ML20128C089
| ML20128C089 | |
| Person / Time | |
|---|---|
| Site: | Vogtle, 05000426, 05000000, 05000427 |
| Issue date: | 06/12/1973 |
| From: | Kniel K US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Mitchell I GEORGIA POWER CO. |
| Shared Package | |
| ML19292B772 | List:
|
| References | |
| FOIA-84-624 NUDOCS 8505280182 | |
| Download: ML20128C089 (22) | |
Text
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ATOMIC ENERGY COMMISSION
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WASHINGTON. O.C. 20$45 Docket Nos. 50-424 June 12, 1973 N
50-425 50-426 50-427 Georgia Power Company ATTN:
Mr. I. S. Mitchell, III Vice President and Secretary P. O. Box 4545 Atlanta, Georgia 30302 Gentlemen:
In order that we may continue our review of your application for a license to construct the Alvin W. Vogtle Nuclear Plant.
Units 1, 2, 3, and 4; additional information is required.
The specific information required is listed in the enclosure.
It is grouped by sections that correspond to the relevant sections of the Preliminary Safety Analysis Report.
Our tentative review schedule is based on the assumption that this additional information will be available for our review by July 27, 1973.
If you cannot meet this date, please inform us within 7 days after receipt of this letter so that we may revise our scheduling.
Please contact us if you desire any discussion or clarification of the material requested.
Since re ly,
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-tu.4.k, Karl Kniel, Otief Pressurized Water Reactors Branch No. 2 Directorate of Licensing Enclosure s Request for Additional Information ces Listed on page 2 l
85052 182 841015 PDR IA SHOLL 84-624 PDR
o Ceorgia Power Company,
ces:
Southem Services. Inc.
ATTN Mr. Ruble A. Thomas P. O. Box 2625Vice President Birminaham, Alabama 35202 Shaw, Pittman, Potts & TrowbridM e
91017th Street, W ge Washington, D. C.
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- o 2-8 REQUEST POR ADDITIONAL INFORMATION CEORCI A POWER COMPANY ALVIN W. VOCTI.E NUCI. EAR PLAST UNITS 1. 2. 3. AND 4 DOCKET N09 50-424, 50-425, 50-426 AND 50-42 7 2.0 SITE CHARACTERISTICS 2.23 Provide those boring logs of holes drilled in the proposed plant area that were not submitted in the PSAR. For example, the logs for drill holes 101B,1078,146,148,149,150, 151,183, and 184, which are located in the containment buildings locations, were not provided in the PSAR.
2.24 Discuss the significance and magnitude of possible subsidence resulting from fluid withdrawal by means of the proposed water wells that will supply normal make-up water and cooling water during emergency shutdown conditions. Verify your estimate of the magnitude of subsidence by providing the appropriate analyses.
2.25 In order to complete the geologic and tectonic framework for the proposed site, describe and discuss the geology to the north of the Savannah River, using as guidelines the
" Seismic and Geologic Siting Criteria" and the " Standard Format and Content of SARs for Nuclear Power Plants."
Show in an appropriate figure the extent and locations of the nearby Triassic Basin and clastic dikes mentioned on page 2.5-6.
2.26 The Regulatory Postion regarding the site foundation exploration for the Vogtle Nuclear Plant is as follows:
In the Vogtle site exploration program a large number of borings were deployed over a large area of investigation.
Although we and our consultants, the Corps of Engineers, agree that the general geologic conditions at the site are basically as described in the PSAR, we believe that more specific details of the foundation conditions, pertinent to design, may have been overlooked due to the wide spacing of the borings covering the critical structure locations.
o 2-9 Specifically, some solutioning of the calcareous clay bearing stratum cannot be ruled out, because of its heterogeneous inte'rlay'e' ring of sand and f ractured lirestone which could have created solution channels, as evidenced at times by partial or complete loss of drilling fluid in the mar 1.
We and our consultants therefore recommend that, for the containment buildings locations, additional borings be placed on a minimum grid of 50 feet on centers and that they penetrate at least 40 feet of fresh, unweathered marl stratum. O th e r, less heavy Category I structures should have an average boring spacing of not more than 100 feet on centers, depending upon the reliance that can be placed on geologic interpreta-tion between borings. An appropriate number of samples should be recovered from these borings and tested to demon-strate the high bearing capacities represented in Table 2.5-2 of the PSAR, the low compressibility characteristics, and that the upper 15' to 20' of the marl can adequately support the heavy structures.
