ML20054G983
| ML20054G983 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/11/1982 |
| From: | Parker W DUKE POWER CO. |
| To: | Adensam E, Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8206220468 | |
| Download: ML20054G983 (40) | |
Text
W, Dunn Powen COMPANY Powra flertnswo 422 SouTu Cuencu SturzT, CuAuMTTE. N. C. 26242 W I L LI AM O. PA R M E R, J R.
'/ic e Port siot N Y TELEP=oNt:AaEA 704 S e r a s. Paoouc tio~
June 11, 1982 3 7 3-4 3 e 3 Mr. liarold 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:
In order to facilitate the completion of the review of the Catawba FSAR, Duke Power Company is transmitting herewith responses, revised responses, or partial responses to the following FSAR questions:
220.21 410.17 440.56 220.22 410.20 440.85 220.25 410.21 460.8 220.30 420.5 480.16 220.46 430.69 730 (A-40) 241.9 430.92 730 (A-46) 410.12 430.102 These responses will be included in FSAR Revision 6.
l Very truly yours, Lo.
/
/
l William O. Parker, Jr.
ROS/php i
Attachment cc:
Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission 00 Region II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. P. K. Van Doorn NRC Resident Inspector l
Catawba Nuclear Station B206220468 820611 DR ADOCK 05000413 PDR
Mr. liarold R. Denton, Director June 11, 1982 Page 2 cc:
Mr. R. Guild, Esq.
Attorney-at-Law 314 Pall Mall Columbia, South Carolina 29201 Palmetto Alliance 2135h Devine Street Columbia, South Carolina 29205 Mr. J. L. Riley Carolina Environmental Study Group 854 llenicy Place Charlotte, North Carolina 29207 Mr.11. A. Presler, Chairman Charlotte-Mecklenburg Environmental Coalition 943 llenley Place Charlotte, North Carolina 28207
CNS th Constant for the i mode.
C
=
9 th Natural frequency of the i mode.
W
=
9 th Constant for the i mode.
N
=
g yj Maximum velocity from the response spectrum for a single-degree-of-S
=
th freedom system for a given value of W for the i mode.
9 N
=
Number of modes considered.
Equipment located in a structure is used as input to the seismic analysis.
The equipment is comparatively rigid by virture of its mass when compared to the mass of the structure.
The equipment is connected to the structure in a rigid fashion.
Q The equipment masses and locations are used as input to the center of mass 220.21 calculation for an individual location where mass will be lumped for use in the seismic model.
The individual mass points are connected to the seismic model by rigid members and response spectra are generated for each point.
The response spectra as generated are then used in the individual equipment design.
3.7.2.1.1.2 Containment Interior Structure, Nuclear Service Water Intake Structure, and Auxiliary Building, Including Diesel Building, Exterior Doghouse, and Upper Head Injection Building The seismic loads on the Containment Interior Structure, Nuclear Service Water Intake Structure and Auxiliary Building, as a result of a base excitation, are determined by a dynamic analysis.
The dynamic analysis is made by idealizing the structure as a series of lumped masses with weightless elastic columns acting as spring restraints.
The base of the structure is considered fixed.
The procedure used to lump masses for the seismic structural model is dependent upon the actual mass distribution and structural characteristics of the structure.
Mass locations are established at elevations in the structure where there are concentrations of mass such as floor slabs and/or equipment.
Mass locations are also established when there are changes in structural properties such as moments of inertia, shear area or elastic properties.
Q Sufficient mass locations of uniform distribution are established for the 220.25 Diesel Building and Nuclear Service Water Pump Structure to assure considera-tion of all significant modes with frequencies less than 20H.
Structures with 7
fundamental frequencies greater than 20H are designed as rigid structures with Z
a constant acceleration equal to the acceleration corresponding to 20H Z
the response spectrum.
The mass of the equipment is lumped at the elevation at which it is supported such as !ateral supports for the steam generators, reactor vessel, reactor coolant pumps, pressurizer, and polar crane.
When equipment is supported on a floor slab, the equipment mass i, lumped with the structural mass of the slab.
3.7-5 Rev. 6
CNS The structural connection between equipment and structure is considered rigid for the seismic analysis of 1.he structure.
A response spectrum has been generated as defined in Section 3.7.2.5 at mass locations where equipment or piping is supported.
This response spectrum is used for the seismic design of equipment and piping as defined in Section 3.7.3.8.
The mass of vertical structural members is distributed to the adjacent mass locations.
Between these mass locations, vertical structural member properties Q
are calculated for moments of inertia, cross-sectional area, effective shear 220.25 area, and length.
For those elevations where the center of mass and center of structural rigidity do not coincide or where tne centers of structural rigidity change locations, infinitely stiff horizontal members connect these points.
3.7-Sa Rev. 6 Carry Over
CNS For each of the structures listed above, it has been determined that the structure moves with the ground motion during an earthquake to account for the soil-structure interaction effect (see Section 3.7.3.12).
3.7.2.5 Development of Floor Response Spectra Figures 2.5.2-5 reflect the time-history spectra and site design spectra.
The synthetic earthquakes used to generate the time-history spectra in Figures 2.5.2-5 are used to generate response spectra at elevations in structures that house systems and components which are required to be designed for seismic excitation.
The analytical technique used to generate the response spectra at specified elevations in a structure is the time-history method.
The acceleration time-history of each elevation is retained for the generation of response spectra reflecting the maximum acceleration of a single-degree-of-freedom system for a range of frequencies at the respective elevation.
Vertical response spectra are not generated.
The floor slabs in Category 1 g
220*30 structures were examined for flexibility and were found to be sufficiently stiff to justify the assumption of negligible amplification in the vertical direction.
Damping values for the structural model are selected from Section 3.7.1.3,
" Critical Damping Values."
TIME-HISTORY ANALYSIS The time-history of the specified mass points is determined by the modal method in which the responses in the normal modes are determined separately, then superimposed to provide the total response to a specified base input motion.
3.7-11a Rev. 6
Table 3.8.1-2 (Page 7) effect of these concentrated loads, provided there will be no loss of func-tion of any safety related system.
Load Combinations And Acceptance Criteria For Category I Foundations In addition to the load combinations and acceptance criteria referenced above, all Category I foundations will also be checked against sliding and overturning due to earthquakes, winds, and tornadoes and against floatation due to floods 4
in accordance with the following:
Minimum Factors of Safety Load Combination Overturning Sliding Floatation a.
D+H+E 1.5 1.5 b.
D+H+W
- 1. 5 1.5 c.
D + H + E' 1.1 1.1 d.
D+H+W 1.1 1.1 t
e.
D + F' 1.5 f.
D+L+H 1,1 1.1 1.1 The minimum factor of safety against flotation of all Category 1 structures used the following procedures:
1.
Uplifting /bouyant force groundwater assumed to rise to grade elevation (593.5).
Q241.9 2.
Dead Load - The total weight of the structure was determined including weights of the major pieces of equipment.
The uplifting /bouyant force was then compared to the total dead load.
i l
1 i
Rev. 6 i
l
[
CNS 9.3 PROCESS AUXILIARIES 9.3.1 COMPRESSED AIR SYSTEM 9.3.1.1 Design Bases The Compr0ssed Air System consists of the Instrument Air, Station Air, and Breathing A r Systems.
The Instrument Air System supplies clean, oil free, i
dried air to all air operated instrumentation and valves.
The Station Air System supplie; compressed air for air operated tools, miscellaneous equipment, and various mai.itenance purposes.
The Breathing Air System supplies clean, oil free, low pressure air to various locations in the Auxiliary Building and in the Containment for breathing protection against airborne contamination while performing certain maintenance and cleaning operations.
