ML19317G643

Forwards Request for Addl Info Re FSAR Sections on Introduction & Summary Reactor,Rcs,Design Bases for Structures & Equipment,Auxiliary & Emergency Sys,Steam & Power Conversion Sys, & Radwaste & Radiation Protection

ML19317G643
Person / Time
Site:	Rancho Seco
Issue date:	08/03/1972
From:	Deyoung R US ATOMIC ENERGY COMMISSION (AEC)
To:	Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT
References
NUDOCS 8003250698
Download: ML19317G643 (20)

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s 1g Docket No. 50-312 AUG 3 1972 Mr. E. K. Davis, Gene m l Counsel Sacramento Municipal Utility District 6201 S Street, P. O. Box 15830 Saca mento, California 95813 y

Dear Mr. Davis:

l We find that we need additional information to complete our review of your application for an operating license for t%o Rancho Seco Nuclear Genenting Station. The specific information required is described in the enclosure and has been categorized into groups which generally correspond to applicable section headings in the Final Safety Analysis Report.

We have not completed our review of the subject matters covered in this request for additional information. At a later date we vill request n

additional informtion, if necessarf, on these subject matters and others not addressed herein.

In order to maintain our licensing review schedule we will need a coin-plately adequate msponse by August 28, 1972. Please inform us within 7 days after receipt of this letter of your confirmation of the schedule app or the date you will be able to meet. If you cannot meet our specified date or if your reply is not fully responsive to our requests it is highJy likely that the ovem11 schedule for completing the licensing' review for this project will have to be extended. Since. reassignment.of the staff's l efforts will require courpletion of the new assignment prior;

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to returning to this project, the extent of extension will most likely.

en be greater than the extent of delay in your response.

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Please contact.us if you have any questions' regarding' the enclosedkunsts; '.1

@hf'..$h Sincerely,

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Request for Additional Information cc:

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' l REQUEST FOR ADDITIONAL INFORMATION SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR POWER PLANT DOCKET NO. 50-312 1

1.0 INTRODUCTION

AND

SUMMARY

1.3 With respect to the Quality Assurance Organization (FSAR Sections 15.1.1 and 1B.2.1) provide the following information:

1.3.1

.The organization chart (FJgure 1B-1) and the responsibilities of each department should be clarified and expanded as follows:

(a). Describe the overall responsibilities and function of the General Manager and Assistant General Manager / Chief Engineer.

(b) Identify and describe each QA/QC position within the SMUD'and Bechtel organizations. The organizational charts and description of responsibilities as presented are not detailed enough.

1.3.2 Specify the projected number of QA/QC personnel required by SMUD and Bechtel to implement the QA/QC program during initial plant operation and the number required by SMUD when full responsibility of' plant operation is assumed.

1.3.3 Identify where the QA Director and QA/QC personnel will be located, i.e., at SMUD's home office or at Rancho Jeco.

J 1.3.4 Describe the qualification and training requirements of each QA/QC position.

1.3.5 What direct access do the QA/QC Engineers have with upper management concerning QA/QC problems?

1.3.6 Describe the overall responsibilities of Bechtel's Startup and Quality Assurance Manager.

1.3.7 Describe the communication and technical direction interfree between

the SMUD and Bechtel organizations with special emphasis given to the QA/QC interface.

1.4 With respect to the Quality Assurance Program (FSAR Sections lE.1.2 and 1B.2.2) provide the following information:

.1.4.1 Describe how the SMUD and Bechtel documents set forth in their OA Manuals are reviewed, approved, revised, distributed and evacr.aed Indicate how SMUD and Bechtel assure that the appropriate d. p tmi and organizations properly implement these documents.

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1.5 With respect to Design _ Control (FSAR Section 1B.1.3) provide the following information:

1.5.1 What procedural measures are used to assure that the SMUD and Bechtel design drawings and specifications are reviewed by QA/QC engineering?

1.5.2 Describe in more detail the QA/QC responsibilities of SMUD and Bechtel for the review and approval of design drawings and specifications.

Is the quality engineer required to evaluate design characteristics to determine if they can be inspected and controlled?

1.5.3 Describe the procedure for controlling design interfaces between SMUD, Bechtel and suppliers.

(a) Are formal design review meetings held with participation by the responsible interface organizations?

(b) How are design interface characteristics identified so they can be easily recognized?

(c) Describe how design interface charteristics are controlled.

