ML19319D736
| ML19319D736 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/05/1972 |
| From: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Davis E SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 8003250738 | |
| Download: ML19319D736 (23) | |
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' DISTRIBUTION Chiefs AEC PDR Local PDR E.use, DRS Docket RRMaccary, DRS DRL Reading DFKnuth, DRS PWR-4 Reading RWKlecker, DRL SHHanauer, DR DKartalia, OCC FSchroeder, DRL CO (3)
TRWilson, DRL BBuckley Docket No. 50-312 RSBoyd, DRL NBrown RTedesco, DRL RCDeYoung, DRL DJSkovholt, DRL HRDenton Mr. E. K. Davis, General Counsel Sacramento Municipal Utility District 6201 S Street, P. O. Box 15830 Sacramento, California 95813
Dear Mr. Davis:
We find that we need additional information to complete our review of your application for an operating license for the Rancho Seco Nuclear Generating Station. The specific information required is described in the enclosure and has been categorized into groups which generally correspond to appli-cable section hesdings in the Final Safety Analysis Report.
L We have not completed our review of the subject matters covered in this request for additional information. At a later date we will request
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additional information, if necessary, on these subject matters and others not addressed herein.
To maintain the present Teview schedule, we will need your teply by June 5, 1972. Please inferm us within seven (7) days af ter receipt of this. letter as to the date when you will be able to submit the requested inption to us so that we may revise our schedula, if necessary.
Please contact us if you h we any q'usstions regarding the enelosed requests.
4Simeerely,
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.s THIS DOCUMENT CONTAINS
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, R. C. DeYoung, Assistant Director for Pressurized Water Reactors 3;
Division of Reactor Licensing Enclosures Request for Additional Information
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9 en REQUEST FOR ADDITIONAL INFORMATION SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GE:iERATING STATION DOCKET No. 50-312
1.0 INTRODUCTION
AND
SUMMARY
1.1 With respect to seismic quality assurance, describe the design control measures which have been instituted to assure that adequate seismic input (including any necessary feedback from structural and system dynamic' analyses) is specified to vendors of purchased Class I (seismic) components and equipment.
Identify the responsible design groups or organizations who assure the adequacy and validity of the analyses W
and tests employed by vendors of Class I (seismic) components and equipment. Describe the review procedures utilized by each group or organization.
1.2 Figure 1.1-1 of the FSAR titled " Sacramento Municipal Utility District Service Area Boundary" shows the number of gas wells.
Indicate any oil or gas lines passing through the reactor site.
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. 2.0 SITE q
2.3 Provide an analysis of the fastest mile of wind with recurrence intervals of 50,100,1000 and 10,000 years using Thom's method.
2.4 Provide the joint frequency distributions of wind direction and wind speed by stability class using the AT method. Use the wind direction and speed at the 50-foot level of the tower and the vertical tempera-ture difference between the 200- and 5-foot levels. The AT intervals for each stability class should be ec 2 parable to those currently accepted by the AEC. Seasonal and average annual joint frequency distributions should be provided. If the data recovery rate is reasonably uniform during the entire period of record, use two years of data to establish these distributions. If it is not uniform use the one year period with the best data recovery. We anticipate better than 90% data recovery. Denonstrate that the period of record selected represents normal meteorological conditions.
2.5 Provide a table or graph showing periods of misaing onsite data during the period of record used in establishing the joint frequency dis tribu-tions. Do this for the AT method and the turbulence method (presen:ad in Appendix 2B of the FSAR).
2.6 Describe the site location by specifying the latitude and longitude of the reactor to the nearest second, and the Universal Transverse Mercator coordinates to the nearest 100 meters.
2.7 Figure 2.2-2 of the FSAR titled " Site Location" shows the 2100 foot exclusion radius and also the property line. It is noted that the exclusion area is traversed by a highway. Describe the arrangements made to control traffic in the event of an emergency.
