaminated areas.

J. We believe that the following criteria pertaining to certain contain-ment piping penetrations should be applied to the design of your facility. These criteria are:

"1. Lines which penetrate containment and are open to the external atmosphere or to systems designed for less than containment design pressure shall be protected by redundant, autcTatic*

isolation valves if they fulfill any of the following conditions:

(a) They are connected to the primary system.

(b) They are normally open to containment atmosphere.

7 (c) They connect to closed sys tems within the containment which are not designed to withs tand accident forces.

Exception: Lines which must remain open subsequent to MCA shall be protected by redundant valves, one or both of which shall be remote-manual.

2. Ventilation lines shall be isolated upon receipt of signals which initiate emergency core cooling.
3. Lines which have a low probability of rupture during MCA (e.g. , certain secondary system lines) shall be protected by at least one automatic valve external to containment.

Exceptiont Lines which must remain open subsequent to MCA shall be protected by one automatic valve or one remote-manual valve external to containment.

4. Circuits which control redundant automatic valves shall be redun-dant in the sense that no single failure shall preclude isolation."

We would appreciate your comments on the above considerations, and a description of any modifications to the valve arrangements, that would be proposed as a result of considering these criteria.

VI. gnaineered Safety Systems A. General _

1. State your design and performance criteria for the emergency core cooling system with regard to core geometry, clad melting, and metal water reaction for the range of primary system piping break sites considering all break locations. Has the ultimate reactor power level been considered in sizing all systems?
2. Provide a performance bar chart showing the extent of coverage to protect the core for any design basis accident provided by various core cooling systems.

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i B. Accumulator Design

1. Provide drawings showing the location of the system piping, vessels, and valves within the containment. Locate the connection to the deluge system pipe.

4 2. What design consideration is being given to operation with an

accumulator out of service or with the stop valves closed?

Describe the stop valves including the type, reliability, and time to-open characterisites.

3. Provide the free volume of the reactor vessel for various water levels with reference to the core.

C. Sa fety In f ec tion and Core Deluge Sys tems Design

1. Please provide a discussion to substantiate the adequacy of the single refueling water storage tank. What consideration has been given to providing redundancy by utilizing an alternate storage tank and piping?
2. State the approximate boron concentration of the water which will

' be stored in the accumelators and refueling veter storage tank.

Plot the concentration (in ppm) of boron vs. percent shutdown (no rods in) for various coolant temperatures from 250 F to 50'T for the worst time in core life. List the water volumes of the primary j

system and each of the steam generators.

3. Discuss the type of coating to be used on surfaces in the contain-ment to prevent dilution of boron by chemical action between boric acid and various materials (such as steel or concrete).

D. External Recirculation Coolina System

1. Provide the design criteria and considerations for the containment sump in terms of debris size which could pass through, Relate this to the spray nozzle diameter and pump requirements.
2. Provide the preliminary design (including elevations) for the pit in which the residual heat removal pumps are le sted. State the atting of the sump pumps in this pit and reists to the leak rate which could occur in the event of a pump seal failure. Where j

would overflow from the pit and the discharge from the sump pump 1 be routed?

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3. Locate all piping and components associated with recirculation

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and safety injection systems on a drawing of the facility plot '

plan.

4. What volume of borated water must be injected into the contain-ment before recirculation can be initiated?
5. What flexibility for the maintenance of safety components is allowed for the designt Would redundancy of function be available for all safety systems which could require maintenance during '

reactor operationt

6. Describe the ability to manually isolate the tube or shell side  ;

of the resudual heat exchangers or the residual heat removat pump inlet or outlet af ter an accident (include radiological effects and time).

> 7. Discuss the provisions, in the vicinity of the recirculation loop, for leakass detection and state the ability to remotely isolate a '

leak from any portion of the system.

8. Provide a curve of dose at the site boundary vs. leakage from the recirculation loop which can be accommodated (af ter an accident) by the ventilation systems. List all assumptiors and include experimental backup concerning the ability of fission products to become airborne from the spilled recirculation water. Describe i

the location of, and shielding for, the filter banks in the auxi- -

11ary building ventilation system.

