ML19319B759
| ML19319B759 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 02/12/1970 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Sampson G TOLEDO EDISON CO. |
| References | |
| NUDOCS 8001270222 | |
| Download: ML19319B759 (34) | |
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February 12, 1970 Docket No. 50-346 The Toledo Edison Company 420 Madison Avenue Toledo, Ohio 43601 Attention:
Mr. Glenn J. Sampson Vice President, Power Gentlemen:
We need additional information to complete our review of your appli-cation for a construction permit for your proposed Davis-Besse Nuclear Power Station. The specific information required is indicated in the enclosure to this letter.
Some of the information requested may be available in the public record in the context of our regulatory review of similar features of other facilities.
If such is the case, you may wish to incorporate the information by reference in your application.
Your reply should be submitted as an amendment to your application.
If you wish, you may respond by revising pages or sections to the Preliminary Safety Analysis Report, rather than submitting answers to our questions as a separate addendum; however, if you choose the former, please provide cross-references.
In your application you propose an acceleration of 0.06g for the operating basis earthquake (0BE) and 0.15g for the design basis earth-quake (DBE). The adequacy of these values was discussed in our November 7,1969 meeting with your representatives. At this meeting we stated that the OBE should be at least one-half the value of the DBE.
It is our conclusion that the acceleration selected for the DBE, 0.15g, is adequate but the value selected for the OBE should be increased to 0.08g.
Please contact us if you have any questions regarding this request.
Sincerely, i
Peter A. Morris, Director Division of Reactor Licensing
Enclosure:
Request for Add'l Info a
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The Toledo Edison Company -February 12, 1970 cc w/ enc 1:
Leslie Henry, Esquire
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Fuller, Seney, Henry & Hodge 800 Owens-Illinois Building 405 Madison Avenue Toledo, Ohio 43604 n
George F. Trowbridge, Esquire s
Shaw, Pittman, Potts, Trowbridge & Madden 910 17th Street, N.W.
h*a;hington, D. C.
20006 Donald H. Hauser, Esquire The Cleveland Electric Illuminating Company P. O. Box 5000, Room 610 Cleveland, Ohio 44101 Distribution:
AEC PDR Docket File DR Reading DRL Reading RPB-2 Reading C. K. Beck M. M. Mann R. S. Boyd R. L. Tedesco L. Kornblith, CO (2)
- 5. Levine (14)
Branch Chiefs, RP H. Steele (2)
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ADDITIONAL INFORMATION REQUIRED DAVIS-BESSE NUCLEAR POWER STATION THE TOLEDO EDISON COMPANY, ET AL..
DOCKET No. 50-346 1.0 General
- 1.1 Table 4-10 of the PSAR established that the Code specified for the design and fabrication of the reactor coolant piping and valves is USASI B31.7 including Errata up to June 1968. However, the interfaces between seismic Class I piping systems are not identified in terms of which systems are designed to B31.7, Classes I, II, or III as applicable and which systems are designed to'UcASI B31.1.
Provide this additional information in a format similar co that which was provided as drawing 4.1-1 of Amendment 5 to the Midland application, Docket Nos. 50-329 and 50-330.
4 1.2 In section 1.6 of the PSAR the asterisked items of. areas of concern indicated in Advisory Committee on Reactor Safeguards letters are dis-cussed for pressurized water reactors. Provide an updating of these matters of concern cs referenced in ACRS letters on pressurized water reactors through January 1970, and discuss how they will be considered i '
in the Davis-Besse plant design.
1.3 List those systems which contain an interface between Class I and a lower class or a transfer of responsibility-between the nuclear steam system supplier and A&E.
Indicate the location of the inte. faces and discuss the manner in which they are considered in the design.
1.4 Provide details of the intake and discharge canal structure and indicate which sections of 'these canals will be Class I.
Include the design criteria and a discussion of the capability to achieve and maintain safe shutdown of the facility under all conditions in the event access to j
Lake Erie is not available due to failure of the canals.
1.5 With respect to the discussion of brittle fracture centrol for ferritic steels in compliance with General Design Criterion 35 published on July 10,1967, discuss the extent to which your design criteria conform to the following statenents:
a.
Those-pipes with wall thicknesses less than 1/2 inch need not have material property tests (such as Charp, V-nctch) if (1) they are austenitic stainless steel, or (2) the ferritic material is normalized (heat treated), or (3) the ferritic material has been fabricated to " fine-grain practice."
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Pipes with wall thicknesses greater than 1/2 inch must have a nil-ductility transition temperature 60*F below anticipated temperature when the system has a potential for being loaded to above 20%
of the design pressure. Ferritic material
- h an NDTT of -20*F or austenitic stainless steel will also fulfill the requirements.
1.6 Previous development efforts (such as those under PVRC auspices) have been devoted to the determination c f stress listributions and effee'ive elastic constants of perforated plates under thermal inplane and trau_-
verse loadings. Although a major application of this theory is to the analysis of the tube sheets in heat exchangers, these efforts, to date, have included no serious attempt to determine the stress state and effective elastic constants in an actual structure, consisting of tubes swaged and welded into the perforated plate. The ASME Section III stress analysis procedure is based upon the perforated plate method described above.
In light of this background, discuss the analytical treatment of the tube-tube sheet complex (as an integral structure) of the steam generators for the Davis-Besse Station.
1.7 Proposed amendment to 10 CFR Part 50 published November 25,1969 would require that piping and fittings within the reactor coolant pressure boundary meet the requirements for Class I piping of USASI B31.7 dated February 1968 and Errata dated June 1968, including the requirements of Appendix IX - Quality Control and Non-destructive Examination Methods, of the 1968 edition of ASME Code,Section III as mandatory supplement to 331.7.
Compare the quality control requirements specified in Section 4.5 of the PSAR with the requirements of Appendix IX.
Specify all significant differences between the two sets of requirements for the reactor coolant piping.
1.8 Will any Class I components, in whole or in part, be designed and/or fabricated in a foreign country? If so, which components will be fabricated by whom?
1.9 To which edition of the ASME Code Section III and Addenda will applicable Class I components be designed and f abricated?
1.10 Appendix 2A of the PSAR discusses the type of activities which are expected to be carried out in the restricted areas adjacent to the site. Provide a discussion and analysis of the margin of safety to be provided in the facility's structural design to protect vital equipment and components against possible missiles generated from the restricted areas. The analy-sis should include all areas which require missile protection to assure safe shutdown of the f acility is not jeopardized and no uncontrolled release of radioactivity will result.
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~3-2.0 Site 2.1 In order that we may determine the appropriate atmospheric diffusion factors for the design basis accident analysis, provide seasonal wind-roses for each atmospheric stability condition, which include wind directions and speeds at a 20-foot height and describe the ranges of temperature dif ferences between the 5-foot and 145-foot elevations which were used to determine the various stability conditions. Describe the location of the meteorological tower and the heights and distances of objects that may affect the diffusion conditions.
2.2 We are reviewing your analysis of the high water protection level for the plant, and thus far we have determined that the following information is needed.
a.
An analysis of the maximum probable water level at the plant site
'rather th'n the once-in-100-years water level. The analysis should a
include static (e.g., seiche) and dynamic (wave) water levels.
b.
