ML19319D754

Forwards Request for Addl Info Re FSAR Sections on Site,Rcs, Design Bases for Structures & Equipment,Engineered Safety Features,Instrumentation & Control Sys,Electrical Power Sys, & Safety Analysis

ML19319D754
Person / Time
Site:	Rancho Seco
Issue date:	01/10/1972
From:	Deyoung R US ATOMIC ENERGY COMMISSION (AEC)
To:	Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT
References
NUDOCS 8003250758
Download: ML19319D754 (26)

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'80 UNITED STATES

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JAN 10 "S72 a,..

Docket No. 50-312 E. K. Davis, General Counsel Sacramento Municipal Utility District 62)1 S Street, P. O. Box 15830 Sacramento, California 95813

Dear Mr. Davis:

In reviewing your application for an operating license for the Rancho Seco Nuclear Generating Station, we find that we need additional information to complete our review of the facility as described in your Final Safety Analysis Report. The specific information required is described in the enclosure. The additional information requested has been categorized into groups which generally correspond to applicable section headings in the Final Safety Analysis Report.

We have not completed our review of the subject matters covered in this request for additional information. At a later date we will request additional information, if necessary, on these subject matters and others not addressed herein.

Our review schedule is based on the assumption that this additional infor-mation will be available for our review by March 5.1972.

If you cannot meet this date, please inforn us within 7 days of receipt of this letter so enat we may revise our scheduling.

Please contact us if you have any questions regarding the enclosed requests.

Sincerely, R. C. DeYoung, Assistant Director for Pressurized Water Reactors Division of Reactor Licensing

Enclosure:

Request for Additional information ec: David S. Kaplan, Secretary and Attorney 6201 S Street P. O. Box 15839 Sacramento, California 95813

e REQUEST FOR ADDITIONAL INFORMATION SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DOCKET NO. 50-312 2.0 SITE 2.1 With respect to the onsite reservoir, provide results of an analysis to show the capability of this reservoir to withstand the effects of a local probable maximum flood (PMT) as defined by the Corps of Engi-neers. The estimates of the effects on the dam from the PMF should include inflow, outflow, and pool elevation hydrographs and outlet rating curve (s). Sunerimpose the effects of any possible coincident wind wave activity.n the PMF maximum reservoir level as a result of 45 mph wind from the critical direction, and discuss the ability of the dam to withstand the coincident event without loss of function.

The probable maximum precipitation (PMP) pattern (estimated from U. S.

Weather Bureau Hydrometeorological Report No. 36, 1961, " Interim Re-port - Probable Maximum Precipitation in California") is tabulated below:

i Elapsed Time Since Accumulated Start of PMP*

PMP (Hours)

(Inches) 6 0.99 12 2.24 18 4.26 19 8.30 21 10.96 24 12.90 If the reservoir cannot withstand a local PMF, discuss the effects on the plant.

2.2 Provide a more definitive description of the groundwater environment, including the following specific information:

2 2.1 A map which shows the location of all existing and known potential future wells within two miles of the site, and any public groundwater supplies to the south and west which would intercept water which has passed under the site.

• Arranged to maximize runoff.

. 2.2.2 For all wells listed in 2.2.1 above, tabulate the owner, depth of well, maximum pumping rate, water elevation (under pumping con-ditions) and type of well.

2.2.3 Provide separate estimates of the horizontal and vertical perme-abilities of the foundation down to the permeable zones, and of the permeable zones themselves, in addition to average values presented in Section 2.4.6.1 of the FSAR.

2.2.4 Describe the location of onsite wells, and whether the top of the casings could provide pathways to the groundwater environment in the event of an inadvertent release of radioactive liquids.

In this respect, state whether all borings will be sealed prior to plant startup.

4.0 REACTOR COOLANT SYSTDI 4.1 To enable the regulatory staff to evaluate the adequacy of the proposed heatup and cooldown limits for this plant, provide the following information:

4.1.1 For all pressure-retaining ferritic components of the reactor coolant pressure boundary whose lowest pressurization cemperature* will be

• below 250*F, provide the material toughness properties (Charpy V-notch impact test curves, dropweight test NDT temperature, or others) that have been reported or specified for plates, castings, forgings, piping, and weld material.

Specifically, for each component provide the following data for materials (plates, pipes, forgings, castings, welds) used in the construction of the component, or your estimates based.on the available data:

(1) The highest of the NDT temperatures obtained from dropweight

tests, (2) The highest of the temperatures corresponding to the 50 ft-lb value of the C fracture energy, and

• Lowest pressurization temperature of a component is the lowest-

emperature at which the. pressure within the component exceeds 25 per-cant of the system normal operating pressure, or at which the rate of temperature change in the component material exceeds 50*F/hr., under normal operation, system hydrostatic tests, or transient conditions.

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1 (3) The' lowest of the upper shelf C energy values for the " weak" direction (WR direction in plates) of the material.

4.1.3 Identify the location and the type of the material (plates, pipes, forgings, castings, welds) in each component for which the data listed above were obtained. Where these fracture toughness parameters occur in more than one plate, pipe, forging, casting or weld, provide the information requested in 4.1.1 for each of them.

4.1.3 For reactor vessel beltline materials, including welds, provide the in-formation requested in 4.1.1 and 4.1.2 and in addition specify:

(1) The highest predicted end-of-life transition temperature corre-sponding to the 50 ft-lb value of the Charpy V-notch fracture anergy for the " weak direction" of the material (WR direction in plates) and (2) The minimum upper shelf energy value which will be acceptable for continued reactor operation toward the end-of-survice life of the vessel.