The applicant has assigned the mechanism causing surface depressions (sinkholes) to the erosion of the shall bed above the earl bearing stratum. The applicant should provide assurance that the control machanism for creating sinkholen is due to this entirely, and not in part to deep seated leaching and consolidation of the soils below the bearing stratum.
We, there fore, recommend that the applicant correlate in detail the geometry, locations, and amount of depression of the sinkholes with the extent and thickness of the shell bed; and provide a detailed discussion of the geomorphology of the area. The additional borings recommended above will provide valuable data in this regard, also.
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REQUEST FOR ADDITIONAL INFORMATION GEORCIA POWER COMPANY ALVIN W. V0GTLE NifCLEAR PLANT UNITS 1. 2, 3. AND 4 DOCXIT NOS. 50-424, 50-425. 50-426 AND 50-427 3.0
[ESIGN OF STRUCTURES. COMPONENTS. EQUIPMENT AND SYSTEMS 3.32 The material in Section 3.6 covering protection against postulated pipe n.pture inside containment is not definitive enough to deter-mine wherber the protection criteria is adequate and how protection t
will actually be" accomplished. The AEC staff position on protection against pipe whip inside containment is contained in Regulatory Guido 1.46.
Provide criteria consistent with Regulatory Guide 1.46.
3.33 Provide the design criteria to be esployed to assure that h'gh energy fluid piping systems outside containment will comply with Ceaeral Design Criterion #4. An acceptable method for compliance would be physical sep.aration or isolation of high energy piping systems from other systene., structures or compenents important to safety.
Indicate how Vogtle Units 1, 2, 3 and 4 will achieve compliance.
3.34 With regard to the techniques to be utilized for determination of required protection against pipe whip inside containment providet 3.34.1 A description of the methods used to calculate the time functions ots (a) the jet thrust force en the ruptured pipe, and (a) the jet impingement force on a distant object.
3.34.2 A suussary of the dynamic analysis methods used to:
- (a) Vetify the design adequacy of pipe whip restraints, and (b) Verify that the m eion of unrestrained, ruptured piping will not damage to an unacceptable degree structures, systems or components which are important to safety.
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3-20 3.35 Clarify your intention to perform a confirmatory type preoperational vibration test for reactor internals on all units of the Vogtle plant using Indian Point 2 as the designated prototype.
Identify any design differences between Vogtle and Indian Point 2 which may lead to different response behavior of the reactor internal structures under flow-induced vibration. In addition, provide a description of the preoperational vibration test program' intended for the Vogtle plant.
If the elements of the intended test program differ substantially from those recommended in AEC Safety Guide 20, submit the basis and justification for these differences.
3.36 Clarify your intention to provide a description of the dynamic system analysis methods and procedures that will be used to confirm the structural design adequacy of the reactor coolant system (unaffected loops) and the reactor internals (including fuel element assemblies, control rod assemblies and drives) under the LOCA loading, or identify the applicable topical report that provides this description.
3.37 The design stress limits for ASME Class 1, 2 and 3 components listed in Section 3.9.2 (Subsections 3.9.2.3, 3.9.2.4 and 3.9.2.9),
the stress limit and load combination criteria of Tables 3.6-8 and 3.6-9 of the PSAR and Sections 3.9 and 5.2 of RESAR-3 are not accept-able. Regulatory Guide 1.48 provides a summary of limits which are acceptable to the regulatory staff. Unless you propose to adopt these design limits, provide the bases for using any limits that exceed those listed in Regulatory Guide 1.48 and demonstrate the adequacy of the design safety margins proposed.
3.38 The position stated in Section 3.9.2.5 of the PSAR regarding the use of Code Cases for ASME Class 2 and 3 components is not acceptable.
It is noted that RESAR-3 provides no indication as to the use of Code Cases for ASME Class I components. The use of ASME Code Cases for all classes of construction requires specific approval by the Conaission in accordance with 10 CFR 50.55a (refer to (a)(2) 11 and footnote 6 of the regulation). Provide a list of the Code Cases
. desired to be used.