A hydrological or mathematical model study which determines the greatest increase in the max 1=um probable water level that could be caused by the convergence of water via the gap between the intal:e and discharge canal.
c.
An analysis of the greatest probable effect of a Toussaint River flood on the maximum probable water level at the plant site.
d.
An analysis of the maximum probable flood levels of the Tcussaint River at the plant site.
2.3 Describe the systems which will assure that radioactive liquid waste will be diluted in circulating water discharge, or confirm that the radioactive liquid waste released into the discharge canal will not exceed the limits of 10 CFR 20 for unrestricted areas. Your description should include the following information:
a.
The instruments and equipment that will prevent the discharge of radio-active liquid waste while circulating water is not being discharged.
b.
The minimum circulating water discharge rate (gpm) that will permit the discharge of radioactive liquid waste, c.
The maximum discharge rate (gpm) of radioactive liquid waste.
d.
A drawing showing how the radioactive liquid waste will be mixed with circulating water discharge.
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2.4 #rovide an analysis of the maximum average and instantaneous concentra-tions of radionuclides in Lake Erie at the discharge canal and at potable water intakes that could result from normal operat'ons or an instantaneous rupture of a tank. The analysis should include the following information and all factors, assumpticns, and references'used to obtain the information.
a.
A list of the radionuclides and the maximum quantity of each that would be released to Lake Erie in a year of normal operations.
b.
A list of all tanks that would contain radioactive liquids and the maximum volume and concentrations of radionuclides that would flow into the lake in the event of the rupture of any tank.
c.
The maximum average and instantaneous concentratiens of radio-nuclides in Lake Erie at the discharge canal and potable water intakes.
d.
An explanation of how the temperature, volume, velocity, and direction of the released water were considered in determining the lake dilution factors, i
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1 l 3.0 Reactor Design 3.1 Section 3.1.2.4.2 of the PSAR,which delineates the stress and strain limits for fuel assemblies under nor=al and abnormal operating conditions,does not sufficiently define those limits nor the manner and extent to which the cited limits provide an assured margin of safety against failure under these loadings.
Provide the following additional information.
Confirm that the type of stresses referred to in Paragraph (a) are in the a.
'brimary" category as defined in Article 4 of ASME Code,Section III.
DescrAe the basis for establishing 75% of the stress rupture life of the material as a numerical limit and indicate whether that limit is con-structed upon the average stress or the minimum stress to produce rupture at the end of 100,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.
b.
Clarify whether the type of stresses referred to in Paragraph (b) are in the secondary category in the same context as above. Where stresses exceed yield, are they calculated on an equivalent elastic basis, i.e.,
pseudo-elastic basis as in Section III? Identify the source of the fatigue curves used for each material of concern, e.g., Article 4,Section III.
Where fatigue data are employed which are not included in any codes or standards, specify whether a basic data or design curve is used or a design curve which incorporates design / correction factors, e.g.,
2 on stress, 20 on cycles and correction for maximum effect of mean stress. The statement that strain limits will be set using no more than 90% of the material (s) fatigue life i= plies that you may use less.
Clarify this statement and in addition, outline the exact procedure (s) used in setting the strain limit (s).
Specify the number and type of cycles that have been established for design purposes and indicate the margin of safety that exists over the expected number and type of operational cycles to be experienced.
c.
For the combination of stresses in (a) and (b), above, specify the stress limit (s) that apply, e.g., 3 S or S.
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d.
What are the stress and/or deformation limits which are specified for the fuel assemblies under emergency and faulted conditions, i.e., normal plus pipe rupture loads, normal plus =aximum earthquake loads, and the simultaneous occurrence of these leads?
3.2 Discuss the status of the develop =ent of possible means for inservice monitoring for vibrations or detection of loose parts in the reactor pres-sure vessel and other parts of the pri=ary system.
. 3.3 Provide a discussion and analysis to show that the Ca11ure of the control rod assembly pressure housing will not cause failure of adjacent control rod assemblies.
Indicate the minimum control rod assembly pressure housing rupture size which would result in ejection of the control rod.
3.4 Provide a discussion and any new analyses which have been completed to justify the increase in the core design power from 2452 to 2633 megawatts thermal. Include any experimental information which supports the power level increase.
3.5 In reference to the discussions of the reactor internals in Section 3 of the PSAR and the criteria of Appendir A, list the components for which buckling is a possible mode of failur when considering the case for the combined concurrent design basis earthquake and the postulated loss-of-coolant accident. For the most critical items, i.e., those closest to failure, state the analytical method and give the margin between the condition considered to constiiute failure and the as-calculated condi-tion for the combined loadings. Relate this margin to the degree of uncertainty involved in the loads used to perform this analysis.
The uncertainties of the selection of the basic seismic ground motion need not be discusscd in this context.
3.6 Identify the extent, method and findings of the analyses of thermal stresses which would result in the core barrel and core support structure in the event of loss of coolant and subsequeat operation of emergency core cooling equipment.
3.7 Describe the analytical and/or test procedure that will be used to ensure the functional integrity of the reactor internals in the event of a loss-of-coolant accident.
Indicate the loading functions and mathematical models that will be used in the analysis.
3.8 The PSAR does not present sufficient infor=ation concerning the specific nondestructive examinations and inspections to be performed for the reactor internals. List all such examinations and inspections and identify the codes or standards which apply in each case.
3.9 The stress limits for the loading combinations as given in Table SA-2 require further clarification and more exact definition. The brief discussion of those limits in section 4.1.2.4 does not provide this clarification. We have identified several items of concern which are i
described as follows:
a.
Case II - Design Loads Plus Maximum Hypothetical Earthquake Loads is defined as a faulted condition and reference is made to paragraph N-412(t)(4) of ASME Code,Section III. We consider this particular loading combination as an energency rather than a faulted condition.
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'An~ additional loading case along with corresponding stress limits is required for the occurrence of the loads due to pipe rupture (loss-of-coolant accident) plus normal design loads.
c.
One of dhe Case III loading combination stress limits utilizes ultimate strengtu curves published by U. S. Steel which 'are adjusted to minimum ultimate strength values by using the ratio of ultimate strength given by Table N-421 of Section III at room temperature to the room tempera-ture strength given-by U. S. Steel. This ultimate strength ratio, which is calculated at room temperature, cannot be verified as a i
conservative estinate of the actual ratio (or margin) at the temperature of concern unless a comparison to the value of the minimum code ultimate strength of each material at temperature is obtained.
Discuss the above items in detail and provide the follcwing specific information:
Clarify whether the stresses to be compared to 2/3 of the ultimate a.
strength of material limit for loading Case III are calculated on an elastic basis,- i.e., pseudo-elastic stress calculation as in Section III. Provide the elastic stresses corresponding to this limit for each of the materials of concern. Furnish the corresponding strain limits for each material.
b.
It is stated that the Case III design stress limits will be based on either the 2/3 of ultimate strength criteria or paragraph N-417.ll of
- Section III. Discuss the basis for using either set of limits addressing:
(1) the reason for the choice of limits, (2) to what 7
components or areas thereof each would apply and (3) the basis of comparison between the two limits including the margins of safety
.that each-provides.