4.1.4 Furnish the proposed heatup and cooldown curves which will be used to control the pressure and temperatures to which the ferritic material of the reactor coolant pressure boundary will be exposed during opera-tion of the plant until the scheduled removal of the first material capsule.

4.2 Describe the plans which were followed to avoid partial or local severe sensitization of austenitic stainless steel during heat treatments and welding operations for core' structural load bearing members, component parts of the reactor coolant pressure bo"adary and component parts.of the ECCS and other safety related systeca. Describe welding methods, heat input, and the quality controls that were employed in welding austenitic stainless steel components.

If nitrogen is to be added to stainless steel types 304 or-316 to cahance its strength (as permitted by ASME Code Case 1423 and USAS Case 71),

provide justification that such material will not be susceptible to sensitization at heat affected zones during welding.

4.3 Since the process of electroslag welding will be used in the fabrica-tion of components within the reactor coolant boundary (i.e., steam generator shell and pressurizer shell), describe the process variables 9

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and the quality control procedures applied to achieve the desired material properties in the base metals, heat affected zones, and welds.

4.4 Indicate whether electroslag welding will be used in fabrication of any other components, particularly those made from stainless steel.

4.5 Section XI of the ASME Boiler and Pressure Vessel Code recognizes the problems of examining radioactive areas where access by personnel will be impractical, and provisions are incorporated in the rules for the examination of such areas by remote means.

Because the equipment used to perform these examinations may require development, the future examination of these areas is dependent upon providing the access and space requirements as dictated by the latest equipment development programs. In this regard, provide the following:

4.5.1 Describe the remote equipment planned or under development to perform the reactor vessel and nozzle inservice inspections for your plant.

4.5.2 Describe the system to record and compare the data from the baseline inspection with the data which will be obtained from subsequent in-service inspections.

4.6 Provide the following information regarding the reactor coolant leakage detection system:

4.6.1 Describe the methods that will be used to detect leakage from the reactor coolant system. Provide sufficient detail to indicate that redundant systems of diverse modes of operation will be installed in the plant.

4.6.2 Describe the methods used to provide positive indications in the control room of leakage of coolant from the reactor coolant system to the containment.

4.6.3 Describe the adequacy of the proposed leakage detection systems to dif-

.ferentiate between identified and unidentified leaks from components within the primary reactor containment and indicate which of these sys-tems provide a means for locating the general area of a leak.

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Provide the sensitivity (in gpm) and the response time of each system.

For the containment air activity monitors, provide the sensitivity and the response time as a function of the percentage of failed fuel rods e

or of the corrosion product activity in the reactor coolant, as applicable.

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+ 4.6.5 Discuss the adequacy of any system which depends on reactor coolant activity for detection of changes in leakage during the initial period of plant operation when the coolant activity may be low.

4.6.6 Specify the proposed maximum allowable total leakage rate for the reactor coolant pressure boundary, and the basis for the proposed limit.

Furnish the ratio of the proposed limit to:

(1) The normal capacity of the reactor coolant makeup system.

(2) The normal capacity of the containment water removal system.

4.6.7 Estimate the anticipated normal total leakage rates and major leakage sources on the basis of operational experience from other plcnts of similar design.

4.6.8 Specify the proposed maximum allowabis leakage rate from unidentified sources in the reactor coolant pressure boundary, and the basis for the proposed limit. In this regard provide the following information:

(1) The length of a through-wall crack that would leak at the rate of the proposed limit, as a function of wall thickness.

• (2) The ratio of that length to the length of a critical through-wall crack, based on the application of the principles of fracture mechanics.

(3) The mathematical model and data used in such analyses.

4.6.9 Provide the time interval in which the reactor will be shut dowr. if either the total or unidentified leakage rate limit is exceeded.

4.6.10 Describe the proposed tests to demonstrate sensitivities and operability of the leakage detection systems.

4.7 For the predicted NDT temperature shift of 250*F (FSAR, page 4.3-9) at least 5 capsules are required by the AEC proposed " Reactor Vessel Material Surveillance Program Requirements," 550.55a, Appendix H, pub-lished in the Federal Register on July 3, 1971. Each of these sur-veillance capsules should include specimens from the base metal, heat-affected zone and the weld metal, as recommended in the ASTM E-185, Section 3.3.

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.Section 4.3.3.4 of the FSAR refers to the report BAW-10006 for the description of the surveillance program consisting of 6 capsules, only 3 of which contain weld metal specimens.

1 Provide the proposed capsule withdrawal schedule, and describe how the.

fracture toughness of the weld metal will be monitored throughout the life of.the reactor vessel in view of the small number of capsules con-taining weld metal specimens.

4.8 Provide the following information regarding the primary coolant pump flywheels:

4.8.1 The type of material to be used for the pump flywheel and its minimum specified yield strength. Indicate the nil-ductility transition (NDT) temperature specified for the material, as obtained from dropweight-tests (DWI),.the minimum acceptable Charpy V-notch (C upper shelf energy level in the " weak" direction (WR orientation En) plates), and the fracture toughness of the material at the normal operating tempera-ture of the flywheel.

4.8.2 The design stress specified for the flywheel as a percentage of the-minimum specified yield strength, for the normal operating speed and the design overspeed condition.

4.8.3 A discussion as to whether the calculated combined primary stresses in the flywheel at the normal operating speed, will include the stresses due to the interference fit of the wheel on the shaft, as well as the stresses due to centrifugal forces.