The Compressed Air Systems are shown on Figures 9.3.1-1 thru 9.3.1-8 and the system component design parameters are given in Table 9.3.1-1.
9.3.1.2
System Description
9.3.1.2.1 Instrument Air System Instrument air is supplied by three non-lubricated instrument air compressors.
Q410.21 The compressors' intakes are in the Service Building basement.
This area is free of corrosive contaminants and hazardous gases.
Downstream of each air compressor, the hot compressed air flows through an aftercooler and water separator before discharging into an instrument air receiver.
The aftercooler cools the hot compressed air to within 10 F of the conventional low pressure service water temperature, and the water separator removes any water condensed in the cooling process.
The air receivers smooth out any pressure surges.
Downstream of the air receivers, the instrument air is dryed to a dewpoint of 35 to 39 F by four refrigerated air dryers.
In ad-dition, dessicant air dryers are provided on the lines going outside the building to dry the air to a design dewpoint of -40 F.
After the refrig-Q410.21 erated air dryers, the air is passed through filters which filter out par-ticles larger than 3 microns.
Downstream of the filters, the instrument air headers supply instrument air throughout the plant.
At each air oper-ated valve or instrumentation the air is filtered again through a filter-regulator.
In the event of low instrument air pressure, the station air system will auto-matically supply air to the instrument air system.
This air will be supplied through two oil removal filters to the instrument air compressors discharge header.
The bulk air supply to the Unit 1 and 2 condensate polishing demineralizers will come off the instrument air compressors discharge header upstream of the instrument air dryers.
Two check valves with a trap between them will be provided in this supply line to prevent the backflow of water into the Instru-ment Air System.
9.3-1 Rev. 6
l CNS Heater Extraction Source A
H-P turbine B
H-P turbine C
H-P turbine exhaust D
L-P turbines E
L-P turbines F
L-P turbines b
G L-P turbines Provided in extraction lines A, B, C, 0, and E are piston-assist spring-closed swing check valves.
The piston-assist spring-closed actuators are designed to overcome friction and allow the valves to rapidly close on tur-bine trip.
(See Table 10.2.2-2 for closure times) Bleed lines F and G are not provided with check valves since installation in the condenser neck would be impractical.
However, heaters F and G are provided with anti-flash baffle Q430.92 plates located below the tube bundles and above the water volume.
The extr-action design was approved by the turbine manufacturer in regard to turbine overspeed protection.
Postulating turbine trip and failure of a single ex-traction check valve, the manufacturer concluded that unacceptable turbine overspeed would not occur.
Each of the two generators is a 1450 MVA, 1800 rpm, direct connected, 3 phase 60 cycle, 22,000 volt conductor cooled synchronous generator rated at.90 P.F,.50 SCR at a maximum hydrogen pressure of 75 psig.
Generator rating, temperature rise, and class of insulation are in accordance with IEEE stan-dards.
Excitation is provided by a shaft driven alternator with its output rectified.
10.2-la Rev. 6
Table'10.2.2-2 Turbine Speed Control System Valve Closure Time Valve Closure Time (Seconds)
Main stop 0.12 Main control 0.19 Intermediate stop 0.20 Q430.92 Intercept 0.17 Extraction check 2 (maximum) i e
Rev. 6
<)
c e
o l
CNS 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM 10.4.1 MAIN CONDENSER 10.4.1.1 Design Bases The main condenser is designed to condense turbine exhaust steam for reuse in the steam cycle.
The main condenser also serves as a collecting point for various steam cycle vents and drains to conserve condensate which is stored in the condenser hotwell.
The condenser also serves as a heat sink for the Turbine Bypass System and is capable of handling 40 percent of rated main steam flow.
Rejected heat is removed from the main condenser by the Condenser Circulating Water System.
10.4.1.2
System Description
The main condenser consists of three surface type deaerating condenser shells with each shell condensing the exhaust steam from one of the three low pres-sure turbines.
The condenser shells are of conventional shell and tube de-sign with steam on the shell side and circulating water in the tubes.
Each condenser shell is joined to the turbine by a rubber belt type expansion joint.
Provisions have been made for mounting two low pressure feedwater heaters in the neck of each of the three condenser shells.
The combined hotwells of the three condenser shells have a water storage capability equivalent to approximately 7.5 minutes of full load operation.
The inter-nal condenser design provides for the effective condensing of steam, scaveng-ing and removal of noncondensable gases, and the deaeration of the condensate.
Q430.100 The condenser tubes are protected from failure due to high temperature drains and blowdown by spray headers and impingement baffles.
The circulating water side of the main condenser is a triple pass arrange-ment having two vertically divided water circuits.
Each of the two water circuits can be isolated for the repair of leaking tubes by closing the motor operated butterfly valves on the condenser inlet and outlet connec-tions.
The main condenser will maintain back pressures of 2.4, 2.9, and 3.7 In. Hg. Abs in the three condenser shells when operating at rated turbine output with 89 F inlet circulating water temperature and 95 percent clean Q430.101 tubes.
Loss of condenser vacuum due to the accumulation of non-condensable gases is prevented by the steam air ejectors described in Section 10.4.2.
Q430.102 However, if vacuum loss should occur, high back pressure signal (s) from tur-bine exhaust instrumentation will trip the turbine, low vacuum signal (s) from condenser instrumentation will block the condenser steam dump valves (see Section 10.4.4), and the condenser will in effect be isolated from all direct sources of main steam.
Loss of condenser vacuum does not affect op-eration of the main steam isolation valves.
The condenser tubes and compo-nents are constructed of corrosion /errosion resistant materials.
In addition, a continuous tube cleaning system (Amertap) is provided to keep the inside tube surfaces clean and the condenser operating at peak performance.
The main condenser can accept a bypass steam flow of approximately 40% of rated main steam flow without exceeding the turbine high backpressure trip point with design inlet circulating water temperature.
This bypass steam 10.4-1 Rev. 6
CNS 11.2.2.7.1.5 Floor Drain Tank Subsystem Operation The water in the floor drain tank is sampled to determine the degree of pro-cessing required.
Normally the contents of the floor drain tank can be sent to Waste Monitor Tank B after filtration.
If further processing is required or it is desirable to reclaim distillate, the laundry and hot shower tank subsystem, waste evaporator, and water evaporator condensate demineralized Q460.8 can be utilized.
Unless it is recycled, ultimately the contents of the floor drain tank should be returned to Waste Monitor Tank B for recirculation, sampling, and discharge into the low pressure service water discharge via the Nuclear Service Water System at a rate determined by the dilution flow-rate available.
Upon accumulation of 23000 gallons of waste in the FDT, valves are aligned for recirculation and the FDT Pump started.
At 24000 gallons, a sample is l
taken to gain a rough estimate of activity and concentrations of glycol, boron, iron, detergents, and decon chemicals.
The rough sample results in-dicate the type of treatment needed according to the level of contamination as well as identify the source in the event of unusual leakage.
The waste is pumped to WMT B, since WMT A will normally be used for receiving treated LHST waste.
For example:
a.
Very low activity and low concentration of non-toxic chemicals, i.e.
meets discharge requirements:
pump through FDT filter to WMT B, recirc.,
sample, analyze, and discharge through the WL System discharge radiation monitor and automatic control valve.
Approximately 30% of the water Q460.8 from the floor drain tank is diverted upstream of Waste Monitor Tank B for processing using the waste evaporator and waste evaporator conden-sate demineralizer to reclaim distillate.
e b.
Low activity detergent-complexed radioactive contaminants:
pump through LHST filters and LHST carbon filter to WMT B and recirc., test, and discharge.
c.
Low activity ionic radioactive contaminants (salts):
pump through LHST filters, carbon filter, and WMT demineralizer to WMT B.
d.
