ML20136E712
| ML20136E712 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Vogtle |
| Issue date: | 01/09/1984 |
| From: | Liaw B Office of Nuclear Reactor Regulation |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082840446 | List:
|
| References | |
| FOIA-84-663 NUDOCS 8401250459 | |
| Download: ML20136E712 (4) | |
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Docket No: 50/424/425 v" w.:.c.
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MEMORANDUM FOR:
M.r Licensing Branch #4 g,' '..-
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SUBJECT:
. GEORGIA POWER COMPANY, V0GTLE ELECTRIC GENERATING PLANTS, UNITS 1 AND 2 REQUEST FOR ADDITIONAL
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Plant Name:
Vogtle Electric Generating Plant, Units 1 and 2 p.-
Supplier: Westinghouse; Bechtel u.
Licensing Stage:
OL Docket Numbers:
50-424/425
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-;ResponsibleBranch.andProject~ Manager:iLB-4,M.cMiller',.
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- Requested Completion Date: Aanuary 1,1984
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. Review Status: Applicant's Response Required.
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The Materials Application Section of the Materials Engineering Branch, i-(/-
. Division of Engineering, has reviewed the available information in the
'FSAR.
In order to provide our input to SER Sections 5.3.1, 5.3.2, and 1
5.3.3, we have prepared the attached request for additional information which must be supplied by the applicant.
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ATTACHMENT REQUEST FOR ADDITIONAL INFORMATION V0GTLE UNITS 1 AND 2 MATERIALS APPLICATION SECTION
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MATERIALS ENGINEERING BRANCH 251.1 Appendices G and H, 10 CFR Part 50 were revised in the Federal Register on May 27, 1983 and became effective on July 26, 1983.
a.
Identify ferritic reactor coolant pressure boundary materials that do not comply with the fracture toughness requirements of Section 50.55a and Appendices G and H of 10 CFR Part 50.
b.
For materials which cannot meet the fracture toughness requirement of Section 50.55a and Appendices G and H of 10 CFR Part 50, provide alternative fracture toughness data and analyses to demonstrate their equivalence to r
the requirements of 10 CFR Part 50.
c.
To demonstrate conformance to Appendices G and H, 10 CFR Part 50:
(1) Provide pressure temperature limit curves for hydrostatic pressure and leak tests, heat-up, cooldown and core operations.
(2) Identify the withdrawal schedule, lead factor, test samples and materials in the Reactor Vessel Materials Surveillance Program.
(3) Indicate the reference temperature, RTNDT' I""
materials in the reactor vessel closure flange region and the beltline regions.
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2-(4) Indicate the chemical composition (copper, nickel and phosphorus), unirradiated upper-shelf energy, and projected end-of-life RT and upper shelf NDT energy for all beltline materials.
RTNDT projec-tions are to be estimated using the "Guthrie Formula" in Commission Report SECY-82-465.
Upper-shelf energy projects are to be estimated using Regulatory Guide 1.99, Rev. 1.
These projects are to be for the end-of-life neutron
. fluence at the 1/4T and ID reactor vessel locations.
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UNITED STATES 8
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Docket Nos.:
50-424 and 50-425 MEMORANDUM FOR:
Elinor G. Adensam, Chief Licensing Branch No. 4 Division of Licensing FROM:
George Lear, Chief Structural and Geotechnical Engineering Branch Division of Engineering
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SUBJECT:
REVIEW QUESTIONS - STRUCTURAL ENGINEERING Plant Name:
Vogtle Electric Generating Plant, Units 1 and 2 Licensing Stage:
OL Docket Number:
50-424/425 Responsible Branch:
Licensing Branch No. 4, M. Miller, LPM We have reviewed Sections 3.3, 3.4.2, 3.5.3, 3.7 and 3.8 of the Vogtle Electric Generating Plant (VEGP), Units 1 and 2 FSAR submitted by Georgia Power Company'in support of their application for an Operating License for VEGP.
On the basis of this review we have identified the additional informa-T tion needed to complete our safety evaluation.
The enclosed questions prepared by S. P. Chan (x29534), Structural Engineering Section A, Structural and Geotechnical Engineering Branch, Division of Engineering, have been prepared for your transmittal to the applicant, dog Georg( Lear, Chief StrucYural and Geotechnical Engineering Branch Division of Engineering
Enclosure:
As stated cc:
J. Knight T. Novak G. Lear L. Heller D. Jeng J. Kane S. Chan i
M. Miller M{&!{
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ENCLOSURE k
V0GTLE ELECTRIC GENERATING PLANT, UNI'TS 1 AND 2 Docket Nos. 50-424/425 REQUEST FOR ADDITIONAL INFORMATION, FSAR REVIEW STRUCTURAL AND GEOTECHNICAL ENGINEERING BRANCH STRUCTURAL ENGINEERING SECTION A 220.0 STRUCTURAL ENGINEERING 220.1 With regard to tornado load combinations identify the SRP 3.3.2.II controlling load combinations used for design of structures FSAR 3.3.2.2 or structural elements.
Provide example of design calcula-tions covering the controlling load combination.
220.2 The FSAR stated that for flood protection construction SRP 3.4.2.II joints are provided with waterstops.
Are the waterstop FSAR 3.4.1 materials properly selected and designed so as to resist possible deterioration due to potential environmental effects such~as time, heat, radiation, and chemicals?
Provide details of the, materials used, their expected service environment, and their, expected resistance to same.
f, 220.3 There are discrepancies in the tornado missile spectrum SRP 3.5.3.II.la between Table 2 of SRP 3.5.3 and Table 3.5.1-5 of the FSAR.
FSAR 3.5.1 In particular, the design missile velocity of automobile and 1.8 is somewhat low in the FSAR.
Explain the descrepancies and demonstrate that these descrepancies would not signifi-cantly affect the outcome of design.
220.4 Are there any openings in the walls or roofs of Category I SRP 3.5.3.II structures which could allow a tornado missile to pass?
FSAR 3.5.3 If so, what protection is provided to protect safety related components or systems which may be located in the way of the missile passage.
220.5 For concrete structural components designed to resist SRP 3.5.3.II impactive or impulsive loads, provide a comparison of the FSAR 3.5.3 design r:riteria you used for allowable ductility ratios and the criteria outlined in Appendix C of ACI 349 as modified by USNRC Regulatory Guide 1.142.
Also provide an explanation for any deviation in criteria which may lead to unconservative results.
220.6 For steel structural components designed to resist'impactive SRP 3.5.3.II or impuslive loads, provide a comparison of the design FSAR 3.5.3 criteria you used for allowable ductility ratios and the criteria outlined in Appendix A of NUREG-0800, SRP Sec-tion 3.5.3.
Also provide an explanation for any deviation in criteria which may lead to unconservative results.
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04/12/84 1
... s 220.7 The current SRP Section 3.7.1 Rev. 1 and Appendix A to SRP 3.7.1 10 CFR 100 require that for seismic analysis of structures, FSAR 3.7.B.1 the design motion is applied at the foundation level of Seismic Category I structures regardless of depth embedment.
The applicant is required to comply with this position and
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provide necessary analyses for all Category I structures including the containment structure.
220.8 Show that the artificial time histories as demonstrated in SRP 3.7.1.II Figures 3.7.B.1-5 and 3.7.B.1-6 will produce response spectra FSAR 3.7.B.1.2 enveloping the corresponding design response spectra of Regulatory Guide 1.60 and meet the SRP requirements for all damping values.
220.9 Damping values of structural systems and subsystems higher SRP 3.7.1.II than those listed in Regulatory Guide 1.61 may be used in a FSAR 3.7.B.1.3 dynamic analysis if documented test data are provided to support them.
These higher damping values should be identified in the FSAR and subject to NRC review and approval.
Specifically, the background information of damping values for cable trays and supports as shown in Figure 3.7.B.1-7 should be provided.
220.10 The current SRP Section 3.7.2 Rev. 1 requires that the SRP 3.7.2 enveloped results of both half-space and finite boundary FSAR 3.7.B.2 methods of modeling should be used for all seismic Cate-O gory I structures, deeply embedded or otherwise.
The j
applicant is required to comply with this position and provide necessary analyses for all Category I structures including the containment structure.
220.11 It is stated in the FSAR that in the confirmatory study, SRP 3.7.2.II the response spectra calculated from the finite element FSAR 3.7.B.2.1 method of soil structure interaction using the VEGP design p*ocedure were compared with those obtained using the impedance (h.tlf-space) method.
Provide additional infor-mation of the analyzed results from comparison of floor response spectra and show that they satisfy the acceptance criteria of Section 3.7.2 of the SRP Rev. 1 (7/81).
220.12 The applicant has identified in FSAR Section 1.8 the SRP 3.7.1.II differences with the Standard Review Plan 3.7.1 (Rev. 1) and 3.7.2.II and 3.7.2 (Rev.1) triat:
FSAR 3.7.B.2.1 1.
For deeply. embedded seismic Category I structures, and 1.8 Vogtle applies the design motion at the grade level instead of the foundation level, and 2.
Vogtle soil-structures interaction analysis uses finite elements methods for deeply embedded structures and half-space methods for shallowly Embedded struc-tures while SRP requires enveloped results of both methods regardless of the depth of embedment.
r 04/13/84 2
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Provide justification and technical bases for these differences and prove that the safety of structures would h
not be compromised.
220.13 Provide significant natural frequencies for major Seismic SRP 3.7.2.II Category I structures as required by the Standard Format FSAR 3.7.8.2.2 and Content of Safety Analysis Reports for Nuclear Power Plants.
220.14 Wh'at is a complex response time-history method? How does SRP 3.7.2.II it apply to soil-strucutre interaction analysir?
If model FSAR 3.7.B.2.2 ing of the soil-structure system involves firite element 3.7.B.2.3 method for soil media and lumped masses for buildings, how are the equations of motion formulated and how are the damping problems resolved?
220.15 Define and describe " transmitting boundary." Describe the SRP 3.7.2.II physical significance of a transmitting boundary.
Provide FSAR 3.7.8.2.4. 1 justificatfon of its use at the VEGP site.