4.8.4 The highest anticipated overspeed of the flywheel and the basis for this assumption.

t 4.8.5 The estimated' maximum rotational speed that the flywheel attains in the event the reactor coolant piping breaks in either the suction or Ldischarge side of the pump.

4.8.6-A discussion of the results of any studies directed towards:

(1) determining the maximum speed' the pump or motor can reach due to physical limitations (e.g., the speed at which the pump -impel-1er seizes in the wear rings due to growth from centrifugal. forces or.the speed at which motor parts come loose and grind or bind to prevent further increase in speed);

'(2): establishing speed and torque for various pipe break sizes; I

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. (3) devising means to disengage the motor from the pump in the event of pum,' overspeed; (4) verifying that pump fragments generated at maximum speed do not penetrate the-pump casing and that any missiles leaving in the blowdown jet do not penetrate containment; (5) establishing failure speeds for motor parts and whether they will penetrate the motor frame and if so with what energy; and (6) defining a minimum rotor seizure time.

4.8.7 The rotational speed that will be specified for the preoperational overspeed tests of the flywheel.

4.8.8 The frequency of inservice inspections, and the acceptance criteria for the flywheels.

5.0 DESIGN BASES - STRUCTURES AND EQUIPMENT 5.1 Several Class I (seismic) tunnels, underground piping and underground cables are connected to the main structures. The seismic response of-buried elements is different from the response of the main structures.

Describe the methods of design and the actual arrangement of the de-tails of connections between these appendages and the main structures.

Provide information to verify that stresses due to dynamic inter-action meet the criteria for allowable design stresses.

5.2 Although the foundations of the main Class I (seismic) structures are not interconnected, soil interaction will exist between them during OBE and DBE disturbances. Explain the method used to account for this condition in the design and construction of the foundations.

5.3 The bottom of the base slabs of Class I (seismic) structures are well below finished grade. Explain the dynamic interaction between soil and structures for the OBE and DBE and demonstrate that critical stresses generated in the soil and in the walls and base slabs of the Class I (seismic) structures during these occurrences meet the criteria for allowable design stresses.

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. 5.4 The ACI 318-63 Code in Sections 2603(a) and 2603(b) indicates that:

" Stresses and ultimate strength shall be investigated at service conditions and at all load stages that may be critical during the life of the structure from the time prestress is first applied, and

" Stress concentrations due to the prestressing or other causes shall be taken into account in the design."

In the commentary on this Code, the same Committee states, in Section 2603:

"...the dasign investigation should include all load stages that may be significant."

The ACI Committee 334 in " Concrete Shell Structures Practice and Commentary" states in Sections 202(d) and 202(e) that:

" Equilibrium checks of interns 1 stresses and external loads are to be made to ensure consistency of results.

"An ultimate strength analysis may be used only as a check on the adequacy of the design.

It is not to be used as a sole criterion for design, except where it can be proven to be applicable."

In its commentary the same Committee 334 says in Part 4 that:

"...the analysis must be made at design and at ultimate loads to ensure both proper behavior at design loads and an adequate overload capacity."

In this regard, describe whether the design and the checking of the design have been made in accordance with the provisions indicated above, and whether all service loads wera considered. Also, indicate whether the compatibility of strains was checked at all critical locations.

5.5 The accuracy and reliability of the computer programs is generally verified by selecting appropriate elasticity problems for which the solution can be validated b-ther acceptable methods of analysis and then comparing that solution with the computer solution.

Explain whether this validation of the computer programs which have been applied has been done for problems involving structures of similar geometric shape, similar size and materials and with similar loadings

.and boundary conditions.

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5. 6 The buttresses have a moment of inertia approximately 3.5 times as large as the wall and their stiffness may introduce aa important perturbation in the behavior of the axi-symmetric structure.

In this regard, discuss the error introduced in the design by the assumption of an axi-symmetric structure.

5. 7 Present a table listing separately the programs that have been used for analyzing axi-symmetric loads and no ' syrmetrical load distribu-

.l tions for different parts of the containment. ' Indicate the matching lines when different programs have been used for the same structure, i

5. 8

' Discuss whether the computations for the structural analysis of the containment are based on ACI-505 to account for the effects of con-crate cracking, or if a finite-element program which accounts for cracking was utilized. If a cracked-section finite-element program l

was used, refer.to an available technical document which describes the 4

' method or ' provide a ' detailed description on the exact techniques used within the program which justifies the design' adequacy of the program

.with respect to the effects of concrete cracking.

5.9 Indicate the areas in which modifications of the modulus of elasti-city are required as a result of the nonlinear stress-strain relation

'at zones of high compressive stresses. Specify the magnitude of the maximum compressive stresses and demonstrate that the modifications of the modulus of elasticity adequately take into account the non-i 1

linear stress-strain. distributions in the design of the specified areas.

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5. 10

~The ACI 318-63 Code is essentially a code for framed structures where stresses are mostly uni-axial. ;In the containment, stress distribution is mostly'tri-axial. Experimental evidence exists that when one or two of the three principal stresses are tensile stresses, the ultimate compressive strength of concrete decreases by a large amount. There-fore, the different stress limits established by the code such as:

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' 0.45 f' ; 0.60 f' ; 6 /ET; 3.5 /pr 2/pr 1,1/yr may not be applicable, unless"some red 6ction of thesE; values; was c8n,sidered.

c In this regard, l

provide information to verify ~that the containment structure as'de-signed and built, has sufficient safety margins, despite the fact that these corrections have not been considered.