3.39 Section 3.7.2.1.1.3 of the PSAR implies that dynamic analysis alone is one of the methods of evaluation to qualify mechanical equipment.
Provide the specific criteria that will be used to guarantee operability of mechanical equipment under faulted condition loads when a dynamic analysis without parformance testing is employed in the design of this equipment.
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3.40 Section 3.9.1.1 of the PSAR presents a partially acceptable basis to. confirm the structural design of the. piping and piping restraints.
Attachment I, entitled."Preoperational Piping Dynamic Effects Test
- Program,'? presents the basis for a program which is complete and which would be acceptable at the Operating Licensing stage. State.
your-intentions to develop such a program for submittal in the iSAR for.the Vogtle Nuclear Plant, Units 1, 2, 3 and 4.
3.41 Provide a detailed description' of an operability assurance program
-for confirming that ASME Class 1, 2 and 3 active
- valves 2 inches and greater in nominal pipe size will function properly under normal, emergency.and faulted plant conditions.** This program may include either the in-situ application of vibratory devices to superimpose the. vibratory loadings 1xt the valve operator and associated mounted.
devices,'or laboratory or shop testing under equivalent simulated loadings'that will ensure valve operability. The test program may be based upon selectively testing a representative number.of active valves in the piping system according to valve type, accident load.
level, size, etc. on a prototype basis.
3.42 Sections 3.9.2.8 of.the PSAR and 5.2.2 of RESAR-3 covering the design of pressure. relieving stations. in seismic Category 1 piping systems are not acceptsble in that design for dynamic effects is not cove red. Your response should include the method of determining the discharge thrust load including all dynamic effects and the method of stress evaluation for open and closed systems. Design should be performed using a standard dynamic hydraulic / structural analysis
'or, alternatively, the equivalence of the method used should be justified and the adequacy of the design safety margins that are proposed should be demonstrated.-
3.43 Provide the basis for the selection of allowable stresses as listed in Table 5.2-6 of RESAR-3 for ASME Class I component supports and
-Table 3.9-4 of RESAR-3 for ASME Class 2 and 3 component supports.
Include information for faulted, emergency and normal / upset opera-ting conditions, particularly for those situations involving supports
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- Active valves are those whose operability is relied upon to perform a safety function such as safe shutdown of the reactor or mitigation of the consequences of a postulated pipe break in the reactor coolant pressure boundary.
- Normal, Emergency and Faulted Plant Conditions relate to the loadings, and environment under which the valve is required to open or close during normal operation, emergency incidents and postulated faults (accidents) which affect the system in which the valve is installed.
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3-22 to active pumps or valves. Indicate whether RESAR-3 applies for the design of Class 2 and 3 component supports for Vogtle since there is apparently no coverage in the PSAR.
3.44 The seismic qualification criteria and the testing performed or to
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be performed for Category I electrical equipment are indicated.in Section 3.10 of the PSAR and the referenced Westinghouse topical report, WCAP-7817, and its associated supplements. Attachment 2, entitled " Electrical and Mechanical Equipment Seismic Qualification Program," presents the-basis for a program which is complete and which would be acceptable at the Operating License stage. Docu-ment-your intention to develop such a program for submittal in the FSAR for the Vogtle Nuclear Plant.
ATTACIDENT 1 PREOPERATIONAL PIPING DYNAMIC EFFECTS TEST PROGRAM Preoperational piping vibrational and dynamic effects testing should be conducted during startup functional testing on piping systems and restraints classified as ASME Class 1 and Class 2 components, The purpose of these tests is to confirm that these components have been designed to withstand the dynamic loadings from.cperational transient conditions that will be encountered during service as required by ASME Code Section III, par.
Nn-1622.3 and NC-3622.*. An acceptable test program to confirm the ade-quacy of the designs should consist of the following:
a.
A listing of the dif ferent flow modes of operation and transients such as pump trips, valve closures, etc. to which the components will be subjected during the test.**
For example, the transients associated with the Reactor Coolant System heatup tests should include, but not necessarily be limited to:
(1) Reactor coolant pump start (2) Reactor coolant pump trip (3) Operation of pressure-relieving valves b.