220.16 The following requests of additional information refer to SRP 3.7.2.II Figures 3.7.B.2-3 and 3.7.B.2-4:
FSAR 3.7.B.2.4.1 (a)
Provide horizontal and vertical distances of the soil-structure sy' stem model, and sizes of elements, (b) Provide mass, stiffness and damping information of stick models of structures, T((g (c) Describe the boundary condition at the bottom (El. l')
of the soil-structure model, s ". ; -
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(d)
Provide a summary of calculated motions at all boundaries of the soil media, including horizontal and vertical components of displacements, accelerations and reaction forces at'significant nodal points.
220.17 Provide a tabulation of the actual structural gaps between SRP 3.7.2.II Category I structures along with an adjacent tabular listing FSAR 3.7.B.2.8 of the worst computed gaps between structures.
Discuss the basis for the selected structural gap.
Also demonstrate that adequate physical separations exist between Category I structures, considering the variability and uncertainties associated with parameters used in the analysis.
220.18 Describe in detail the methods used for seismic design and SRP 3.7.3.II analysis of Category I tunnels.
Also, provide a description FSAR 3.7.8.2.4.3 of pertinent design criteria and results of design / analysis,
3.7.B.3.12 used for the buried Category I tunnels.
220.19 Provide details of a seismic instrumentation inservice SRP 3.7.4.11 surveillance program.
The staff's position is outlined in FSAR 3.7.B.4 Nuoc1-0800, SRP Section 3.7.4.11.5.
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04/12/84 3
220.20 A concrete containment design report should be prepared and SRP 3.8.1.11.4.1 made available f~or review during the structural design audit
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to be performed by the staff at a later date.
A suggested format is included in Appendix C to SRP Section 3.8.4, but as long as substantial structural design information is included in its content, some deviation from that format will be acceptable.
220.21 With respect to the allowable stress for tangential shear SRP 3.8.1.II.4 in the concrete, SRP Section 3.8.1.II.5 stated that under FSAR 3.8.1.4.5 no conditions shall the tangential shear carried by the FSAR 3.8.1.5.2 concrete exceed 40 psi and 60 psi for the load combinations representing abnormal / severe environmental and abnormal /
extreme environmental conditions respectively.
The FSAR should address compliance with this position or provide justification for deviat'ons.
220.22 Identify all discrepanciez between BC-TOP-1 and Sub SRP 3.8.1.II.4 article CC-3600 of the ASME Code in steel liner plate and FSAR 3.8.1.4.7 anchorage system design.
Provide justification for these discrepancies.
220.23 Identify all items of materials, quality control and special SRP 3.8.1.II.6 construction techniques that do not comply with, or do not FSAR 3.8.1.6.1-9 meet the requirements of, SRP Section 3.8.1.II.6 and its referenced regulatory guides.
Provide explanation and
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justification for such non-compliances.
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220.24 The long-term surveillance program of the post-tensioning SRP 3.8.1.II.7 system should be modified to agree with the current staff FSAR 3.8.1.7.2 position which has been described in the Proposed Revision 3 to Regulatory Guide 1.35, April 1979.
Liftoff testings are required for both containments at a site.
220.25 The SRP specifies that interior structures of containment SRP 3.8.3.II.2 should be designed in accordance with the requirements of FSAR 3.8.3.2.1 the ACI 349 Code as augmented by Regulatory Guide 1.142.
3.8.3.2.4 The Vogtle interior structures are designed in accordance with the requirements of ACI 318-71 Code including the 1974 supplement.
Identify and justify all deviations of the interior structural. design from the applicable require-ments of the ACI 349 Code as amended by Regulatory Guide 1.142.
220.26 The SRP specifies that Category I structures shall be SRP 3.8.4.II.2 designed in accordance with the requirements of the ACI 349 FSAR 3.8.4.2.1 Code as augmented by Regulatory Guide 1.142.
The Vogtle Category I structures are designed in accordance with the ACI 318-71 Code.
The applicant should identify and justify his Category I structural design from the r'equirements of ACI 349 Code as amended by Regulatory Guide 1.142.
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220.27 Provide design reports for future strucutral design audit SRP 3.8.3.II.4.e work covering SRP Sections 3.8.3, 3.8.4, and 3.8.5.
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SRP 3.8.4.II.4.d suggested format is shown on Appe'ndix C to SRP Section 3.8.4.
SRP 3.8.5.II.5.e As long as the design reports provide sufficient structural design information, some deviation from that format is acceptable.
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NUCLEAR REGULATORY COMMISSION o
5 WASHINGTON, D. C. 20555
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APR 171984 i
l MEMORANDUM FOR: ' Elinor Aden m, Chief -
Licensing ranch No. 4, DL THRU:
James P. Knight, Assistant Director for Components & Structures Engineering, DE Robert E. Jackson, Chief FROM:
0 Geosciences Branch, DE
SUBJECT:
GEOLOGY AND SEISM 0 LOGY REQUEST FOR FURTHER INFORMATION IN THE OL-SER REVIEW 0F THE V0GTLE SITE Attached are questions relating to Sections 2.5.1, 2.5.2, and 2.5.3 of the FSAR for the Vogtle OL-SER review.
It is suggested that a meeting between the GSB reviewers and the applicant be arranged at headquarters after the applicant has had an opportunity to study the questions. The meeting is to clarify the questions and discuss feasibility, time-table and applicant's plans for responding to the questions.
CI The reviewers, Dr. Ina B. Alteman (Geologist) and Dr. phyllis Sobel (Seismologist while Dr. Ibrahim is on travel) may be reached on x27856 and x24416, respectively, for further infomation.
obert E.
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, Chief Geoscie s Br ch Divis n of gineering
Attachment:
As stated cc: w/ attachment S. Brocoum L. Reiter M. Miller A. K. Ibrahim I. B. Alterman P. Sobel R. Jackson J. Knight G. Lear
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r Vogtle Electric Generating Plant Seismology Questions 230.1 Update Table 2.5.2-1 and Figure 2.5.2-1 to include recent (SRP 2.5.2.1) events contained in Bulletins of the Southeastern U.S.
Seismic Network. Wherever possible note the hypocentral depth of the earthquakes. Evaluate seismic data acquired for the review of the nearby Savannah River Plant.
230.2 Provide supporting bases and assess the impact of Charleston (SRP 2.5.2.2) seismicity on the site in light of the new thformation and v s.
data obtained from various funded investigations and the lOV}
various working hypotheses that have emerged from these studies.
For example, it has been suggested that Triassic Basins may be a possible source for generating a Charleston type earthquake (Marine and Siple, 1974, BGSA). Provide an evaluation of this hypothesis and its impact upon ground motion at the VEGP site, which lies in a Triassic Basin.
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V0GTLE GE0 LOGY QUESTIONS s
231.1 Figure 2.5.1-10, a site map, shows an area in a rectangle enclosure with a note referring the reader to Figure 2.5.1-6 for more detail of the area. The figure actually being referred to appears to be Figure 2.5.1-11.
Please check and correct.
231.2 The discussion of the tectonic model for the development of the Southern Appalachians used the model described by Hatcher in 1972.
In the twelve years since that time more geophysical and geologic information has resulted in the development of much newer concepts. The simple subduction model has given way to decollements, exotic terranes, A
j suspect terranes, etc. Please update the section on the tectonic development of the Southern Appalachians.
(NUREG-0800,Section2.5.1,p.2.5.1-4,6andSection 2.5.3.5).
Some recent references are suggested, but your update should not be limited to them.
~
Selected References Cook, F. A. and Oliver,1981, The late Precambrian-early Paleozoic continental edge in the Appalachian Orogen: Amer.
Jour. Sci. v. 281, p. 993-1008.
Williams, H. and Hatcher, R.D., Jr., 1982, Suspect terranes and accretionary history of the Appalachian origin, Geology,
- v. 10, p. 530-536.
9
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231.3-1 Section 2.5.1.1.3.3.1.4 (p. 2.5.1-11, 12), which describes the Hawthorn Formation and its characteristic occurrences of clastic dikes is very much out of date. Since Siple's 1967 paper, several investigations and papers have dealt with details of the clastic dikes.
The more recent information should be obtained, reviewed, and used in responding to the remaining questions on the dikes, which fol. low. A suggested reference list will be found at the end of this set of questions. The staff further suggests the following course of action in obtaining the infomation:
(a) In obtaining a copy of Secor's 1979 report to Allied Chemical on the Chem-Nuclear waste disposal site at
}
Barnwell, the applicant should request the original photographs or copies, as the reproduction process used for the report was unsuccessful in providing useable copies of the photos. The report depends heavily upon the photos in the descriptions of relationships.
~
(b) The applicant should contact P. Talwani of the University of South Carolina about recent discoveries of seismically induced sand blows associated with the Charleston earthquake, for comparison with the dikes in the site region, and those described by Heron (1971).
(NUREG0800,Section2.5.1,1).
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231.3-2 Considering the information in the referenced papers, including (1) the absence of solution cavities at Barnwell (Final Environmental Statement,1976), (2) three cross-cutting sets of dikes with preferred orientations (Secor,1979),(3)theapparentsimilaritybetweenTalwani's seismic sand-blow feature and a clastic dike in Barnwell County drawn by Heron (1971), and (4) repor-ts of seismic liquefaction phemmena up to 100 mi from Charleston during the 1886 earthquake in or near Barnwell County (Seeber and Armbruster, 1981,1983) evaluate the impact of these findings on your interpretation that the dikes resulted from a combination of dessication-and-infilling and collapse into
~
solution basins, and not the result of seismic ground shaking. (NUREG 0800, Section 2.5.1, I).
231.3-3 a.
What is the relative proportions of sand, silt and clay in the three sets of dikes?
b.
What are the ages of the dikes? Can the ages be bracketed or absolute ages be determined by any presently used methods?
(NUREG0800,Section2.5.1,I).
- c. Based on the ezi; sting literature, how far southwest and northeast of Vogtle along the Coastal Plain are the dikes known to occur, i.e. do they continue northward tf North Carolina into Virginia and beyond, and southward through Georgia f
and Alabama.
Are there known limits to their distribution?
q Suggested References Allied-General Nuclear Services,1980, Geological Investigation at the Chem-Nuclear Waste Storage Site, Barnwell, South Carolina-February 24, 1980, Summary report by Dr. Paul G. Mayer; Report by Dr. Donald T.'