2.8 Provide a drawing which clearly defines the boundary lines on which epecifications limits on release of gaseous effluents are based. This boundary line [which may or may not be the same as the plant property lines or the exclusion area boundary line) demarcates the area, access to which will be actively controlled for purposes of protection of individuals from exposure to radiation and radioactive materials. The degree of access control required is such that the licensee is able to fulfill his obligations with respect to requirements of 10 CPR Part 20,
" Standards for Protection Against Radiation." Distances from plant effluent release points to the boundary line should be clearly shown.
. 2.9 A new map should be provided which shows variations in population on a seasonal basis and, where appropriate, variations in population distribution during the working day should be discussed, particularly where significant shifts and population or population distribution may occur within the low population zone.
2.10 Figures 2.2-7 and 2.2-8 of the FSAR shcw the, current land"used for field crops, pasture and range land, and dairy cattle within a 50 mile radius of the site. The principal food products, acreage and yields should be indicated. The nearest location suitable for dairying should be identified. Sufficient data should be provided regarding food crops in conjunction with estimated releases of radioactivity and gaseous effluents to permit estimates of the range and maximum potential annual radiation doses to individuals and to the population resulting from the principal radionuclides in the discharged effluents.
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3.0 REACTOR 3.1 The FSAR states that the results of vibration tests performed on other reactors of similar design to Rancho Seco will be used in evaluating the adequacy of the design of the internals to withstand the effects of flow-induced vibration.
In this regard, identify the prototype reactor (i.e., the initial reactor of the same design, size, and con-figuration) from which applicable test data will be obtained to determine the design adequacy of Rancho Seco core internals to sustain these effects. Provide a detailed comparison of the applicable design parameters for the Rancho Seco and prototype units that verifies that no significant design or fabrication differences exist between the subject reactors that could materially affect the vibrational response characteristic of the reactor internals. Also, discuss your plans, should the vibration tests of the prototype reactor be delayed or if the results of these tests are incomplete or unsatisfactory.
3.2 Provide a discussion of those preoperational test program elements described in Safety Guide 20, Vibration Measurements on Reactor Internals.
In the event elements of the proposed program differ substantially from Safety Guide 20, provide the basis and justifica-tion for these differences.
- 3. 3 Describe your program for monitoring the reactor coolant system for loose parts during operation.
3.4 There have been several instances reported in other nuclear power plants where deviations in fuel enrichments have occurred. From your discus-sion on page 3.2-16 e of the FSAR, it is not clear whether the quality control checke referred to include 100% gamma scanning of the fuel rods prior to fuel loading.
Since the potential for fuel enrichment devia-tions exists for Rancho Seco, provide the following information:
1 (1) State whether 100% gamma scanning of the fuel will be conducted.
(2) Discuss the threshold of detection of fuel enrichment deviations using existing instrumentation.
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(3) Discuss the effects on the results of the accidents analyzed for this facility if such enrichment deviations exist. State the assumptions and compare the results with those listed in Section 14 of the FSAR.
. 4.0 REACTOR COOLANT SYSTEM 4.11 The list of transients specified in Table 4.1-1 of the FSAR appears to be incomplete.
Identify all design transients and their number of cycles, such as control system or other system malfunctions, component malfunctions, transients resulting frem any single operator error, and inservice hydrostatic tests, etc., which are specified in the ASME Code-required " Design Specifications" for the components of the reactor coolant pressure boundary. Categorize all transients or combination of transients with respect to design Cases I through IV of paragraphs 4.1.2.5.1 and 4.1.2.5.2 of the FSAR (comparable to the operating condi-tion categories identified as "nomal," " upset," " emergency," and
" faulted" as defined in the Summer 1968 Addenda of the ASME Section III Nuclear Vessel Code). Describe the program which will be used to record and maintain an accounting of significant transients occurring during plant operation and to compare the service number of transients accumulated with those specified as the permissible number of transients for which the plant is designed.
4.12 Provide a list of the ASME and ANSI code case interpretations that have been applied :o components within the reactor coolant pressure b e tr.la ry.