A list of selected locations in the piping system that will be subjected to visual inspection and measurements (if needed) as performed by the piping designer during these tests. For each of these selected locations, the allowable deflection (peak-to-peak) criteria that will be applied to establish that the stress limita are within the design levels, c.
If vibration is noted beyond the acceptance levels set by the criteria of b. above, corrective restraints should be designed and installed.
If during the test, the piping systems restraints are determined to be inadequate or damaged, corrective restraints should be installed and another test should be performed to determine that the vibrations have been reduced to an acceptable level.
- Ref erence ASME Code Section III, " Nuclear Power Plant Components"
- Additional guidance for the selection of such transients is provided in the AEC Guide for Planning of Initial Startup Programs'.' December 7, 1970
o ATTACHMENT 2 ELECTRICAL AND MECHANICAL EQUIPMENT SEISMIC QUALIFICATION PROGRAM I.. Seismic Test for Equipment Operability 1.
A test program is required to confirm the functional oper-ability of all Seismic Category I electrical and mechanical equipment and instrumentation during and after an carthquake of-magnitude up to and including the SSE.
-2.
The characteristics of the required input motion should be specified by one of.the following:
(a) response spectrum (b) power spectral density function (c) time history Such characteristics, as derived from the structures or systems scismic analysis, should be representative of the' input motion at the equipment mounting locations.
3.
Equipment should be. tested in the operational condition. Oper-ability should be verified during and after the testing.
4.
The actual input motion should be characterized in the same manner as the required input motion, and the conservatism in amplitude and frequency content should be demonstrated.
5.
Scismic excitation generally have a broad frequency content.
Random vibration input motion should be used. However, single frequency input, such as sine beats, may be applicable
.provided one of the following conditions are met:
(a) The characteristics of the required' input motion indicate that the motion is dominated by one frequency (i.e., by structural filtering effects).
(b) The anticipated response of the equipment is adequately represented by one mode.
(c) The' input has suf ficient intensity and duration'to excite all modes to the required magnitude, such that the testing response spectra will envelope the corresponding response spectra of the individual modes.
6.
The input motion should be applied to one vertical and one l
principal (or two orthogonal) horizontal axes simultaneously unless it can be demonstrated that the equipment response along the vertical direction is not sensitive to the vibratory l
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In the case of single frequency input, the time phasing of the inputs in the vertical and horizontal directions must be such that a purely rectilinear resultant input is avoided.
7.
The fixture design should meet the following requirements:
(a) Simulate the actual service mounting (b) Cause no dynamic coupling to the test item.
II. Seismic Design Adequacy of Supports
'I. Analyses or tests should be performed for all supports of electrical and mechanical equipment and instrumentation to ensure their structural capability to withstand seismic excitation.
2.
The analytical results must include the following:
(a) The _ required input motions to the nounted equipment should be obtained and characterized in the manner as stated in Section I.2.
(b) The combined stresses of the support structures should be within the-limits of ASME Section III, Subsection NF -
" Component Support Structures" (draf t version) or other comparable stress limits.
3.
Supports shouldlbe tested with equipment installed.
If the equipment is inoperative during the support test, the response at the equipment mounting locations should be monitored and characterized in the manner as stated in Section I.2.'
In such a case, equipment' should be tested separately and the actual input to the equipment should be more conservative in amplitude and frequency content than the monitored response.
4.
The requirements of Sections I.2, I.4, I.5, I.6 and I.7 are applicable when tests are conducted on the equipment supports.
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o4. 0 REACTOR 4.1 Provide a summary ;of the results of the prototype test. program for the control rod drive mechanisms applicable to Vogtle Units 1, 2, 3 ' and 4, indicating any significant differences in design materials, tolerances and fabrication techniques between prototype and production units and their importance in determining the need to repeat the basic tests with production units. Discuss the production unit test program and acceptance criteria to be applied.
4.2 Provide information on whether results of prototype and operational test programs indicate that the design bases and operational
- requirements of the reactivity control system components specified in Section 4.2.3.1.4 and other areas of Section 4.2.3 of RESAR-3 have been achieved.