Secor, Jr.; Related Correspondence.
Cox, J., and Talwani, P., 1984, Discovery of a paleoliquefaction site near Charleston, South Carolina, Geol. Soc. America Southeastern and North Central Sections Annual Meeting, Abstracts with Programs, p.130.
9 A
V)
Heron, S.D., Judd,' J.B. and Johnston, H.S., Jr.,1971; Clastic dikes
.c associateddith soil horizons in the North and South Carolina Coastal PlaingGeol Soc. America Bull., v. 82, p. 1801-1810.
Inden, R.F. and Zupan, A.J., W.,1975, Normal faulting of Upper Coastal
~.
Plain sediments, Ideal Kaolin Mine, Langley, South Carolina, Geologic Notes, So. Carolina State Development Board. Division of Geology, v.19,
- p. 159-165.
Seeber, L. and Armbruster, J.G.,1981, The 1886 Charleston, South Carolina Earthquake and the Appalachian Detachment, Jour. Geoph. Res.,
v.,86, p. 7874-7891.
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Seeber, L. and Armbruster, J.G.,1983, Large strain effects of the 1886 South Carolina Earthquake, U.S.G.S. Charleston Earthquake Workshop, Proceedings of Conference XX, Charleston, S.C., p. 142-149.
Zupan, A-J. W. and Abbott, W.H., 1975, Clastic dikes: Evidence for post-Eocene tectonics in the upper Coastal Plain of South Carolina; Geologic Notes, So. Carolina State Development Board Division of Geology, v. 19, p. 14-23.
231.4 Section 2.5.1.2.3.1 (P. 2.5.1 - 8) presents a brief summary of the USGS Open File Report 82-156, in which the Millett and Statesboro Faults are postulated, and references the Aj Bechtel Millett Fault Study in response. A summary of the fault study and evidence for the conclusions should be incorporated into the FSAR, in accordance with SRP guidelines that call for evidence that a capable fault does not exist within the site vicinity.
(NUREG0800,Section
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2.5.3,1).
231.5 In Section 2.5.1.2.8.2, the statement is made that there is no evidence that surficial or subsurface materials have been affected by prior earthquake activity.
Discuss in detail the kinds of evidence that would indicate prior earthquake
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activity.
(NUREG0800,Section2.5.1-I).
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231.6 Provide an interpretation of seismic profile A between points 15 and 20, and seismic profile D between points -10 and -5 on Figure 2.5.4-2 (sheets 2 and 3).
(NUREG0800,,
Section 2.5.1-1).
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. UNITED STATES 8
NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 f
April 12,1984 W
s Docket Nos. 50-424 i
and 50-425 MEMORANDUM FOR: Thomas M. Novak, Assistant Director for Licensing Division of Licensing Office of Nuclear Reactor Regulation FROM:
J. Nelson Grace, Director i
Division of Quality Assurance, Safeguards, and Inspection Programs Office of Inspection and Enforcement
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION:
V0GTLE QUALITY ASSURANCE Applicant:
Georgia Power Company (GPC)
Licensing Stage:
-i Responsible Branch:
LB #4 Project Manager:
Melanie Miller
'lOp-i Review Status:
Q-1 complete except for QA-list (See Q 260.61)
The Quality Assurance Branch has reviewed the FSAR section describing the Quality Assurance Program (Section 17) submitted in support of the Operating License application for the Vogtle Electric Generating Plant (VEGP). We
- viewed the FSAR versus Section 17.2 of Revision 2 of the Standard Review lan (NUREG-0800).
Additionalinformation(seeenclosure)shouldberequestedofGPCtosatisfy the staff's current requirements. The last question results from the NRR branches' review of the safety-related items controlled by the QA Program.
If there are any questions regarding this review, please contact Jack Spraul on Ext. 24530.
o We suggest a meeting with GPC after they have drafted a response to the enclosure.
The meeting should include a tour of the plant site.
i' Q, N
- ~
J. Nelson Grace, Director Division of Quality Assurance, Safeguards, and Inspection Programs _
Office of Inspection and Enforcement
Enclosure:
- t.
Request for Additional Information n,
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Q Request for Additional Information Vogtle Electric Generating Plant 260.0 Quality Assurance (QA) 260.1 Provide an organization chart which clearly differentiates between (1A5)*
the onsite and offsite organization elements which function under the QA program controls.
Provide an organization chart of 4
the GPC QA Department. Describe the criteria for determining the size of the QA organization under the:Vogtle QA Manager and the QC organization under the QC Supervisor.
Indicate the approximate number of technical personnel. planned for these QA and QC organizations during normal operations.
260.2 Describe how QA and QC personnel (183) a.
Identify quality problems b.
Initiate, reconnend, or provide solutions c.
Verify implementation of problem solution 260.3 FSAR Section 17.2.1.3.2 indicates the Vogtle QA Manager has (184) stop-work authority. Clarify that this authority is in writing g
and includes the authority to control further processing, delivery, and _ installation of nonconforming material.
Clarify whether QC personnel and personnel reporting to the Vogtle QA Manager have the same or similar authority.
260.4 FSAR Figure 13.1.2-1 shows that the VEGP Quality Control (1081)
Supervisor is a member of the plant staff. Describe GPC's QA overview of the QC function.
For example, QA should review and approve QC procedures, should audit QC activities, should test and certify QC personnel, etc.
260.5 Describe measures which assure that designated QA and QC (1B6) personnel are involved in day-to-day safety-related plant activities (i.e., the QA and QC organizations routinely attend and participate in daily plant work schedule and status meetings to assure that they are kept abreast of day-to-day work assignments throughout the plant and that there is adequate QA/0C coverage relative to procedural and inspection controls, acceptance criteria, and QA/QC staffing and qualifi-cation of personnel to carry out QA/QC assignments).
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260.6 Describe GPC's qualification requirements for its GMQA and (IC2)
Vogtle QA Manager.
(Refer to the NUREG-0800 reference for guidance.)
260.7 Describe measures which assure that (2B1) a.
quality-affecting procedures required to implement the QA program are consistent with QA program commitments and corporate policies and are properly documented, controlled, and made mandatory through a policy state-ment or equivalent document signed by a responsible official.
Identify, by position title, who has responsibility for the policy statement.
b.
GPC's QA organization reviews and documents cencurrence with these quality-related procedures.
260.8 Section 1.9 of the VEGP-FSAR lists numerous exceptions, (2B3) alternatives, and clarifications.go.the NRC QA guidance provided in the Regulatory Guides listed on pages 17.1-26, 17.1-27, and 17.2-6 of the Standard Review Plan (NUREG-0800).
a.
For each of these Regulatory Guides, the VEGP Position should include words to the effect that "GPC commits to implement the Regulatory Position of this Regulatory Guide during the operation phase of VEGP except as noted below."
b.
Each of the notes should quote the specific guidance that is being clarified, that an alternative is being provided for, or that an exception is being taken to.
c.
Each note should then list GPC's position and identify the position as a clarification, an alternative, or an exception.
Each exception should be justified, and clarifications and alternatives should be discussed as appropriate.
d.
For each note, provide GPC's assessment whether GPC's position meets the guidance or standard, exceeds it, or.
fails to meet it.
260.9 Describe measures which assure that, for structures, systems, (2B3) and components covered by the ASME Boiler and Pressure Vessel Code Section III (Classes 1, 2, and 3), the Code QA require-ments will be supplemented by the specific guidance addressed in the regulatory positions of the applicable Regulatory Guides.
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260,10 Describe measures which assure that GPC's QA organization and (2B3) the necessary technical organizations determine and identify the extent QA controls are to be applied to specific safety-related structures, systems, and components during the operations phase of VEGP.
260.11 Identify QA procedures which reflect that Aopendix B to 10 CFR (2B4) 50 will be met during the operations phase af VEGP.
If any of these procedures are not yet issued, indicate the planr.ed date of publication.
260.12 Describe how GPC's Executive Vice President Power Supply regularly (2C1) assesses the scope, status, adequacy, and compliance of the QA program to 10 CFR Part 50, Appendix B.
These measures should include:
a.
Frequent contact with program status through reports, meetings, and/or audits.
b.
Perfor'mance of a preplanned and documented annual assess-ment. Corrective action is identified and tracked.
h 260.13 Describe measures which assure that:
V (20)
Proficiency tests are given to personnel who perform and a.
verify activities affecting quality and that acceptance criteria are developed to determine that these individuals are properly trained and qualified.
b.
Certificates of qualifications clearly delineate (a) the specific functions personnel are qualified to perform
~
and (b) the criteria used to qualify personnel in each function.
Proficiency of personnel who perform and verify activities c.
affecting quality is maintained by retraining, reexamining, and/or recertifying.
260.14 Item F on FSAR page 17.2.3-2 addresses design interface control.
(3D)
Clarify that this control assures that structures, systems, and components are compatible geometrically, functionally, and with processes and the environment.
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260.15 Describe procedures which require that drawings and specifica-tions be reviewed by the QA organization to assure that the (3E2)'
documents are prepared, reviewed, and approved in accordance with company procedures and that the documents contain the necessary QA requirements such as inspection and test require-Li ments, acceptance requirements, and the extent of documenting inspection and test results.
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260.16 Describe guidelines or criteria established by GPC for deter-(3E3) mining the method of design verification (design review, alternate calculations, or tests).
260.17 Regarding design verification, describe measures which assure (3E4) that:
a.
The verifier is qualified and is not directly respon-sible for the design.
b.
Design verification, if other than by qualification testing of a prototype or lead production unit, is completed prior to release for procurement, manufactur-ing, installation, or to another organization for use in other design activities.
In those cases where this timing cannot be met, the design verification may be deferred, providing that the justification for this action is documented and the unverified portion of the design output document and all design output documents, based on the unverified data, are appropriately identi-fied and controlled.
In all cases, verification is 3
complete prior to relying upon the item to perform its 4
function.
c.
Procedures differentiate between design documents that are verified by design review teams and those which can be verified by review by a, single individual.
d.
Design documents subject to QA controls include items such as drawings, specifications, calculations, computer programs, and system descriptions.
e.