4.13 Describe the vibration operational test program required by paragraph 1701.5.4 of the ANSI B31.7 Nuclear Power Piping Code which will be used to verify that the piping and piping restraints within the raactor coolant pressure boundary have been designed to withstand dynamic effects due to valve closures, pump trips, etc. Provide a list of the transient conditions and the associated actions (pump trips, valve actuations, etc.) that will be used in the vibration operational test program to verify the integrity of the system.
Include those transients introduced in systems other than those comnrising the reactor coolant pressure boundary that will result in significant vibration response of reactor coolant preasure boundary systems and components.
4.14 Specify whether the design criteria which have been used to examine the effects of pipe rupture have considered postulated pipe breaks to occur at any location within the reactor coolant pressure boundary, or at limited areas within the system. Provide confirmation that both longitudinal and circumferential type ruptures were evaluated and describe the basis for your design approach.
. 4.15 Provide a more detailed description of tha measures that have been used to assure that the containment liner and all essential equip-ment within the containment, including ec:penents of the primary and secondary coolant systems, engineerad safety features, and equipment supports, have been adequately protected against blowdown jet forces, and pipe whip. The description should include:
(1) Pipe restraint design requirements to prevent plastic hinge formation.
(2) The features provided to shield vital equipment from pipe whip.
(3) The measures taken to physically separate piping and other components of redundant engineered safety features.
(4) A description of the analyses performed to determine that the failure of lines, with diameters of 3/4 inch, will not cause failure of the containment liner under the most adverse design basis accident conditions.
i (5) The analytical methods which were used in (1) and (4) above.
4.16 Provide the design loading combinations, the associated stress or deformation limits, and the design codes or standards applicable to the principal reactor coolant system component supports that have l
been used, (i.e., supports, restraints, " snubbers," guides, etc., as applied to vessels, piping, pumps, and valves).
If the specified limits permit plastic deformation of supports for these components, describe the manner in which the design approach includes the inelas-tic strain compatibility in the supports and supported components (i.e., combined dynamic system analysis for all systems where the stress and strain limits apply).
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4.17 Reported service experiences of PWR steam generators have indicated that flow induced vibration and cavitation effects can cause tube thinning, and corrosion and erosion mechanisms both from primary and secondary side may contribute to further structural degradation of the tube integrity during the service lifetime. The failure of a group of weakened tubes as a consequence of a design basis pipe break in the reactor coolant pressure boundary could impair the capability of emergency core cooling systems to perform their intended function.
In order to evaluate the adequacy of design bases used to prevent such conditions from developing in the steam generator during service, the following additional information is required:
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. (1) State the design conditions and transients which were specified in the design of the steam generator tubes, and the applicable design stress intensity limits associated with Cases I through IV of paragraphs 4.1.2.5.1 and 4.1.2.5.2 of the FSAR (comparable to the " normal," " upset," " emergency," and " faulted" operating condition categories). Justify the basis for this selection.
(2) Specify the margin of tube-wall thinning which could be tolerated without exceeding the allowable stress limits iden-tified in (1) above, under the postulated condition of a design basis largest pipe break in the reactor coolant pressure boundary during reactor operation.
(3) Describe the inservice inspection which will be employed to examine the integrity of steam generator tubes as a means to detect tube-wall thinning beyond acceptable limits and whether excess material will intentionally be provided in the tube wall thickness to accommodate the estimated degradation of tubes during the service lifetime.
4.18 To facilitate our review of the bases for the pressure relieving capacity of the reactor coolant pressure boundary, submit a copy of the summary technical report on overpressure protection which has been prepared in accordance with the requirements of paragraph N910.2 of the ASME Section III Nuclear Vessel Code.
4.19 List the analytical methods and criteria used to evaluate stresses and deformations in all pumps and valves within the reactor coolant pressure boundary including safety and relief valves. For design conditions other than those explicitly addressed by the applicable Code (e.g., design condition categories comparable to Cases I through IV of FSAR Table 4.1-2 for which code limits have not been developed, geometries not included, etc.), provide a summary of each analytical method and the associated acceptance limits. Where empirical rela-tienships and methods determine the design, provide the bases for extrapolating these methods or experience to all loading conditions specified for each component.