Low to high activity particulate radioactive contaminants:
treat as in (a) or through LHST filters, if analysis shows Su filtration necessary.
e.
Low activity and moderate levels of elemental chemical contaminants (boron, chromium, iron):
pump through LHST primary, secondary filters, and WMT demineralizer to WMT B.
f.
Low activity and moderate levels of organic contaminants (glycol, deter-gents):
treat as in (b).
g.
High activity in the form of soluble contaminants (detergent-complexed and/or ion pair):
pump to MST in 600 gallon batches, add chemicals to adjust pH and prevent foaming, and pump to the WEFT for processing in the evaporator.
11.2-30 Rev. 6
CNS 4)
Valve stem leakoffs flow directly to the recycle holdup tank at a rate of 200 gpd and at approximately reactor coolant concentration.
5)
Based on the above values, the recycle holdup tank has collection time of 29.17 days, assuming 40% capacity.
Recycle holdup tank discharge is to the reactor makeup water storage tank through the boron recycle evaporator and evaporator condensate demineralizer, which have a collective DF of 10E3 for. iodines ard 10E3 for other isotopes.
Process time is 4.15 days.
Ten percent (292,000 gal.) of the boron recycle input is assumed to be released to the environment.
This value is considered conservative enough to include any deliberate discharges for tritium control.
6)
Waste collected by the waste evaporator feed tank is processed through the waste evaporator and the waste evaporator condensate demineralizer, which have a collective DF of 10E4 for iodine and 10E5 for other isotopes.
Collection time is 2.1 days assuming 40% capacity of the waste evaporator feed tank, while process time is 0.2 days for the 7.5 gpm evaporator.
Ten percent of the stream is released to the environment.
7)
Floor drain tank contents are normally sampled, then processed through the floor drain tank strainer and filter.
Approximately 30% of the influent to the floor drain tank is processed in the waste evaporator ard waste evaporator condensate demineralizer to reclaim distillate.
Q460.8 The remaining effluent is transferred to Waste Monitor Tank B where it is resampled and, if within specifications for release, the entire batch is discharged from the station.
Further processing is performed on the batch if it is out of specifications.
This is discussed in Section 11.2.2.7.1.5.
8)
Laundry and hot shower tank contents may be discharged through the waste monitor tank demineralizer, as required.
However, since use of the demineralizer is not continuous, no credit for reduction of laundry waste activity is taken.
9)
Blowdown occurs at a rate of 280 gpm through the steam generator blowdown tank, where liquids are directed upstream of the condensate polishing demineralizers and gases are vented to the "D" heaters.
There is no release of wastes from the blowdown system.
11.2.3.1 Release Points All discharges of detectable radioactivity are through the Low Pressure Service Water System into Lake Wylie.
The location of the discharge can be seen on Figure 2.1.1-4.
11.2.3.2 Dilution Factors low pressure service water will provide dilution for liquid wastes with a flow that will vary depending, among other things, on the station power output and Lake Wylie water temperatures.
For the purpose of dose evaluations, an average dilution with 54,000 gpm is assumed.
Estimates of near-field and far-field dilution are discussed in Chapter 5 of the Catawba ER-OLS.
The rate of radioactive discharges will be based on the available dilution and the concentrations of 10CFR20, Appendix B, Table II.
11.2-36 Rev. 6
CNS 1
steel barriers in your analyses.
The staff's position on allowable ductility ratios is contained in Attachment 3.
Indicate your com-pliance with the staff's position in this matter or provide justifica-tion for any deviations therefrom.
Response
See revised Section 3.5.3.2.
220.20 For concrete and bolted steel structures, you have used damping (3.7.1.3) values which are slightly greater than those given in Regulatory Guide 1.61.
Provide the assessment of impact if you have to con-form to Regulatory Guide 1.61 values.
Also, provide justification for use of higher values and correlation between stress levels and these values.
Response
An assessment was made to determine the impact should Catawba be required to conform to the lower damping values recommended by Reg-ulatory Guide 1.61.
This guide recommends damping values of 4% and 7% for OBE and SSE load cases, respectively, when designing concrete and bolted steel structures.
Catawba designs utilize damping values of 5% for both earthquake load cases.
Evaluations indicate that a stress increase of approximately 7% would be realized in the OBE load condition if the 4% damping values are used in lieu of the 5%
values.
The SSE load condition, however, controls the design of our concrete and steel structures.
The SSE load cases create stresses approxi-mately 85% higher than the 08E load cases.
Since the controlling SSE design load case utilizes a more conservative damping value than that recommended by Regulatory Guide 1.61, conformance to this guide would have no impact on the Catawba design.
220.21 In Section 3.7.2.1.1.2 of the FSAR you have stated that tile mass of (3.7.2.1) the equipment is lumped at the elevation at which it is supported and the structural connection between equipment and structure is considered rigid for the seismic analysis of the structure.
Your decoupling criterion between system and subsystem is not clear from above statement.
Provide clear definition of your decoupling cri-terion.
One acceptable criterion is outlined in SRP Section 3.7.2.
Response
See revised Section 3.7.2.1.1.1.
220.22 Clarify, whether or not, your lumped mass model of the containment i
(3.7.2.1) interior structure contains detailed representations (stiffness, t
220-6 Rev. 6
CNS damping, and mass) of major equipment such as reactor vessels and steam generators.
If not, provide justification for not doing so.
Response
The lumped parameter model used in the seismic analysis of the Catawba Nuclear Station, Reactor Building interior structure contained only the masses of the NSSS components.
The stiffness characteristics of the individual items were not used in the model.
This method of treating these components was state of the art at that point in time and was excepted when the PSAR was submitted.
In spite of the above, additional work was undertaken.
(See the response to SEB Action Item 5, which was transmitted by letter of April 8, 1982, W. O. Parker, Jr. to H. R. Denton).
The results of this work indicate that if the NSSS System were to be included in the structural model, two basic changes would result.
One, the sys-tems force response would drop from 10% to 15%, indicating that the building is adequately designed and two, the resulting response spectra would decrease but would show a frequency shift of from approximately 10% to 15% toward the low frequency end of the curve.
The resulting spectra should fall within the present spectra.
In view of the above it is concluded that the simplification used in the original calculation is justified.
220.23 Provide the description of your procedure used in modeling the hydro-(3.7.2.1) dynamic phenomena for the buildings (e.g., Fuel Pool and Fuel Handling building, Nuclear Service Water Pump Structure) where this is a con-sideration.
Response
See revised Section 3.7.2.1.1.3.
220.24 Describe your procedure to compute dynamic lateral earth pressure (3.7.2) and hydrodynamic groundwater pressure during seismic event.
Response
See revised Section 3.7.2.4.
220.25 For structures described in Section 3.7.2.1.1.2 demonstrate that (3.7.2.1) adequate numbers of masses or degrees of freedom in dynamic models are considered.
A criterion acceptable to the staff is described in SRP Section 3.7.2.II.1.a(4).
Response
See revised Section 3.7.2.1.1.2.
220-6a Rev. 6
CNS 220.29 Provide description of your procedure to account for three components (3.7.2.5) of earthquakes in generation of floor response spectra.
Response
See revised Section 3.7.2.6.
220.30 Provide description of the procedure used in allowing for vertical (3.7.2.5) flexibility of floors in generation of vertical response spectra.
Response
See revised Section 3.7.2.5.
I 220.31 In this section, you have stated that the earthquake ground motions i
(3.7.2.6) are assumed to act in one of the horizontal directions and vertical direction simultaneously.
It is not clear how you have combined the responses due to these two motions.