The responsibilities of the verifier (s), the> areas and features to be verified, the pertinent considerations to be verified, and the extent of documentation are identified in procedures.
260.18 Describe measures which assure that the following provisions (3ES)
- are included if the verification method is only by test:
Procedures provide criteria that specify when verifica-a.
tion should be by test.
b.
Prototype, component, or feature testing is performed as early as possible prior to installation of plant equipment, or prior to the point.when the installation would become irreversible.
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l 260.19 Describe measures which assure that procedures are established (3E6) to assure that verified computer codes are certified for use and that their use is specified.
260.20 Describe how responsible plant personnel are made aware of (17.2.3.2) design changes / modifications which may affect the performance of their duties.
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260.21 Part B of the second paragraph of FSAR Section 17.2.4 indicates (4A1) that qualified personnel knowledgeable in the QA requirements will review procurement documents.
Clarify that these personnel are organizationally independent of personnel who initiated the procurement documents.
Typically, the Cucuments are originated by engineering, health physics, operating, or maintenance personnel and reviewed by QC and/or QA personnel.
If this review is done by personnel outside the QC and/or QA organization, describe the training and qualifications of these personnel.
260.22 Describe organizational responsibilities (including the involve-(481) ment of GPC's QA organization) for:
(1)procurementplanning; (2) the pre aration, review, approval, and control of procurement f}
documents; 3) supplier selection; (4) bid evaluations; and C
(5) review and concurrence of supplier QA programs prior to initiation of activities affected by the program.
260.23 FSAR Section 17.2.5 states that activities affecting quality (SA) shall be prescribed by and accomplished in accordance with documented instructions, procedures, or drawings. Describe GPC organizational responsibilities for assuring that this is true.
+-
260.24 The last sentence of the first paragraph of FSAR Section (58) 17.2.5.1 begins with "As applicable." Define or delete this expression.
It also states that the documents listed will be reviewed and concurred with by a person knowledgeable in
" quality requirements." Typically, this person should be from
~
- GPC's QA and/or QC organization. The last sentence of the first paragraph of FSAR Section 17.2.5.2 also uses the expression i
" quality requirements" in lieu of " quality assurance require-ments." Clarify.
~
260.25 The first paragraph of FSAR Section 17.2.6 lists documents to be (6A1) controlled per GPC's QA program.
It appears that the following documents should be included in the list. Clarify.
a.
Design documents (e.g., calculations, specifications,
[:-
and analyses) including documents related to computer codes.
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Instructions and procedures for such activities as fabrication and installation.
c.
As-built documents.
d.
e.
Topical reports.
f.
Nonconformance reports.
260.26 Describe measures which assure that someone from GPC's QA (6A2) organization, or an individual other than the person who generated the controlled document but qualified in QA, reviews and concurs with the document with regard to QA-related aspects.
260.27 Item D on FSAR page 17.2.6-2 states that master status lists (682) identify document revisions.
Identify the documents so listed, the periodicity of updating, and the personnel (by organization ti-ties) who receive the lists.
260.28 Describe measures which assure that maintenance, modification, C
(17.2.6.2) and inspection procedures are reviewed by qualified personnel knowledgeable in QA disciplines (normally the QA organization) to determine:
a.
The need for inspection, identification of inspection personnel, and documentation of inspection results.
b.
That the necessary inspection requirements, methods, and acceptance criteria have been identified.
260.29 Describe GPC organizational responsibilities, including inter-(7A1) faces, for the control of purchased material, equipment, and services.
260.30 Clarify that procurement of spare and replacement parts is (7A4) subject to QA program controls in place when the procurement is made and to technical requirements equal to or better than the original technical requirements, or as required to preclude the repetition of defects.
260.31 Item D on FSAR page 17.2.7-2 refers to " Items accepted or l
-(7B2) released...." Change "or" to "and" or justify not doing so.
260.32 Describe measures which assure that the supplier furnishes the (783) following records to the purchaser:
a.
Documentation that identifies the purchased item and the specific procurement requirements (e.g., codes, standards, and specifications) met by the item.
- i b.
Documentation identifying any procurement requirements that have not been met.
c.
A description of those nonconformances from the procurement requirements dispositioned " accept as is" or " repair."
Describe how GPC reviews and accepts these documents.
260.33 Part A of FSAR Section 17.2.8 states that procedures provide for (882)
" verification that items received at VEGP are properly identified and can be traced to the appropriate documentation...." Then an exception is taken for off-the-shelf items.
Clarify that the exception does not negate the quotation above.
l 260.34 Describe measures which assure the verification and documentation (883) of correct item identification prior to release for fabrication:,
assembly, shipment, and installation.
260.35 Provide as complete a listing of special processes as possible (9A1) rather than repeating the 4 examples given in NUREG-0800.
260.36 Since it appears that QA's only involvement with special process 2
O (9A2&981) control is audits, describe the QC organization's responsibili-
, hf ties in this area.
260.37 The second paragraph of FSAR Section 17.2.9 addresses personnel, 1
(9B1) equipment, and procedures in the first sentence. The second sentence addresses qualification of personnel and procedures only and does not address qualification of equipment.
Clarify.
260.38 Describe measures which assure that procedures are established (982) for recording evidence of acceptable accomplishment of. special processes using qualified procedures, equipment, and pcrsonnel.
260.39 Describe measures which assure that procedures provide criteria (10A) for detemining the accuracy requirements of inspection equipment and criteria for determining when inspections are required or define how and when inspections are performed. Also clarify that GPC's quality control specialists and quality control inspectors are within the Quality Control Supervisor's organization.
260.40 The fifth paragraph of FSAR Section 17.2.10 states that the QC (1081) supervisor is responsible for administering and implementing tests and inspections " assigned" to the QC department.
Identify any organization (s) other than the QC department with responsi-bility for inspection. Describe measures taken to assure that the inspection procedures, personnel qualification criteria, i[
and independence from undue pressure such as cost and schedule are reviewed and found acceptable by the QA organization prior to initiation of any inspection activity.
. 260.41 The last sentence in the sixth paragraph of FSAR Section 17.2.10' (1081) states:
" Procedures containing inspection criteria shall be reviewed by qualified personnel to ensure that adequate inspection hold points are included and inspection methods are adequate."
Clarify that the reviewing personnel are from GPC's QA or QC organization. Also discuss the review of procedures which do not contain inspection criteria but which, perhaps, should.
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. 260.42 Inclusion of the words " nondestructive examination" in the (1082) last sentence in the fifth paragraph of FSAR Section 17.2.10 unacceptably limits the comitment. Delete these words or describe in similar detail the qualifications dnd certification of inspectors for inspections other than nondestructive examination.
260.43 Describe measures which assure that procedures provide criteria (10C1 & 11A1) for determining the accuracy requirements of inspection and test equipment and criteria for determining when a test is required or how and when testing activities are performed.
260.44 Describe measures which assure that test prerequisites are met.
(11 Big) h@
260.45 Since it appears that QA's only (involvement with the control
-(12.2) of measuring and test equipment M&TE) is audits, describe the QC organization's responsibilities in this area.
260.46 Identify the organization (s) responsible for the preparation, (12.3) review, and documented concurrence of calibration procedures.
260.47 Item B in FSAR Section 17.2.
states that installed process (12.5) instrumentation at VEGP will not be tagged or labeled with calibration due date. Describe how GPC will assure that such instrumenation is not past due for calibration.
260.48 Item F and the third paragraph of FSAR Section 17.2.12 refer to (12.6)
"an authorized level of management." Clarify what this means.
- Also clarify or delete the expression, "when applicable," at the end of the third paragraph.
260.49 Describe measures which assure that inspection and tests are (12.9) repeated on items determined to be suspect because of questionable calibration status of M&TE.
260.50 Clarify or delete the following expressions from FSAR Section (13)-
17.2.13 a.
as necessary
(-
b.
as required c.
to the extent required by thest special handling instructions
Also identify the GPC organization responsible for the inspections and surveillance referrred to in item A of this FSAR section.
260.51 Clarify that inspection stamps and weld stamps are controlled (14.2) as per the procedures noted in item A on FSAR page 17.2.14-1.
260.52 Describe procedures to control altering th'e sequence of safety-(14.3) related tests, inspections, and other operations such that the alterations are subject to the same controls as the original.
260.53 Describe QA, QC, and other organizational responsibilities for (14.4 & 15.2) the definition and implementation of activities related to nonconformance control.
Identify the organization (s) responsible for identifying an'd documenting nonconforming, inoperative, and malfunctioning) items to prevent inadvertent use.Identify the organization (s with authority and responsibility for the disposition of nonconforming items and for the review of the disposition and closecut.
260.54 Clarify that the nonconformance controls of FSAR Section 17.2.15 (15.1) are applicable to computer codes.
260.55 Describe measures which assure that nonconformances are corrected (15.3) or resolved prior to the initiation of the preoperational test program on the item.
260.56 Identify the GPC organization responsible to independently review (15.5) and analyze nonconformance reports and report quality trend information as stated in FSAR Section 17.2.15 (page 17.2.15-2).
260.57 Clarify whether GPC's QA or QC organization reviews and documents (16) concurrence with corrective action procedures, is involved in the documented concurrence of the adequacy of the corrective action, verifies proper implementation of corrective action, and closes out the corrective action in a timely manner.
260.58 Describe measures which assure that inspection and test records (17.3) include (in addition to the items listed in FSAR Sections 17.2.10 and 17.2.11):
a.
Information related to conditions adverse to quality (inspection records only).
b.
Action taken to resolve any discrepancies noted.
260.59 Clarify if the GPC QC organization at the VEGP is responsible (17.2.18.2) for any auditing and/or surveillance.
If so, describe measures
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which assure that:
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a.
QA reviews and concurs in the schedule and scope of these activities by QC.
b.
Results of these activities by QC are provided to QA for review and assessment.
260.60 Either delete or clarify what is meant by the word (18)
" selectively" when used in front of the word " audit" in the
~
last paragraph of SAR Sections 17.2.8, 17.2.11, and 17.2.13.
Also, since the last paragraph of SAR Sections 17.2.4 through 17.2.17 refers to GPC's audit program, include a similar reference as the last paragraph of SAR Sections 17.2.2 and 17.2.3 or justify not doing so. Describe measures used to audit conformance to SAR commitments.
