ML20023D419

Forwards Revised Pages to FSAR Addressing Open Item 15, Confirmatory Items 2,5 & 10 & License Condition 4,per Ser. Revised Pages Will Be Included in Rev 8

ML20023D419
Person / Time
Site:	Catawba
Issue date:	05/16/1983
From:	Tucker H DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8305200476
Download: ML20023D419 (11)

Text

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DUKE POWER GOMPANY P.O. HOX 30180 CHARLOTTE, N.O. 28242 HALB. TUCKER retzenoxz vuos rassment (704) 373-4531

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May 16, 1983 Mr. Harold R. Denton, Director

' Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414

Dear Mr. Denton:

The Safety Evaluation Report for Catawba identified a number of Open Items, Confirmatory Items and License Conditions. The attached Catawba FSAR pages have been revised to address the following items:

1 - Open Item 15, Inadvertent operation of fire protection system in diesel generator buildings - Addressed in revised response to Question 430.91.

2 - Confirmatory Item 2 - Sediment accumulation in SNSW pond intake structure -

Additional information provided in revised Section 2.4.8.

3 Confirmatory Item 5 - Dynamic stability of the SNSW pond dam under extreme loading conditions - Additional information provided in Section 2.5.4.8.4.

4 - Confirmatory Item 10 - Listing of ASME Code Cases used in the construction of Section III, Class 1 components within RCPB - Response provided in Section 5.2.'.2.1.

5 - License Condition 4 - Control and shutdown rods surveillance requirements -

Response provided in Section 4.2.2.3.1.

The attached revised Catawba FSAR pages will be included in Revision 8.

Very truly yours, st/ / 0. d#6 ~

Hal B. Tucker %f ROS/php Attachment c

8305200476 830516 PDR ADOCK 05000413 E PDR

. Mr. Harold R. Denton, Director May 16, 1983 Page 2 cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 -

Atlanta, Georgia 30303 Mr. P. K. Van Doorn NRC Resident Inspector Catawba Nuclear Station Mr. Robert Guild,-Esq.

Attorney-at-Law P. O. Box-12097 Charleston, South Carolina 29412 Palmetto Alliance 2135 Devine Street Columbia, South Carolina 29205 Mr. Jesse L. Riley Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28207 Mr. Henry A. Presler, Chairman Charlotte-Mecklenburg Environmental Coalition 943 Henley Place Charlotte, North Carolina 28207-L

ChS The forward velocity in the pond is taken conservatively as 0.1 ft/sec, and the depth of settling zone is taken to be 20 ft. The fall velocity of the sediment is found from Stoke's Law v = ,g Where:

0 = diameter of particle p = viscosity of water (2.359 x 10 5 l b. 5 /f',2 )

y = density of sediment (165 lb/ft3) s y = density of water (62.4 lb/ft3)

Using the grain size analysis for site soils, it is calculated that approxi-mated two-thirds of the sediment entering the pond would settle rapidly (L < 310 f t) . This would lead logically to the formation of delta deposits in the head water areas of the pond. This conclusion is substantiated by data from existing resevoirs, presented in Reference 20. The one-third or approximately 1.7 ac-ft of sediment which does not deposit in the upper regions of the pond can be assumed to distribute uniformly in the form of bottom set deposits in the vicinity of the dam and SNSW intake structure.

This amount of deposition should have no adverse impact on the operation of the SNSW intake structure. Soundings will be taken around the SNSW in-take structure prior to fuel loading and at 5 year intervals thereafter to assure that sediment deposits will not adversely affect the operation of

the NSW System.

With the exception of approximately 30 acres, the entire 450 acre drainage basin of the SNSW pond is owned by Duke Power Company.

l The 30 acre tract of private property is located along the western edge of the drainage basin, over 3,000 feet from the pond. The effect on sediment load delivered to the SNSW Pond due to disturbance on this property would be minimal because of its relatively flat terrain and its remote location with respect to the pond.