5.11 Provide information to verify that the earthquaks shears can safely bo ' carried by the containment structure concrete without special shear reinforcing.

Include in the discussion the influence of.the buttress plate, the manner in which the vertical seismic shears are transmitted through the buttress plate from one part of the buttress to the other 1part and substantiation that with zero friction the structure will be I'

stable.

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. 5.12 Tests on models (see the paper by Isao Toriumi "Model Analysis of Aseismic Design of a Nuclear Reactor Building" Nuclear Structural Engineering 2 (1965) pp. 301-305) indicate that differential radial motion, the so-called "lobur motion" and not symmetrical modes may be of importance. A similar conclusion is indicated in " Dynamic Stress Analysis of Axisymmetric Structures Under Arbitrary Loading" by Sukumar Ghosh, et al. report #EERC 69-10 September 1969, University of California. Indicate-to what extent the stresses and deformations due to these modes were considered in the design of the containment wall and dome. If they were omitted, justify this omission.

i 5 13 Prasent the results of a study which analyzes the stresses in the containment structure at end of service life. This study should in-clude the effect of structural and thermal creep and shrinkage on stress redistribution, and should be based on a realistic evaluation of the modulus of elasticity E, and Poisson's ratio. The 93e of a uniform reduced E is not considered satisfactory. The purpose of this study should be to demonstrate that after repeated reactor shut-downs and reactor startups with resultant thermal cycles, to excessive tensile stresses will exist in the concrete at the liner where there is no reinforcing to control cracking.

5.14 Justify the omission of radial reinforcing in the wall and in the dome of tha containment building. Prestressing generates tensile stresses and tensile strains in the direction perpendicular to the tendon, especially where tendons are curved. Demonstrate that the absence of radial reinforcing will not jeopardize the long range stability of the containment.

Indicate the ultimate tensile strength of the concrete if the design relies upon it.

Where radial reinforcing is provided, justify its design.

5. 15 For the base slab the following considerations are generally required to establish the adequacy of design:

_the slab carries a load parallel to its plane.

- the slab carries a heavy thermal load due to transient and steady state thermal gradients.

- during a seismic disturbance the slab carries loads which are not distributed'in an axi-symmetrical pattern.

- the slab is supported on an elastic foundation, for which the elastic constants are known only in an approximate way.

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I Present the results of a stress analysis of the containment building i

slab which demonstrates how these considerations have been factored into the design and include a justification for the omission of any shear reinforcing. Also, describe the construction procedure for the slab.

Indicate on a sketch the location of the reinforcing bar splices and-on the same sketch the location of the maximum stresses in concrete and in steel (for concrete: ' compression, tension, and shear stresses).

1 5.16 Explain what provisions are made for the seismic design of the con-tainment equipment, personnel, and escape locks.

5.17 The 1/4-inch thick liner is attached to the concrete by means of an anchor grid system with the unpainted liner face in contact with concrete. Under normal operating conditirns considering prestress load, concrete creep, concrete shrinkage and thermal load, the com-pression stress in the liner may reach the yield point.

In the_ event some plates buckle and deflect towards the inside of the containment, voids which may therefore occur between concrete and liner may be connected with the outside through cracks in the concrete. In this

-regard explain what provisions are made to prevent the unpainted liner from corroding under those conditions and what surveillance measures could be used to detect this condition.

5.

3 The liner is not backed up by concrete at openings and behaves partly as an element of a pressure vessel.

Indicate the provisions incor-porated in the structure for the safe transmission of membrane and bending stresses-existing in these transition pieces into the rest of the structure. Provide information to verify that critical stresses or strains occurring at these locations meet the' criteria for allow-able design stresses and strains.

5.19 The use of the ASME Pressure Vessel Code Section III for the design of-the liner is questionable since the Code applies to a structure supporting mechanical loads, such as pressure. The liner is loaded by strains transmitted to it by concrete and must follow the concrete 4

deformations. -In this case the thickening of the liner at the openings does not help to reduce the stresses but only increases the local stiffness of the liner, which, '.n turn, may increase local stress con-centrations. Provide inform.cion to sut stantiate the manner by which liner failure or leakage -211 be preclud'd since the ASME Code rules-may not;adequatelv over this situation.

5.20 Explain the design methods used for the expansion joint bellows for

. axial and lateral displacement, especially for the fuel transfer tube iw n

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5.21 Describe the provisions made in the design and' construction of Class II (seismic) structures in order to prevent them from damaging adjacent 3

Class I (seismic) structures under OBE or DBE loads.

5.22 Where Class I (seismic) structures are directly connected to Class II (seismic) elements such as equipment and piping systems, indicate the manner in which the influence of seismic activity of the Class II (seismic) elements on the Class I (seismic) structures has been con-sidered in the Class I (seismic) designs, such that damage or excessive movement of Class II (seismic) systems will not adversely affect Class I (seismic) structures.

5.23 The factored load equations used for the design of Class I (seismic) structures are not in accordance with the equations presented in the ACI 318-63 Code.. Justify their use.

Specifically, compare the safety factors actually provided with the safety factors resulting from the use of the ACI Code.

5.24 Describe in detail and provide sample sketches of the details of the anchoring and reinforcing arrangement for the Class I (seismic) auxiliary building, control room, fuel handling building, contain-ment interior structure, and the equipment supports for the reactor, steam generator, pressurizer, and main pumps.