260.61 Section 17.1.2.2 of the standard format (Regulatory Guide 1.70) requires the identification of safety-related structures, systems, and components controlled by the QA program.
You are requested to supplement and clarify the Vogtle FSAR in accordance with the following (additional clarification may be required when the Mechanical Engineering and Power Systems Branches complete their FSAR review):
(Ps) a.
The following items do not appear on FSAR Table 3.2.2-1.
Add the appropriate items to the table or justify not doing so.
1.
Fuel assemblies 2.
Underground Category 1 piping and conduits 3.
Site drainage system alterations 4.
Roof scuppers (Category I structures) 5.
Dikes around the refueling, reactor make-up, and condensate tanks.
6.
Fuel building radiation monitors 7.
Accident-related meteorological data collection equipment 8.
Radiation protection systems (including necessary equipment and supplies) a)
Radioactive contamination measurement and analysis
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b)
Personnel monitoring (internal, e.g., whole body counter and external, e.g., TLD system).
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Instrument storage, calibration, and maintenance program d)
Decontamination facilities, personrisl, !nd equipment e)
Respiratory protection equipment (including testing) f)
Contaminatior, control 9.
Safety-related masonry walls (see IE Bulletin No. '80-11) 10.
Expendable and consumable items necessary for the functional performance of safety-related structures, systems, and components (i.e., weld rod, fuel oil, boric acid, snubber oil, etc.)
11.
PORVs, block valvesa and their actuators 12.
Control rods O~
b.
Clarify FSAR Table 3.2.2-1 as noted below or justify not doing so.
1.
Sheet 90, item 33 shows the NSCW tower valve house as not being Q-listed. Clarify that this structure shall be subject to the pertinent provision of the Vogtle operational QA program or justify not doing so.
2.
As part of the Control Room HVAC System (Sheet 59),
F-clarify that the hydrogen sulfide, chlorine, and l
radiation monitors for the air intakes are subject to the pertinent provisions of the Yogtle operational QA program or justify not doing so.
3.
Pr6 vide a commitment that'the safety-related l
instrumentation and controls (I&C) described in Sections 7.1 through 7.6 of the-FSAR plus safety-related I&C for safety-related fluid systems will be subject to the pertinent requirements of the FSAR QA program.
This can be done by footnote.to Table 3.2.2-1.
- c. of NUREG-0737, ' Clarification of TMI Action Plan Requirements' (November 1980) identified numerous items that are safety-related and appropriate for OL application and therefore should be on Table 3.2.2-1.
These items are listed below. Add the appropriate items s_
to Table 3.2.2-1 and provide a commitment that the remaining j
items are subject to the pertinent requirements of the i
FSAR operational QA program or justify not doing so.
l
- n.
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- r - l tk-NUREG-0737 Clarification Item 1.
Reactor coolant system vents II.B.1 2.
Plant shielding II.B.2 3.
Valve position indication II.D.3 4.
Auxiliary feedwater system II. E.1.2 -
initiation and flow 5.
Emergency power for II.E.1.2 pressurizer heaters 6.
Dedicated hydrogen penetrations II.E.4.1 7.
Containment isolation II.E.4.2 dependability 8.
Accident monitoring II.F.1
]
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Containment Water Level Monitor b)
Containment Pressure Monitor 9.
Instrumentat' ion for detection II.F.2 of inadequate core cooling l
10.
Power supplies for pressurizer II.G.1 relief valves, block valves, and level indicators
- 11. Automatic PORY isolation II.K.3(1)
- 12. Automatic trip of reactor II.K.3(5) coolant pumps 13.
PID controller II.K.3(9)
- 14. Anticipatory reactor trip II.K.3(12) l on turbine trip L
15.
Power on pump seals II.K.3(25)
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equipment) plans (andrelated 16.
Emergency III.A.1.1/III.A.2
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NUREG-0737 Clarification Item
- 17. Equipment and other items III.A.1.2 associated with the emergency support facilities
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Inplant I2 radiation monitoring III.D.3.3
- 19. Control room habitability III.D.3.4 O
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'APR 121984 MEMORANDUM FOR: 3Elinorl:G 7Ade
,cChief Licensing B ch #4 Division Licensing FROM:
Victo enaroya, Chief Chemical Engineering Branch Division of Engineering
SUBJECT:
FIRE PROTECTION REQUEST FOR INFORMATION FOR V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 & 2 Plant Name:
Vogtle Electric Generating Plant, Units 1 & 2 Docket Nos.:
50-424/425 Milestone No.:
N/A Licensing Branch & Project Manager:
LB #4; M. Miller CMEB Reviewer:
R. Eberly Requested Completion Date:
April 16, 1984 Review Status.* -Q-l's completed Enclosed is our request for the additional information CMEB needs on the fire protection program to complete our review. The primary items of concern
,_ f apply throughout the plant.
We will need the information by July 1,1984 to
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meet the schedule.
t Victor Benaroya, Chie.
Chemical Engineering Branch Division of Engineering
Enclosure:
As stated
Contact:
R. Eberly x24302 cc:
R. Vollmer S. Pawlicki D. Eisenhut T. Sullivan W. Johnston R. Eberly R. Ferguson S. Ebneter, RO:I
- 0. Parr T. Conlon, R0: II T. Wambach C. Norelius, R0: III M. Miller E. Johnson, R0: IV D. Kirsch, R0: V 645443 Xh u
U Ch:mical Engin:ering Branch Fire Protection Section 1
Request for Information Vogtle Electric Generating Plant, Units 1 & 2 Docket Nos. 50-424/425 280.1 Your fire protection program will be reviewed to the guidelines of BTP CMEB 9.5-1 (NUREG-0800), July 1981.
Provide a comparison that shows conformance of the plant fire protection program to these guidelines.
Deviations from the guidelines should be specifically identified.
A technical basis should be provided for each deviation.
Verify that a plant fire brigade, the brigades personnel and 280.2 minimum equipment, and training intervals will be provided in accordance with BTP CMEB 9.5-1, Section C.3.~b.
280.3 Verify that all fire barriers have been tested and approved by an independent laboratory in accordance with BTP CMEB 9.5-1, Section C.5'.a.
280.4 Verify that all openings in rated fire barriers will be sealed to provide a fire resistance rating at least equal to that of the barrier in conformance with BTP CMEB 9.5-1, Section C.S.a.
280.5 Provide a design description of the types of penetration seals used, including materials of construction.
Verify that tests have been conducted to qualify the resistance of the seals in accordance with BTP CMEB 9.5-1, Section C.S.a, including the maximum allowable temperature of 325 F on the unexposed side of the test assembly.
Verify that the seals will be install'ed in accordance with the manufacturer's instructions.
280.6 Verify that door openings in fire barriers will be protected with equivalently rated doors, frames, and hardware.
Verify that a nationally recognized independent testing laboratory has tested and labelled this equipment in accordance with BTP CMEB 9.5-1, Section C.5.a.
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Ve'rify that all fire barrier ductwork penetrations will be sealed 280.7 by fire dampers having a fire resistance rating at least equal to that of the barrier.
Verify that such dampers have been tested and approved by a nationally recognized laboratory in accordance'
~ithSection,c35.aofBTPCMEB9.5-1.
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Verify that fire protection has been provided for safe shutdown 280.8 so that one train of systems necessary to achieve and maintain hot shutdown conditions from either the control room or emergency control station (s) is free of fire damage and that systems necessary to achieve and maintain cold shutdown from either the control room or the emergency control station (s) can be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Provide an analysis which shows that one redundant train of equipment structures, systems, and cables necessary for safe shutdown can be maintained free of fire damage by either:.
Separation of cables and equipment and associated circuits a) of redundant trains by a fire barrier having a 3-hour rating.
Structural steel forming a part of or supporting such fire barriers should be protected to provide fire resistance equivalent to that required of the barriers; Separation of cables and equipment and associated circuits b) of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards.
In addition, fire detectors and an attomatic fire suppression system should be installed in the fire area; or Enclosure of cable and equipment and associated circuits of c) one redundant train in a fire barrier having a 1-hour rating.
In addition, fire detectors and an automatic fire suppression system should be installed in the fire area.
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' f 280.9 Identify those areas of the plant that wil not meet the guidelines
.,(of_ Sect;onC.5.bofBTPCMEB9.5-1and,thusalternativeshutdown
.).# y will be provided.
Additionally, provide a statement that all other f
areas of the plant will be in compliance with Section C.5.b of a
BTP CMEB 9.5-1.
280.10 Verify that the fire pumps, motors, and controllers will be listed by an independent testing laboratory for the service
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intended and that the fire pump installation will be in accordance with NFPA STD 20, referenced in BTP CMEB 9.5-1, Section C.6.b.
280.11 VEGP FSAR Appendix 9B, page 48 states that " Detection systems are located in areas containing safety-related equipment (
It is our position that areas which present a fire exposure to safety-related equipment also be provided with detection systems.
Verify that detection systems are provided for all areas that contain or present a fire exposure to safety-related equipment, to comply with BTP CMEB 9.5-1, Section C.6.a(1).
Verify that, as a minimum, fire detection systems installed in 280.12 the plant comply with the requirements of Class A systems as l'
defined in NFPA 72D and Class 1 circuits as defined in NFPA 70, to comply with the provisions of BTP CMEB 9.5-1, Section C.6.a(2).
Describe those instances in which the design of installed fire detection systems will not meet these minimum criteria and provide the basis for such deviations.
l 280.13 Verify that, as a minimun, automatic sprinkler systems installed intheplantcomhwiththerequirementsofNFPA13andNFPA15, to comply with the provisions of BTP CMEB 9.5-1, Section C.6.c(3).
Describe those instances in which the design of_ the installed systems will not meet these minimum criteria and prc' tide the basis for such deviations.
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. 280.i4 Ve'rify that, as a minimum, interior standpipe and hose systems installed in the plant comply with the requirements of NFPA 14, to comply with the provisions of BTP CMEB 9.5-1, Section C.6.c(4).
Describe those instances in which the design of these systems will not meet these minimum criteria and provide the basis for such deviations.