Operation of the pond as the ultimate heat sink is described in Section 9.2.8.

i 2.4.9 CHANNEL DIVERSIONS The source of cooling water for Catawba Nuclear Station is Lake Wylie. There are seven reservoirs on the Catawba River upstream from Wyli e Dam,'all of which are owned and operated by Duke Power Company. Within limitt. set by the operat-l ing licenses of these dams and dam leakage, the minimum disc iarge to Lake Wylie i

is controlled by Duke Power. No present means exist to divert or reroute other l than minor amounts of water used for municipal supply.

l In the event of the loss of Lake Wylie, the Catawba Station could be safely shut down using the Standby Nuclear Service Water Pond.

l 2.4-15a Rev. 8

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CNS 2.5.4.8.3 Residual Soil The low consistency (N=9) soil encountered in boring A-152 is a micaceous silty fine to coarse sand and is residuum. The N=9 condition occurs in a single SPT sample. (This boring is located 50 ft perpendicular from the axis of the pipeline). Because of its location, the low consistency residual soil in boring A-152 is not of concern to the NSW pipeline. The residual soil directly below the pipeline in borings A-138, A-153, and A-154 is partially 4 weathered rock having N-values of 100 or more and silty fine to coarse sand-saprolite having N-values of 20 or more.

2.5.4.8.4 Alluvial Soil In order to evaluate the liquefaction potential of the pocket of alluvial soils encountered in boring A-138 (a single sample within the interval from 33 to 30 ft), a soil column analysis using the program SHAKE is performed to compute the cyclic shear stresses induced by the SSE (Figure 2.5.4-14A).

The shear wave velocity of the compacted fill is assumed the same as from the SNSW Pond Dam (Section 2.5.6). The shear wave velocity of the alluvium is estimated from the standard penetration test value, overburden pressure, and information contained in Anderson, et al., (1978) (Reference 109). The shear wave velocity of the partially weathered rock was obtained from in-situ measurements of shear wave velocity in similar material at the Catawba site.

The cyclic shear strength of the compacted fill is obtained from tests on the compacted fill for the SNSW Pond Dam, Figure 2.5.6-14. The cyclic shear strength of the alluvial sand is estimated from Figure 24 of Seed (1979)

(Reference 110). As can be seen from Figur? 2.5.4-14A, the cyclic shear strength of the compacted fill exceeds the induced cyclic shear stresses by -

a safety factor of 1.91 (minimum). The cy::lic shear stresses exceeds the cyclic shear strength in the pocket of alluvial sands; thus, this localized

zone of alluvial soil has a potential for pore pressures to become equal to

- the confining pressure during the SSE (consisting of the synthetic time l histories).

Borings with continuous sampling are drilled 20 ft each way along the axis of the pipeline and perpendicular to the pipeline (50 ft downstream in the i pre-construction drainage feature) to explore the lateral extent of the alluvial soil encountered in one SPT sample of boring A-138. These borings do not encounter any of this soil; therefore, the alluvial soil at boring ,

A-138 occurs as a pocket of limited lateral extent (maximum of 40 ft along the pipeline) and is confined on all sides by soil (residuum and compacted earth fill) that does not liquefy or undergo excessive deformation during a seismic event.

If the local alluvial soil does experience pore pressures equal to the initial effective confining pressure in this material during the SSE, the relatively small volume and constrained nature of the alluvial zone would not cause a mass soil mover..ent in the flat ground; rather the effect would be localized settlement. Lee and Albasia (Reference 1) measured the volumetric strain of laboratory specimens on reconsolidation after liquefaction. They reported 2.5-44a Rev. 8

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CNS consideraole scatter in their results, but when averaged and plotted as Figure 7 in their paper, the data showed a " fairly well defined trend with volumetric strains decreasing as the relative density increases." For 50 percent relative density, they showed 1.5 percent maximum average strain; for a 6 ft.

thick liquefied soil mass, these results indicate about 1 to 1-1/4 inches of settlement potential. The settlement of the pipeline subgrade should be less than the potential compression of the sand pocket due to the 24 ft. of cover of compacted earthfill above the layer and below the pipelines. No estimate of the reduction in settlement is made due to the relatively insignificant amount of settlement potential.