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Provide information on whether results of the prototype and operational test programs indicate that the design bases of the fuel assembly structure and supports indicated in Section 4.2.1.1.2.of RESAR-3 have been achieved under faulted conditions.
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5-1 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.1 The material in Section 5.2.1.3 of RESAR-3 on faulted design analysis and procedures provides only general information with little or no detail. Most of the information presented has been extracted from Appendix F of Section III of the ASME Code. If a particular method is desired to be used, complete detailed procedures must be provided.
For example, if you choose to do an inelastic stress analysis, loadings should be developed using inelastic dynamic analysis methods, and details for developing true stress-strain curves must be submitted. The general conditions of instability and potential for unstable crack growth under faulted conditions require more detailed treatment to indicate that you have adequately considered the phenomenon.
Provide the specific procedures to be applied for faulted design analysis for Vogtle Units 1, 2, 3 and 4.
5.2 Identify all Seismic Category I components whose design is based on experimental stress analyses (Appendix II of the ASNE Code,Section III), and provide a summary of the analytical and experimental testing procedures to be used to demonstrate compliance with the code. Submit a brief description cf the mathematical or test models to be used and the methods of calculations or tests.
5.3' Section 5.2.1 of RESAR is referenced for the Design Criteria of the Reactor Coolant Pressure Boundary for Vogtle. Section
. 5.2.1.3.3 of RESAR states that the faulted condition stress limits need not be satisfied if it can be shown from the test of a prototype or model that the specified loads do not exceed 80% of the ultimate load or load combination used in the test.
If this alternative is used in the design of any component or component support, provide the specific test procedure that will be utilized during the testing. The test procedure should demonstrate that the loads imposed on the prototype or model
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duplicate the effects of all design loads and/or temperatures.
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5.4 Provide a brief description of the analytical methods used to determine the stress levels of ASME Class 1 components. This discussion should include computer programs and classical hand methods used, and any unique methods developed.
5.5 Submit a list of computer programs that will be used in dynamic and static analyses to determine mechanical loads and deforma-tions of Seismic Category I components and equipment and the i
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d 5-2 t-analysis co determine stresses' in ASME Code Class 1 components.
. For ~each program, include la brief description of the ' theoretical' 1 basis, the assumptions and references used, and the extent of l
Sits application.
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- Describe -the' design control measures ac required by Appendix B,
'10 CFR Part 50 that will be employed to demonstrate the applic-ability and validity of the above computer programs by any of the following criteria or procedures 1(or other equivalent
' procedure s).
(a) The computer program is a recognized program in the public.
domain. and has had sufficient history of use to justify its applicability and validity without further demonstration.
The dated program version that will be used, the 'sof tware or operating. system, and the computer-hardware configuration must be specified to lue accepted by virtue of its history of use.
(b) The computer program's solutions to a series of test problems with accepted results, have been demonstrated to be substan-tially identical:to those obtained by a similar, independently written program in the public domain. The test problems should be demonstrated to be similar to or with the range of applicability for the problems analyzed by the computer program to justify acceptance of the' program.
.(c) The program's solutions to:a series of test problems are substantially. identical to those obtained by hand -calcula-tions or.from accepted experimental test or analytical results
. published in technical' literature. The test problems'shculd be demonstrated to be similar to the problems analyzed to justify acceptance of the' program.
5.7
. Provide a summary comparison of the results obtained from each
-computer program with either the results. derived-from a similar program in the public domain, on a previously: approved computer program or results' from the test problems. Include typical static-
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and/or dynamic response loading, stress, etc. comparisons, prefer-ably in graphical form.
5.8' In -addition to the specifications already listed in the PSAR for
- principal pressure retaining ferritie f materials and austenitic stain-less steels intended to be used for components that are part of 'the reactor coolant pressure boundary '(RCPB), provide RCPT specifications w--
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5-3 for the weld materials for fabrication and assembly of the components.
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With' respect to ferritic materials (including welds) of the reactor pressure vessel beltline, provide information regarding the specifica-tions for these-materials showing additionally Luposed limits on residual elements (reportable and nonreportable) by specification requirements which are intended to reduce sensitivity to irradiation embrittlement in service. Any additional or special requirements should also be indicated.