In particular, describe the dry standpipe system designed to operate after an SSE from a Category 1 water source.
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280.15 Describe how fire protection features of the control room complex compy with BTP CMEB 9.5-1, Section C.7.b with respect to the f'ollowing:
Location and operation of automatic smoke dampers in a.
ventilation system openings between the control room and peripheral rooms Smoke detectors in control room cabinets and consoles
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V Carpeting in the control room c.
d.
Protection of peripheral rooms Cables in the ceiling
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280.16 VEGP Appendix 98, page 61 indicates that the seismic analysis on the reactor coolant pump oil collection system has not yet been completed.
This analysis is required in order to comply with the provisions of BTP CMEB 9.5-1, Section C.7'.a(1)(e).
Indicate the date by which this analysis will be provided.
Describe how hydrogen concentrations in the battery room will 280.17 be maintained below 2 volume % in conformance with BTP CMEB 9.5-1, Section C.7.g.
Verify that the loss of ventilation is alarmed C
in the control room.
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280.18 Verify that self-contained 8-hour minimum capacity, battery powered emergency lighting units are installed in all areas needed for remote shutdown, and in access and egress routed thereto in conformance with BTP CMEB 9.5-1, Section C.S.g.
280.19 Verify that hydrogen lines in safety-related areas are protected in accordance with our guidelines in Section C.S.d of BTP CMEB 9.5-1.
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APR.I 6 1984 kElinor ensam, Chief, Licensing Branch #4 MEMORANDUM FOR:
Div on of Licensing FROM:
Brian W. Sheron, Chief. Reactor Systems Branch Division of Systems Integration
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - V0GTLE ELECTRIC GENERATING STATION Plant Name:
Vogtle Electric Generating Station, Units 1 and 2 Docket No.:
50-424/425 Licensing Status:
OL Responsible Branch: Licensing Branch #4 e
Project Manager:
M. Miller Review Status:
Request for Additional Information Enclosed with this' letter is a set of questions concerning the Vogtle plant. These questions are a result of a review of those sections of 6.3 of the FSAR for which Reactor Systems Branch has primary review responsibility. RSB is continuing its review and will submit additional p) questions as the evaluation proceeds through the other areas for which i -kr~
we are responsible.
O-Brian W. Sheron, hief Reactor Systems Branch Division of Systems Integration
Enclosure:
As stated cc:
R. W. Houston M. Miller CONTACT: M. Wigdor, x27592 f"'
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REQUEST FOR ADDITIONAL INFORMAIT0N GEORGIA POWER CORPORATION V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 424/425 440.49(6.3)
Because of freezing weather conditions, blocking of the vent line on the RWST has occurred on at least one operating plant.
Describe the features you have incorporated into the design that preclude this condition from occurring in the Vogtle plant or otherwise discuss how your ECCS performance analysis accounts for the possibility of this condition occurring. (6.3.2.2.9) 440.50(6.3)
Recent plant experience has identified a potential problem regarding the long-term reliability of some pumps used for long-term core cooling following a LOCA.
For all pumps that are required to operate to provide long-term core cooling, describe how you established the period of time the pumps must remain operational following a LOCA, and provide justification that the pumps are capable of operating for this required period of time.
This justification could be based on previous testing or on previous operational experiences of identical pumps.
Differences between expected post-LOCA conditions and the conditions during previous testing or operational experience cited should be justified (e.g., water temperature, debris, water chemistry). (6.3.4)
7 2
Ls 440.51. (6.3)
So that we may evaluate the dependence of the ECCS equipment on the plant auxiliaries, provide, or reference in the FSAR the following:
(1) A list of all of the primary auxiliary systems required to directly support each ECCS component.
(2) A brief description of the support function performed by the primary auxiliary systems. This should include the ECCS components that are supported and the associated trains.
(3) The method of initiating the primary auxiliaries to provide support to the ECCS.
(4) The additional secondary auxiliaries required to directly
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support the primary auxiliary specified in (1).
(5) A brief description of this supporting function performed by the secondary auxiliary.
(6) The method of initiating this secondary auxiliary.
(7) For those primary and secondary auxiliary systems required
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to directly support each ECCS component, discuss the i
classification you assign to the system (i.e., is it a
safety-related system or component and is it designed to safety-related standards?) and your rationale for this assignment.
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Also, discuss the potential for damage to ECCS equipment as a result of an auxiliary system transient such as overpres-surization or overheating. (6.3.2) 440,52(6.3)
Table 6.3.2-2 lists the capacity of the accumulator relief valves as 1500 SCFM. Verify that this capacity is adequate to relieve all possible RCS backleakage and _that it is adequate to prevent accumulator overpressurization during level adjustments, assuming equipment malfunction or operator error while adding water to the accumulators. Show the relief valve fluid flow t
rate and temperature assumed in this calculation. (6.3.2.2.14) 440.53(6.3)
Westinghouse has indicated a potential problem associated with the volume control tank level instrumentation and level control system.. In some designs a potential single failure could cause loss of suction and subsequent damage to all safety injection pumps.
Provide a discussion of this potential problem for the Vogtle plant, and what design modifications ycu have made to your system to prevent this single failure situation.
What is the safety classification of the volume control tank level control system? In the event of a failure of this system, describe the alarms and procedures that direct the operator to
assure an adequate water supply is maintained to the charging pump.
(6.3.2) 440.54(6.3)
During our reviews of license applications we have identified concerns related to the containment sump design and its effects on long term cooling following a Loss of Coolant Accident (LOCA).
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These concerns are related to (1) creation of debris which could potentially block the sump screens and flow passages in the ECCS andthecore,(2)inadequateNPSHofthepumpstakingsuction CL lD2;j from the containment sump, (3) air entrainment from streams of water or steam which can cause loss of adequate NPSH, (4) formation of vortices which can cause loss of adequate NPSH, air entrainment and suction of floating debris into the ECCS and (5) inadequate emergency procedures and operator training to enable
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a correct response to these problems.
Preoperational recircu-lation tests performed by utilities have consistently identified the need for plant modifications.
We require the following actions to provide additional assurance that long term cooling of the reactor core can be achieved and maintained following a postulated LOCA.
f 1.
Establish a procedure to perform an inspection of the containment, and the containment sump area in particular, to identify any materials which have the potential for
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t, becoming debris capable of blocking the containment sump when required for recirculation of coolant water. Typi-cally, these raate'ials consist of:
plastic bags, step-off r
pads, health physics instrumentation, welding equipment,
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f scaffolding, metal chips and screws, portable inspection lights, unsecured wood, construction materials and tools as well as other miscellaneous loose equipment.
This inspection should be performed at the end of each shutdown as soon as practical before containment isolation.
1 2.
Institute an inspection program according to the require-ments of Regulatory Guide 1.82, item 14. This item ad-dresses inspection of the containment sump components including screens and intake structures.
e 3.
Discuss possible actions for the operator to take for both a vortex problem (with consequent pump cavitation) and sump blockage due to debris. These should address all likely scenarios and should list all instrumentation available to
.the operator (and its location) to aid in detecting prob-lems which may arise, indications the operator should look for, and operator actions to mitigate these problems.
4.
Pipe breaks, drain flow and channeling of spray flow released below or impinging on the containment water surface in the area of the sump can cause a variety of A
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Describe any changes you plan to make to reduce vortical flow in the neighborhood of the sump.
Ideally, flow should approach uniformly from all directions.
5.
Evaluate the extent to which the containment sump (s) in your plant meet the requirements for each of the items previously identified; namely, debris, inadequate NPSH, air entrainment, vortex formation, and operator actions.
O The following additional guidance is provided for perform-ing this evaluation.
1)
Refer to the recommendations in Regulatory Guide 1.82 (Section C) which may be of assistance in performing this evaluation.
2)
Provide a drawing showing the location of the drain sump relative to the containment sumps.
3)
Provide the following information with your evaluation of debris:
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(a) Compare the size of opening in the fine screens with the. minimum dimensions in the pumps which take suction from the sump, the minimum dimension
.t in any spray nozzles and in the fuel assemblies
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in the reactor core or any other line in the recirculation flow path whose size is comparable to or smaller than the sump screan' mesh size in order to show that no flow block' ge will occur at a
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any point past the screen.
.s (b-) Estimate the extent to which debris could block
--[j the trash rack or screens (50 percent limit).
If a blockage problem is identified, describe the corrective actions you plan to take (replace s
insulation, enlarge cages, etc.).
(c)
For each type ri fr, a al insulation used in the containment, provide the following information:
(i) type of material including composition and density, (ii) manufacturer and brand name, (iii) method of attachment, (iv) location and quantity in containment of
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each type, (v) an estimate of the tendency of each type to form particles small enough to pass
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8 through the fine screen in the suction lines.
(d) Estimate what the effect of these insulation particles would be on the operability and perfor-mance of all pumps used for recirculation cool-ing. Address effects on pump seals and bearings.
440.55(6.3)
Provide a discussion of procedures and administrative controls for manually resetting SIS following a LOCA. Specifically address the minimum time after actuation that the SI signal can i
be reset, and procedures to be followed if a reset were to be followed by a loss of offsite power. (6.3.2) 440.56(6.3)
Certain automatic safety injection signals and certain safety system components, such as accumulators, charging pumps and/or SI pumps, are blocked to preclude unwanted actuation of these systems during normal shutdown and startup operations. Describe the alarms available to alert the operator to a failure in the primary to secondary system for which these blocked systems would be required to mitigate the effects of the failure during this phase of operation and the time frame available for the operator to take the necessary actions to mitigate-the consequences of such an accident.
If applicable, provide or reference sensitivity studies to demonstrate that these cases are bounded by existing analysis.
(6.3.2)
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440.57(6.3)
Provide a discussion on excessive boron concentration in the reactor vessel and hot leg recirculation flushing related to long term cooling following a LOCA. During the hot leg recircu-lation, what will be the minimum expected flow rate in the hot leg and what is the required rate to match boil-off? (6.3.5.4) 440.58(6.3)
Discuss the design provis' ions for prevention of post-LOCA vortex formation in the containment sump. Discuss any anti-vortex criteria which was utilized during the sump design.