4.20 Describe the qualification test program that has been performed to verify that all valves whose operability is relied upon to perform a safety function or shut down the reactor will operate under the transient loadings specified for the design service life.
9-4.21 Describe the design and installation criteria for the mounting of the pressure-relieving devices (safety valves and relief valves) within the reactor coolant pressure boundary. In particular, specify the design criteria which have been used to take into account full discharge loads (i.e, thrust, bending, torsion) imposed on valves and or. connected piping in the event all the valves are required to tscharge. Indicate the provisions made to accommodate these los a.
4.22 For components that are to be constructed in accordance with Section III of the ASME Code, Subsection NB, a summary of the typical analytical results or experimental testing performed to demonstrate compliance with the Code should be provided. A brief description should be submitted of the typical mathematical or test models and the methods of calculation or tests.
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5.0 DESIGN BASES - STRUCTURES AND EQUIPMENT 5.49 For all listed Class I (seismic) structures, systems, and components identify the methods of seismic analysis (=cdal analysis response spectra, modal analysis time history, equivalent static load, etc.)
used for each of the Class I (seismic) items including the reactor core support structure. Include applicable stress or deformation criteria and descriptions (sketchea) of the mathematical models used.
If empirical methods (tests) are used in lieu of analysis, provide the criteria and acceptance basis used to confirm the integrity of the structures, systems, components and equipment. Describe all seismic methods of analyses used.
5.50 Provide the criteria used to lump masses for the seismic system analyses (system mass and compliance to component or bay character-istics and floor mass and compliance to equipment characteristics).
5.51 Provide the criteria which were used to compute shears, moments, stresses, deflections and/or accelerations for each seismic-excited mode as well as for the combined total response, including the criteria for combining closely spaced modal frequencies.
Submit the basis for the methods used to determine the possible combined horizontal and vertical amplified response loadings for the aeismic design of structures, systems and components including the following:
(1) The possible combined horizontal and vertical amplified response loading for the seismic design of the building and floors.
(2) The possible combined horizontal and vertical amplified response loading for the seismic design of equipment and components, including the effect of the seismic response of the building and floors.
(3) The possible combined horizontal and vertical amplified response loading for the seismic design of piping and instrumentation, including the effect of the seismic response of the building, floors, supports, equipment, component, etc.
5.52 Describe the measures taken to consider the effects on floor response spectra (e.g., peak width and period coordinates) of expected varia-tions of structural properties, dampings, soil properties, and soil-structure interactions. Provide the structural material properties l
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and soil-structure interactions which were used in the seismic system analyses including the bases for the selection of these properties.
5.53 The use of both the modal analysis response spectrum and time history provide a check on the response at selected points in the station structure. List the responses obtained from both methods at selected points in the Class I (seismic) structure to provide n basis for checking the seismic system analysis.
5.54 Describe the method employed to consider the torsional modes of vibration in the seismic analysis of the Class I (seismic) building structures.
If static factors are used to account for torsional accelerations in the seismic design of Category I structures, justify this procedure in lieu of a combined vertical, horizontal, and torsional multi-mass system dynamic analysis.
5.55 The methods of Bechtel topical report, B-TOP-4, are referenced in the FS AR as pertaining to the seismic analysis of Class I (seismic) struc-tures. Specify in detail the extent of the applicability of B-TOP-4 to the seismic system design analysis of all Class I (seismic) items listed in the FSAR.
5.56 Damping values of one and two percent for Class I (seismic) piping may not be adequately conservative (p. 5B-7 of the FSAR). Provide I
the bases for the use of these high damping values in lieu of more conservative values of one-half and one percent for seismic design.
j 5.57 Clarify the discussion in paragraph 5B.4.D.2.C.2 (p. 5B-ll of the FSAR) to describe the specific employment of the scaling factors and their bases.
1 5.58 Provide the criteria which were used to determine the number of modes i
considered in the response spectrum modal analyses performed for Class I (seismic) items.
5.59 Provide the dynamic methods and procedures used to determine Class I (seismic) structure overturning moments. Include a description of I
the procedures used to account for soil reactions and vertical earth-quake effects.