3.10 It has been indicated in Appendix 5A that " reactor vessel internals must satisfy defor=ation limits which are more restrictive than the stress limits." The deformation limits, though referenced in other sections of the PSAR, are not presented. Provide:
a.
The deformation Itnits for reactor internals.
b.
The deformation limits that correspond to the stress lisits in -
' Table 5A-2 for essential reactor internals.
I The deformation limits at which " loss of function" for essential c.
.rcactor internals is anticipated to occur.
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. 4.0 Reactor Coolant System Design 4.1 Discuss the means to be used to determine if Class I mechanical components qualify for service under seismic loading conditions; e.g., analysis or shake table tests. Relate the methods used to the frequency spectrum and amplitude calculated to exist at the equipment mounting. State the basis for assuming that items such as emergency core cooling pumps and drives will start and run, if needed, under seismic loading. Relate your response to any tests or analyses to be performed on equipment in the running mode as well as the static mode.
Provide a summary of all such pieces of equipment, such as tanks and pumps.
4.2 Describe in more detail the analysis procedures that will be used to determine that the nuclear steam supply system will meet seismic Class I criteria (Section 3.1.2.4, page 3-3 PSAR).
include in this discussion the following:
A detailed description and sketch of the proposed mathematical model(s) a.
of the system, including a discussion of the degrees-of-freedom and methods of lumping masses and determining section properties.
b.
Dae mathematical model(s) to be used for the reactor vessel internals.
c.
A discussion of the analytical procedures to be used, including the methods of computing periods, mode shapes, design accelerations, displacements, shears and moments.
d.
An explanation of which " actual earthquake records" are to be used in the time-history analyses and a comparison of the response spectra from these earthquakes and the spectra postulated for the site (PSAR Figures 111-5 and 111-6, pages 2C-47 and 2C-48).
e.
An explanation of how it will be determined that the LOCA and maximum earthquake time-histories are conservatively applied such that the maximum structural response is obtained. Are the LOCA and earthquake time-histories assumed to start at the same time? If so, would it be possible to obtain greater response if the earthquake were started at some increment of time, such as 10 seconds, either before or af ter the start of the LOCA?
f.
A listing of the damping values to be used.
4.3 How will flow induced vibration loads be considered in the design of the primary system? State the extent, methods and findings of the analyses or tests which will be made. In this statement include responses to the following specific considerations:
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a.
Will both normal and emergency modes of operation be considered?
b.
What design limits, amplitude and frequency apply to these conditions?
c.
Discuss the tests which are considered necessary for the Davis-Besse plant.
In these discussions include comments on the type and extent of instrumentation planned.
4.4 Provide the following information concerning the design and design criteria for the reactor vessel:
a.
Special requirements, if any, imposed by local or state regulation on the reactor vessel design.
b.
The acceptance standards intended for nondestructive testing procedures for the reactor vessel.
Indicate if these meet or exceed the require-ments of Section III of the ASME Boiler and Pressure Vessel Code, 1968 edition.
c.
Indicate whether transients such as loss of reactor vessel flow (one or two loops) and loss of load will be considered in the transient stress analysis. If not, provide a justification for your position, d.
The design conditions for the core flooding water nozzle.
e.
A list of stainless steel component parts in the reactor vessel and the reactor coolant system that will become furnace-sensitized during the fabrication cycle.
f.
Provide a summary discussion and enumeration of results of transient stress analyses, illustrated by sketches showing points of analysis, and a list of associated cumulative fatigue usage f actors for the reactor vessel.
g.
Discuss item 2 in Table 4-9 with respect to design cycles versus actual cycles (the design cycles appear to be larger by a factor of 10 than for previous plants).
h.
Discuss the magnitude of the stress in the reactor vessel membrane induced by gamma ray heating.
- i. Will ring forgings be used for reactor vessel shell sections other than the closure flanges?
- j. Provide summary results of Charpy V-notch and Drop Weight tests for the reactor vessel plates and forgings.
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. 4.5 Discuss the extent to thich electroslag welding will be used in the fabrication of Class I systems.
If electroslag welding is to be used, describe the process, its variables, and the quality control procedures to be employed.
4.6 _ Section 4 of the PSAR describes your plans for inservice inspection of the reactor coo.lant system. Compare your proposed program with the ASME Code for Inservice Inspection of Nuclear Reactor Coolant System, 4~
If any of the reactor coolant system pressure boundary area initial baseline examinations called for in this code are to be omitted, identify the dreas and discuss your reasons for not testing them. Should any areas required in the code ba" precluded from an inservice inspection due to inaccessibility, discuss.the reason for such inaccessibility.
4.7 Describe your inservice inspection program for the Class I (seismic) mechanical systems outside the primary system pressure boundary, including items to be inspected, inspection schedule, and types of inspection.
Some items te be considered are primary system components, support, primary pump flywheels, and Class I (seismic) mechanical components in the engineered safety features.
4.8 Describe the provisions used to protect the reactor primary system, other vital systems, and structural supports for these systems from missile hazards. Include a discussion of design criteria, missile shields, missile size and masses, missile velocities and the penetration formulas used for design.
4.9 Reactor Vessel Material Surveillance Program You state that you are participating in t.te B&W integrated surveillance program as presented in the topical repore, BAW 10006, " Reactor Vessel Material Surveillance Program." We understand that extensive revisions are being made to this program.
Provide the following in regard to the Davi's-Besse vessel:
a.
The total number of capsules and capsule locations related to the core midplane.
b.
The number of capsules of each type to be withdrawn and tested over:the life of the vessel.
~. A listing of the archive material reserved. Indicate if sufficient c
archive material hes been set aside to fabricate enough specimens for a minimum of two. capsules.
d.
A listing of the chemical content of the vessel materials, including the residual element content in weight percent to the nearest 0.01%.
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, 4.10 With regard to the capability for detection of leaks from the primary system, provide additional discussion of the following:
a.
The proposed leak detection instrumentation, including a discussion of sensitivity, response time, control room alarms, diversity and redundancy.-
b.
The maximum leak rate from an identified or unidentified source that will be permitted during operation, and the predicted crack size that can be related to the allowed leak rate from an unidentified source. Include a discussion of the bases for the selected leak rate, and a description of the analytical methods used to establish the relationship between the unidentified leak reta and the crack size.
The sensitivity of the leak detection system for the primary c.
coolant pressure boundary.
d.
The leak detection systems provided for other Class I fluid systems. List those Class I fluid systems for which no special leak detection system is provided.
4.11 Several statements in the PSAR indicate or imply that Class I systems will be protected against damage from failure of other systems. Specify the design criteria which will be applied to protect against damage of these. Class I systems by pipe whipping.
4.12 Failure of a primary pump flywheel could result in the generation of missiles capable of severely damaging equipment within the containment.
-Provide the results of an evaluation assessing the potential consequences from possible missiles generated by failure of a flywheel. Describe the program to be followed to minimize the probability for experiencing a flywheel failure, including the consideration to be given to material selection, design margins, fabrication, failure analyses, acceptance testing, inservice inspection requirements, and other special quality assurance measures. What practical measures can be taken to provide missile protection to vital equipment that could be damaged by missiles generated by failure of the flywheel?