(6.3.2.2.9)
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440.59 (6.3)
The staff will require verification that no vortexing tendencies exist in the containment sump during recirculation phase of a LOCA. Discuss the full scale preoperational tests which will show that under prototypical post LOCA conditions, no adverse flow conditions will occur which could degrade ECCS pump performance.
In lieu of full scale in-plant tests, a scale model sump test may be acceptable to the staff.
If you chose to conduct a scale model test, provide details of the test program.
Include information of the model size, scaling principles utilized, comparison of model parameters to expected post LOCA conditions, and a discussion on how all possible flow conditions and screen blockage will be considered in the model tests. Due to scaling problems, the staff will require that model tests show that considerable margin is available in respect to
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vortexing tendencies.
Rotational flow patterns and surface dimples, which might be acceptable in full scale tests, may not be acceptable in a model program.
(6.3.2.2.9) 440.60(6.3)
What is the minimum elapsed time following initiation of LOCA before the operator must initiate switchover from injection to recirculation.
What is the minimum time available to the operator to complete the switchover from injection to recirculation following a LOCA considering the most limiting single failure? How much D
time is required for the automatic switchover actions? Indicate the time required to complete each manual action identified in Table 6.3.2-7 of the FSAR. Also indicate any other duties the operator would be responsible for at this point in the postulated scenario.
(6.3.2.2.9.2) 440.61(6.3)
Describe the instrumentation available for monitoring ECCS performance during post-LOCA operation (injection mode and recirculationmode).
Include a description of the instrument location, power supply, and ranging as well as environmental qualification and safety characterization.
(6.3.5).
10.62(6.3)
Describe the means provided for ECCS pump protection including monitoring of overcurrent, overspeed, overtemperature and high vibration conditions.
(6.3.5)
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440.63(6.3)
The Standard Review Plan (NUREG-0800) indicates that ECC testing should include delivery of coolant to the vessel during shut-downs for refueling.
Provide or reference a discussion of proposed ECC testing during refueling.
(6.3.4) 440.64 (6.3)
Certain operator actions are required for the various modes of operation of the ECCS to mitigate the consequences of certain events (i.e., steam line break, small LOCA, large LOCA).
For each of these modes of operation, list, along with the required operator actions, the alarms / indications available that would lead the operator to take the appropriate actions.
Discuss the lP-time interval assumed in the FSAR analyses between the time the 1
V operator is alerted to a condition by these alarms / indications and the time that the operator is assumed to perform the action.
(6.3.2.8)
What would be the consequences of perfonning the required manual action in an incorrect order or accidently omitting one of the sequential actions,?
440.65(6.3)
A minimum flow bypass is provided on each safety injection (SI) pump discharge to recirculate flow to the refueling water storagetank(RWST)intheeventthepumpsarestartedwiththe normal injection flow paths unavailable. Nonnal injection paths
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could be unavailable for the situation of inadvertent actuation
12 of safety injection while the RCS is at normal operating pres-sure or in the event of a small LOCA during the period when RCS pressure remains above the shutoff head of the pumps.
The minimum flow bypass line for each pump contains a single motor-operated valve. Downstream of these motor-operated valves the minimum flow bypass lines join and are connected to a single line which terminates in the RWST.
In this single line is a single motor-operated valve (8813).
If valve 8813 should close while SI pumps are running with the normal injection flow paths unavailable, both SI pumps could be damaged as a result.
Demonstrate that no pump damage will occur as a consequence of the closure of this valve or modify the design of the minimum flow bypass lines. Any proposed design must ensure that (1) no single failure can result in the loss of degradation of both SI
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pumps and (2) no single failure results in not being able to isolate the RWST during the recirculation phase following the postulated LOCA. Please note that if you rely on the operator to observe the failure and take corrective action, you must provide.
1.
A description of the alanns and their setpoints that will alert the operator to the failure.
13 2.
The minimum amount of time after receipt of the alarm that the operator has to correct the situation.
3.
A description of the corrective actions that would need to be taken.
Note that if the failure occurred during SBLOCA, an acceptable action would not be to stop the HPI pumps, but rather to manually open the valve.
In this case we need to know valve j
accessibility and the amount of time needed to dispatch an operator to the valve and take the necessary corrective action.
We will also need justification why pump damage will not occur during this time internal.
(6.3.2.2.9.2) 440.66(6.3)
Please describe the function of PSV-8852, a relief valve in the Boron Recirculation System.
Compare the pressure at which this valve lifts and the design pressure of the piping up to the isolation valves with the head of the charging pumps. (6.3.2) 440.67(6.3)
Section 6.3.3.1 discusses the failure of a single steam dump.
In your design, could a single failure in the steam dump control circuitry cause more than one steam dump to fail open or
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inadvertently open? (6.3.3.1)
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14 440.68(5.4.7)
Section 5.4.7.2.4 of FSAR states that the RHRS suction side reliefs have a set pressure of 450 psig.
It also states
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that the RHRS is not isolated from the RCS until a pressurizer bubble is fonned and prior to increasing RCS pressure to 600 psig and that the isolation' valves receive an automatic close signal at 750 psig.
Provide an explanation as to how the RHRS can be kept in service above 450 psig.
440.69(6.3)
The FSAR states that there are manual valves "which could Ct.
through mispositioning, potentially degrade ECCS performance".
Please list these valves and describe the effect their misposi-tioning would have on the ability to cool the core. Describe the administrative controls, surveillance frequencies and posi-tion indication used to ensure and verify proper valve position.
(6.3.2.2.17) 440.70(6.3)
Section 6.3.2.5 states that the most ECCS components can be tested on line. Which components are not testable on-line and for which power is not locked out? (6.3.2.5) 440.71(6.3)
Please explain what is meant by "alann for group monitoring of component" as is stated in Table 6.3.2-5.(6.3.2.2)
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15 440.72(6.3)
What precautions will be taken during the recirculation mode to protect both trains of ECCS pumps when a passive failure in a common suction line is considered? An example is a gross failure of a seal for valve HV880dB which is on the suction line to both SI pumps from RHR train B heat exchanger. How much time is available to the operator to isolate the failure and how much time will it take to detect the failure? (6.3.2.2) 440.73(6.3)
Section 6.3.5.1 lists the temperature indication for the ECCS,
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however the Refueling Water Storage, Tank is not included.
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.G Please provide a discussion of the temperature indication for the RWST and its associated piping located outside of the auxiliary building or any other indication that would indicate freezingofthewater.(6.3.5.1)
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440.74(6.3)
Identify any length of ECCS piping which have normally closed valves and do not have pressure relief in the piping section betweenthevalves.(6.3.2.2.14) 440.75(6.3)
How does the water temperature of the refueling water storage tank assumed in your ECCS performance analysis compare with the maximum expected temperature? Do you propose any tech spec l(
limit on the maximum refueling water storage tank temperature?
If not, what assurances do you provide that the maximum temperature will not exceed that assumed in your ECCS analysis?
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
h WASHINGTON, D. C. 20555 L
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MAR 3 01984 Docket No:
50-424/425 MEMORANDUM FOR: Elinor Adensam, Chief Licensing Branch #4 Division of Licensing FROM:
L. G. Hulman, Chief Accident Evaluation Branch Division of Systems Integration
SUBJECT:
ACCIDENT EVALUATION BRANCH QUESTIONS ON THE FINAL SAFETY ANALYSIS REPORT FOR THE V0GTLE ELECTRIC GENERATING STATION Enclosed are the Accident Evaluation Branch questions on the Vogtle docket. These questions are based upon our review of the radiological consequence evaluations of Chapter 15, the proposed fission product control systems (Sections 6.5.1 and 6.5.2), and the control room habitabilityenvelope(Section6.4).
The AEB lead reviewer for this project is F. Akstulewicz (X24993) and
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any questions regarding this input should be directed to him.
.[:::
L. G."Hulman, Chief Accident Evaluation Branch Division of Systems Integration
Enclosure:
As stated cc:
R. Mattson D. Muller T. Novak B. Sheron T. Marsh M. Miller
- 0. Parr M. Wigdor
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FSAR REVIEW QUESTIONS FOR V0GTLE UNITS.1 & 2 450.1 List the areas in the zone serviced by the control room emer-(6.4) gency ventilation system and show ventilation patterns within the control room emergency zones and areas adjacent to the control room emergency zone.
450.2 Identify the source of unlimited offsite bottled air replen-(6.4) ishment capability for the control room envelope and the estimated delivery time.
450.3 Provide the assumptions & bases used to calculate the control (6.4) room X/Q values, including separation distance from release point (s) to the control room air intake, building width or diameter, source type (diffusion or point source) and building projected area.
450.4 Radiation release point 3, shown on Figure 6.4.2-2 of the FSAR, (6.4) appears to be less than 100 feet from the control room air intakes.
Provide an evaluation of control room operator doses for those design basis accidents that result in radiation release from release point 3.
450.5 For the evaluation of radiation doses to control room personnel (6.4) following design basis accidents, provide the following information for FSAR Section 15A.3.1:
1.
Recirculation rate; 2.
Control room air intake rate; 3.
Exhaust rate; 4.
Control room unfiltered inleakage rate; and 5.
Filter efficiencies for recirculation and intake flow rate.
450.6 In FSAR Section 6.5.2.3, you state that transfer from the spray (6.5.2) injection mode to the spray recirculation mode will be made manually. This design is not in accordance with Acceptance Criteria c.1.a of SRP Section 6.5.2, Revision 1 (NUREG-0800) which states that the spray system should be designed to transfer automatically from the injection mode to the recirculation mode.
Provide a revised spray system design which meets all acceptance criteria of SRP Section 6.5.2, or provide justification of the design which utilize operator actions by using the guidance of ANSI /N660, " Time Response Design Criteria for Safety-Related Operator Actions".
450.7 In Section 15.4.8 of the Vogtle Final Safety Analysis Report, App. A) you provided an analysis of the rod ejection accidentwhich assumes a c (15.4.8 products between containment leakage and secondary side releases. While the staff recognizes in SRP 15.4.8, Appendix A,
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that the radiological consequences could occur as a result of the two release pathways, the SRP section indicates that the two release pathways be evaluated independently (see Review
5 Procedure 3) to bound the potential offsite consequences.
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Therefore, provide the following information:
- 1) Provide an analysis of the offsite radiological consequences for the case where all the fission product activity is released to the containment.