The effect of the consolidation of the sand pocket is assumed analagous to the effect of tunnelling in soil, for which there is empirical evidence that the surface settlement trough takes the form approximated by a normal probability curve (Reference 2). To estimate the shape of the potential settlement trough that would result on the ground surface (and at the pipe subgrade in the absence of the pipe) due to reconsolidation of the sand pocket after lique-faction, a trough shaped like a normal distribution curve centred at boring A-138 is assumed. It is assumed that the 1-1/4 inch settlement potential is also the settlement of the ground surface at boring A-138, and thus forms the .naximum ordinate for the normal curve. It is further assumed the normal curve passes through settlement values that are 61 percent of the maximum value at 20 ft. away, and 13.5 percent at 40 ft. away. Then, using the properties of the normal distribution curve, the maximum radius of curvature anywhere along the settled ground surface is calculated as 48,350 inches, occurrring at boring A-138 as a concave upwards shape. This is a conservative estimate of the radius of curvature, considering that the sand pocket may feather out laterally rather than end abruptly as assumed for these calculations. The radius of curvature lengthens quickly to either side of the bottom of the settlement trough and is about 70,000 inches at 10 ft. away from the center. Because of their shallow depth of burial (and thus low soil confining pressure), the 42-inch diameter NSW pipes may not deform as much as the settled subgrade rpofile, However, for the sake of calculating the maximum for potential stress in the pipe caused by the deflection, the maximum radius of curvature computed and stated above is assumed to be imposed on the buried pipes, and the induced stresses from this curvature moment are computed. The stress resulting from the above curvature, the internal pressure, and an SSE seismic loading is calculated to be 16.3G ksi. This is less than the ASME allowable of 16.44 ksi for a combination of sustained and occasional loads. The material for the NSW pipe is SA-155, Class 2, Grade C55 with a minimum yield stress of 30 ksi, a minimum ultimate tensile stress of 55 ksi and a pipe wall thickness of .435 inches.

2.5.4.9 Earthquake Design Basis The earthquake design basis is discussed in Section 2.5.2.

2.5-44b Rev. 8

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CNS 107. Hardin, B. 0., " Shear Modulus of Gravels," UKY TR74-73-CE19, Soil Mechanics Series No. 16, University of Kentucky, College of Engineering Dept. of Civil Engineering, September, 1973.

108. Newmark, N. M., Blume, J. A. and Kapur, K, K. (1973), " Seismic Design Spectra for Nuclear Power Plant" Proceedings of Power Division of ASCE, Vol. 99, No. P02, November 1973, pp. 287-303.

109. Anderson, D. G., Espana, C. and McLamore, V. R., (1978) " Estimating In-Situ Shear Moduli at Competent Sites," Proceedings of the ASCE Geotechnical ,

Engineering Division Specialty Conference on Earthouake Engineering and Soil Dynamics, Vol. 1, page 181-197, Pasendena, CA, June 19-21, 1978.

110. Seed, H. B. (1979), " Soil Liquefaction and Cyclic Mobility Evaluation for Level Ground During Earthquakes," Journal of the Geotechnical Engineering Division 105, No. GT2, ASCE, February 1979.

111. Serf, N. Seed, H. B. Makdisi, F. I. and Chang, C. Y., " Earthquake Induced Deformations of Earth Dams" Report No. EERC 76-4, College of Engineering, University of California, Berkeley, California, September, 1976.

112. Newardk, N. M. , " Effects of Earthquakes on Dams and Embankments,"

Geotechnique, Volume 15, No. 2, January 1965.

113. Franklin, Arley G. , and Chang, Frank K. , " Permanent Displacements of Earth Embankments by Newark Sliding Block Analysis," WES Miscellaneous Paper S-71-77, November 1977.

114. Seed, H. B., and Martin, G. R., "The Seismic Coefficient in Earth Dam Design," Pecceedings of ASCE, SM & FD Journal No. SMa, May 1966.

115. Makdisi, F. I., and Seed, H. B., "A Simplified Procedure for Estimating Earthquake - Induced Deformations In Dams and Embankments," Report No. -

UCB/EERC 77-19, College of Engineering, University of California,

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[ Berkeley, August, 1977.

l 116. Lee, K. .

L., and Albasia, A., "Carthquake Induced Settlements in Saturated l

Sands," Journal of the Geotechnical Division, ASCE, No. GT4, April, 1974.

t 117. Hudson, J. A., Attewell, P. B., Atkinson, J. H. and O'Reilly, M. P.,

" Understanding Ground Movements Caused by Tunnelling," GROUND ENGINEERING, April 1976.

2.5-84 Rev. 8 New Page

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CNS The absorber rods are fastened securely to the spider. The rods are first threaded into the spider fingers and then pinned to maintain joint tightness, after which the pins are welded in place. The end plug below the pin position is designed with a reduced section to permit flexing of the rods to correct for small misalignments.

The overall length is such that when the assembly is withdrawn through its full travel the tips of the absorber rods remain engaged in the guide thimbles so that alignment between rods and thimbles is always maintained. Since the rods are long and slender, they are relatively free to conform to any small misalignments with the guide thimble.

After each refueling, prior to startup, control rod worth measurements are lperformedonthecontrolandshutdownbanks. Greater than expected worth loss would be detected by this surveillance.