4.13 Provide a summary description of the reactor vessel stress analysis which includes simple sketches showing the location and geometry of areas of discontinuity or stress concentration.
Identify the controlling critical loading conditions.
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f, 4.14 Provide a summary of the maximum intensities and cumulative damage usage factors calculated for the steam generators and pressurizer vessel accompanied by sketches illustrating the points of analysis.
4.15 The PSAR states that the reactor coolant pump casings will be designed and fabricated to meet the intent of ASME Code,Section III, Class A vessels as applicable. Outline the stress analysis procedures to be used for the pump casing, furnishing references as appropriate, and provide a summary of stress intensities and cumulative damage usage factors obtained. Specify any deviations frem Code requirements other than lack of Code stamping.
4.16 With respect to formulation of the operating pi e sure-temperature relationship limits insofar as a brittle fracture mode of failure, discuss the adequacy of using a stress concentration factor of 4 4
(page 4-19) on assumed flaws in calculating stresses, particularly in the presence of a crack or cracklike defect (either present initially or developing in service).
4.17 State the relief capacity of the spring-loaded safety valves in the main steam lines.
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- 5.0 General Structural Criteria and Design Loads 5.1 With regard to the design basis loads for consideration of containment stability, indicate the extent to which a combination that includes seismic loads but excludes accident pressure loads will be considered.
5.2 Specify the design criteria for the flexible closure of the space between a penetration and the shield building. Does this closure meet the single failure criterion? What design basis loads (earthquake, maximum accident pressure, tornado pressure differentials, etc.) is it designed to accommodate?
5.3 With regard to reinforcement lap splicing, provide the minimum lapping in terms of bar diameter that will be used, the splice stagger that will be employed, and data to show that the splicing proposed will be sufficient to develop reinforcement ultimate tensile strength especially for areas of high seismi'. tensile forces.
5.4 The design criteria for containment penetrations are not as explicit as is desired. Provide the design criteria:
a.
to be applied for forces on the penetrations resulting from pipe rupture or relative displacement of internal or external systems.
b.
for design for or isolation against vibrational loads.
5.5 With regard to a pipe break in the annular space between the containment and the shield building, provide:
The maximum local pressure load that can be developed.
a.
b.
The maximum local stress that will result therefrom.
c.
The margin of safety with regard to steel shell stability from such a local load.
5.6 Provide a description of the protective coatings and paints to be used within the contaimnent. Include the following items:
a.
Identification of material to be used, location, and function.
b.
Physical and chemical characteristics.
c.
Performance under accident conditions including washdown, radiation, steam, tenperature, and jet impingement ef fects.
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. 5.7 With regard to the shield building and connected penetration rooms, provide:
a.
The manner in which negative pressure will be assured in the shield building base slab area.
b.
The type of connection that is being used for the penetra'. ion room / shield building connection.
c.
A discussion of specific lines penetrating the steel containment and shield building structures which will not be enclosed withi:.
the containment system complex. Describe the manner in which potential leakage through these lines is assessed and factored into dose estimates for accident conditions.
5.8 The liquefaction potential for the site and the margin of safety against its occurrence under the specified seismic ground motion (s) has not been
_ presented. Provide:
a.
Identification of those structures that will be located on v
foundation soils and, hence, subjected to potential displace-ment due to soll liquefaction.
b.
An evaluation of the stress or strain intensity for which seismic liquefaction is estimated to occur for the various soil strata at the site. Include the extent to which variability in relevant soil properties, foundation preparation, ground water variations, and structural loads has been considered.
The analytical procedures used to predict stress intensity under c.
seismic loading, the preliminary results achieved and the margins of safety with respect thereto.
5.9 With regard to the extent to which the structures of the facility may be subjected ' to differential settlement due to seismic ground motion, provide:
a.
The differential displacements predicted as the result of design basis ground motions.
b.
The safety margins that will be incorporated in the design of vital structures of the facility for acco=modation of such displacements.
5.10 with regard to site slope stability under seismic ground motion, provide:
The seismic ground motions that are being considered as input a.
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A sketch or sketches showing the extent of embankment and slope I
areas at the facility and their slope ratios.
c.
The criteria for the design.
d.
A detailed description of the design methods used, including an example calculation illustrating each procedure used.
l e.
A tabulation of' results achieved for each vital soil structure to include factors of safety under the design seismic ground
-mot 1ons.
f.
Specific details concerning possible failure modes considered, the extent to which pore pressure effects were considered in the analy-4 l
sis, and' indication as to whether vertical earthquake excitation was included concurrently with lateral excitation.
5.11 The description in the PSAR indicates a possibility for relative motion-between the various structures of the facility. Provide a listing of adjoining and interconnected structures, components, and systems where differential motion is of possible safety consequence. Provide a i
description of the analysis methods through which possible motions at these locations will be quantitatively evaluated and the design margins 7~
that will be employed with respect to the computed differential motions.
With special regard to penetrations bridging between the containment and shield building, what differential motions are computed and what l
differential motions are tolerable without loss of function?
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exterior piping runs due to static and seismic. differential settle =ent and relative seismic' motion is not described. Provide:
i The design procedures and/or detailing that are being used to a.
accomplish the design of interconnects.
b.
A typical example of how these design procedures and/or detailing are being accomplished.
5.13 Class II structures, systems, and equipment are defined on page SA-3 of 4
j the PSAR as those whose failure would not result in the release of radio-
- activity and would not prevant safe reactor shutdown. Provide a listing of all Class II systems _which may contain radioactive material and
' discuss how they ' meet the above criterion.
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5.14 Your definition of Class I structures, systems and components on page 5A-2, of the PSAR ' states that certain portions of these systems may be composed of Class.II componente Clarify your intention in this regard. Provide a
. list of. Class II components of - Class I systems.
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% 5.15 Describe the containment building vacuum breaker system and provide the following information:
a.
Criterion used to establish the system's capacity.
b.
A failure analysis of the system.
c.
Would the vacuum breaker system be adequate for a pipe break in the shield building annulus or penetration rooms?
5.16 Discuss the criteria for design of the pressure vessel cavity shield walls.
Include the following information:
a.
Discuss design provisions that will be made to reflood the reactor pressure vessel esvity following a loss-of-coolant accident.
b.
Indicate any changes in design which result from the pressure vessel being supported from the inle' and outlet nozzles.
c.
Describe the maximum capability of the cavity to withstand a pressure transient and indicate largest rupture of primary system considered.
Include a discussion of the loading condi-tions considered and the associated stress levels in the steel and concrete.
d.
Discuss the heat removal requirements for the cavity shield wall.
5.17 Provide a listing of all valves which will be required to operate on a containment isolation signal and include the bases for establishing requirements for valve closure times.
Include the steam line isolation valves and feedwater system valves.
5.18 Describe the criteria that will be used in the collapse analysis of those Class II structures that enclose Class I equipment / systems or othar Class I structures.
5.19 Describe the instrumentation (such as strong-motion accelerographs and relative displacement measuring devices) to be provided to assess potential damage in the event of strong earthquake ground motion.
5.20 Specify the containment design leak rate and conditions that will be met during plant preoperational testing. Various construction tests requiring leak rate determination at various stages should be identified as well as the acceptance criteria established for each test.