E. Component Coolina Loon ,

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1. Please submit a revised drawing with regard to changes made to 4

Figure 9 4 showing all components serviced by the component i

cooling loop and indicate what portions of this system would be

located outside of the auxiliary building and discuss provisions for isolation of connecting lines. l
2. Provide the volume of the surge tank, the normal free volume above  ;

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the water in the tank, and the approximate volume of the loop.

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3. Discuss design provisions to monitor for leakage to the salt water [

system. What size leak could be detected and would be acceptable

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4 If the residual heat removal loop develops a considerable leak to the component cooling loop, where and how is the excess water in the component loop routed to prevent contamination of poten-tially inhabited areast

5. In view of the importance of core and containment cooling, discuss the capability and provisions made in the design for providing emergency backup cooling water to the residual heat exchangers and fan-coolers, such as from the salt water system.
6. Which components inside the containment will not be automatically isolated from the loop after a LOCA accident? Discuss effects of failure inside containment and ability to isolate.
7. At what rate can water be made up to the surge tank? Compare this rate with the leakage which could occur due to failure of lines considered in 6 above and include makeup source and emergency power availability.
8. Provide the flow rates in each component required for the accident loads and indicate the size and number of pumps necessary to meet these requirements.

F. Provide changes made to the service water system design and include approximate discussion of safety significance.

G. Auxiliary Salt Water Systes

1. We understand that changes have been made to the salt water cooling system. Pelase describe these changes and provide an updated Figure 9-7.
2. Provide the preliminary pump characteristics for the salt water pumps. How many must operate during an accident?

H. Fan-Cooler Design

1. Describe the cooling coils to be used and provile a detailed discussion of the heat transfer model used to determine that the coolers will be capable of removing the design amount of heat from the containment accident environment. .
2. Provide a detailed quantitative discussion of the ef fect that a '

superheated atmospehre could have on the assumed heat removal capacity. Cite reference material in support of this discussion.

3. Provide a plot of assumed fan cooler heat removal rate as a function of containment pressure (assuming the atmosphere is superheated starting from various pressures).
4. Provide the flow rate, inlet and outlet cooling water tempera-tures, and pressures under normal and accident conditions.

Discuss the ef fects of the cealing water reaching saturation temperatures.

5. What provisions will be made for leakage detection during opera-State the tion and periodic leak testing of the cooling coils?

leakage rate to the containment if one cooling tube is severed, assuming the containment is essentially at 0-psig.

6. Provide information on the known performance capability of the motors chosen which indicates that they can operate under the accident conditions for several days without failure.
7. Describe the particulate filters to be used and state the effi-ciency for removing particles less than 0.3 microns in diameter.

Provide a discussion (and reference $ to justify expected particle size in the containment atmosphere following an accident.

8. Describe the provisions which will be made to assure that the absolute filter material will retain its strength when wet.
1. Containment Spray and lodine Removal System
1. We believe that there should be two valves in parallel in the discharge line from the containment spray pumps to provide further assurance that the containment spray and iodine removal system will function properly. Please discuss your position in this matter.
2. Please justify the proposed location of the spray headers for assurance of ef fective distribution of the sodium thiosulfate spray.
3. Provide the pre 1Lainary design of the sodium thiosulfate storage tank with regard to the inlet system and diffuser which will enable the solution to be forced out of the tank with little mixing.
4. Describe the tests which will be conducted to demonstrate the effectiveness of the thiosulfate spray system. What is the mini-num acceptable removal rate? Also, discuss tests which will be performed to demonstrate compatibility of the solution with the

'long ters' cooling equipment.

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5. Describe quantitatively the ef fect that a superheated atmosphere would have on the ef fectiveness of the heat removal capacity of the spray system. Plot heat removal rate from the containment as a function of pressure, assuming the atmosphere is superheated starting from various pressures.

J. Regarding the boron injection or chargirig systes, provide the followingt

1. The operator's control sequence for initiation of boration from the chemical and volume control system.
2. Provisions for automatic actuation of the system.
3. The concentration in the boric acid storage tanks.