5,25 For the containment interior structure, describe the static and dynamic design used to ascertain that this structure meets Class I (seismic) design criteria.

Explain the methods used to compute the jet forces for the structural design of the containment interior structure.

In-dicate the pressure and temperature safety margine resulting from the stress analyses of the interior structures.

5.26 Describe the method used to evaluate the amount of radiation-generated heat in concrete immediately adjacent to the reactor and the criteria and methods ~used for the design of the reinforcement at this location.

5.27 Indicate the design criteria, materials, allowable stresses and strains, and load combinations considered for the design of the reinforced concrete and cast-in steel for the structural supports for the-l' reactor, steam generators, pressurizer, main coolant pumps, and safety injection tanks.

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5.28 Describe the tie down ' arrangements made for all Class I (seismic) equipment to resist seismic forces. Restraints provided for equipment usually have gaps between the restraints and the equipment.

Indicate whether the seismic analysis considered these gaps and if the impact forces due to the gaps acting on the equipment, the restraints, and the

" Dyna ic Analysis of Mechani-structures were evaluated.

(Refer to:

m cal Systems with Clearances" Part I - Formation of Dynamic Model and Part II - Dynamic Response by Dubowsky, C. and Furdenstein, F.,

Jour.

Engin. Industry Trans. ASME 93(1) Feb.

• 1.)

5.29 Present a list of typical missiles whien have been considered in t!M design, and indicate the nature of the missile, the weight, mass tJ cross-section ratio, shape, assumed point of impact, assumed impac$

velocity, and where and how originated.

Indicate the criteria and the method of analysis used for checking the structure at the point of impact. Indicate whether all Class I (seismic) structures or parts of structures have been checked for impact of missiles.

Indicate

'whether the crane can generate internal missiles which may endanger the containment structure.

1 Describe the provisions made to tie down all removable slabs, blocks, 5.30 or partitions to prevent them from becoming missiles.

15.31 With respect to the spent fuel pool, discuss whether cracking from thermal stressas was considered in the design of the pool walls and the provisions made to control cracking in this structure. To sub-stantiate whether or not cracking is a potential problem, state the predicted maximum thermal stresses which can be developed in the pool walls under the most severe anticipated thermal conditions, and l

considering the combined stresses due to the worst anticipated loading conditione.

1 5.32 Where yielding is permitted in structures, define the allowable limits and list the ' structures,. parts of structures, and elements where yielding may occur and under what load combinations. Provide infor-l mation to show how these limits meet acceptable design strain limits.

5.33-If Part 2 of the AISC Specification of February 12, 1969, was used in the design of any Class I (seismic) structural components, define the components on which these criteria were used.

l 15.34 For the design of the containment structure and other Class I (seismic) u structures,- describe the technique used for the proportioning of the reinforcement after completion of the computer analysis.

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5.35

- For several of the critical locations of Class I (seismic) structures, including the' containment, indicate separately the seismic stress contributions as a portion of the total stress.

5.36 The Proposed Revisions of 1970 for the ACI 318-63 Code define a more conservative approach for computing the flexural cracking load at a section. Explain whether this conservatism of the 1970 proposed revisions was incorporated in the design.

5.37 It is noted that the Class I (seismic) reinforced concrete structures are designed in accordance with Building Code Requirements for Rein-forced Concrete, ACI 318-63, including the 1970 Proposed Revisions.

Describe what structural components are designed for shear using the components of Section 11.15, Shear Friction, of the 1970 Proposed Revisions. In addition, indicate the origin of the loads that are being designed for by this technique. For those components designed by this concept, indicate the value of p, the coefficient of friction that is assumed.

5.38 During an ode and a DBE, torsional loads will be applied even to sym-metrical structures (see the paper by N. M. Newmark " Torsion in Symmetrical Buildings" Fourth World Conference on Earthquake Engineer-ing, Santiago, Chile, 1969).

Indicate whether torsional effects were considered for the containment, the internal structure, and other symmetrical and nonsymmetrical Class I (seismic) structures.

Indicate the structural elements which carry these loads and provide information to verify that the corresponding critical stresses meet the criteria 4

for allowable design stresses.

5.39 A high degree of assurance is needed that the spent fuel pool is ade-quately designed and constructed to withstand the effects of a dropped fuel cask. In this regard, provide the following:

(a) The assumptions and results of an analysis which evaluates the design adequacy of the fuel pool floor and walls to withstand, without leakage which would uncover the fuel, the impact of a dropped fuel cask from the maximum height to which it can be lifted by the crane.

(b) Describe applicable test information that would be available to support the analytical assumptions and results of (a) above.

Also discuss the extent to which the analysis and any available test information applies to the cask contacting the pool floor at various possible angles.

m

. 5.40 Describe the sampling program used for fresh concrete. ASTM C-172 does not. indicate the location of sampling except to state that it is normally performed as the concrete is delivered from the mixer to the conveying vehicle. To permit an evaluation cf the adequacy of the system and controls exercised to assure the quality and proper place-ment of the poured in-place concrete, provide information on the type of concrete delivery systems used, such as a central mix plant with agitator trucks, conveyors and pumps or other combinations.

In-dicate at what location concrete compression and slump test samples

. were taken during construction.

5. 41 For the tendon assemblies:

(a) Describe the method of erection of the tendon anchor, especially the means by which a good concrete bearing of the tendon bearing plate has been achieved and what quality control has been pro-vided in order to assure good bearing.