5.60 Provide a more detailed description of the simplified seismic analysis methods and procedures used for seismic designs, including the criteria used to account for the predominate input frequencies produced by the response of buildings, supports and components to the earthquake input.
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i 5.61 Describe the procedures used to account for the number of earthquake cycles during one seisn.
event, and specify the number of loading cycles for which Class I (seismic) systems, components and equipment are designed, including the expected duration of the seismic motions or the nu=ber of major motion peaks.
5.62 Describe the analbical procedures applicable to piping that take into account the relative displacements between piping support points, i.e., floors and components, at different elevations within a build-ing and between buildings.
5.63 Provide the criteria employed to account for the torsional effects of valves and cther eccentric masses (e.g., valve operators) in the seismic piping analyses.
5.64 With respect to Class I (seismic) piping buried or otherwise located outside of the containment structure, describe the seismic design criteria employed to assure tha* allowable piping and structural stresses are not exceeded due to differential movement at support points, at containment penetrations, and at entry points into other structures.
5.65 Describe the evaluatf on performed to determine seismic induced effects of Class II piping sys* ems on Class I (seismic) piping.
5.66 Provide the criteria employed to determine the field location of seismic supports and restraints for Class I (seismic) piping, piping system components, and equipment, including placement of snubbers and dampers. Describe the procedures followed to assure that the field location and characteristics of these supports and restraining devices are consistent with the assumptions made in the dynamic analyses of the system.
5.67 Describe the analyses, testing procedures, and seismic restraint measures employed to establish the seismic design adequacy of Class I (seismic) electrical equipment supports such as cable trays, battery racks, instrument racks, control consoles and mechanical components such as pumps and fans. Provide the criteria used to account for the possible amplification of the seismic floor input by the frames, racks, and supports of this equipment. Include the criteria and verification procedure employed to account for the possible amplified design loads (frequenc'/ and amplitude) for vendor-supplied components.
5.68 It is indicated on page 5.5-2 of the FSAR that the cooling towers are not expected to fall on any critical equipment. Provide a further discussion of this postulated incident and evaluate the potential for damage to equipment or structures important to reactor safety in the event one or both of these towers do collapse.
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~ 6.0 ENGINEERED SAFETY FEATURES 6.4 Discuss the design capability to determine and adjust the pH of the post-accident cooling solution in containment.
6.5 Explain the conservatism used in the design and sizing of the reactor building emergency cooling units as discussed in paragraph 6.4.2.3.3 of the FSAR.
6.6 Discuss the net positive suction head (NPSH) requirements for the ECCS and the Reactor Building Spray Pumps, and provide information to show the margin between the required and the available NPSH at extreme operating conditions.
6.7 Describe the inservice inspection program for safety-related fluid systems other than those composing the reactor coolant pressure boundary.
Include items to be inspected, accessibility requirements, and the frequency and types of inspection. The fluid systems to be considered are applicable engineered safety systems, reactor shutdown systems, cooling water systems, and the radioactive waste treatment sys tems.
6.8 Identify any portions of the engineered safety feature piping that will or may be " field run."
Describe the testing to be performed to verify that the design and installation of this piping are in accordance with assumptions made for safety analysis purposes.
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7.0 INSTRUMENTATION AND CONTROL SYSTEMS 7.19 Discuss the seismic instrumentation provided and compare the proposed seismic instrumentation program with that described in AEC Safety Guide
- 12. " Instrumentation for Earthquakes." Submit the basis and justifica-tion for elements of the proposed program that differ substantially from Safety Guide 12.
7.20 Provide a description of the seismic instrumentation, such as peak recording accelerographs and peak deflection recorders, that will be installed in selected Class I (seismic) structures and on selected l
Class I (seismic) components. Include the bases for selection of these structures and components and the bases for location of the instrumenta-tion thereon.
7.21 Describe how the control room operator will obtain the value of the peak acceleration experienced in the basement of the reactor containmcnt structure within a few minutes af ter the earthquake. Include the bases for establishing the predetermined values for activating the readout of the accelerographs.