4 A statement including bases of the maximum cooldown rates which can be accommodated by the charging system when aligned to the boric acid tanks and when the pump is operating from the refueling water storage tank.

K. Emergency On-Site Power

1. Describe the design criteria for the diesel housing, protective barriers , fire sys tems , etc. i
2. State the time required to (a) start the diesels, (b) accept load, and (c) full load condition.
3. Describe the starting systems indicating the redundancy and fail-safe features.

4 Describe the cooling systems to be used for these units if the cooling is available including the length of tbme the diesel equipment can operate without cooling.

5. Describe the fuel storage and supply for the diesels including the fuel pump arrangement, day tanks and amount of fuel to be stored at the site. Are there any auxiliaries common to both diesels?

L. Emergency Feedvater Supply

1. Provide a plot (as a function of time) of the feedwater necessary to shut the plant down from 3391 Mw(t).
2. State the capacities of all emergency feedwater pumps.

f rom the condensate s torage tank How much feedwater .- ,.6.1 aE '.e 3.

and other sources?

VII. Accident Analysis A. , Containment Design Basis Accident i

1. We understand that you have developed an evaluation method which can be easily applied to estimate the capability of the contain-ment to withstand energy releases without exceeding design pressure. Please submit the information you have developed for this purpose.
2. Provide the containment pressure vs. time history resulting from a double ended break of the largest primary coolant pipe for the no core cooling case assuming that the metal-water reaction pro-ceeds according to the parabolic rate equation (unlimited steam).

Assume that when the clad or fuel slumps, the stored energy is State all assumptions in detail added to the containment as steam.

and discuss sensitivity of varying parameters.

3. Plot the integrated energy release to the containment as a function of time for each energy source assumed in item (2) above. Each energy source considered should be clearly identified.
4. It is apparent that the proposed containment design is essentially identical to other similar PWR reactor plants in terms of pressure rating, volume, and heat removal capability whereas the core power density is of the order of 6 to 18% greater. Accordingly, provide full justification including design criteria for the adequacy of the proposed containment and cooling system as well as complete discussion of the relative margins. Your response should be in terms of a sensitivity analysis using power density and containment design parameters.

B. Core Thermal Transient

1. Provide a plot of the pressure, flow, end heat transfer coefficient in the core during and following blowdown for various break sizes and indicate core water level.
2. Plot the water level in the vessel and core as a function of time for various break sizes assuming (a) three accumulators operate, and (b) two accumulators operate (along with minimum safeguards).

Discuss the conservatism of these calculations and indicate how much accumulator water (for each case) is calculated to be carried out of the vessel during blowdown. Describe how the effect of resistances to core flooding due to steam bubble, line losses, loop seal, etc. are accounted for.

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3. Provide a core thermal transient sensitivity study for the following parameters for various break sizes including the "design" basis breakt (a) Reduction of chemical energy from metal-water reaction by the energy required to heat up the water or steam to reaction temperature.

(b) Emissivity used in the radiative heat trans fer model.

(c) Metal-water reaction rate.

(d) Variations in the heat transfer coefficient.

(e) Heat sink initial temperature.

(f) Combined effect of all the above.

4. Plot the following (a) Hot spot temperature as a function of time for the adiabatic case (with unlimited steam metal-water r ea c tion) .

(b) The hot spot temperature as a function of time on the same plot as (a) using your model calculation assuming that 3 accumulators operate and secondly that 2 operate (for a spectrum of break sizes). Continua this plot below a temperature of 1800 F.

(c) The total percent metal-water reaction as a function of break size for 3 and 2 accumulators operating.

(d) The maximum tempgrature as a function of break area down to 0.005 ft .

(e) The ef fective cooling time, and time for 0% and 1%

cladding melt as a function of break area.

(f) Number of fuel rod cladding at various temperatures as a function of time for various break sise.