(b) It is understood that no stress analysis has been made for the different parts of the tendon anchor and that the only knowledge of safety margins in the prestressing system is based on loading tests up to the ultimate strength of the tendon. Describe what real safety margin may exist against over-loading and other unexpected contingencies.

5. 42 If fly ash has been used in any concrete work, provide a typical chemical analysis of the fly ash and provide information to show that it will not increase the probability of corrosion of the prestressing tendons.

5. 43 Describe how the seal welds between the crane bracket and thickened liner plate are checked for leaktightness. These are working welds under normal operation and may have a substantial probability of failure' relative to typical liner plate seam welds.

5. 44 Indicate whether any welding has been used on reinforcing steel.

If so, describe the design criteria for the weld and the quality control used during welding operations. Also state whether any protective 1

measures were taken during welding near unprotected prestressing tendons.

5. 45

. Describe the permanent provisions made for access to the exterior upper l

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, *% -parts of the containment structure to facilitate periodic inspection

_and testing.

5. 46 The acceptance pressure testing should provide a check,on design and construction of the containment. In this context, justify the number of meridians used for taking measurements during pressure testing, and substantiate that the information thus collected will be

sufficient to fulfill the test purpose. Indicate the tolerances on test acceptability..

5. 47 Since it has not been' established how many surveillance tendons are sufficient or that the frequency of testing is adequate to assure continued performance ~ capability of the containment, provide the following:

(a) The program for surveillance of the containment structure con-l crete, tendons and liner during the lifetime of the plant.

(b) Discussion of tendon inspection on the basis of a sampling pro-cedure as specified in the MIL-STD-105D or MIL-STD-414 or similar standards.

t (c) A description of the bases on which the decision is made to select the tendons for surveillance.

(d) Infctmation as to whether the individual wires will be checked for breakage-on the surveillance tendons and on other tendons.

If they will be checked, indicate the procedure which will be followed.

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5. 48 Discuss what provisions have been made for the cathodic protection of the steel liner, reinforcing bars, tendons, and the tendon steel 2

casings, and the protection such a system may offer, especially to the prestressing tendons.

Include a discussion of the detrimental effect the cathodic system may have if small faults, such as voids in tendon protective grease and discontinuities in the tendon ducts,.

should. occur.

6.0 ENGINEERED SAFETY FEATURES 6.1 Indicate the degree of compliance of the containment leakage testing

. program with the AEC proposed " Reactor Containment Leakage Testing for Water Cooled Power Reactors," 559.54(o), Appendix J, published in the Federal Register on August 27, 1971.~

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6.2 Describe the design features.of the containment airlocks that will permit testing of the airlocks at the calculated peak containment pressure. Describe the test method that will be used to verify leak tightness of air lock doors, door penetrations and door gaskets.

6.3 Identify which containment surfaces are covered with the protective coating listed in Table 14.4-2 of the FSAR, and the thickness of this coating.

~

- 7.0 INSTRUMENTATION AND CONTROL SYSTEMS 7.1 Provide the following information with regard to the protection systems which actuate reactor trip, engineered safety feature action, and other safety related systems:

7.1.1 A list of those systems designed and built by Babcock and Wilcox (B&W) that are identical to those of the Oconee Nuclear Station (as documented in the FSAR) and a list of those that are different, with a discussion of the design. differences.

7.1.2 A list of those systems and their suppliers that are designed and/or built by suppliers other than B&W.

7.1.3 An identification of, and justification for, those features of the design that do not conform to the criteria of IEEE-279 and the applicable portiva of the Commission's Gener;l Design Criteria.

7.2 Provide the following information with regard to the control systems designed by B&W:

7.2.1 An identification of the major plant control systems (e.g., primary temperature control, primary water level control, steam generator l

water level control) that are identical to those in the Oconee Nuclear Station.

7.2.2 A list and a discussion of the design differences in those sys-tems-that are not identical to those used in the Oconee Nuclear Station. This discussion should include an evaluation of the safety significance of each design difference.

7.3 Describe the seismic design criteria for the reactor protection sys-tem, engineered safety feature circuits, and the emergency power system. The criteria ahould address the capability to initiate a protective action during the design basis earthquake and the capability 9

. of the engineered safety feature circuits to withstand seismic dis-turbances during post-accident operation. Describe the qualification testing requirements that have been or will be used to assure that the seismic design criteria for the above cited systems are satisfied, the means by which these requirements are being imposed on equipment suppliers, and a summary of the successful completion of qualification tests for each type of equipment.

7.4 Describe the criteria and their bases that established the minimum requirements that werd used for preserving the independence of redun-dant reactor protection systems, engineered safety feature systems and Class IE* Electrical Systems through physical arrangement and availability during any design basis event. Your response should include a discussion of the administrative responsibility and control to assure compliance with these criteria during the design and instal-lation of these systems. As a minimum the criteria and bases for the installation of electrical cable for these systems should address:

(a) Cable derating, maximum percentage fillT in trays, and minimum spacing between trays.

(b) Cable routing in containment, penetration areas, cable spread-ing rooms, control rooms and other congested or hostile areas.

(c) Sharing of cable trays with non-safety related cables or with cables of the same system or other systems.

(d) Fire detection and protection in the areas where these cables are instal:41.

(e) Cable and cable tray markings.

(f) Spacing of wiring and components in control boards, panels, and relay ra ks.