Is it by absolute suia method or SRSS method? Also, provide a comparison of your method of combining two components of earthquakes with the currently acceptable procedure of combining all three components with SRSS method.
This comparison need only be done for one structure which is essentially non-symmetric; e.g., internal structure.
Response
See revised Section 3.7.2.6.
220.32 The present technical position of the staff requires that the acci-(2.7.2.11) dental torsion, based on eccentricity of minimum 5% of the base dimen-sion, be included in the design of structures.
This is in addition to that which results from the actual geometry and mass distribution of the building.
Either indicate your willingness to comply with this position or provide justification for not doing so.
Response
Standard Review Plan 3.7.2, Seismic Systems Analysis, Part III, Sec-tion II, states that an acceptable method for accounting for acci-dental torsion is to add an additional 5% of the maximum building base dimension to the eccentricity that exists naturally in the building.
This was not done at Catawba because the requirement did not exist at the time of the initial analysis.
The effect of incorporating this requirement into the Catawba analysis has been evaluated using the Auxiliary Building as the case study.
The Auxiliury Building was selected because it exhibits the largest existing eccentricities as well as the largest plan dimensions.
?20-8 Rev. 6
~
w._
Response
220.43 Provide analysis details of the ice condenser floor and the lower (3.8.3) support structures.
Demonstrate that the applicable code provisions are fully met in your design.
Response
See revised Section 3.8.3.1.10, 220.44 In FSAR Section 3.8.3.4.4 you stated that the crane wall was analyzed (3.8.3) as a space frame.
Provide more analysis details, such as computer code, governing loads and load combinations, critical design forces and the design of reinforcing bars.
Response
See revised Section 3.8.3.4.4.
120.45 With respect to FSAR Section 3.8.4.1, Spent Fuel Building, describe
'(3.8.4) in detail your design and analysis of spent fuel building structures and fuel pool racks.
Enclosed is a copy of staff position on "the minimum requirements for design of spent fuel pool racks" (Attachment 5).
Indicate compliance with this position or justify deviations therefrom.
Response
The design and analysis of the Spent Fuel Building structures is described in Sections 3.8.4.4, 9.1.2.1, 9.1.2.2, and 9.1.2.3.
The design and analysis of the spent fuel racks is described in Section 9.1.2.2 and 9.1.2.3.
Spent fuel rack plans and details are provided in Figures 9.1.2-7 and 9.1.2-8.
The spent fuel pool liner plate is described in Section 3.8.4.1 (A.2).
The spent fuel storage racks are described in Sections 9.1.2.1 (8), 9.1.2.2, and 9.1.2.3.
Refer also to revised Sections 9.1.2.1 (1), 9.1.2.2, and 9.1.2.3.
220.46 In FSAR Section 3.8.4.4 you stated that masonry construction is de-(3.8.4) signed and reinforced to remain functional under the applicable load-ing conditions.
Enclosed is a copy of design criteria for safety-related masonry wall evaluation (Attachment 6).
Identify any dif-ference in requirements of materials, testing, analysis, design and construction between Catawba design and the staff position.
Provide O
justification for these differences or indicate your compliance with 220-11 Rev. 5
CNS them.
Provide a general description of the masonry walls at Catawba indicating number of walls, usage, buildings in which they are located, types (e.g., single wythe, double wythe), construction practices, loads and load combinations, and general design procedures.
Provide sample calculations for each different type of wall.
Response
A response was provided in W. O. Parker's letter of April 8, 1982 to H. R. Denton in response to SEB Action Item 34.
220-11a Rev. 6
CNS static loads and uplift in the design of subsurface walls and foundation mats are justified and conservative.
(b) Provide material gradation curves for the in-situ and backfill soils, and fine and coarse filter materials, and also confirm (with calculations) that the filter criterion has been met be-tween:
(a) insitu rock and back fill soils; (b) fine and coarse filter materials, and (c) coarse filter material and the perfo-rated pipe drain.
(c) Provide the settlement (total and differential) data and piezo-metric pressure readings taken on the Standby Nuclear Service Water (SNSW) dam, and compare the measured data with estimated values.
Response
See revised Sections 2.4.13.5, 2.5.4.13, and 2.5.6.10.
241.8 Describe the procedures to be used in performing periodic inspections (2.5.6.8) of the SNSW Dam and discharge facility to conform to the provisions of Regulatory Guide 1.127.
Provide a report on the results of any recent dam inspection which may have been performed, particularly re-garding erosion, excessive seepage, ' piping,' etc.
Response
See revised Section 2.5.6.8.2.
241.9 In Table 3.8.1-2 (page 7) in FSAR, Vol. 4, the minimum factor of (2.5.6.8) safety against floatation for all Category I foundations is given as 1.1.
Describe briefly the procedure used in determining this factor of safety.
Response
See revised Table 3.8.1-2.
240-10 Rev. 6
.)
CNS reading will indicate approximately 2200 lbs (assembly plus mast-wet condition) and for the dry condition, ap-o proximately 2500 lbs (assembly plus mast total weight).
Should the assembly movement fall outside the range of the suspended weight 1200 lbs, movement should stop.
(iii) Once the guide cable becomes slack, the operator verifies the correct position of the assembly, using a z-z tape which identifies the depth of the assembly.
Verification is made within ik inch according to operational procedures.
Ib.
The following precautions are taken to prevent loading an Oconee fuel assembly into a spent fuel rack designated for Catawba fuel (without spacer):
(i) A visual examination will verify that the rack is empty.
(ii) As stated in the McGuire and Oconee operating procedures and shown in the applicable drawings, to separate the fuel handling tool from the fuel assembly, a roll pin must be turned by hand, in a counterclockwise direction 350.
Should the handling tool and guide line be under any form of tension, the roll pin would be difficult to turn the necessary rotation.
As such, the bridge operator can observe and correct the situation.
(iii) The B&W fuel assembly can be inserted into the Catawba spent fuel rack.
However, due to the design and necessary clearance required to remove the grapples, it is considered impossible to remove the handling tool.
Thus, the assembly cannot disengage in the spent fuel rack.
2.
As per the operational procedures, the Reactor Engineer or his designee, is given the completed procedure process record form, labeling exactly which spent fuel racks are equipped with spacers.
3.
As stated in Section 9.1.2.4 the storage of the Oconee spent fuel in the Catawba spent fuel pool is within the Safety Evalua-tion presented in Section 9.1.2.3.
Sections 9.1.2.3.1 and 9.1.2.4 state that the analysis of an infinite array of Oconee assemblies (3.3 wt% - U-235) stored in Catawba's spent fuel pools will have a K of less than 0.95 (1).
Oconee fresh eff fuel will never be stored in the Catawba pool; therefore, the criticality analysis is conservative.
410.12 Provide information on the spent fuel pool water temperature fol-9.1.3) lowing the loss of one cooling train assuming the maximum heat load condition with Catawba fuel and with non-Catawba fuel.
Present an analysis of temperature vs. time for these two conditions.
Describe 410-6a Rev. 4 Carry Over a
CNS the disposition of the heat load to all heat sinks including the spent fuel pool heat exchanger, the spent fuel pool water purification sys-tem, any waste or drains, and the environment of the fuel building.
Discuss the effects of the heat and humidity load on the fuel build-ing ventilation system.
Response
An analysis of temperature versus time for Q = 39.0 x 108 BTU /hr (Catawba fuel) and Q = 42.7 x 106 BTU /hr (non-Catawba fuel) was performed using only one spent fuel pool heat exchanger and the pool water as heat sinks.
The results of this analysis are pre-sented in Figure Q410.12-1.