The leakage specifi-cations for penetrations and valves should be included in your response.
5.21 In section 5.2.2 of the PSAR, it is stated that the pressure in the shield building cavity will be vented to the atmosphere in the event the pressure exceeds six inches of water. Provide a discussion and analysis to show the amount of leakage that would be vented following a design basis accident with subsequent heating of the shield cavity air volume.
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r 6.0 Engineered Saferv Features l
6.1 Identify the available and required net positive suction head (NPSH) for all emergency core cooling system pumps and spray pumps following a loss-of-coolant accident. Describe the manner in which the available NPSH was calculated and provide a breakdown of the factors and assumptions used to assure minimum NPSH is met even with containment pressure 3 psi below atmospheric pressure as stated in section 6 of the PSAR.
6.2 With regard to the calculations of the sump coolant temperature following a loss-of-coolant accident, state the time af ter shutdown when (1) the heat removal capability of one decay heat exchanger equals the reactor decay heat and (2) the capability of the two heat exchangers equals the decay heat.
6.3 Indicate the maximum pri=ary system leakage rate which could be accommodated by the normal charging system without initiation of the emergency core injection system.
Relate this leakage rate to that resulting from the double-ended rupt are of any small lines connecting to the primary system.
6.4 To permit determination of the adequacy of isolation provided between the high pressure primary system and the low pressure portions of the emer-gency core cooling system, list those portions of the ECCS which are not designed for reactor operating pressure and temperature and indicate the means by which isolation is provided.
6.5 State your criteria regarding the sizing, design, and construction of the borated water storage tank. What concentration of boron will be maintained in the borated water storage tank and fuel storage pool?
6.6 Provide the analysis which shows that the 4 psig containment pressure signal for ECCS initiction provides adequate backup for the 1500 psig primary system pressure signal for ECCS initiation for the spectrum of primary break sizes of 0.05 f t2 to 14,1 gg2 Initiation by either of these diverse signals, assuming f ailure of the other signal, should provide the required ECCS capability.
If the analysis of the effective-ness of the Davis-Besse safety injection design takes credit for a reactor trip, show how the initiation by only the high containment pressure signals meets this requirenent.
6.7 Provide the analysis and discuss the results which support the reduction from 1800 psig to 1500 psig for the primary system pressure ECCS initia-tion signal.
Includeadiscussiantoshowtheeffectsofthispresgure 2 to 14,1 5:
reduction for primary system rupture sizes from 0.05 f t
se.. 6.8 Provide the design bases for the shield building emergency ventilation system. Include the following information:
a.
Expected conditions (such as temperature and humidity) of the air in the annulus and penetration rooms.
b.
Capacity and type of activated charcoal filters.
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Consideration to prevent possibility of fire, d.
Failure analysis of the system.
6.9 The makeup tank is available to makeup-injection pumps during emergency l
conditions. Demonstrate that the cover gas in the makeup tank cannot override the borated water head and feed gas into the makeup pump suction.
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o 70 Instrumentation, Centrol and Power 7.1 In regard to the protection syste=s which actuate reactor trip and engineered safety feature action, the following information is requested:
a.
A listing of these syste=s designed aad built by Babcock and Wilecx that are identical to those of the Three Mile Island Unit #2 (as documented in the PSAR); discuss any design differences; b.
Identification of the supplier for these syste=s that are designed and/or built by suppliers other than Babcock and Wilcox, l
and c.
Identification of those features of the design which differ fres the criteria of IEEE 279 and the Commission's General Design l
Criteria. Explain the reasons for any differences.
7.2 In regard to the Babcock and Wilcox designed control syste=s, the i
following information is requested:
a.
Identification of the major plant control systems (e.g., pri=ar,y te=perature control pri=ary water level control, steam generater water level centrolh which are identical to those in the Three Mile Island Unit #2; and I
b.
A listing and a discussion of the design differences in those systems not identical to those used in the Three Mile Island Unit Y2. This discussion should include an evaluation of the safety significance of each design change.
73 What are your seis=ic design bases for the reactor protectica system (as described in Section 1 of IEEE 279) and the emergency electric power system?
Will the systems be designed to be capable of actuating reactor trip er engineered safety feature action during the maximum peak acceleration? If a seismic disturbance occurred after a najor accident, would energsicy core i
cooling be interrupted? What tests and analyses vill be perfor ed to assure that the seismic design bases are met? What seissic specifications are included in the instrucentation and control purchase orders?
7.h Describe the_ quality control procedures which apply to the equipment in the reactor protection systes (as described in Secticn 1 cf ists 279) and the energency electric power system. This descriptien should include, but not necessarily be limited to:
(a) quality centrol procedures used during equipment fabrication, shipment, field stcrage, field installatien, and system component checkout; and (b) records pertaining to (a) abcve.
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+ 75 Submit your cable installation design criteria for preserving the independence of redue. ant reactor protection system and ccgineered safety feature circuits (instrumentation, control and pcVer). Fcr the purpose of cable installation, the protection syste circuits should be interpreted in their broadest sense to include sensors, signal cabics, control cables, power cables (both a.c. and d.c.),and the actuated devices (e.g., breakers, valves, pumps).
Cable separatien should be considered in terms of space and/or a.
physical. barriers between redundant cables. Please address (1) the separation of power cables frc: these used for control and instrumentation, (2) the intermixing of control and instrument cables within a tray (or conduit, ladder, etc.), (3) the intermixing within a tray of cables for different protectica channels, and /L) the intermixing of ncn-vital cables with protecticn systa cables.
b.
What are your criteria with respect to (1) the separation of penetration areas, (2) the grouping of penetraticna in each area, and (3) the separation of penetrations which are rutually redundant?
c.
Discuss cable tray leading, insulatten, derating, and evericad protec-tien for the various categories of cables.
d.
Discuss your criteria with respect to fire stops, protection of cables in hostile environments, temperature =cnitoring of cables, fire detection, and cable and vire-way markings.
e.
Discuss the administrative responsibility for, and control over, a2or the foregoing (a-dl during design and installaticn.
f.
Discuss your design criteria for locating the process instrumentation inside containment to include (1) separatica cf redundant senscrs and sensing lines, (2) protection provided to sensors and sensing lines, and (3) protection provided to cables between senscrs and electrical penetrations.
7.6 What are your design criteria for reactor protectica system an1 engineered safety feature related electrieni and acchani:s1 equip =ent located in the containment or elsewhere in the plant which take into account the potential effects of radiation en these ecmponents due to either nornal operation or accident conditions (superimposed on 1cng-ter normal cperation)? Eescribe the analysis and testing perferred to verify ec pliance with these design criteria.
7.7 Identify all equipment and cc ponents (e.g., motors, cables, filters, pump seals) located in the prinary centain=ent which are required to be operable during and subsequent to a 1 css-of-coolant or a steam-line break accident.
Eescribe the qualifications tests which have been er will be perferred on each of these items to insure their availability in a ec bined high te perature, pressure and humidity enviren=ent.