7.22 Provide the criteria and procedures that will be used in the event of an earthquake to compare measured responses of Class I (seismic) structures with the results of the system dynamic analyses. Include consideration of different underlying soil conditions or unique structural dynamic characteristics that may produce different dynamic responses of Class I (seismic) structures at the site.
7.23 Describe the means for maintaining adequate communications between areas of the station as necessary to cope with emergency conditions.
Provida the criteria for these systems related to assuring their capability to operate under emergency conditions.
-. 9.0 AUXILIARY SYSTEMS 9.1 Describe the equipment and methods used to monitor the boron content of the reactor coolant during power operation, during startup and shutdown, and during refueling operations. Show the location of the equipment on a flow diagram.
9.2 With respect to the failed fuel monitoring system provide the follow-ing information:
9.2.1 The time interval between annunciation of the failed fuel monitor alarm and receipt of confirmatory radiochemical analysis.
9.2.2 The full range capability of the detector in percent failed fuel and the correlation between percent failed fuel and reactor coolant activity.
9.2.3 The indication, alarm, and readout locations associated with this system.
9.2.4 The action, automatic or administrative, taken upon determination that " gross" fuel failure has occurred. Provide the maximum time after the event has occurred in which this action will be initiated.
9.2.5 The action to be taken upon outage of this system.
9.3 Provide radioactivity level or other criteria for determining when the letdown flow should be diverted to the waste disposal system.
State the bases for these criteria.
9.4 Paragraph 9.3.1.3 of the FSAR discusses sampling and analysis of the noncondensibles in the pressurizer. Describe the sampling procedure and the basis for shutdown limits, if any.
9.5 Discuss the significance of the detection of radioactivity within the component cooling water system. The discussion should consider the sensitivity of the radiation monitors, the leakage rate of the radio-active fluid into the component cooling water system necessary to pro-duce the minimum detectable concentration, and the rate of release of
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radioactive material to the environment prior to a protective action to isolate the release path.
Indicate the nature of protective action taken.
9.6 Describe how the failures in the nuclear service raw water system can be recognized, diagnosed and proper action taken by control room operators.
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- 16 9.7 Discuss the basis for the spent fuel pool purification system.
Identify the sources of impurities and the rate at which the impurities enter the fuel pool water.
9.8 State your intent related to the recommendations set forth in Safety Guide 13, Fuel Storage Design Basis. Describe and justify any exceptions to the recommendations therein. Describe means for pro-tecting the stored fuel from objects entering the pool from unusual sources, such as handling equipment failure.
9.9 Provide the basis for single valves in the supply and return lines for the radiation monitoring sacpling assembly as shown in Figure 9.7.1 of the FSAR.
(See Safety Guide 11) 9.10 Paragraph 9.7.2.1.5 of the FSAR indicates that the pressure equalizing system isolation valves do not close on high radiation signal. Provide the basis for this criterion.
9.11 What precautions are taken to preclude freezing of containment fan cooler cooling water during purge during freezing temperatures?
9.12 Describe the alcrm and suppression aspec:: of :he fire procac:1on system. Indicate areas of auto =atic coverage. Consider a fire in one of the on site oil storage tanks.
9.13 Provide the design criteria and bases for sizing the filter bank in the emergency ventilating system provided for the control room and computer room in the auxiliary building.
9.14 In paragraph 9.7.3.1 of the FSAR you state that motorized dampers in the control room ventilation system close automatically upon a high radiation signal.
Provide the criteria for use of the emergency control room ventilation system and state the radiation levels at which it will be implemented.
9.15 Discuss the seismic design of the spent fuel storage facility ventilation system and the bases therefor.
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. 11.0
- RADIATION WASTE AND RADIATION PROTECTION 11.1 Subsection 11.1.1.3 of the FSAR indicates one percent defective fuel cladding to be the basis for the source term.
Expand this section to include all items specified in Section 11.1 of the " Standard Format and Content of the Safety Analysis Reports for Nuclear Power Plants" issued February,1972.
i 11.2' Identify and justify the source term of radioactivity for each licuid radwaste input stream. Specify the concentration by nuclide and flow rate for the input waste streams to the various subsystems of the radioactive liquid waste system.