5.9 Provide a description of the compatibility of the reactor coolant pressure boundary. materials of construction with insulating materials and with the environmental atmosphere in the event of coolant leakage.
5.10 To allow evaluation of the adequacy of the proposed heatup and cooldown limits for this plant, as dictated by fracture toughness of the ferritic materials, provide the following information:
5.10.1 Identify criteria by which the initial upper shelf fracture energy levels for the materials of the reactor vessel beltline (including welds), as determined by Charpy V-notch tests on specimens oriented in the " weak" direction of the material, will be established.
'5.10.2 It is not clear which revision of ASTM E-185 will be used, because 5.4.3.1 references ASTM E-185, whereas-B.~1.27 references ASTM E-185-70.
State if the program will conform with ASTM E-185-73.
'5.11 For all austenitic stainless steel used for components that are part of:
(1) the reactor coolant pressure boundary, (2) systems required for reactor shutdown, (3) systems required for emergency core cooling, (4) reactor vessel internals required for emergency core cooling, and (5) reactor vessel internals relied on to permit adequate core cooling for any mode of normal operation or under postulated accident conditions, provide the following information (Reference Regulatory Guide 1.44):
5.11.1 Describe the procedures that will be used to ensure that the material is suitably cleaned and protected against contaminants capable of a
5-4 causing stress corrosion cracking throughout the fabrication, shipment,
. storage, construction, testing, and operation of components and systems.
5.11.2 Provide a description of materials, processes,~ inspections, and tests that will be used to _ ensure freedom from the increased susceptibility to intergranular stress corrosion caused by sensitization. This should include the following:
a.
If special processing or fabrication methods are used that subject-the material to temperatures between 800*F and 1500*F, or that involve slow cooling from temperatures.over 1500*F, provide justification that'such treatments will not cause increased susceptibility to intergranular stress corrosion.
b.
Indicate special requirements on chemical analysis for any materials that, during normal operation, will be exposed to water environments containing over 0.10 ppm dissolved oxygen when at temperatures over 200*F.
c.
If the presence of delta ' ferrite is relied on to prevent sensitiza-tion of welds or castings, describe the methods that will be. used to ensure the presence of at least 5% delta ferrite.
5.11.3 Describe the procedures and requirements that will be employed to avoid hot cracking of austenitic stainless steel welds, especially addressing filler metal compositions, welding procedure qualifica-tions, and methods for ensuring adequate delta ferrite content of production welds.
5.12 The inservice inspection program, described in the Technical Specifica-tions 'and PSAR, should incorporate the provisions of the 1972 Winter Addenda for Hydrotest and Class 2 system preservice and inservice inspection, except that the pressure and temperature in paragraph ISC-261(a) of the Winter Addenda should be changed to 275 psig and.
200*F. This change will make Section XI requirements confoon to the -
proposed Regulatory Guide, " Inservice Inspection of ASME, Class 2 and 3 Nuclear Power Plant Components." Consideration should also be given to future Class 3 system examinations as well as preservice and operational pump and valve tests, i
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12-1 12.0 RADIATION PROTECTION 12.1 Describe the operating procedures that will assure that personnel exposures will be kept as low as practicable during plant operation. and maintenance. Include a description of (1) administrative methods such as prejob planning and preparation, use of exposure allocations, use of radiation work permits, area access control methods, specific job training, and job debriefings, and (2) methods of reducing radiation levels such as removal of unnecessary sources of radiation, installation of temporary shielding, and decontamination. The criteria used for implementing various exposure reduction methods should also be presented.
12.2 Identify the maximum and average airborne radioactivity levels for normal operation, including anticipated operational occurrences, that will be allowed in areas within plant structures and within the restricted area where plant personnel, construction workers, or site visitors are permitted.
12.3 Describe the process equipment sources which contribute significantly to plant radiation levels and which constitute the basis for shielding design. In general, any piece of equipment in which radioactive materiala may be filtered, demineralized, concentrated, or stored should be listed.
For each component listed, describe (1) the approximate physical size and shape, (2) the total curie quantity of each of the principal nuclides expected in the component for both normal operacion and expected operational occurrences, and (3) the expected radiation levels at the surface.
Th e information presented should be in tabular form and the listing should be grouped by system function as much as possible.