- 7.8 What criteria have you established relative to assuring that less of the air conditioning and/or ventilation system vill net adversely affect operability of safety related control and electrical equipment located in the control room and other equipment roces? Describe the analysis perforced to identify the worst case environment (e.g., tcmperature, humidity). What is the limiting conditicn with regard to te=perature that would require reactor shutdown, and hev vas this determined?
Describe any, testing (factory and/or onsite) which has been or will be performed to confirm satisfactory operability of centrol and electrical equiptent under post-accident envirennental eccditions.
7.9 rescribe hev reactor protectica system and engineered safety feature equip =ent will be physically identified as safety equipment in the plant.
7.10 Cn Fage 8-8 of the PSAR the electrical and physical independence of the diesel-generators is discussed. Provide the design criteria for the diesel-generator cooling water systen wht:h vill insure a fully redundant ecoling water system.
7.11 Ident'.fy the emergency diesel-generator protective interlocks and discuss the basis for their selection.
7.12 The text states that during nor al operation station pcVer loads are supplied from the auxiliary transfer:er. Upon failure of this pcVer scurce, startup transformers 01 and C2 provide the required loads in what appears te be a " split bus" concept. Eescribe the design cencepts utilized to protect this " split bus" concept and show that the design is net cc promised by any cf the following:
a.
Multiple power sources to switchgear buses A and 3.
b.
Cross feedt to switchgear buses CD, F3 and E3 c.
The use of transformer between cuses C1 and Dl.
7.13 Paragraph 8.2 3 3 of PSAR discusses but does not cicarly i:lentify the diesel atarting signals. Identify the diesel start signals and clarify whether the circuit is an "and" circuit.
7.lk Provide the design criteria fer the switchyard 125 vdc system and show that a single failure will not negate the ability to supply effsite pcVer.
7.15 Frevide a loading table for the essential buses for the safe shutdown and for the accident conditions. This table should provide the timing sequence for starting uhe leads and the resulting =argin for the wcrst case lead with reference tc the diesel continucus rating.
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-. 7.16 Perform a system stability study and shev that neither loss of this unit nor icss of the largest generating unit on the system (either CAPCO cr Toledo Edison) will negate the ability to provide offsite pcVer.
7.17 Paragraph 7.1.1.23 of the PSAR discusses the use of isolation amplifiers.
Pro'<tde the design criteria for these amplifiers.
7.13 1he neutron detectors which are part of the Plant Protection Systes are temperature limited at 1750F, Provide the design criteria for detector cooling and temperature =cnitoring which allevs the detectors to =eet the require ents of IEEE 279, Paragraph 3(g).
l T.19 Paragraph 7.6.h of the PSAR describes the cet=unication systems as having independent power supplies. Describe these sources of pcver.
'# cat radio co==unications do you have from the station to your syste:
(offsite) in the event of an emergency?
7.20 Provide the design criteria for supplying pcwer to the radiatica monitoring subsyste=s, specifically these which monitor releases to the at=csphere and these which provide control functions.
7.21 Provide the design criterien for sizing the containment air recirculating fan motors. Do the required ratings represent nameplate ratings or naceplate times a service factor?
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7.22 Discuss explosion prcof require =ents for instrumentation (local acunted detectors and transmitters) and electrical equipment (=otors, switches I
and other arcing devices) which are located in potentially hazardous areas. (A hasardous area is defined as one in which fla==able or explosive concentrations or caterial =ay exist).
7.23 Generally the design criterien followed for instru=ent air prcvides guidance in the design of a reliable source of clean dryedr. Past experience has indicated that the use of instru=ent air valve operators (pisten operators) can result in sticking or galling due to lack of lubrication.
State your design criterica regarding air aupplied to pisten cperators.
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r a 8.0 Conduct of Operation 8.1 Discuss your plans regarding the development of an overall s!te emergency plan including scope, organization, protective measures, let ical bases, action of offsite agencies, medical support arrangements, training of personnel, recovery following accident, and other applicable material.
8.2 Describe what provisions will be made to ensure plant security from unauthorized entry both during construction and operation.
Indicate the extent of perimeter' fences, lighting, guards, employee screening procedures, visitor control, control of containment access, and other site surveillance methods.
8.3 Describe the locations of hospitals, schools and detention institutions within the low population zone (LPZ) and state the number of beds in the hospitals, the number of students in the schools and the number of individuals in the institutions.
9.0 Auxiliary and Emergency Systems 9.1 On page 9-1 of the PSAR, you have provided a listing of the fifteen codes and standards used in the design, fabrication, and testing of components, valves, and piping of the auxiliary and emergency systems. Provide a cross-reference identifying the components to which the various codes and standards apply.
9.2 Makeup and Purification System 9.2.1 Discuss the possibility of the leakage of hydrogen frem the makeup tank.
If local concentrations of hydrogen in excess of two volume percent can occur in the auxiliary building external to the makeup tank, describe the means ta' gen to prevent Ignition of the hydrogen.
9.2.2 Discuss your criteria regarding the prevention of crystallization in those portions of the system containing concentrated boric acid. Discuss the design criteria for the tracing system proposed considering (1) the type of tracing proposed, (2) redundancy, (3) the minimum temperature margin between fluid temperature and the saturation temperature for the concentra-tion involved, (4) means used to assure adequate coverage of critical areas (e.g., valve bodies and elbows), and (5) plans for preoperational testing. Discuss the consequences of solidification at any single point in this system, indicating if cold shutdown could be achieved.
9.3 If the component cooling water system fails downstream of the reactor building isolation valve, the function of the shield coolers, reactor coolant pump coolers, and the letdown coolers are lost. Since under this situation the makeup system could not be used for bleed-and-feed
A operations owing to the loss of the letdown coolers, discuss the ability to achieve cold shutdown.
Indicate the margin between the volume con-traction in the primary system during cooldown and the volume of borated water required to compensate for the reactivity gained during cooldown.
9.4 Service Water System 9.4.1 Identify the number of auxiliary feedwater pumps needed to dissipate decay heat Lnmediately after shutdown.
9.4.2 Provide a process diagram of the auxiliary turbine-driven feedwater pump steam supply system.
9.5 List the seismic design classification of the various components of the fire protection system. Indicate to what extent this system can function with any single failure. To facilitate understanding, provide a diagram of the system.
Identify those portions of the fire protection systems designed to Class II seismic standards whose failure could da= age Class I structures and components. Would failure of a Class II portion of the system prevent fire protection to any Class I structures or components?
9.6 In Section 9.7.2 the normal liquid and gas sampling capability is describ ed.
Provide a discussion and similar listing of the sampling capability following a loss-of-coolant accidenti include the capability to monitor hydrogen gas concentrations throughout the facility.
10.0 Steam and Power Converaion Ecuipment 10.1 Discuss the consequences of inadvertent opening of the feedwater startup valve during hot zero power operation.
Indicate the maximum reduction in primary coolant system temperature which would be experienced.
10.2 Describe the provisions made for steam generator blowdown and secondary system cleanup, indicating the magnitude of the blowdown flow and the disposal of the secondary coolant discharged.
11.0 Radioactive Waste Disposal System 11.1 Provide an estimate of the amounts of radioactivity on an isotopic basis (including tritium) that would be released daily and annually from the proposed Davis-Besse facilities. Consider both liquid and gaseous wastes, assuming that the activity of the primary coolant is that corresponding to operation with an activity level to be stated in the Technical Specifications. List all assu=ptions made and explain their bases.