11.2.1 Specify the concentrations and quantities for both normal operation and for conditions resulting from anticipated operational occurrences.
11.2.2 Identify vents, drains and secondary flow paths for each system.
Identify all bypasses through which liquid waste could circumvent process equipment and be released to the environment.
11.2.3 Provide a process flow diagram indicating those field run lines which contain significant radioactivity, 11.2.4 Cite pertinent previous experience to justify the decontamination factor for each isotope for each piece of equipment.
11.2.5 Describe the performance tests that will be used on a periodic basis to verify the decontamination factors and ather aspects of a given design for each piece of equipment.
11.3 Specify the design objectives of the gaseous waste systems in terms of expected annual activity releases (by nuclide) and exposure to individuals and the population, in light of the requirements of 10 CFR Parts 20 and 50.
Include iodines, noble gases and airborne particulates.
11.3.1 Identify and justify the source term of radioactivity for each gaseous input stream. Specify the concentration by nuclide and flew rate in the gaseous waste input streams in the various subsystems of the radioactive gaseous waste system.
Specify the concentrations for both normal operation and for conditions resulting from anticipated opera-tional occurrences.
11.3.2 Identify vents, drains and secondary flow paths for each system.
Identify all bypasses through which waste could circumvent process equipment and be released to the environment.
~19 11.3.3 Provide a process flow diagram indicating those lines which contain significant radioactivity which are to be field run.
11.3.4 Cite previous experience to justify the decontacination factors for filters used during purging of the gaseous holdup tanks and contain-ment.
11.3.5 Specify the v. y. 2 releases from the gaseous waste systems in curies for each system. The expected releases should per year pet cover normal c '
md anticipated operational occurrences.
11.3.6 Identify sll release points from the gaseous waste systems to the environment on process flow diagrams, on general arrangement drawings, and on a site plot plan.
11.3.7 State and justify all dilution factors which are used in evaluating the release of gaseous radioactive effluents.
11.4 Estimate the following doses that would be received by the general public as a result of releasing the radioactive effluents by the paths and with the dilution factors mentioned above:
The maximum whole body dose to an individual; a.
b.
The maximum organ dose to an individual from halogens and particulates; c.
The whole body dose to the population.
Il'. I State the design objectives of the radiological monitoring systems for normal opsration and anticipated operational occurrences in relation to the requirements of 10 CFR Parts 20 and 50 and AEC General Design Criterion 64.
Distinguish the differences between the design objectives for these situations and those for accident situations.
11.5.1 For each location subject to continuous montoring provide:
(a) the basis for selecting the location; (b) the expected concentrations or radiation levels; (c) the quantity to be measured (e.g., external radiation level, gross concentration, isotopic concentration); (d) the detector i
type, sensitivity and range, considering items (a), (b) and (c) above, j
and, for remote devices, the type and arrangement of the sampler and estimates of sampling line interferences or losses; (e) the type and locations of power sources and recording and indicating devices; (f) setpoints and the bases for their selection; and (g) the type and locations of annunciators and alarms, and the system or operator actions which they initiate.
. 1 11.5.2 For each location subject to periodic sampling, provide:
(a) the basis for selecting the location; (b) expected composition and concentrations; (c) the quantity to be measured (e.g., gross or isotopic concentrations);
(d) sampling frequency and procedures; (e) analytical procedure and sensitivity; and (f) influence of results on plant operations.
i 11.5.3 Describe the arrangements for obtaining independent audits and verifi-cation with respect to calibration and maintenance of the radiological monitoring equipment.
11.6 Specify the design objectives of the solid radwaste system in terms of volumes, forms and activities, and the radiation levels that can be accommodated.
11.6.1 Provide the derivation and justification used to determine the expected volumes of solid wastes, the associated curie content and the principal nuclides that will be shipped from the site.
11.6.2 Provide a detailed flow diagram for the total solid radwaste system.
Identify the solid input streams on a detailed process flew diagram.