12.4 Describe design criteria for the erection and dimensions of shield walls, for penetrations through shield walls, and for acceptable radiation levels in the control room, at valve stations, sample stations, and other areas likely to be occupied during normal operational and maintenance activities.
Provide justification of the thickness of shielding to be provided, including the geometric and physical models, and assumptions and data used.
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.17.0 OUALITY ASSURANCE 17.1 The Georgia Power Compar.y (GPC) Quality Assurante (QA) Program for the design and construction of the Vogtle Nuclear Plant (VNP) is not ade-quately described in the PSAR. Provide more detailed descriptions and clarifications as follows:
17.1.1 Section 17.1 is unclear because the individual activities of GPC, Southern Services, Inc. (SSI), and the Bechtel Corporation (BC) cannot be identified and evaluated. Clearly describe GPC's QA. activities in Section 17.1 including all assigned responsibilities and authorities for the major organizations, contractors, and vendors.
17.1.2 Clearly define the QA interfaces and attendant responsibilities within GPC for onsite and offsite organizations. Also, describe the QA interface between GPC and SSI and BC.
17.1.3 Provide organizational charts which clearly show those GPC individuals and groups responsible for implementing the 18 criteria of 10 CFR Part 50, Appendix B, for the design and construction of VNP.
17.1.4 Identify those individuals or groups responsible for establishing and implementing OA related policies, procedures, and instructions.
17.1.5 Identify who reviews and approves the QA Program and Manuals for GPC and for the major contractors.
17.1.6 Identify and describe the authority and independence of those indi-viduals or groups who perform the QA functions of inspection and auditing.
17.1.7 Describe the qualification requirements for the positions responsible for managing and directing the QA Program.
17.1.8 List and describe the duties of the Quality Assurance Committee, Quality Assurance Engineer (QAE) and QA Field Representatives (QAFR).
17.1.9 Identify and provide a brief description of the purpose and scope of the QA Program procedures which pertain to the design, construction, and pre-operational testing of the VNP and which demonstrate compli-ance with all applicable criteria of 10 CFR Part 50, Appendix B.
17.1.10 Provide a listing or cross index table which shows each QA Program procedure and the related criterion of 10 CFR Part 50, Appendix B, which is addressed.
17-2 17.1.11 Describe the' formal indoctrination and/or training program that has b'een or.will be established for all those personnel performing QA related activities.
~ 17.1.12 _ Describe 1the controls which assure that QA policies, manuals, proce-dures, and. instructions, including changesLthereto are received and
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implemented by responsible individuals or groups.
17.l.13 Identify and describe those audits which evaluate the QA program policies, activities, and procedures. Include the frequency, identify the audit personnel, and list distribution of reports.
.17.~ 1.14 Describe those audits performed by GPC' personnel which provide a com-prehensive verification and evaluation of all phases.of the QA Program activities, to _ assure that the QA Program for GPC is effective and meaningful.. State the distribution of reports, frecuency, and identify-audit personnel.
17.2 The QA Program which covers all of the responsibilities deleaated to Southern Services, Inc. for design, procurement, insoection, contractor,-
and vendor activities pertaining to the VNP has not been described'in the PSAR.. We_ require the followirg additional information relevant-to SSI's QA activities.
17.2.1-Identify those major centractors, vendors, consultants, and third parties for which SSI has the responsibility, relative to the design, construction, and preoperational testing of the VNP.
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17.~ 2. 2 Identify and describe the organizational interfaces between SSI and the parties identified in 17.2.1 above.
= 17.2.3 Provide complete organizational charts which show the SSI organiza-
-tions,. contractors, vendors, third party inspectors and consultants
. responsible for establishing and implementing quality related activi-ties. All interfaces must be clearly shown.
17.2.4-Provide clear identification and descriptions of duties showing ade-Lquate authority and independency of those individuals and/or groups responsible for-formulating, establishing, and implementing QA related policies,-procedures, and instructions within the areas of responsi-bility of SSI.
17.2.5 Describe the organizational authority and independence of those individuals or groups responsible for the QA functions of inspection and audit.
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17-3 17.2.6 Describe the qualification requirements.for the position of QA Maaager and list his specific responsibilities.