7.5 In regard to the design bases related to the capability of the engi-neered safety features (electrical and mechanical equipment) and

-reactor protection system to perform their intended functions in the combined post-accident environment of temperature, pressure, humidity and radiation, provide the following:

• Class IE electrical systems and design basis events are defined in the IEEE Criteria Jor Class IE Electrical Systems for Nuclear Power Generating Stations (IEEE-308).

l e 7.5.1 Identify all safety related equipment and components (e.g., motors, cables, filters, pump seals, shielding) located in the primary containment and elsewhere that are required to function during and subsequent to any of the design basis accidents.

7.5.2 Describe the qualification tests and analyses that have been or will be performed on each of these items to assure its availability in a combined high temperature, pressure, humidity and radiation environment. Include the specific values of temperature, pressure, humidity and radiation, noting that the accident conditions should be superimposed on the long term environment to which she equip-ment in question is normally exposed.

7.5.3

-A summary of the successful completion of qualification tests for each type of equipment.

7.5.4 The expected margin with regard to the temperature rating, con-sidering the maximum environmental temperature, the design tempera-ture rise and the class of insulation used.

7.6 Describe the methods planned for periodic inservice testing of reactor

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protection system response times for the various trip parameters.

In-clude a discussion of the relationship of channel response time to safety limit and identify the margin in terms of time.

7.7 Describe the methods planned for periodic inservice testing of engi-neered safety feature instrumentation and control equipment. We interpret IEEE-279 to require the same high degree of online testability for engineered safety feature actuation as is required for the reactor trip system.

7.8 With respect to the display of information discussed in Section 7.4.2 of the FSAR, delineate the surveillance readouts or indications that will be provided to the operator for monitoring conditions in the reactor, the reactor coolant system, and in the containment over the full operating condition of he plant. Include the design criteria, the type of readout, and the lumber of channels provided including their range, accuracy and location.

7.9 Discuss the extent to which the operational bypasses described in Section 7.1.2 of the FSAR meet the' intent of Paragraph 4.12 of IEEE-279 (1971).

7.10 Describe the criteria and design bases pertaining to the heat tracing.

requirements, temperature control, monitoring, and power requirements

. for the boric acid tanks, borated water storage tanks, containment spray additive tanks (if any), and the piping associated with these systems. Discuss the consequences of a single failure in the heat tracing of each of the.above mentioned systems.

7.11 The FSAR states that pump interlocks assure agains ccidental startup of a cold loop if reactor power is greater than 22 percent.

In this regard, discuss the criteria to which these interlocks are iasigned and the consequences of an interlock failure.

7.12 With respect to the reactor protection system, identify any reactor trip set point adjustments that must be made when operating in a restrictive mode such as power operation with one or more pumps not operating. Describe how these trip set point adjustments will be made, and verify that these adjustments will be made in accordance with Section 4.15 of IEEE-279. Also, discuss whether the operating transient-limits delineated in. Table 7.2-1 should be restricted when one or two pumps are not operating, and describe how these more restrictive limits are automatically inserted.

7.13 Identify those trip set points (if any) of the reactor protection sys-tem and engineered safety feature system that are within 5% of the high or low and of the calibrated range. To verify that the required output signal is always conservative when viewed from a safety standpoint, provide the results of an error analysis.for each such case.

7.14 Describe the radiation monitoring system in more detail including the design confidence level, the calibration methods to be employed, and the in-service testability provided. Discuss the redundancy and independence of the process radiation monitors, and provide a block /

logic diagram of the systems that include the following:

(a) Alarms and alarm logic (b) Control interlocks (c) Computer outputs

_ (d) Number and type of detectors

.(e) Power requirements and sources 7.15 The proper functioning of_the core flooding tanks is required to con-trol the consequences resulting from postulated loss-of-coolant

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. accidents. If the motor-operated isolation valve in the line connect-ing the flooding tanks to the primary coolant system should be closed inadvertently prior to or during an accident, the ECCS might fail to perform in an acceptable manner.

An acceptable degree of protection would be provided if the control circuit for the motor-operated isolation valves between the flooding tanks and the primary coolant system were designed to meet the intent of IEEE-279 and to incorporate the following features:

(a) Automatic opening of the valves when (1) primary coolant system pressure exceeds a preselected value (specified in the Technical Specifications) or (2) a safety injection signal has been initiated.

(b) Valve position visual indication that is actuated by sensors on valve ("open" and " closed").

(c) An audible alarm, independent of item "b," which is actuated by a sensor on the valve when the valve is not in the fully open position.

(d) Utilization of a safety injection signal to automatically remove (override) any bypass feature that may be provided to allow a motor-operated valve to be closed, for short periods of time, when the primary system is at pressure (in accordance with the provisions of the Technical Specifications).

Provide a discussion of the design of the control circuit for those motor-operated isolation valves, and indicate your plans and schedule to modify the design either to provide the above features or to con-form to other criteria that provide an equivalent degree of protection.

7.16 The FSAR states that flexibility in controlling rate of reactivity change is provided by a patch panel that allows individual rods to be switched from one group to another. Discuss the interlocks and/or administrative procedures that will be used to assure that rod pattern changes made in this manner will be acceptable.

7.17 The decay heat removal (DHR) system contains motor-operated iso-lation valves in the letdown line connecting the relatively low pres-sure DHR system to the high pressure primary coolant system.