(3) Amount of fuel rod cladding at various temperatures as a function of time for various break sises. Compare this to the same information for a core with 18% lower power density, i

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5. Assuming three accumulators and two high head pumps operate, what is the total metal water reaction and maximum clad temperature for various break sizest
6. Assuming three accumulators operate, how much time would be available to start the diesels prior to clad setting?
7. At what clad temperature (and clad heating rates) would your assumed heat transfer fail to reduce the temperature transient?

C. Fuel Clad Tailure

1. Present a discussion of the ef fects of fuel clad fatture on the ability to cool the core following a loss-of-coolant accident.
2. Discuss the condition of the fuel upon clad failure. Wha t amount of oxide would you expect could leave the element?
3. The effects of fuel migration were discussed in a recent progress ,

report from the Argonne National Laboratory (Section V F(1) of ANL-7317), dated March 1967. In this report it was indicated, ,

on a prelimianry basis , that fuel migration or redistribution could af fect the physical and chemical properties of U02 abd thereby affect cled integrity. Accordingly, please provide a discussion in terms of an evaluation of this phenomenon as related to your fuel design and clad damage limits.

D. State the maximum fuel enthaply and the extent of fuel damage which you calcula te could occur as a result of a rod ejection accident for 4

this core. State the value of the moderator reactivity coef ficient used.

E. Steam Line Break Accident

1. Provide an estimate of the sequence of events in the core which would follow a steam line break accident including pressures and water levels in the pressuriser. Assume worst case conditions and that one rod remains out of the core af ter scram. Discuss the

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number of rods which would be in DNB, the amount of clad at various temperatures, the amount of fuel f ailure which could occur, and any inaccuracies in the methods used to estimate these effects. Provide a sensitivity evaluation en break size and isolation valve closure time.

2. For DNB calculations used to evaluate this accident, discuss which correlations are used for the lower pressures experienced in this type of accident.
3. Assuming that the af fected steam generator leaks at a rate of 10 gpm during normal operation, what would the leak rate be during and af ter depressurization of the secondary as a function of time? Provide a dose calculation listing all assumptions for the worst case conditions and assuming the break occurs between the containment and the isolation valve.

F. Tube Ruptures

1. Plot the pressure in a steam generator as a function of time following a loss-of-coolant accident assuming (a) that all valves leak at the design rate and the operator takes no action, (b) same as (a) but the operator takes action to cool down the : team generator at maximum feedwater rate, ? nd (c) that a tube failure occurs.
2. Evaluate the consequences of a tube rupture occurring during a loss of-coolant accident. Discuss detection capability and action to prevent leakage through the steam line and valves.

G. Dose Calculations

1. The PSAR states that some Class 11 structures and/or components have the potential for the release of non substantial amounts of radioactivity. State the specific components and discuss the potential hazard if failure were to occur.
2. Describe the environmental conditions in which the operational and area radiation monitoring systems will be expected to operate during routine and accident conditions. Discuss the specifica-tions which will be applied to the purchase of this equipment relative to the environmental conditions discussed above.
3. Provide the model used in calculating the fission product inven-tory in the primary coolant based upon 1% failed fuel, including (1) the derivation of extent of dif fusion of the fission products

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i to the primary system, taking into account burnup and operating fuel temperatures for the averase and hottest fuel rods, a..d (2) the proposed operation of the primary coolant cleanup system.

Include general formulae, assumptions, and justifications for these calculations. Based upon this model, provide a breakdown i of the maximum concentrations of the various fission products in the primary coolant, the gaseous decay tanks and the volume control tank.

J j 4 Provide the basis for the estimated 10 curies of Xe-133 released as a result of the fuel handling accident which damages 15 fuel rods. Provide the bases for (a) the amount of fisstun products j

in the gap, and (b) the separation factors used in determining ,

iodine release.

5. Please recalculate all pertinent dose calculations based on the 4

maximum power level of 3391 Mw(t) which we interpret as being i the ultimate power at which the reactor would ever be operated.

The dilution calculations should be based on a volumetric source with a shape factor of 1/2 instead of the virtual source dilution factors calculated in the PSAR. ,

H. Provide design bases and criteria for the concrete structure surrour.d. l ing the reactor vessel. Consider the effects of a vessel r. orale l rupture in the region inside the concrete structure.

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