17.2.7-Identify the responsible individuals who originate, review, and approve SSI's QA Manual,-procedures, and instr'uctions.
Identify the SSI management final approval; level for the OA Manual.
17.2.8 Briefly describe the major aspects of SSI's QA program procedures relative to VNP which demonstrate conoliance to apolicable criteria of 10 CFR Part 50, Appendix B.
17.2.9 Provide a listing or cross index table which relates each SSI QA pro-gram procedure to the applicable criteria of 10 CFR Part 50, Appendix B.
17.2.10 Describe the control systems or activities which assure that SSI's QA program policies, procedures and manuals, including changes or revisions thereto, are distributed to all responsible parties and properly implemented.
17.2.11 - Describe SSI's training program for those personnel who perform QA activities that wil'. assure proper implementation of QA procedures.and requirements.
If a training or indoctrination program does not exist at the present time, state implementation date.
17.2.12 Provide an indepth discussion of those audits performed by SSI manage-ment which provide independent verification and evaluation of the effectiveness of the QA program and procedures relative to compliance with SSI policies and with 10 CFR Part 50, Appendix B.
State the frequency of audits, state the distribution of audit reports and identify audit personnel.
-17.2.13-Describe those audits performed by SSI personnel or SSI's representa-tives~which provide for verification and evaluation of SSI's QA pro-gram, procedures and activities to assure that they are effective and in compliance. Provide audit schedules, identify audit personnel and 1ist audit report distribution.
17.3 The Bechtel Corporation QA Program description is unclear in the areas of organizational responsibility and authority, QA activities, vendor activities, and field activities at VNP. We require the following clarification to complete our evaluation.
17.3.1 Provide an organizational chart which clearly shows the organizational structure, line authority, and responsibility for each BC organization performing QA related activities on the VNP project.
17-4 17.3.2. LIdentify and describe the responsibilities, authority, and independence
- of.those, individuals. or groups responsible for establishing and imple-menting QA policies and procedures relative to those activities associ-ated with'the VNP program.
Specifically, describe and clarify;the dual responsibilities of the position of Manager, Start-up and Quality Assurance. The duties of Start-up and Quality Assurance appear to be in conflict.
'17.3.3
' Describe the authority and independence of those individuals or groups responsible for QA activities in the areas of. design' review, procure-ment review, inspection, audit activities.
c 17.3.4 Describe the qualification requirements for those positions that manage and supervise the QA Program activities.
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BC's QA Manual, procedures, and instructions. Identify the final 17.3.5' Identify the responsible individuals who originate, review, and approve review and approval authority level in BC's management for the QA Manual.
4' 17.3.6
. Provide' a brief description of the important aspects of BC's QA pro-cedures which demonstrate compliance to the applicable criteria of 10 CFR Part 50, Appendix B.
17.3.7 Provide a cross index table or listing of BC's QA procedures, showing
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4 those criteria of 10 CFR Part 50, Appendix B, which. the individual procedures address.
117.3.8 - Describe the controls which assure that BC's QA program policies, activities, procedures, manuals, and changes thereto are distributed to responsible personnel and properly implemented.
~17.3.9-Describe' BC's formal indoctrination and training program for those personnel performing quality related activities to assure proficiency in implementing the QA procedures in review, inspection' and audit activities.
17.3.10 Describe design reviews and design change reviews and state the distri-bution of the related reports.
< 17.3.11 Provide a more indepth description of the calibration and maintenance activities for measuring and testing equipment.. Indicate the criteria for calibration, the frequency, and the procedure.
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17-5 17.3.12 Describe in more detail the process for control of nonconforming mate-rial, parts, or components.
17.3.13 Describe those audits performed by corporate or top level management which provide independent verification and evaluation of the QA pro-gram policies, procedures, and activities to assure they are meaningful and effective and are complying with corporate policy and 10 CFR Part 50, Appendix B.
Include the audit report distribution, identify audit personnel, and state audit frequency.
17.3.14 Identify and describe those audits performed by'BC personnel in the areas of design, procurement, vendor control, and inspection that provide a comprehensive verification and evaluation of the QA program.
procedures and activities to assure a meaningful and effective program.
State the audit schedules, identify the audit personnel, and list distri-bution of audit reports.
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