The additional heat and humidity load for these two cases would not affect the fuel building ventilation system.
l l
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410-6b Rev. 6
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i 170-6 Q = 42.7 x 10 BTU /hr i
165 C
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=
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I I
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I I
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5 10 15 20 25 30 35 40 45 50 TIME (hrs)
SPENT FUEL P00L-TEMPERATURE VERSUS TIME E
CATAWBA NUCLEAR STATION Figure Q410.12-1 L
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. v
'CNS
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.l 410.15 Previde a listing of all light loads (those of weight less than one (9.1.4) fuel assembly) carried over the open reactor vessel or the spent fuel 93 pool. including their kinetic energy on impact with spent fuel and dis
,. ~
cuss the consequences of dropping of these loads on stored fuel. 'It
~
is our position that dropping of these loads not result in release of radioactivity in excess of that assumed in the design basis fuel han-dling accident.
Response
i t
c i
n e
1-
'/%
_ ) 410.16 Figure 9.5.4-10 indicates that the nuclear service water yard piping (9.2.1) crosses the non-safety-related condenser circulating water system piping.
Describe the protective features provided for the nuclear service water piping at these crossover points against the effects resulting from failed condenser circulating water pipe in a SSE.
It l
is our position that the integrity of-the nuclear service water sys-l tem be maintained in a SSE.
Response
See revised Section 9.2.1.2.1.
410.17 Your component cooling water system has a single common supply line (9.2.2) and a single common return line to the reactor building components cooled by this system.
These lines contain electrically-operated containment isolation valves.
Cooling water to all four reactor coolant pump (RCP) seals and motor bearings which require continuous cooling during all modes of operation would be lost on failure of one of these valves.
Loss of cooling could result in motor bearing failure and locking of all pump rotors.
Indicate the consequences of the loss of motor bearing and seal thermal barrier cooling to'all pumps and show that, without operator intervention', this event will not cause fuel damage or damage to.the reactor coolant system pres-sure boundary with consequences greater than those assumed in the design basis RCP locked rotor accident.
410-8 Rev. 3 New Page
- f.
s
If this can not be demonstrated, then it is our position that you verify 'by test that the RCPs can function satisfactorily for 30 minutes lwithout component cooling water flow and provide redundant safJty g'rade indication of the loss of component cooling water to the pumps in order to assure that the operator will have sufficient time to trip them.
Alternatively, you may provide redundant com-ponent cooling water supply and return lines to/from the RCPs, or,
an automatic trip of the RCPs on loss of component cooling water
' flow.
Response
Redundant safety grade indication of the loss of component cooling water to the reactor coolant pumps (RCP) will be provided to assure thaC the operator will have sufficient time to trip them.
Two safety grade transmitters will be provided on the common header supplying the reactor coolant pump motor upper and lower bearings, and the reactor coolant pump thermal barriers.
These transmitters will alarm on. low filw in the control room.
It has been verified, by testing l that the pumps can function satisfactorily for 10 minutes without component cooling water flow.
If component cooling water flow can-not be rd-established to the RCP's within 10 minutes following a low flow alarm, the reactor and then the RCP's will be tripped by the operator.
410.18 Indicate the effect of including the maximum worst case spent fuel
~
(9.2.5) pool. cooling load on your analysis of the heat rejection capabilities of the standby nuclear service water pond (SNSWP).
It is unclear if this load has been considered in your analysis of the 30-day heat inputs to the SNSWP from the station auxiliary systems.
Also pro-vide d drawing Showing the physical locations of the SNSWP ii?.ake and discharge structures, the nuclear service water pump house, and the interconnecting pipe routing.
Respopce:
t The fuel pooi cooling heat exchangers are on the non-essential KC header'and are isolated from the KC and RN Systems following a l
safety injection (SS) signal.
The fuel pool is designed to reject decay heat during this time by boiling off and is therefore not included in the 30 day heat rejection load on the nuclear service water pond.
The location of the SNSWP intake and discharge structures and the nuclear service water pump house is shown on Figure 2.4.1-1.
Inter-connecting pipe routing is shown on Figures 9.5.4-10 and Q410.18-1.
/
6 410-9 Rev. 6
a CNS 410.19 Provide drawings that indicate the relationship of the auxiliary (9.2.6) feedwater condensate storage tank to the condensate storage system and to the auxiliary feedwater system.
FSAR Figures 9.2.6-3 and 10.4.9-1 concerning this tank and system do not agree.
Response
See revised Figure 9.2.6-3.
'410.20 Provide additional information concerning the discussion in FSAR Sec-(9.3.1) tion 9.3.1.3 which describes the need for air operated valve operation in the event of a control room evacuation coincident with a station blackout in order to bring the station to a hot standby condition from the auxiliary shutdown panel.
Describe any special provisions made to assure the reliable delivery of instrument air to these valves under this coincident event.
Identify the valves requiring this capability.
Response
As discussed in Sections 9.3.1 and 10.3 a non-safety related air sup-ply and control system is sufficient and acceptable for the steam generator PORV's based on the following:
1.
The steam generator PORV's are equipped with redundant safety related solenoid valves (Train A & B) which are deenergized to vent air off the spring loaded PORV's to close them upon receipt of a main steam isolation signal.
Continued heat rejection fol-lowing PORV closure will be provided by the main steam safety valves.
2.
Once hot standby is reached there will be ample time to call in additional personnel or expertise to assess the situation and take the necessary corrective action.
The plant can then be taken to a cold shutdown condition by manually operating the steam generator PORV's using local handwheels in the event in-strument air is not available and cannot be restored.
3.
Instrument air can be provided by any of the three instrument air compressors or either of the twa station air compressors which automatically back up instrument air.
The instrument air compressors and dryers can be manually loaded on the black-out bus during sequence #13 after 12 minutes in the event of a station blackout.
Based on the above, complete unrestorable loss of instrument air is very unlikely and manual operation of the PORV's is acceptable.
4.
Catawba is a Class 2 plant and limited operator action is al-lowed outside the control room in BTP 5-1.
410-10 Rev. 6
d CNS 410.21 Describe the means provided for assuring that instrument air quality (9.3.1) is within the necessary limits to assure proper functioning of all air operated valves and instrumentatic;,in safety related systems.
Response
Section 9.3.1.3 was revised to clarify that the Instrument Air System is not safety related.
Failure of this system will not render any safety system equipment or its function inoperable.
However, a means of operating the Instrument Air System during a station black-out has been provided to facilitate shutdown.
See revised Sections 9.3.1.2.1 and 9.3.1.5 for further discussion on instrument air quality.
410.22 Provide an analysis to demonstrate that drainage of leakage water away (9.3.3) from safety related components or systems is adequate for worst case flooding resulting from pipe breaks or cracks in high or.noderate-energy piping or postulated failure in all non seismic Cacegory I piping near these safety related components or systems.
The analysis must show that drainage by natural routes such as stai wells of equip-ment hatches or by the non-seismic Category I drainage system under failed conditions is adequate to prevent the loss of function of safety-elated components and systems.
Indicate how interconnected drains serving redundant saftey-related equipment or cubicles can be pre-vented from allowing leakage from one failed redundant train from backflowing and flooding out the other train.
In those cases where separate drains are provided for redundant safety related components or systems, provide an analysis that demonstrates that the component and/or area drains serving these components or systems have been sized for maximum leakage flow conditions.
I
)
410-10a Rev. 6 Carry Over
CNS 420.5 On November 7, 1979, Westinghouse notified the Commission of a po-tential undetectable failure which could exist in the engineered safeguards P-4 interlocks.
Test procedures were developed to detect failures which might occur.
The procedures require the use of volt-age measurements at the terminal blocks of the reactor trip breaker cabinets.
1 Concern:
In order to minimize the possibility of accidental shorting or ground-ing of safety system circuits during testing, the staff believes that suitable test jacks should be provided to facilitate testing of the P-4 interlocks.