1.6 With respect to Instructions, Procedures and Drawings (FSAR Section 13.1.5) indicate:

1.6.1 What provisions in the QA Program assure that activities affecting quality are defined by documented instructions, procedures and drawings?

1.7 With respect to the Control of Special Processes (FSAR Section 1B.1.9) provide the following information:

1.7.1 Describe how special process procedures are controlled.

1.7.2 Describe how changes and/or revisions to special process procedures are generated, reviewed, approved and controlled.

1.8 With respect to Inspection (FSAR Section 1B.1.10) provide the following infornscion:

1 1.8.1 What are the SMUD and Bechtel qualification requirements for dimensional, ultrasonic, liquid-penetrant and radiographic inspectors?

1.9 With respect to Test Control (FSAR Section 1B.1.11) indicate:

1.9.1 Where in the QA Program does SMUD provide for a well defined, controlled and documented test program to deternine that structures, systems and components perform satisfactorily it, service?

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1.9.2-What are the responsibilities of QA/QC for preparing, reviewing, approving, controlling and implementing test control plans?

1.10

-With respect to the Quality Assurance Program for Station Operations (FSAR Section 1B.2) provide the following information:

1.10.1 Provide more detail concerning the functional responsibilities of the Quality Assurance Director.

1.10.2 Identify and describe each QA/QC position within SMUD's operating organization.

1.10.3 Describe in more detail the QA/QC functions during plant operations as

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related to the 18 criteria of Title 10 CFR 50 Appendix B.

Include the impact of these activities on plant tests and surveillance operations, calibration of test equipment, maintenance, modifications of systems and equipment, and purchase of material.

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3.0 Reactor 3.7

-Supplement your response to Request 3.1, Amendment 12 to FSAR, by providing a detailed comparison of key parameters, such as the configuration of reactor internal structures and the flow characteristics, between Oconee-1 and Rancho Seco-l.

In the event the Oconee-1 is qualified as the prototype plant, verify your intention to conduct a confirmatory preoperational vibration test or visual inspections in accordance with Safety Guide 20.

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4.0 REACTOR COOLANT SYSTEM 4.23 To assure that ferritic materials of pressure-retaining components of the reactor coolant pressure boundary will exhibit-adequate fracture toughness under normal reactor operating conditions, system hydrostatic tests, and during transient conditions to which the system may be subjected, provide the following information:

4.23.1 A statement that proposed operating limitations during startup, shutdown, and hydrostatic tests of the reactor coolant _ system will use as a guide Appendix G, " Protection Against Non-Ductile Failure," of the recently revised ASME Code Section III fracture toughness rules (Code Case 1514).

4.23.2 Discuss the extent to which you have reviewed the design of affected systems and components to determine that annealing of the reactor vessel will be feasible'should it be necessary because of radiation embrittlement af ter several years of operation. State the maximum reactor vessel temperature that can be obtained using an in-place annealing procedure.

4.24 Provide information regarding your proposed capsule withdrawal schedule and state the degree of conformance with that of the AEC proposed " Reactor Vessel Material Surveillance Program Requirements," $ 50.55a, Appendix H, published in the Federal Register on July 3,1971.

4.25 Regarding preoperational mapping'of the reactor vessel by ultrasonic examination, to meet the requirements of IS-232 of Section XI of the ASME Code, state the acceptance standards that were used to establish acceptability of the vessel for service.

4.26 Describe the vibration operational test program which will be used to verify that the piping and piping restraints within the reactor coolant pressure boundary have been designed to withstand dynamic effects resulting from transient condition such as valve closures, pump trips, etc. Provide the bases for the acceptable vibration amplitudes established to confirm the structural integrity'of the piping and piping restraints under operational transient loadings.

m 4.27 Identify the dynamic testing procedures used in the design of Category I mechanical equipment (such as fans, pump drives, valve operators, heat exchanger tube bundles) to withstand seismic, accident and operational vibratory loading conditions.

Include the methods and procedures employed which consider the frequency spectra and amplitudes calculated to exist at the equipment supports. Where tests or analyses do not include' evaluation of the equipment in the operating mode, descrJbe the bases for assuring that this equipment will function if subjected to seismic accident loadings and vibratory loadings.

4.28 Supplement the response to Requests 4.14 and 4.15, Amendment 10 to FSAR, by providing the following information concerning che criteria and measures that have been used to assure that the containment vessel and all essential equipment within the containment, including components of the reactor coolant pressure boundary and main steam lines, have been adequately protected against the effects of blowdown jet forces, and pipe whip resulting from postulated rupture of piping com-ponents located either inside or in close proximity to the outside of containment:

4.28.1 Identify the systems (or portions of systems) in which design basis piping breaks are postulated to occur.

4.28.2 Specify the design loading combinations and criteria including the associated design stress limits that will be applied to the postulated rupture of unrestrained piping and pipe whip restraints.