The containment should be assumed to leak at its design leak rate. Additional assumptions should be consistent with Regulatory Guide 1.77, Appendix B.
- 2) Provide an analysis of the offsite radiological consequences for the case that the rod housing is not punctured and all the fission products released from the damaged fuel are retained in the primary coolant. This case should assume that activity in the primary system is leaked to the steam generators at.the technical specification leakrate.
Primary to secondary leakage should be assumed until equalization of pressure between the primary and secondary systems can be expected.
For both types of analyses, the cases analyzed should be the most severe from the standpoint of fission product releases to the environment.
450.8 To verify that break flow to and releases from the affected m
(15.6.3)steamgeneratorcanorcannotbeterminatedwithin30 minutes of accident initiation, provide an analysis of the design basis SGTR which uses the Westinghouse emergency operator guidelines for operator actions to mitigate SGTR events as appropriate for your plant. This analysis should follow the guidance of ANSI /N660, " Time Response Design Criteria For Safety-Related Operator Actions" for each operator action. This includes the assumption of the first operator action at 5 minutes following reactor scram and one minute between successive operator actions (manipulation). The analysis should assume the loss-of-offsite power at the time of reactor scram and should be carried out to the time at which no additional releases from the affected steam generator are required. The following time-dependent parameters should be provided in graphical form:
1.
the primary system pressure and reactor coolant temperature (hot leg);
2.
the primary system mass; 3.
the tube rupture flow rate and integrated break ~ flow; 4.
the secondary liquid mass in the faulted and non-faulted steam generators; 5.
the steam generator pressure for the faulted and
non-faulted steam generators; k
the integrated mass release through the safety / relief 6.
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valves for the faulted and non-faulted steam generators; 7.
the integrated mass release through the atmospheric dump valves for the faulted or non-faulted steam generators; and 8.
the pressurizer water volume.
450.9 Provide or reference a discussion to confirm that ECCS (15.6.5 components located exterior to the reactor containment are housed in a structure which, in the event of leakage from the App. B). ECCS, permit venting of releases through iodine filters designed in accordance with Regulatory Guide 1.52.
Identify the f.SF ventilation system used for this purpose and for maintaining the ECCS pump rooms less than 0.25 inches W. G.
Vacuum with respect to the environment.
450.10 Relative to your analysis of a fuel handling accident inside (15.7.4) fuel building, supply the following information:
- 1. radiation detection time;
- 2. isolation damper closure time;
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- 3. time radioactive material takes to travel from a detector
(,;)
to the isolation damper; and
- 4. whether automatic or manual action is needed to switch the fuel handling building ventilation system to the emergency mode and exhaust any potential release through the ESF filters.
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e UNITED STATES 8
NUCLEAR REGULATORY COMMISSION s
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April 13,1984 a
i cket Nos. 50-424/425 MEMORANDUM FOR: Elinor Adensam, Chief Licensing Branch No. 4 Division of Licensing FROM:
Dennis L. Ziemann, Chief Procedures and Systems Review Branch Division of Human Factors Safety
SUBJECT:
V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION - OPERATING AND MAINTENANCE PROCEDURES We have reviewed Sections 13.5.2 and 15.8 of the Vogtle Units 1 and 2 FSAR.
Please transmit the enclosed request for additional information to the applicant. This review was based on the criteria in NUREG-0800, " Standard Review Plan," Sections 13.5.2 and 15.8.
A Procedures Generation Package (PGP) is to be submitted in accordance with Supplement I to NUREG-0737 "Requirenents for Emergency Response Capability"
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(Generic Letter No. 82-33). We will conduct this review prior to the applicant beginning formal operator training on the Vogtle emergency operating procedures. The applicant estimates that the PGP will be submitted about 28 months before fuel loading.
Sam Bryan (X29852), Principal Operational Safety Engineer, performed this review and developed the request for additional information.
Dennis L. Ziema
, Chief Procedures and Systems Review Branch Division of Human Factors Safety
Enclosure:
Request for Additional Information cc w/ enclosure:
M. Miller 6
~n u IOS Ol u
- s, REQUEST FOR ADDITIONAL INFORMATION i
OPERATING AND MAINTENANCE PROCEDURES V0GTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 640.01 In the FSAR discussion of E0Ps, immediate action steps (13.5.2-2) contained in E0Ps are identified as having to be memorized by operators.
If other plant procedures contain immediate actions required to be memorized by operators, identify any such procedures.
640.02 Provide a discussion of the methods Georgia Power Company (13.5.2-1) will use to issue and control temporary procedures.
640.03 Identify by title each procedure classified as Plant / Unit (13.5.2-1)
Operating Procedures, System Operating Procedures, and Abnormal Operating Procedures.
640.04 Provide the general format for any of the classes of (13.5.2-1) procedures in 640.03 if different from that identified in Section 13.5.2.1 Operating Procedures of the FSAR.
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JUN 2 01984 Docket Nos. 50-424 and 425 MEMORANDUM FOR:
Elinor Adensam, Chief Licensing Branch No. 2 Division of Licensing, NRR FROM:
J. Nelson Grace, Director Division of Quality Assurance, Safeguards, and Inspection Programs Office of Inspection and Enforcement
SUBJECT:
RAI REGARDING V0GTLE QA LIST We have enclosed an additional RAI for Georgia Power Company regarding what items will be controlled by its QA program during the operations phase of the Vogtle Electric Generating Plant. The RAI reflects the review results of the Mechanical Engineering and Power System Branches of NRR. RAI 260.61 reflects the review results of the other review branches.
It was-transmitted to the applicant by your letter of April 30, 1984.
My memorandum to Darrell Eisenhut of April 12, 1984, identified Vogtle as one "j
of the plants scheduled to start more than 30 days from the date of Board Notification 84-011. This Board Notification concerns the "important to safety" versus " safety-related" issue.
To our knowledge, this issue has not been resolved for Vogtle, and we suggest the following request for additional information also be issued to Georgia Power Company.
4 On January 5, 1984, Generic Letter 84-01 was issued to all holders of operating licenses, applicants for o
operating licenses, and holders of construction permits for power reactors regarding NRC use of the terms "important to safety" and " safety-related." Please confirm that you understand your responsibility for developing and implementing quality assurance programs as described in Generic Letter 84-01.
Any questions concerning this review should be dircated to Jack Spraul on X24530.
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J. Nelson Grace, Director Division of Quality Assurance.
Safeguards, and Inspection Programs Office of Inspection and Enforcement
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Enclosure:
Vogtle RAI 260.62 w~ m ;
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Vogtle Request for Additional Information 260.0 Quality Assurance Branch 260.62 The request for additional information 260.61 indicates that additional clarification may be required when the Mechanical Engi-neering and Power Systems Branches complete their review of FSAR Table 3.2-1.
This review is now complete, resulting in the following:
a.
The following items do not appear on Table 3.2-1.
Add the appropriate items to the table and commit to apply the pertinent QA program requirements to the remaining items during the operations phase or justify not doing so.
1.
Control rods 2.
Control rod drive mechanisms.
3.
Fabricated supports such as Unistrut or Superstrut that are used to' support systems'and components identified in Regulatory Guide 1.29.
4.
Items that are within the scope of Regulatory Position C.2 and C.3 of Regulatory Guide 1.29.
5.
Core support structure.
6.
Containment building polar bridge crane components that perform a safety function.
7.
Steam generator steam flow restrictors.
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8.
Reactor coolant pump seals.
b.
The following items are on Table 3.2-1, but they show "Q-List, No; Safety-Related, No."
Connit to apply the pertinent QA program requirements to these items during the operations phase or justify not doing so.
1.
Shell side of letdown heat exchanger of CVCS (item 12, sheet 7).
2.
Shell side of excess letdown heat exchanger of CVCS (item 13 sheet 7).
3.
Shellsideofsealwaterhe.atexchangerofCVCS(item 14, sheet 8).
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2 os.
4.
Auxiliary component cooling water surge tank (item 1, sheet j
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16).
5.
Auxiliary component cooling water pumps (item 2, sheet 16).
6.
Refuelingmachine(item 5, sheet 91).
7.
Fuel transfer system (item 19, sheet 92).
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JUN 18 BB4 MEMORANDUM FOR ME Licen[gBranch#4,DL FROM:
C. H. Berlinger, Chief Core Performance Branch, DSI
SUBJECT:
CPB QUESTIONS ON V0GTLE 1 & 2 Enclosed are the questions on Section 4.2 of the Vogtle 1 & 2 FSAR.
These were prepared by PNL on contract BP,544 under CPB direction. Please forward these questions to the applicant.
r C. H. Berlinger,' Chief
} g Core Performance Branch, DSI
Enclosure:
As stated cc:
L. Rubenstein M. Miller
Contact:
M. Dunenfeld, CPB:DSI X-28097 O-I
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ENCLOSURE I
1 CORE PERFORMANCE BRANCH QUESTION ON V0GTLE 1 AND 2 490.1 The reference in Section 4.2.1.3 for fuel rod models does not appear to be correct.
Please confirm that it should be Reference 6 rather than Reference 5.
490.2 Verify that the methods and criteria used for the fuel to be used in the Vogtle reactor are those described in WCAP-9400.
I If this is not so, then provide this infonnation as delineated in Section 4.2 of the SRP for the fuel design basis and design evaluation.
490.3 (a) Provide the results of the cladding creep ~ collapse analysis p
for the fuel to be used in Vogtle.
(b) Verify that the fuel rod internal pressure criterion that the pellet-clad gap not increase when the internal pressure is greater than RCS pressure is satisfied.
490.4 The use of the CVCS letdown monitor for detecting fuel rod failures has been explained in the Vogtle FSAR.
Is there a definite commitment and plan for the active use of this system to monitor fuel failures, as per SRP Section 4.2?
490.5 Does the analysis of the fuel handling accident (Section 15.7.4 of the FSAR) take into account that the peak pellet burnup of approximately 50,000 mwd / tonne of uraniun shown on p. 4.2-2 exceeds the value (i.e., approximately 45,000 mwd /t) stated in Footnote a, 3 of Table 15.7.4-2 (Sheet 13 of 13)?
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