Provide a discussion on how the above issue will be resolved for Catawba.
Response
In order to implement the Westinghouse recommended procedures, a voltage indicator will be wired to the reactor trip breaker terminal blocks.
This will allow operating personnel to check the status of the P-4 interlock.
This modification will be completed prior to fuel load.
420-9 Rev. 6 New Page
CNS height these lines are terminated above finished ground grade.
If~
these lines are located outdoors discuss the provisions made in your design to prevent entrance of water into the storage tank dur-ing adverse environmental conditions.
Response
l See revised Section 9.5.4.2.1.
l 430.69 Recent licensee event reports have shown that tube leaks are being (9.5.5) experienced in the heat exchangers of the jacket cooling water sys-tem.
Provide a discussion on the means used to detect tube leakage and the corrective measures that will be taken.
Provide the permis-sible inleakage or outleakage for each subsystem cooled by the cooling water system and the time of operation with leakage to assure coolant or cooled liquid quality can be maintained within safe limits.
Also provide the results of a failure mode and effects analysis to show that failure of a piping connection between subsystems (engine water jacket, lube oil cooler, governor lube oil cooler, and engine air inter-cooler) does not cause total degradation of the diesel generator cooling water system.
(SRP 9.5.5, Part III, Item la.)
Response
See revised Section 9.5.5.2.2.
430.70 FSAR Section 9.5.5.2 refers apparently to two kinds of " standby (9.5.5) operation." The first paragraph states that the engine-driven cooling water circulation pump operates during engine standby and assures that the system is completely filled with water.
The third paragraph states that a keep-warm circulating pump keeps "the engine warm during standby." Describe the meaning of standby operation for both of these cases.
Modify the FSAR accordingly.
Response
See revised Section 9.5.5.2.
430-38 Rev. 6
CNS radioactive release to the atmosphere from the faulty unit.
The recovery procedure can be carried out on a time scale which ensures that break flow to the secondary system is terminated before water level in the affected steam generator rises into the main steam line.
Sufficient indications and controls are provided to enable the op-erator to carry out these functions satisfactorily.
Consideration of the indications provided at the control board, together with the 4
magnitude of the break flow, leads to the conclusion that the is-olation procedure can be completed within 30 minutes of accident initiation.
Included in this 30 minute time period would be an al-lowance of 5 minutes to trip the reactor and actuate the safety in-jection system (automatic actions), 10 minutes to identify the ac-cident as a steam generator tube rupture and 15 minutes to isolate the faulted steam generator.
Immediately apparent symptoms of a tube rupture accident such as falling pressurizer pressure and level and increased charging pump flow are also symptoms of small steam line breaks and loss of cool-ant accidents.
It is therefore important for the operator to deter-mine that the accident is a rupture of a steam generator tube in order that he may carry out the correct recovery procedure.
The ac-cident under discussion can be identified by the following method.
In the event of a complete tube rupture, it will be clear soon after the trip that the level in one steam generator is rising more rapidly than in the others.
Also this accident could be identified by either a condenser air ejector exhaust high radiation alarm or a steam generator blowdown radiation alarm.
The operator carries out the following major operator actions sub-sequent to reactor trip which lead to isolation of the faulted steam generator and minimizing primary to secondary leakage.
1.
Identification of the faulted steam generator.
2.
Isolation of the faulted steam generator.
3.
Subcooling of NC system fluid to 50 below no-load temperature.
4.
Depressurization of the NC system to terminate breakflow, and 5.
Terminating safety injection.
Loss of Coolant Accident:
See Table 440.56-4 No manual actions are required of the operator for proper operation of the ECCS during the injection mode of operation.
Only limited manual actions are required by the operator to realign the system for the cold leg recirculation mode of operation, and, at approxi-mately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, for the hot leg recirculation mode of operation.
lTheseactionsaredelineatedinTable 440.56-4.
440-58 Rev. 6
CNS The changeover from the injection mode to recirculation mode is in-itiated automatically and completed manually by operator action from the control room.
Protection logic is provided to automatically open the two Safety Injection System (NI) recirculation sump isola-tion valves when two of four refueling water storage tank level chan-nels indicate a refueling water storage tank level less than a low level setpoint in conjunction with the initiation of the engineered safeguards actuation signal ("S" signal).
This automatic action would align the two residual heat removal pumps to take suction from the containment sump and to deliver directly to the NC system.
It should be noted that the residual heat removal pumps would continue to op-erate during this changeover from injection mode to recirculation mode.
The two charging pumps and the two safety injection pumps would con-tinue to take suction from the refueling water storage tank, follow-ing the above automatic action, until manual operator action is taken to align these pumps in series with the residual heat removal pumps.
The refueling water storage tank low level protection logic consists of four level channels with each level assigned to a separate process control protection set.
Four refueling water storage tank level trans-mitters provide level signals to corresponding normally deenergized level channel bistables.
Each level channel bistable would be ener-gized on receipt of a refueling water storage tank level signal less than the low level setpoint.
A two out of four coincident logic is utilized in both protection cabinets A and B to ensure a trip signal in the event that two of the four level channel bistables are energized.
This trip signal, in conjunction with the "S" signal, provides the actuation signal to automatically open the corresponding containment sump isolation valves.
The low refueling water storage tank level signal is also alarmed to inform the operator to initiate the manual action required to realign the charging and safety injection pumps for the recircula-tion mode.
The manual switchover sequence that must be performed lbytheoperatorisdelineatedinTable 440.56-4.
Following the auto-matic and manual switchover sequence, the two residual heat removal pumps would take suction from the containment sump and deliver bor-ated water directly to the NC system cold legs.
A portion of residual heat removal pump A discharge flow would be used to provide suction to the two charging pumps which would also deliver directly to the NC system cold legs.
A portion of the discharge flow from residual heat removal pump B would be used to provide suction to the two safety injection pumps which would also deliver directly to the NC system cold legs.
As part of the manual switchover procedure, the l suctions of the safety injection and charging pumps are cross con-nected so that one residual heat removal pump can deliver flow to the NC system and both safety injection and charging pumps, in the event of the failure of the second residual heat removal pump.
440-59 Rev. 6
l Q440.56-4 TABLE LOSS OF COOLANT ACCIDENI IMPACT Of THE OPERATOR'S INSTRUCTIONS GIVEN COMPONENTS AND IMPACT Of FAILURE TO TAKE ACTION klQUIRlD ALARMS 10 ALERI IHE TO THE OPERATOR FOR INSTRUMENTATION SINGLE ACTIVE OR THE OPERATOR TAKING OPikATOR OPERAIOR 10 INiilATE DELAY TIME PERFORMING THE NECESSARY 10 COMPLETE COMPONENT A CLOSELY RELATED BUT ACilON A PARilCULAR ACTION ASSUMED REQUIRED ACTION INDICATED ACil0N FAILURE ERRONEOUS ACTION A.
ihe operatur Refer to Sections long Term 1.
Verify that the con-1.
RWST Level 1.
None A&B The plant emergency must manually 6.3.5.4 and 9.2.6 tainment sump isola-Indicator:
operating procedures complete the Lion valves are open.
(PAMS) 2.
Refer to include instructions c harnro ver Table and verification steps of the ICCS 2.
Verify the isolation 2.
All valves re-6.3.2-5.
to ensure proper manual system from valve in each RHR quired to be realignment of the ECCS the injection suction line from the operated for for recirculation by mode to the RWSi is closed.
the switchover the operator. The cold leg re-sequence.
failure to perform one circulation 3.
Close the isolation step or the performance mode.
valve in each SI of one step out of pump miniflow line.
order, as a single Please Refer failure, should not to Section 4.