4.28.3 Provide a summary of the dynamic analyses performed for Class 1

. piping and associated supports which determine the resulting loadings as a result of a postulated pipe break including:

(a) the locations and number of design basis breaks on which the dynamic analyses are based.

(b) the postulated rupture orientation, such as a circum-ferential and/or longitudinal break (s), for each postulated design basis break location.

(c) description of the forcing functions to be used for the pipe whip dynamic analyses.

Including direction, rise time, magnitude, duration and initial conditions that adequately represent the jet stream dynamics and the system pressure differences.

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~(d) typical diagrams of the mathematical models used for the dynamic analysis.

(e) a summary of-the analyses performed to demonstrate that unrestrained motion of ruptured lines will not sever adjacent impacted piping or pierce impacted areas of containment steel wall. (or liner).

4.28.4 Discuss the pipe restraint design criteria to prevent pipe whip impact, include a description of the typical pipe whip restraints, and a summary of number and location of all restraints used in each system.

4.29 Response 4A.20 (AEC No. 4.21) states that pressure-relieving devices were designed in accordance with Paragraph I-722.6 of ANSI B31.7.

This standard does not address itself to the structural integrity of the valve bodies. Provide the stress or deformation limits which were used for safety valves and justify the basis for their application.

4.30 With respect toResponse 4A.15 (AEC No. 4.16) provide the actual stress limits employed and state whether the stresses are based on an elastic or inelastic analysis. If an inelastic (or elasto-plastic) analysis was used provide a summary of t;he method of analysis and justify the limits used.

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m 5.0 DESIGN BASES - STRUCTURES AND EQUIPMENT 5.74 Response 5A.55 is inadequate. To check the accuracy of the modal response spectrum method, provide a comparison of the responses obtained at selected points in the station structure using both the modal analysis response spectrum and time history method s.

5.75 Response SA.62 is not acceptable.

If static loads equivalent to the peak of the floor spectrum curve are used for the seismic design of components and equipment, justify the use of peak spectrum values by demonstrating that the contribution of all significant dynamic modes of response under seismic excitation have been included.

5.76 Response SA 60 is inadequate. Consideration of only those modes having frequencies lees than 20 cycles per second may not be sufficiently conservative. Provide justification for not including all. modes in the seismic analyses which have a significant contribution to the response.

5.77 With regard to Response 5A.71, provide the basis for establishing the actual damping values which will be used to verify the response of different types of Category I systems and components under seismic excitation.

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9.0 AUXILIARY AND EMERGENCY SYSTEMS 9.18 The letdown flow rate during normal reactor operations will vary between 45 and 140 gpm.

Provide a description of the indications available to the operator that determine these flow rates. Describe how letdown flow is regulated during abnormal reactor conditions.

9.19 The letdown temperature instrumentation in the letdown line, downstream of the coolers, providc; an alarm and an interlock for isolation to protect the purification system. Discuss the effects of failure of the interlock on the purification system.

9.20 Provide an evaluation of the effectiveness of the purification system to provide clean-up capability.

9.21 Section 9.3.2.3 of the FSAR states that portions of the boric acid addition system are designed as seismic and quality Class I.

Further, it is stated that all valves and lines contacting concentrated boric acid solution are provided with redundant heat tracing, controls and alarms. Provide the following additional information regarding this system:

9.21.1 Delineate the portions of the system designed as seismic and quality Class I.

9.21.7 Delineate the lines, valves, and tanks provided with redundant heat tracing.

9.21.3 Describe the alarms that are used with the heat tracing and where they are annunicated. Describe the operator action that will be taken.

9.22 Section 9.4.2.2 of the FSAR dealing with the nuclear service raw water system states that portions of the nuclear service cooling water system are designed as seismic and quality Class I.

Appendix SB states that i

both of these systems are in all aspects designed as seismic Class I.

Section 9.4.2.2 should addresa the entire raw water system. If contrary to Appendix 5B, only portions of that system are designed as seismic Class I, provide an analysis showing that failure of the non-Class I (seismic) portions of the system will not prevent the system from meeting all emergency requirements.