Indicate how a load-following mode of operation will affect releases. Based upon the estimated available annual average dilution factor, previde an estimate of the maximum gaseous release Ibnit for this facility which would be within the itsits of 10 CFR 20.
Compare this estimated release limit with the estimated gaseous releases from this facility.
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, 11.2 Provide a list of the seismic Class I and Class II equipment, components and/or structures in the radwaste system. Summarize the design capacities, pressures and temperatures for each of the waste disposal tanks. Provide an estimate of the maximum radioactivity levels in each of the tanks in the liquid and gaseous radwaste systems and the mini =um available holdup times. Evaluate the consequences of the simultaneous release of all the radioactive liquid contained within the seismic Class II storage tanks.
Indicate whether these amounts will be limited, by design provisions or operational restrictions, to values within 10 CFR 20 limits.
11.3 Describe the Water Radiation Monitoring System, including the type of monitors used and their sensitivity.
11.4 Identify the systems and tanks that will contain radionuclides and will not be designed to prevent the release of radionuclides in the event of the design basis tornado and flood. Lis t the maximum quantity of radionuclides that would be contained in each system and tank. Provide analyses that show the maximum whole body and critical organ doses that could result from the release of radionuclides to unrestricted areas as a result of the tornado or flood. Identify all factors and assumptions used in the analyses.
11.5 Provide the following information regarding the radioactive gas waste system.
The maximum volume of primary coolant that will be bled and/or a.
degassified each year for normal and refueling operations.
b.
A list of the noble gas radionuclides and the maximum quantity of each that will be released from the primary coolant each year.
c.
A list of the fractions and quantity of each noble gas radionuclide that will remain af ter 3, 30, and 60 days holdup in the waste tanks.
11.6 Provide the following information regarding the radioactive liquid waste system.
A list of the maximum primary coolant concentrations of radionuclides a.
that are activated erosion or corrosion products and a statement of all factors, assumptions, and references used to determine the maximum concentrations.
b.
The minimum efficiencies of the primary demineralizer for the removal of the various activation products and the minimum efficiencies of the spent fuel pool, miscellaneous waste, and polishing demineralizers (Figures 9-7,11-1, and 11-2) for the removal of the various fission and activation products.
n.
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c.
The types of resins to be used in the demineralizers and factors, assumptions, and references used to determine the minimum effi-ciencies.
d.
The smallest sieve or mesh designation of the filters (Figures 11-1 and 11-2) that will remove solid matter from the liquid radioactive waste prior to discharge into the lake.
e.
An analysis of the possible accumulation in lake bottom sediments of the radionuclides discharged from the liquid waste system.
12.0 S C tv Analyses 12.1 Reactivity Transients 12.1.1 We note that your design does not incorporate a pressurizer level trip.
Provide the basis for your conclusion that this trip is not required.
State the pressurizer level assumed in your analyses of the startup accident and the accident resulting frem rod withdrawal at rated power.
State the mass of steam discharged from the secondary system during these transients.
In view of the fact that no pressurizer level trip is provided, give assurance that the integrity of the primary system would not be jeopardized if these transients were to occur with the pressurizer full.
Indicate the discharge capability of the safety valves for releasing liquid water.
12.1.2 Discuss the consequences of the startup accident occurring with minimum core flow and indicate how this operating condition will be treated in the technical specifications.
12.1.3 State the basis for your conclusion that a cold-water accident is not credible. Discuss the amount of backflow which may occur through inactive pumps when operating with one or more primary coolant pumps idle.
Indicate the volume and minimum temperature of the water in the idle portion of the loop.
12.1.4 Analyze the consequences of scramming or inadvertently moving one or all of the xenon control rod assemblies.
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. 12.2.1 Calculate the two-hour whole body and thyroid doses at the site boundary and the course-of-the-accident doses at the outer boundary of the low population zone from the rod ejection accident analyzed which results in failure of 4.1% of the fuel rods. Consider (1) ECCS operation, (2) depressurization within thermal stress cooldown limits, (3) leakage to the reactor building from the pressure housing f ailure having the minimum possible flow area, (4) primary-to-secondary leakage occurring at the anticipated technical specification limit prior to the accident and continuing in accordance with orifice flow assumptions thereafter until the decay heat removal system can be placed in operation and (5) loss of offsite power, thus making the condenser unavailable for decay heat removal. Terminate the analysis when the steam generators are no longer required for heat removal.
To aid our understanding of your analysis, supply the following:
a.
Mass of fluid discharged to the reactor building at two hours and at termination of steam generator boiloff.
b.
Mass of fluid transferred to the secondary systen for the time period in 1 above.
c.
Justification for any partition factors employed.
d.
A plot of primary system pressure versus time.
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, 12 3 Eauipment Shlfunction 12 3 1 on page IL-16 of the PSAR, it is i= plied that the auxiliary feedwater pump can be started within 83 minutes folleving a less of effsite pcver.
Assuming relief through the safety valves, what is the calculated maximum pri=ary system pressure during the 83-minute period in which reactor ecolant bo11off is required to dissipate decay heat?
12 3 2 Describe the techniques used to analyse steam generator blevdcwn folleving a steam line failure. Indicate the pressure drop and heat transfer assu=ptions e= ployed.
12 3 3 In the event of a steam line failure, what is the accunt of water which tust be boiled off in the steam generator not affected by the failure in order to depressurice the primary system? Assu=ing that the steam generator with the failed steam line was exps ;encing pri=ary-to-secondary syctem leakage at the anticipated technical specification limit, indicate the volume of primary system water which would be releaced to the at=caphere during this accident. State the basis for your assumption that unit cooldout and leakage termination can be accc=plished in three hcurs.
12.3.h State the =ax1=um stresses experienced by (1) the steam generater tube sheets in the event of a steam line rupture, and (2) the steam generator tubec and tube sheets in the event of a less-of-coolant accident.
12 3 5 Your analysis of a steam generater tube rupture assumes that the affected steam generator vill be isolated as soon as pessible. Discuss the means available to detect which steem generster experienced the failure and estimate the time required to detect and isolate it.
12 3.c Assuming that the unaffected steam generator is experiencing primary-to-secondary system leskage at the anticipated technical specificatien limit prior to the tube rupture, state the volume of primary system
'er transferred to the unaffected steam generator during the ecol.
n of the primary system.
12.3.7 To illustrate the safety targin which exists due to the inherent design of the facility, identify the largest =1ssile criginating frem the turbine which vould not penetrate the containment, ccatrol rcce, or spent fud pccl. To indicate the sensitivity of the analysis tc your assumptions regarding energy abscrption by the turbine casing, present this infermatica assuming casing energy absorptions of 0, 25, 50 and 100 percent of your best esti= ate cf absorption.