4.2.2.3.2 Burnable Poison Assembly Each burnable poison assembly consists of burnable poison rods attached to a holddown assembly. A burnable poison assembly is shown in the composite core component Figure 4.2.2-12. When needed due to nuclear considerations, burnable poison assemblies are inserted into selected thimbles within fuel assemblies.

The poison rods consist of borosilicate glass tubes contained within Type 304 stainless steel tubular cladding which is plugged and seal welded at the ends to encapsulate the glass. The glass is also supported along the length of its inside diameter by a thin wall tubular inner liner. The top end of the liner is open to permit the diffused helium to pass into the void volume and the liner extends beyond the glass. The liner is flanged at the bottom end to maintain the position of the liner with the glass.

l The poison rods in each fuel assembly are grouped and attached together at the top end of the rods to a hold down assembly by a flat perforated retaining plate which fits within the fuel assembly top nozzle and rests on the adaptor plate. The retaining plate and the poison rods are held down and restrained

. against vertical motion through a spring pack which is attached to the plate and is compressed by the upper core plate when the reactor upper internals assembly is lowered into the reactor. This arrangement ensures that the poison j rods cannot be ejected from the core by flow forces. Each rod is permanently attached to the base plate by a nut which is lock welded into place.

The cladding of the burnable poison rods is slightly cold-worked Type 304 stainless steel. All other sturctural materials in the assembly are Types 304 or 308 stainless steel except for the springs which are Inconel-718. The bor-osilicate glass tube provides sufficient boron content to meet the criteria discussed in Section 4.3.1.

4,2.2.3.3 Neutron Source Assembly

• 4.2-13 Rev. 7 4 ..

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a j- CNS same terms as the newer requirements (see fracture toughness information in

Section 5.3). It should be noted that the actual hardware configuration and i material selection would not have been changed by upgrading to a later ASME Code.

. Thus, the Unit 1 reactor vessel, although not in strict accordance with 10CFR 50.55a, is acceptable as built to ASME Code Section III,1971 Edition through~

Winter 1971 Addenda.

The actual addenda of the ASME Code applied in the design of each component is listed in Table 5.2.1-1.

Inservice Inspection will be performed in accordance with 10CFR50.55a(g) to the extent practical. Requests for waiver from the requirements of the ASME Code Section XI-in effect per 10CFR50.55a(g) will be submitted to the NRC for review and disposition. Code cases applicable to preservice and inservice inspection are addressed in Section 5.2.1.2.2.

5.2.1.2 Applicable Code Cases

5.2.1.2.1 Fabrication and Construction Activities Conformance with Regulatory Guide 1.84 and 1.85 is discussed in Section 1.8.

i. ASME Code cases used for Catawba Class 1 components are listed in Table 5.2.1-3.

Code Case 1528 (SA 508 Class 2a) material has been used in the manufacture of.the Catawba steam generators and pressurizers. It should be noted that the purchase orders for this equipment were placed prior to the original issue of Regulatory i Guide 1.85 (June 1974); Regulatory Guide 1.85 presently reflects a conditional NRC approval of Code Case 1528. Westinghouse has conducted a test program which demonstrates the adequacy of Code Case 1528 material. The results of the test- ,

program are documented in Reference 1.. Reference 1 and a request for approval of the use of Coda Case 1528 have been submitted to the NRC (letter NS-CE-1730 dated March 17, 1978, to Mr. J. F. Stolz, NRC Office of Nuclear Reactor Regulation, from i Mr. C. Sicheldinger, Westinghouse Nuclear Safety Department). Responses to NRC questions on their review of this. report (Reference 1) were transmitted to the r NRC (letter NS-TMS-2312, dated September 18, 1980, to Mr. J. R. Miller, Special j Projects Branch, from Mr. T. M. Anderson, Westinghouse Nuclear Safety Department).

5.2.1.2.2 Operation, Maintenance, and Testing Activities Requests for use of Code Cases concerning operation, maintenance, and testing activities will be submitted to the NRC as necessary. Table 5.2.1-2 lists '

the Code Cases whose use is anticipated.