9.22.1 Indicate the extent to which AEC Safety Guide No. 27 " Ultimate Heat

Sink" will be followed. If the Rancho Seco Station design does not meet-this guide, provide the basis and supporting evaluation regarding the acceptability of your design.

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9.23.

State'the operating pressures of the nuclear service cooling water system and the nuclear service raw water system. Discuss the control and detection of radioactivity in the raw water system.

9.24 Describe the operational requirements of the decay heat removal system when it is used to cool the pressurizer by auxiliary spray for cold shutdown.

Include a single failure analysis for this mode of operation., Add this spray function to Figure 9.5-1 showing all values.

9.25

.The borated water storage tank is located outside the reactor and auxiliary buildings.

In light of the safety function requirements of this tank, provide the following additional information:

(a) discuss the effects of a heater failure, (b) describe the requirement for leak detection and leakage control, and (c) describe in-service inspection requirements.

9.26 Provide an analysis of the failure of the supply header of the seismic Class II component cooling water system delineating the effects of the loss of the high pressure injection pump lube oil coolers, the spent fuel pool coolers, and the penetration coolers in' conjunction with respective design basis accidents.

i 9.27 Describe the means for detecting and isolating a decay heat removal loop (reference Figure 9.5.1) in the event of a passive failure in the loop following a design basis accident.

9.28 Our present requirements include the capability of the pump room ventilation system to adequately handle the anticipated humidity loads due to leakages in the areas of ECCS equipment which must function following an accident, and to limit radioactive releases which would result from leakage from pumps and equipment required to operate for prolonged

• periods following an accident. Discuss this capability with respect to the design of your system including a failure analysis of the proposed system.

9.29 Provide an analysis of the failure of the seismic Class II ventilation systems for the electrical equipment roors considering the limiting ambient conditions. - The analysis is required to justify 'the acceptability of the loss of this Class II system on the safety related equipment in the rooms.

9'30

' Provide an analysis of the failure of the seismic Class II diesel l

generator room ventilation system considering limiting ambient i

conditions including assurance that the diesel generator functional

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capability following an accident will not be precluded.

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m 9.31 Appendix SB states that a portion of the spent fuel cooling system

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is designed to seismic Class I requirements. State which portion is

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designed Class I (seismic) and provide just!fication for designing the remaining portions other that Class I (seismic).

1 9.31.1 State the provisions for long term cooling in the event normal spent fuel poel cooling capability is lost.

l 9.32 Discuss measures that will be taken to prevent the accidental opening of the gate valves when the fuel transfer canal is drained (reference page 9.8-4 of the FSAR).

9.33 Verify that dropping a fuel cask more than the 10 CFR 71 design value (30 foot drop) is precluded in the design of your plant. If the 30 foot drop limit can be exceeded, provide an analysis of the structural and radiological consequences.

9.34 Provide a discussion, with the aid of drawings, of the principles of operation of the spent fuel shipping cask crane (Turbine Building Gantry Crane) and the reactor building crane plus any special handling fixtures employed in handling their respective loads such as the reactor

- vessel head, reactor vessel internals and the spent fuel shipping cask.

Describe in detail the applicable codes and standerds used in the design, fabrication, installation and testing of the crane, rails, supporting structures, bridge, trolley, hoists, cables, lifting hooks, special

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handling fixtures and slings. For each, list its design load rating, preoperation test load, maximum operating loads and the tests loads that will be used throughout the life of the facility.

9.34.1 Describe the modes of failure that were cotisidered in the design of the cranes such as breaking of cables, lif ting slings, sheared shaf ts, keys and stripped gear teeth and brake failures. Also discuss the limitations and control that will exist in handling objects over an opened reactor

- vessel.

9.35 Assuming the maximum drop height, discuss with the aid of drawings where appropriate, the consequences of dropping the following:

9.35.1 The reactor vessel head over the open reactor vessel.

9.35.2 The upper plenum assembly within'the reactor vessel.

i 9.35.3 The largest segment of the removable missile shield on the reactor vessel.

9.36 Provide specific reference to the applicable codes and standards to which the fire protection system, described in answer to Response 9A.7, has been designed.

9.37 Provide the seismic design classification of the fire protection system components and piping. Describe, with the aid of drawings, w

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m the design of the system and how it assures that the normal me

' accidental operation of the system does not cause an unsafe condition (i.e., flooding of safety related equipment) and the failure of any parts of the system not de. signed to Class I (seismic) standards would not damage or prevent fire protection to a Class I (seismic) structure, system or component.