The assumed inputa based on volume or weight and isotopic curie content inventories snould be derived and justified. The inventories should be consistent with the source terms presented under Section 11.1 of the " Standard Format and Content of the Safe,ty Analysis Reports for Itueles'r Yower Pl' ants" issued February 1972.
11.6.3 Provide a detailed description of the storage facilities available for packaged solid radwastes including capacity, exact location on a plot plan and general arrangement and details for removal of the solid rad-wastes. State the expected onsite storage period and the decay realized by such storage.
11.6.4 Identify the allowed locations on the site where the shipping containers or vehicles may be stored.
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, ~. 12.0 CONDUCT OF OPERATIONS 12.1 Provide the qualifications, including nuclear and conventional stea=
power plant engineering experience, of peracanal other than plant staff, who will be available to provide engineering support and assistance to the plant staff.
12.2 Referring to the American National Standards " Selection and Training of Nuclear Power Plant Personnel" ANSI N18.1-1971, show that each position category on the plant staff will be filled by a person or persons who meet the minimum qualifications designated in the standard at the time of initial core loading. For any positions to be filled by persons not meeting those qualifications at that time, if any, indicate the source and qualifications of backup support. Resumes of each individual's education, training, and experience background should be provided.
12.3 Provide listings of specific protective action levels and/or indicators (e.g., monitoring instrument levels or alarms) which will be used to give rapid identification of the class of emergency involved, with due allowance to be made for the possibility of a lower clase emergency developing into a higher one.
12.4 Provide the qualifications of EMUD personnel outside of cne plant staff who may be called upon to assist in emergencies in either technical or administrative capacities. Also outline their responsibilities and the means of keeping them informed and up-to-date with respect to these duties.
12.5 Provide evidence, such as copies of letter agreements, that show the existence of clear understsndings between SMUD and each support group, individur.1, or outside agency that may participate in emergencies.
Authorities and responsibilities for specific functions must be clear.
12.6 For a variety of potential accidents within each emergency class, provide listings of specific protective measures to be taken, along with protective action levels, limits, or indicators that will be used to initiate each measura.
12.7 An industrial security plan which is responsive to Safety Guidc 17 and the draft criteria on industrial security should be submitted, taking note of the applicability of 10 CFR 2.790.
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<w 13.0 INITIAL TESTS AND OPERATIONS 13.1 Additional information la requested on your preposed initial test programs. Your attention is directed to 10 CFR 50.34(b)(6)(iii) which refers to two AEC guides relating to Planning for Pre-Operational and Initial Start-up Test Programs. For those portions of these guides which are applicable to the Rancho Seco plant identify any tests or procedures, which either will not be performed or will not be performed in the manner suggested or implied. Provide justification for each such exclusion or variation. Although not =entioned in these guides, information relating to vibration measurements on reactor internals should be provided (see AEC Safety Guide No. 20).
Identify by test name, any additional safety related tests that you plan to perform during the pre-cperational and initial start-up test programs.
i 13.2 Indicate your method of determining that your pre-operational and initial start-up and operations test programs a e complete and adequate to demonstrate that the Rancho Seco plant will be capable of withstand-ing the accidents and transients analyzed in the FSAR and in the design bases of the plant.
13.3 Identify those safety related tests on systecs for which some degree of simulation of anticipated emergency operating circumstances are prudent or desirable for test purposes, and provide justification for the simulation employed.
13.4 Test results at variance with
- hose expected may give rise to the need j
for system modifications or changes in operating or maintenance proce-dures which have already been prepared.
Indicate your administrative procedures for assuring that such test results will be properly reflected in such modifications or changes.
13.5 Throughout initial fuel loading, approach to criticality and the sub-sequsnt start-up test program, the start-up crew mus t include at minimum, one person on every shif t in a superv.isory or consulting-advisory capacity who has had nuclear steam power plant initial start-up experience.
Show how this condition will be met, either by regularly assigned plant staff personnel or vendor personnel, including experience resumes of the persons involved. For those persons, if any, who fulfill this condition and are not regular plant staff personnel, indicate their j
functions, responsibilities, and autaorities.
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