Close the two isola-reduce ECCS recircula-
- 6. 3.
tion valves in the tion capability below crossover line down-minimum safeguards.
stream of the RHR Should the operator heat exchangers.
fall to take any action following automatic 5.
Open the two paral-ECCS switchover initia-lel valves in the tion, the consequences common suction line will be the loss of the between the cent-safety injection and rifugal charging charging pumps. The pump suction and residual heat removal the 51 pump.
pumps will be protected from damage by auto-6.
Open the valve in matic ECCS switchover the discharge line initiation. For small break LOCA in the 440-62b Rev. 6 New Page
Q440.56-4 (Cont'd)
TABLE LOSS Of COOLANT ACCIDENT IMPACT OF THE OPERATOR'S INSTRUCTIONS GIVEN COMPONENTS AND IMPACT Of FAILURE TO TAKE ACTION M QUI R[ D ALARMS TO ALERI THE TO THE OPERATOR FOR INSTRUMENTATION SINGLE ACTIVE OR THE OPERATOR TAKING OFERATOR OPERATOR 10 INillAIE DELA'Y TIME PERFORMING THE NECESSARY TO COMPLETE COMPONENT A CLOSELY RELATED BUT ACilGN A PARilCULAR ACIl0N ASSUMED REQUIRED AC110N INDICA 1ED ACTION FAlluRE ERR 0NEOUS ACTION from the number 1 unlikely event of losing RHR heat exchanger all high head pump de-l to the suction of livery capability this l
the centrifugal situation could lead to charging pumps; core uncovery and in-open the valve in adequate core cooling.
the discharge line This situation is ad-from the number 2 dressed in WCAP-9691 RHR heat exchanger as the loss of the Emer-to the suction of gency Coolant Recircula-l the 51 pumps, tion (ECR) function fol-I lowing a small break.
7.
Close the isola-LOCA Analyses have been tion valve in the performed for loss of Si pump suction high head safety injec-line from the RWST.
tion for small LOCA which are presented in Close the two paral-WCAP-9573. Inadequate tel isolation valves Core Cooling guidelines in the centrifugal are Leing developed charging pump suc-which wuuld instruct tion line from the the operatar on the RWST.
appropriate operator actions to be taken for this event.
Ll. At appruxi-None Required N/A 1.
Insure that the "5" Refer to Section Refer to for large break LOCA mately 24 signals have been
- 6. 3 Table 6.3.2-5 the residual heat re-hours after reset and defeated moval pump delivery the transient, if not previously to the RCS would be 440-62c Rev. 6 New Page S
l 1
l I
I Q44u.56-4 (Cont'd)
IABLE 1055 0F COOLANT ACCIDENT I
l IMPACT OF THE OPERATOR'S INSTRUCTIONS GIVEN COMPONENTS AND IMPACT OF FAILURE TO TAKE ACTION l
RtqulRED ALARM 5 10 AttR! IHE 10 IHE OPERATOR FOR INSTRUMENTATION SINGLE ACTIVE OR THE OPERATOR TAKING l
OPERAIOR OPERATOR 10 INiilATE DELAY IIME PERIORMING THE NECESSARY TO COMPLETE COMPONENT A CLOSELY RELATED BUT ACIJON A PARTICul AR ACIlON A55UMfD REQUIRED ACTION INDICATED ACTION FAltuRE ERR 0NE005 ACTION
)
the operator accomplished.
sufficient to provide i
must manually adequate core cooling i
switch over 2.
Terminate RHR pump during recirculation.
j to the hot fluw to the RCS i
ley ret ir-cold legs and es-cul stion tablish RHR pump l
mode.
flow to the RCS i
hot legs by:
j a) closing the RHR l
cold leg header
}
isolation valves b) opening the two isclation valves in the crossover line downstream of the RHR heat exchangers I
c) opening the RHR hot leo header i
isolation valve l
I 3.
Termina'.e 51 pump 8
i flow ts the RCS cold legs and establ.ish 51 ptap flow to the RCS hot legs by:
a) stop SI pump no.
I and close its corresponding i
440-62d Rev. 6 New Page
i l
l l
Q440.56-4 IABLE LOSS OF COOLANT ACCIDENT IMPACT OF THE OPERATOR'S INSTRUCTIONS GIVEN COMPONENTS AND IMPACT OF FAILURE TO TAKE ACTION ktQulktD ALARM 5 TO ALERI IHE TO IHE OPERAIOR FOR INSTRUMENTATION SINGLE ACTIVE OR THE OPERATOR TAKING OFERAIOR OFERATOR TO INIllATE DELAf TIME PERFORMING THE NECESSARY TO COMPLETE COMPONENT A CLOSELY RELATED BUT Acil0N A PARTICULAR ACTION A55UMED REjulRED ACTION INDICATED ACTION FAILURE ERRONEOUS ACTION cold leg cross-over header iso-lation valve b) open its corre-sponding hot leg header isolation valve c) restart SI pump no. I d) stop SI pump No.
2 and close its corresponding cold leg cross-over header iso-lation valve e) close the SI common cold leg header isolation valve f) open SI pump No. 2's hot leg header isola-tion valve g) restart SI pump No. 2 C.
Check Reactor None Required Ihe operator is instructed C.1 RCP's If a single Discussed in WCAP-9584.
Coolant Pump to trip all RCP's when active failure trip criteria the RCS pressure reaches C.2 RCS pressure prevents trip-a specified pressure and indicators ping RCP's 440-62e Rev. 6 New Page m
Q440.56-4 (Cont'd)
TABLE LOSS OF COOLANT ACCIDENT IMPACT OF THE OPERATOR'S INSTRUCTIONS GIVEN COMPONENTS AND IMPACT Of FAllORE TO TAKE ACTION REQUlktb At ARMS 10 ALERT IHL 10 THE OPERATOR FOR INSTRUMENTATICN SINGLE ACTIVE OR THE OPERATOR TAKING OPFRAICR OPE RATOR TO INITIATE DELAY TIME PERf0RMING THE NECESSARY 10 COMPLETE COMPONENT A CLOSELY RELATED BUT ACTION A PARI!CulAR ACTION ASSUMED REQUIRED ACTION INDICATED ACTION FAlluRE ERR 0NE0US ACTION SI operation is verified.
the impact is C.3 Safety injection discussed in flow indicators.
- lime of RCP trip is a f unctiori of t;reak size.
It is assumed that RCP's are tripped prior to or at the RCS pressure specified in the plant procedures.
440-62f Rev. G New Page E
CNS 40.84 Provide an analysis, with justification, of the extent of DNBR and (15.3.3) fuel damage.
Show that the extent of fuel damage is sufficiently limited so that the core will remain in place and intact with no loss of core cooling capability.
Has the calculation of the Zr-H2O reaction included the reaction on the inner cladding surface when the cladding fails?
Provide an analysis of the percent of Zr-H O 2
reaction for the core and discuss the consequences of H generated.
2 Provide figures for three loops in operation that are consistent with those for four loops in operation, for example, Figure 15.3.3-6 is labelled for two loops in operation.
Response
The locked rotor analysis in Section 15.3.3 shows no fuel failures occuring with calculated peak clad temperatures well below the limit temperature of 2700 F.
With four loops operating initially, a peak clad temperature of 2200 F was calculated.
With three loops operating initially, a peak clad temperature of 2331*F was calculated.
A maximum of 9% of the rods experience DNB ratios less than the limit value.
DNB ratios less than the limit values do not neces-sarily imply fuel failures.
However, for the purpose of conser-vatively calculting environmental radioactivity dose releases only, the percen* age of rods in DNB was assumed to be equal to the per-centage of fuel activity released.