9.38 Describe the extent'to which the safety-related aspects of the fire protection system can operate with any single failure.

9.39 Describe the accessibility (with respect to radiation, toxic combustion products, etc.) of all areas which rely on manual fire protection equipment and identify these areas.

9.40 Describe the construction features of the plant (i.e., fire walls, autamatic fire doors, noncombustible materials, spatial separation) that are incorporated to minimize the potential for fires in plant areas vital to reactor safety.

9.41 Provide a description (inclading a drawing) and the design bases of the plant compressed air system. Provide a listing of any safety-related equipment that the system supplies and describe the functions of each such equipment if this has not already been done.

9.'42 -. State the design bases for the diesel engine closed circuit water.

cooling system and include a P&ID.

9.43 State the design considerations with respect to protection from galvanic corrosion of the buried diesel fuel storage tanks.

9.44 Describe the potable water system for the Rancho Seco plant. Justify that the system, by its failure, can not affect any safety related equipment. Provide the criteria applied to the design of the system that preclude contamination from any radioactive'or. toxic source in the plant.

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7 10.0 STEAM AND POWER CONVERSION SYSTEM 10.7

.escribe the design features provided to minimize the effect of a rupture of a feedwater line. Identify the preoperational quality control and surveillance procedures that will be used to assure the integrity of the main and auxiliary feedwater system.

10.8 Provide a description of your inservice inspection plans for the main steam lines, steam valves, and the turbine-generator. Describe the tests and nondestructive examinations planned for the turbine generator highly stressed parts and rotating members and relate the detectable flaw size to the critical crack size for those members.

10.9 Describe, with the aid of drawings, the bulk hydrogen storage facility and its distribution system. Describe the protective measures taken l

to prevent fires and explosions during operations, such as purging

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the generator, as well as during normal operation.

10.10 Section 10.4 of the FSAR discusses motor operated valves in branch.

lines connected to the main steam line that must be closed to preventan uncontrolled blowdown of both steam generators in the event of.a main steam line break. Describe the means used to close these valves (manual or automatic), the seismic design classification of the sesociated branch lines out to and including the motor-operated valves used for isolation and whether the design precludes such uncontrolled blowdown upon the failure of a single valve in one of these lines.

10.11 Describe control instrumentation on the turbine bypass valved that precludes overpressurization of the condenser during loss of all unit i

a-c power.

10.12 Main steam isolation is accomplished by the piston operated turbine stop valves. With respect to a steam line break outside the reactor building provide the fellowing information:

10.12.1 Describe the operators for the turbine stop valves and verify that they are designed to Class 1 (seismic) requirements also.

10.12.2 Verify that each of the stop valves is individually operated.

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11.0 RADIATION WASTE AND RADIATION PROTECTION 11.8 Describe the offsite monitoring program with respect to its capability 1

to determine, in conjunction with effluent monitoring, estimates of individual exposure by all significant pathways and the population l

exposures resulting from them at the design levels of radiation and radioactive affluents. Any differences between preoperational and operational programs should be delineated.

i 11.9 Enumerate the expected (and measured) background levels of radiation and radioactivity (and their variation in time and space), both from natural and man-made sources. Analysis should be done to determine background levels of the specific nuclides expected to l

be released during reactor operation.

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11.10 Based on the liquid and gaseous source terms already provided, describe all significant pathways of human exposure, both internal i

and external, that result from plant operation.

Discuss the bases i

and mathematical models used to make exposure estimates from the effluent release and environmental monitoring data.

List and justify all assumptions made, or relevant information to be developed such 4

as, biologica1' accumulation factors, food consumption rates, cloud depletion rates, decay factors before exposure, etc.)

11.11 Provide the bases for frequency of sampling in light of Requests 11.9 and 11.10.

11.12 Describe the size and physical characteristics of each type of 4

sample, the kinds of. radiological analyses to be performed and the measuring equipment to be used, and discuss and justify the sample detection sensitivity.

11.13 Describe the kinis of mathematical and statistical analyses to be performed on the resultant data, and give an indication of the type of. format to be used to present the results.