In our view, the valving system should be designed to provide protection against over-pressurization of the DHR system that could result from common mode failures or operator errors. The following design features for the motor operated valves in the letdown line between 'the high

4 9

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t pressure primary coolant system and the relatively -low pressure DHR would provide an acceptable degree of protection:

t (a) Provision of at least.two valves in series, with each valve inter-l i

locked to prevent valve. opening unless.the primary system pressure

-is below the DHR system design pressure.

(b) Interlocks of diverse principles, and designed to meet the intent cf IEEE-279.

(c) Provision for automatic closure of the two valves in series whenever the pressure in the primary coolant system exceeds a selected fraction of the design pressure of the DHR system. These closure

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devices should be designed to the intent of IEEE-279.

ProvidaLinformation to indicate your plans and schedule to mod'fy the j'

design of the DHR system valving either to include these desi h b

. features or to confora to other criteria that provide an equivalent

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degree of protection.

7.18 Since all the nuclear power range channels are ahared by the protec-I tion and control systems, discuss the extent to which the design meets the requirements of Section 4.7 of IEEE-279 (1971). The discussion should include:

(a) The results of tests and analyses which verify that no credible faults within or downstream of the isolation devices can prevent the protection system from meeting its minimum performance requirements, (b) The methods used to ' isolate the averaging circuits from the pro-taction system, and l(c) The results of tests which verify that the isolation capability is not degraded with time.

If these degradation tests have not l been performed, indicate thel in-service methods that will be used

-to verify that unacceptable degradation has not occurred.

i

.8.0 ELECTRICAL POWER SYSTEMS l' ~

l8.1 Identify.all aspects-of.the safety related portion of your standby electrical power system that do~ not conform to. Safety Guides 6 and 9, and IEEE-308. Provide the bases for any deviations.

I 8.2

- Describe'in more detail the auxiliary equipment of the-.mergency diese1~

generator system.' The description should include the fuel storage and I

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. transfer system, the source of power for control, the starting system and number of start attempts provided, method of cooling and warming the engine, the control and protection system including relevant schematic diagrams, logic diagrams depicting interlocks and related constraints, and equipment protective interlocks.

8.3 The tabulation of connected loads to the diesels, as shown in Table 8.2-2 of the FSAR, is insufficient to evaluate the adequacy of the design with respect to Safety Guide 9.

Provide data to include actual starting KVA loads (in addition to the expected loads shown), and the time required for the various loads to reach full speed. To correlate these data with the capability of the diesel generator, provide the following information:

8.3.1 A load profile during a LOCA showing the timing sequence and the time duration of the various loads subsequent to diesel star'.

Superimpose the anticipated starting KVA requirements for the various loads at the appropriate portion of the load profile, and indicate the computed duration of the transients and the calculated system voltage.

8.3.2 Indicate the maximum loads that can be incrementally added to the various block plateaus without exceeding the recommendations of Safety Guide 9.

8.3.3 Indicate the continuous, 2000 hr., and 30 min. rating of the diesel engine.

8.3.4 State the generator's X Xj X"

and SCR and the unit's WR.

In addition,describetheNy,peo,fNx,citationsystemprovidedandthe response time of same for voltage regulation during the various step load changes.

8.4 Discuss the design criteria for the diesel generator rooms with respect to ability to prevent missilca, explosions and fires in one unit from affecting the redundant unit.

8.5 Identify the sources of control power to the 220 KV switchyard breakers.

Provide the results of an analysis to show that no single failure in these power supplies,' control circuits and protective ralaying will negate the ability to provide offsite power to the engineered safety features.

8.6 Describe the monitoring features that are provided to continuously indicate that the capability of the station battery to supply. power is-

. not degraded. Consider the relevance of the monitored par.ameters to the actual charge stored in the battery and discuss the limita-tions of the system to ensure disclosure of battery degradation.

Also, describe the protection that will be provided against overcharging.

8.7 Seismic qualification testing of d-c systems should include tests of the batteries (cells) as well as the auxiliary equipment and battery racks. Provide the results of such tests, including extrapolations that account for degradation due to time.

If these tests have not been conducted, indicate when they will be conducted and the scope of the tests.

14.0 SAFETY ANALYSIS 14.1 To permit us to verify the design capability of the containment with respect to pressure response following the design basis loss-of-coolant accident, provide the following:

14.1.1 A table of blowdown masses and enthalpies versus time for the design

_basis coolant pipe break area.

14.1.2 An accident chronology for the loss-of-coolant accident that includes at least the following:

EVENTS TIME, SECONDS Design Break Occurs 0.0 Core Flooding Tanks Start to Inject Water Containment Reaches Peak Pressure End of blowdown Core Flooding Tanks Empty ECCS Starts

• Refueling Water Storage Tank Empties

---(Minimum Heat Removal)

• Assume no offsite power.

. 14.1.3 An energy balance table that lists how the energy is stored prior to a loss-of-coolant accident, the amount of energy generated and absorbed from t = 0 seconds to the time of the peak pressure, and how the energy is distributed at the time of the peak pressure.

A sample of a typical table is presented bslow:

ENERGY BALANCE TABLE Before At Pe k Item LOCA Ouring Pressare Primary Coolant Internal Energy 4

Core Flooding Tank Internal Energy Energy Stored in Fuel and Clad Energy Stored in Core Internals Reactor Vessel Metal Shutdown Energy and Decay Heat Transferred to Steam Generators Piping, Pumps, Valves, etc.

Steam Generator Metal Secondary Coolant Internal Energy Air in Containment Steam in Containment Energy Transferred to Steel Structures Others (list)-

Energy Transferred to Concrete Structures

- --