12.4 Fuel Handling 12.h.1 Discuss the provisions that vill be made to prevent the drcpping cf the spent fuel element cask into the spent fuel storage pcol. If the spent fuel element cask or other heavy objects must te soved over the spent fuel
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. storage pool, analyse the ccesequences of dropping a cask or other object into the pool. Consider the possibility of (1) fuel clad damage, (2) loss of pool water and ability to continue cooling the spent fuel, (3) damage to other equipment by flooding if the integrity of the pool liner is lost and (4) da= age to the gas decay tanks located beneath the pool if the cask can penetrate the base of the pool.
i 12.k.2 Describe the manner in which containment integrity is =aintained during refueling operations. Describe the mode of cperation of the containment building ventilation system. If the containment building is not isolated during refueling operations, provide the time required to detect and isolate any potential radioactivity releases during refueling. Indicate the potential fraction of the total fissica products reaching the centain-ment building atmosphere that could escape prior to isolation.
12.k.3 Analyze the refueling accidents in the containment and auxiliary buildings assuming damage of an entire fuel assembly, a release to the refueling water of 10% of the iodines and 20% of the noble gases from the hottest fuel assembly, a refueling water retention factor of 10 for iodines, Pasquill Type F meteorological conditions, an at=ospherie diffusion factor determined by 1
F. A. Gifford's equation (Nuclear Safety Vol. 2, No. k, June 1961) and a building shape factor of e = 0.5. Using the results of the analysis, discuss the adequacy of the facility to limit the accident doses to within the guidelines of 10 CFR 100.
12.5 Emergency Ventilation System 12 5.1 In order that we may determine the adequacy of the auxiliary and shield building ventilation systems which may be necessary to reduce the potential doses from several classes of accidents, provide (1) the radiation source terms and heat loads used for the design of air cleaning equipment in these syste=s, (2) a discussion of the ability of the filter and charcoal units to withstend desorption and/cr ignition with loss of air flev, with these heat loadings, (3) the anticipated accident atscspheric conditions to which these systems veuld be exposed, (k) a description of the design of the equipment to withstand these conditions, and (5) the method of testing the equipment to demonstrate that it will function as designed under accident conditions.
12.6 Co ntainment Pressure Response 12.6.1 Heat transfer to the heat sinks from the reacter building atmosphere is determined by a " Modified Tagami" heat transfer coefficient in the COPATTA program. Provide a discussion of the metheds used to correlate this coefficient with the Tagami, Ucbida, and Kolf1st data. Present a plot of the value of the condensing heat transfer ccefficient as a functicn of time.
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12.6.2 Discuss the manner in which the var._;1cn in turbulence in the various compartments of the containment building is conside M. in applying the
" Modified Tugani" coefficient to the entire centainment building.
4 12.6.3 Indicate hev painted surfaces within the reaeter building are censidered in the determinaticn of heat transfer to painted heat sinks. Discuss the effect on the condensing heat transfer coefficient and statc the thickness and ther=al conductivity of the paint layer assumed. List the areas of both painted metal and painted concrete surfaces within the reacter building.
12.6.k To previde n better understanding of the sequence of events in your analysis i
of the design basis accident; provide a chronolcgy indicatin6 the time after the pipe rapture occurs that the folleving events occur:
(1) core ficoding tanks start injection, (2) reactor building pressure reaches peak pressure, (3} primary system blevdcun is ce=cleted, (k) ecre f1 ceding tanks e=pty, (5)emergencyinjectionstarts,(6)containmentspraystarts,(7)airecoling units start, (8) recirculation mode starts, ard (9) containment returns to atmospheric pressure, assuming (a) two air ecoling units and one spray pump operate for containment heat re= oval, and (b) three air ecoling units and two spray pu=ps operate.
12.6.5 Provide a table indicating energy distributien prior to and folleving the
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lose.-of-ecolant accident, the a: cunt of energy generated and abscrbed from the time of pipe rupture to the time of peak pressure, and the distribution of energy at the time of peak pressure. This listing shculd include at least the following for the break sices indicated in the PSAP.: (1) pri=ary coolant internal energy, (2) core flood tank internal energy, (3) energy initially stored in the core', (k) energy stored in core internals, (5) energy 4
stored in reactor vessel tetal, (6) energy generated durin heat, (7) energy transferred to the steam generaters, (3) g shutdown and decay energy stored in piping) valves, and pumps, (9) energy in steam generator metal, (r) secondary coolant internal energy, (11) energy content of water vapor in the reactor building, (12) energy centent of air in the reactor building, (13) energy content of water in the reactor building, (14) energy centent of steel structures, (15) energy content of concrete internal structures and the reactor building valls and do=e, (16) energy re=oved by the air handling units, and (17) energy removed by the containment spray.
12.6.6 Provide plots of the te=perature of the stear-air nixture within the reactor building vs. time and of the reactor building su=p water tenperature vs. tire assuming:
(1) reactor building' heat renoval assuming (a) two air cooling units and one spray pump operate and (b) three air ecoling units and two spray pumps i
operate.
(2) e=ergency cere ecoling assuming (a) one independent pumped injecticn train operates and (b) the two independent pumped injection trains operate.
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. T 12.6.7 In order that we may assess the ability of the air handling units to function as proposed, the following information is requested:
a.
A preliminary design of the heat exchanger surfacts and fan assembly including configuration and dimensions.
b.
Heat tra- ~ar performance of the unit for the spectrum of accident anditions including flow rutes, temperatures, I
pressures and compositions for both steam-air flow and cooling water.
c.
An outline of the analytical procedures used in designing the heat transfer surface and in determining its performance and the basis for these analytical procedures.
Include fin efficiencies and the potential for liquid binding of finned surfaces.
12.6.8 Examine your piping systems to detecnine if any leakage paths from the containment building can exist in lines normally filled with liquid which may be emptied during a loss-of-coolant accident.
For example, following the cessation of high pressure injection, two check valves block the leakage of gases from the open primary system to the vented borated water storage tank.
Indicate if these valves will seat with a slight backpressure of air or if leakage will occur. Examine other valves in potentially open lines to determine if all valves sea t tightly once water is removed.
12.6.9 Provide your analysis of the potential hydrogen evolution to contain=ent during the period following the post-loss-of-coolant accident. Clearly state your assumptions. Describe your plans regarding use of combustible gas control measures, including consideration of measures which do not entail the intentional elease of fission products from the containment to the atmosphere.
12.7 Steam Line Failure 12.7.1 The analysis of the steam line failure accident in section 14.1.2.9 assumes the availability of offsite power. Provide an additional i
analysis assuming loss of offsite power coincident with the steam line failure.
In addition, provide the following information:
a.
Describe the sequence of events throughout the accident assuming the condenser heat sink and primary coolant pumps are lost.
b.
Provide a discussion and an analysis to support your conclusion that a return to criticality will not occur with a stuck rod.
This analysis should consider the equilibrium core which will have no burnable poison rods.
e...
a
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s c.
In the analysis, discuss how the primary temperature and flow in each loop varies during the accident.
d.
On page 14-19 it is indicated that operator action is required to switch the feedwater valve controller to manual to assure the startup and main control valves for the affected steam generator remain closed. Provide a discussion of the time requirements forthis operator action and evaluate potential consequences if the feetiwater valve controller is not switched to manual.
12.8 Steam Generator Tube 'ailure 12.8.1 Provide the analysis to support your assumption that failure of a single steam generator tube will not result in rupture of adjacent tubes.
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