11.14 Discuss and justify, in light of _ the parameters described above, the overall, statistical sensitivity of the program to achieve its

. objectives of estimating the exposures to man which result from normal operation of the plant. The monitoring program should be i

modified to include y-ray analysis for specific radionuclides on j

all samples which now require only gross - Y scans. The activity shown to be present in the samples by this analysis should account for the activity found by gross - S scans. Also, total background doses should be measured monthly rather than quarterly, and milk

.should be sampled and analyzed weekly rather than monthly, i

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3 11.15 In Section 10.2.7 of the FSAR it is stated that monitoring of the steam generator shell-side sample points and the air ejector off-gas will detect contamination and actuate an alarm. Describe these 1

monitors, the manner in which the cumulative release of radioactive material will be able to be determined, and the tests that have been or will be performed to demonstrate that the monitors to be used will reliably detect the appropriate range of possible radioactive releases.

11.16 The statement is made in Section 10.2.4 of the FSAR that radiation shielding around components of the steam power conversion system is not required. Section 11.2.1.1 of the FSAR places the turbine plant service areas in Radiation Zone I.

Assuming operation with the maximum permissible primary system activity and a maximum permissible primary to secondary leakage rate, show by calculation that radiation levels to which operation and maintenance personnel could be exposed while performing i

operations around the main steam lines, turbine generator, feedwater system, steam generator, and letdown systems, including miscellaneous waste processing systems are not in excess of 1 mrem. List all assumptions used in deriving these levels, j

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14.0 SAFETY ANALYSIS 14.3 Discuss the consequences of failure of one of the turbine bypass valves failing to close.

.14.4 Provide an analysis of a main steam line break assuming a concurrent active failure of one turbine stop valve operator (and the. valves it operates if redundant Operators are not used) in the unbroken steam line. Further, for the dose calculations, assume that the pr. mary to secondary leakage and activity are at the maximum values specified i

in the Technical Specifications prior to and during the accident.

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14.5 For.the steam line failure accident:

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14.5.1 Indicate the time sequence for significant events such as initiation of accident, reactor trip, turbine stop valve closure, and isolation of upstream branch lines as appropriate, feedwater valve closure, steam bypass valve operation, safety and relief valve operation, and high pressure injection actuation.

14.5.2 In addition to the tine-history traces of Figure 14.2-1 of the FSAR, provide additional traces of DNBR, pressurizer level, pressurizer pressure, affected steam generator pressure and outlet temperature, and unaffected steam generator pressure and outlet temperature.

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14.5.3 We note that you have analyzed this accident assuming certain operator actions in one case and without operator action in another case. Provide a detailed clarification of the actions which occur and their timing both with and without operator action. If the time-history traces 3

of Figure 14.2-1 of the FSAR are not for the worst case, provide a j

set of. traces for the worst case. Discuss the effects of the dif-

' ferences in action on the _ potential consequences of the accident such as fuel failures and radiological dose.

14.5.4 Evaluate the effect' of the. secondary system pressure on steam generator leak rate. If a constant leak rate was used, revise Table 14D-16 to reflect the increased leak rate due to the reduced secondary system pressure. State the length of time for the primary to secondary leakage to be terminated.

14.5.5. Indicate each term used in the calculation of the doses and release activities in Table 14D-16.

If steady state decontamination factors

'are ueed,' justify their validity under accident conditions.

14.6 Two loss-of-electric-power situations are considered in the FSAR.

Provide tira-history traces of the parameters shown on Figure 14.2.1 of ~the FSAR and specified in Raquest 14.5.2 above for each situation.

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14.7 For the loss of all a-c power discuss the capability of sadeiy~

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shutting down the plant.

Include a list of all necessary components, control systems, instrumentation, and other electrical systems that must be supplied from the station batteries. Discuss how power will be provided from these batteries.

Estimate the power consumption of each component and indicate the maximum time the plant could main-tain this condition safely.

14. 8 For the loss of flow accident:

14.8.1' Specify the Doppler and moderator coefficients used in the analysis.

14. 8.2 Provide assurance that the reactor coolant pumps, piping, and restraints have been designed in such a manner as to prevent the locked rotor accident from initiating a more serious accident.

14.9 State the total amount of steam expected to be released per year of plant operation; include the data base us.d for estimating the number of turbine trips per year. Indicate the maximum quantity of radio-activity that could be released via this path assuming the plant is operating with the radioactivity concentration in the secondary system at (1) the maximum level to be permitted by the technical specifica-

.tions and (2) the anticipated level assuming some tube leakage and fuel failures.

14.10 In Sections 14.1.2.8.3 and 14.1.2.8.4 of the FSAR'it is assumed that heat is released to the atmosphere via the secondary system. A.'though under normal circumstances partial or full operation of secondary system dump valves is likely, no credit will be given for their. operation. Revise your safety analysis as necessary to eliminate credit for these dump valves.

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