5.2.2 OVERPRESSURE PROTECTION RCS overpressure protection at operating conditions is accomplished by the utilization of pressurizer safety valves along with the Reactor Protection System e and associated equipment. Combinations of these systems provide compliance with the overpressure requirements of the ASME Boiler and Pressure Vessel Code Sec-tion III, paragraph NB-7300 and NC-7300, for Pressurized Water Reactor Systems.

l 5.2-2 Rev. 8 i

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TABLE 5.2.1-3 ASME CODE CASES USED FOR CATAWBA UNITS 1 & 2 CLASS 1 COMPONENTS Equipment Unit 1 Unit 2 Steam Generators 1528 1355 1484 1493 1493 1484 1355 1528 1498 Pressurizer 1582 ----

RC Pipe / Fitting Fab. 1423-2 1423-2 t

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CNS 430.91 You state in Section 9.5.8.3 of the FSAR that "A fire within one (9.5.8) diesel room, along with a single failure of the fire protection sys-tem, will be completely contained within that room. The ceabustion products will be exhausted from the room by the ventilation system at the end of the building opposite from the end which contains the intake structure for the redundant diesel. If the fire protection system operates as designed and extinguishes the fire, the gaseous carbon dioxide (extinguishing medium) will be contained in the same matter." We disagree with this statement. A fire within one diesel room along with the failure of the supply ventilation fire damper would allow the products of combustion and/or the carbon dioxide to go out the ventilation inlet which shares the same plenum as the combustion air intake. If the design is as is stated above in 4 Question 430.90 or if the outer air intake structures are separate, i the gaseous products could be drawn into the other diesel generator's air intake. Show by analysis that a potential fire in the diesel generator building together with a single failure of the fire pro-tection system will not degrade the quality of the diesel combustion air so that the remaining diesel will be able to provide full rated power.

Response

A Fire Hazards Analysis of the Diesel Generator Building is found in Duke's " Response ta Appendix A to Branch Technical Position APC5B 9.5-1" (submitted by W. O. Parker, Jr. letter dated October 23, 1981 to H. R. Denton).

Primary fire protection is provided by an automatic carbon dioxide system. The system is activated by temperature detectors (i.e., not smoke) which alarm and annunciate in the control room. Ten detectors per diesel room are provided. The circuit is seismically designed

, and supervised to annunciate control malfunctions.

l Each diesel room is provided with ' electrically separate CO2 actuation systems to prelcude a common malfunction affecting both diesel

rooms. Thermal detectors have a 225 F setpoint and will actuate l on rate of temperature rise. Actuation takes 60 seconds during l which visual / audible alarms are given, the CO 2System master valve opens-charging the supply header, ventilation systems shut down and the hazard selector valve opens discharging CO2-The CO 2 System piping would not remain intact during a seismic event, preventing discharge, although the diesel equipment will function during a CO2 release. A " purge" switch located immediately outside the diesel room will utilize the ventilation system to remove CO 2-430-47 Rev. 8

CNS Products of combustion will be contained by closing the ventilation system dampers early in the CO2 detection / actuation sequence. The intake structures are separated by a 3-hour rated fire wall (approxi-mately 20 ft. height). The volume of CO2 discharged will not fill the intake structure preventing carryover into the adjacent structure.

Excess CO 2 will exit through the lower exhaust structure openings (approximately 10 ft difference). Infiltration into the adjacent room's intake will not degrade diesel performance since the relay contacts, switch contacts and other electro-mechanical devices associated with starting / operation are housed in Class 1E, drip proof, bottom-entry NEMA 12 control panels. Process control devices located external to the control panels are NEMA 4 enclosures. (NEMA 12 enclosures provide protection against fibers, flyings, lint, dust, dirt, light splashing seepage, dripping, and external condensation of non-corrosive liquids. NEMA 4 enclosures provide water-tight protection.

Reference:

NEMA publication No. ISI.1-1977).

Manual hose stations are provided as a secondary fire protection system. Hose station water source is the Nuclear Service Water System essential header. Floor drains are provided to remove fire protection water if secondary means are needed in addition to the CO2 System.

430.92 In the turbine generator section discuss: 1) the valve closure (10.2) .

times and the arrangement for the main steam stop and control and the reheat stop and intercept valves in relation to the~effect of a failure of a single valve on the overspeed control functions;

2) the valve closure times and extraction steam valve arrangements in relation to stable turbine operation after a turbine generator system trip. (SRP 10.2, Part III, Items 3, 4.)

Response

I See revised Section 10.2.2.

430.93 The FSAR discusses the main steam stop and control, and reheat (10.2) stop and intercept valves. Show that a single failure of any of the above valves cannot disable the turbine overspeed trip functions.

(SRP 10.2, Part III, Item 3.)

Response

See revised Section 10.2.4.

430.94 Discuss the effects of a high and moderate energy piping failure (10.2) or failure of the connection from the low pressure turbine to con-denser on nearby safety related equipment or systems. Discuss what j protection will be provided the turbine overspeed control system 430-47a Rev. 8

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