Because no clad failures occur, zirc-water reaction on the inner clad surfaces is not applicable.
With four loops operating initially, 1.4% by weight of the outer clad surface will undergo zirc-water re-action, with three loops initially operating, 1.9% by weight of the outer clad surface will undergo zirc-water reaction.
The maximum amount of H release is calculated to be less than 1 Ft3 Even if all 2
of this non-condensable gas were to migrate and collect at the top of the steam generator U-tubes (volume of U-tubes is + 70 Ft3/SG).
This would not be enough to disrupt core cooling.
This is substantiated by natural circulation work done in WCAP-9600 (Vol. 1), page 2.9-7.
The title for Figure 15.3.3-6 was corrected to read "...Three Loops In Operation, 1 Locked Rotor."
In addition, figures of nuclear power, heat flux, and clad temperature with three loops in operation are added to the FSAR.
440.85 The acceptance criteria in the Standard Review Plan for Sections (15.3.3, 15.3.3 and 15.3.4 state that the rotor seizure and shaft break tran-15.3.4) sients should be analyzed assuming turbine trip and coincident loss of offsite power and coastdown of undamaged pumps.
Describe how your analysis has considered these assumptions.
C 440-82 Rev. 4 New Page h
Response
See response to Question 440.115.
440.86 Demonstrate that a rotor seizure and shaft break in a reactor coolant (15.3.3) pump will not generate a more serious condition or result in a loss of function of the reactor coolant system or containment barriers.
Response
The peak RCS pressure observed during a locked rotor or a shaft break in a reactor coolant pump was 2570 psia.
This is well below 110% of design pressure (2750 psia).
During the short time that RCS pressure is high enough to open pressurizer safety valves, the relief rate was within the maximum capacity of the safety valves.
Thus, a loss of function of the reactor coolant system or contain-ment barriers will not occur.
440.87 You classify the reactor coolant pump shaft break and locked rotor (15.3.3, accidents as ANS Condition IV (limiting fault).
The Standard Review 15.3.4)
Plan for Sections 15.3.3 and 15.3.4 classify these accidents as Con-dition III (infrequent incident).
Show that the transients meet the acceptance criteria for an infre-quent incident.
Response
The reactor coolant pump locked rotor and reactor coolant pump shaft break events are classified according to ANS as Condition IV events -
limiting faults.
Westinghouse follows this classification in Chapter 15 safety analysis.
The results of the Catawba locked rotor shaft break events meet acceptance criteria for an infrequent event.
The peak RCS pressure is maintained well below 110% of design pressure (2750 psia).
The peak clad temperature is well below 2700 F, thus no clad failures are expected to occur.
440.88 Reference or describe the analytical model used for obtaining the (15.4.6) results in Section 15.4.6.2.
Discuss the degree for conservatism incorporated in this model.
Response
440-83 Rev. 6
CNS 460.7 Table 1.8-1 indicates that Regulatory Guide 1.143, Rev. 1, 10/79 (Tables 1.8-1 (formerly Branch Technical Position ETSB II-1, Rev. 1), " Design 11.2, 11.3, Guidance for Radioactive Waste Management Systems, Structures and 11.4)
Components Installed in Light-Water-Cooled Nuclear Power Plants" is not applicable to Catawba.
This is not acceptable.
Compare your design of liquid, gaseous and solid radwaste systems to each position in Regulatory Guide 1.143 and list the items of non-compliance and the justification for it for the purpose of evaluat-ing your design.
Response
See revised Table 1.8-1.
l l
460.8 In Section II.2.3 for estimating the liquid releases, credit is (11.2) taken for waste evaporator and condensate demineralizer in pro-cessing the floor drain tank contents.
From the description of the floor drain tank subsystem operation (11.2.2.7.1.5) it is not clear that evaporator and condensate demineralizers will be used frequently.
Full credit may not be taken unless both of these pieces of equipment are used continuously.
Similarly, the full credit should not be taken for waste evaporator condensate demin-eralizer for processing waste collected by waste evaporator feed tank.
Please justify your assumption.
Response
Catawba Nuclear Station anticipates never having to process the con-tents of the floor drain tank using the waste evaporator or waste evaporator condensate demineralizer, to meet effluent requirements.
However, these components will be used to process approximately 30%
of the floor drain tank's contents to reclaim distillate.
See revised Sections 11.2.3(7) and 11.2.2.7.1.5.
460-3 Rev. 6
CNS Residual containment spray operation is initiated manually when re-quired by the operator and only if the following two criteria are met:
1.
Emergency Core Cooling System and Containment Spray System are operating in the recirculation mode.
2.
At least one hour has passed since the initiation of the Safety Injection signal.
The Residual Spray may be required due to high containment pressure.
High containment pressure would be indicated to the operator via the pressure sensors in the Containment Pressure Control System, which has pressure indicators in the control room.
Refer to FSAR Section 7.6.5 for Containment Pressure Control System.
It is not intended that RHR spray be limited to a maximum value of 1575 gpm.
Therefore, no flow control elements are required to ad-just to a particular flowrate.
480.16 Identify all secondary containment openings such as personnel access (6.2.3) doors and equipment hatches.
Discuss the administrative control to be exercised over these openings.
Response
Access to the annulus ic under the control of the Station Operations Department.
Both upper and lower doors to the annulus are kept locked and the keys are under the control of the shift supervisor.
If an authorized individual must gain entry to the annulus he or she obtains the key from the shift supervisor and is " logged" in.
Anytime someone is inside the annulus another person is placed at the annulus door to prevent unauthorized entry of other individuals.
Station Directive 3.1.2 is the procedural document that establishes the above controls.
There are no remote alarms or indications of annulus door position or activation.
480.17 For each containment penetration, specify whether the line is es-(6.2.4) sential or nonessential, in accordance with the provisions of NUREG-0737, " Clarification of TMI Action Plan Requirements," II.E.4.2, Containment Isolation Dependability.
480-13 Rev. 6
CNS A Seismic Design Criteria NRC regulations require that nuclear power structures, systems, and compo-nents important to safety be designed to withstand the effects of natural phenomena such as earthquakes.
Detailed requirements and guidance regard-ing the seismic design of nuclear plants are provided in NRC regulations and in regulatory guides issued by the Commission.
However, there are a number of plants with construction permits and operating licenses issued before the NRC's current regulations and regulatory guidance were in place.
For this reason, rereviews of the seismic design of various plants are be-ing undertaken to assure that these plants do not present an undue risk to the public.
Task A-40 is, in effect, a compendium of short-term efforts to support such reevaluation efforts of the NRC staff, especially those related to older operating plants.
In addition, some revisions to the Standard Re-view Plan sections and regulatory guides to bring them more in line with the state-of-the art are expected.
The seismic design basis for this facility is described in FSAR Section
- 3. 7.
Pending the identification of new issues by the NRC Staff that could alter the plant design basis, we do not expect the results of Task A-40 to affect our conclusions.
Accordingly, we conclude that Catawba Nuclear Sta-tion can be operated before the ultimate resolution of this generic issue without endangering the health and safety of the public.
l 1
730-10 Rev. 6
s_
CNS A Seismic Qualification of Equipment in Operating Plants The design criteria and methods for the seismic qualification of mechanical and electrical equipment at the Catawba Nuclear Station are presented in Sections 3.7, 3.9, 3.10, and 5.2 of the FSAR.
These criteria and methods are under review by the Staff as a part of the overall operating license review, and any outstanding items.in this area will be resolved to the sat-isfaction of the Staff prior to issuance of the operating license.
There-fore, we conclude that the Catawba Nuclear Station can be operated before resolution of this generic issue without undue risk to the health and safety of the public.
730-15 Rev. 6