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Nov;mber 14, 1934 MEMORANDUM FOR:
T. Novak, Assistant Director for Licensing, DL FROM:
R. Wayne Houston, Assistant Director for Reactor Safety, DSI
SUBJECT:
INPUT FOR DRAFT SAFETY EVALUATION REPORT - V0GTLE, UNITS 1 AND 2 Plant Name: Vcgtle Electric Generating Plant Units 1 and 2
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Docket No.: 50-424/425 Licensing Stage: OL Responsibe Branch: LB #4 Project Manager:
M. Miller Review Status:
Incomplete The enclosed draft Safety Evaluation Report (SER) has been prepared by the Containment Systems Branch (CSB) after having reviewed the applicable portions of the FSAR. The bases used in the review are contained in SRP Section 6.2.1, 6.2.2, 6.2.4, 6.2.5, and 6.2.6. provides a sumary of the outstanding issues. is a brief SALP report.
OstginalSignedBy R.TUayne Hecston R. Wayne Houston, Assistant Director for Reactor Safety, DSI
Enclosures:
As stated cc:
R. Bernero D. Eisenhut E. Adensam M. Miller i
CONTACT:
C. Li, CSB: DSI x29484 DISTRIBUTION:
Docket File CSB R/F CLi JShapaker WButler AD/RS/RF W
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j DRAFT SAFETY EVALUATION REPORT 1
V0GTLE. UNITS 1 AND 2 DOCKET NO. 50-424/425 1;
6.2 Containment Srstems The Vogtle containment systems include the containment structures, containment heat removal systems, the containment isolation system, and the containment n
hydrogen control system.
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p The staff has reviewed the applicant's design, design bases, and safety analyses for the containment and the containment systems provided in the FSAR.
c The acceptance criteria used as the basis for our evaluation are contained in Sections 6.2.1, " Containment Functional Design," 6.2.2, " Containment Heat Removal Systems," 6.2.4, " Containment Isolation System," 6.2.5, " Combustible e
Gas Control In Containment," and 6.2.6, " Containment Leakage Testing," of the Standard Review Plan (SRP), NUREG-0800, dated iluly 1981. These acceptance criteria include.the applicable general design criteria (Appendix A of 10 CFR Part 50), regulatory guides, branch technical positions, and industry codes and standards as specified in the above cited sections of the SRP. The results i
of the staff review are discussed below.-
6.2.1 Containment Functional Design The containment and associated systems all function to minimize the release of~
radioactive fission products which might be released to the environment following postulated design basis accidents (DBAs).
6.2.1.1 Containment Structure The Vogtle containment is a steel-lined, reinforced, prestressed concrete cylinder with a net free volume of 2,750,000 cubic feet. The containment
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designed for an internal pressure of 52 psig.
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6.2.1.2 Containment Analysis fj 6.2.1.2.1 Maximum Containment Pressure and Temperature Analysis
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i The applicant has perfomed containment analyses on a spectrum of postulated l
reactor coolant system and secondary system pipe ruptures to verify the contain-I ment functional design pressure and temperature and to establish the pressure and temperature conditions for envirormental qualification of safety-related equipment located ins.ide containment. The containment functional analyses included the assumpt' ion of the most limiting single active failure and the 1
availability or unavailability of offsite power, depending on which resulted
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in the highest containment temperatures and pressures.
Loss of Coolant Accidents The applicant's spectrum of breaks in the reactor coolant system (i.e., loss-of-coolant accidents) included a double-ended guillotine break in the hot leg,
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I double-ended guillotine breaks in the cold leg at the reactor coolant pump suction and discharge, and a 0.6 double-ended break and a 3-square-foot split
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break in the pump suction line.
For the double-ended guillotine pump suction break,bothminimumandmaximumemergencycorecoolingsystem(ECCS) flow 9
cases were considered. The design basis LOCA for peak containment pressure was detemined to be the double-ended guillotine break in the hot leg. The design basis LOCA for the long-term containment pressure transient was deter-e mined to be the double-ended pump suction guillotine break with minimum ECCS flow. The analyses for both of these design basis LOCAs assumed the loss of l
offsite power and, for the limiting single active failure, assumed the concurrent failure of one train of the containment spray system and two of the
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four containment f,an coolers (i.e., one diesel generator failure).
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The metthodology, used by the applicant to compute the mass and energy release. rates
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for postulated Loss-of-Coolant Accidents (LOCA) for the containment functional j
analysis is documented in the Wertinghouse Topical Report, WCAP-8312A. The staff 4
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has previously reviewed and accepted this methodology, as discussed in the
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Safety Evaluation Report transmitted to Westinghouse by a letter dated i'
March 12, 1975.
- i The applicant performed all of the containment functional analyses using the Westinghouse C0CO computer code.
Initial conditions and input data, including passive and active heat removal system parameters, were conservatively chosen j
to produce the highest containment pressures and temperatures.
The applicant's analysis of the design basis LOCA for peak containment pressure gave a maximum i
pressure of 38.9 psig versus the containment design pressure of 52 psig.
The analysis of the design basis LOCA for the long-tenn containment pressure l
transient demonstrated, as required, that the containment pressure is reduced to less than 50 perce' t of the peak calculated pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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The staff has performed a confinnatory analysis using the computer code CONTEMPT-LT/28.
Initial conditions and input data, including passive and active heat sink parameters, were similarly chosen to produce the highest containment pressure. The highest calculated containment pressure was 35.7 psig which occurred for the design basis LOCA identified above.
9 We are, however, unable to confirm the acceptability of the long-term containment pressure transient. A more detailed discussion.of long term containment heat removal capability is needed to resolve this matter. This is a confirmatory item.
Secondary System Breaks The spectrum of secondary system breaks analyzed by the applicant included various sizes of double-ended and split breaks of the main steam line at five different power levels from 0 percent to 102 percent. Main feedwater line breaks (MFWLBs) were not included since the pipe break mass flow for the MFWLB is limited by the steam generator internals design and the break effluent is of a lower specific, enthalpy, and thus, main feedwater line breaks are not as
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. conservatively assumed (1) the availability of offsite power to maximize
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energy transfer from the primary to the secondary system, and (2) reduced containment heat removal system effectiveness (one of two spray trains and j
one of two fan cooler trains) due to the loss of offsite power and a diesel-generator failure. The applicant found the design basis pipe break for i
containment peak pressure to be a 0.14 ft2 split break at hot shutdown; the peak calculated containment pressure was 41.9 psig. The design basis pipe 9
j break for containment peak temperature was a 0.80 ft2 split break at 102 l
percent power; the peak calculated containment temperature was 352*F.
The applicant calculated the mass and energy release rate data for the postu-lated MSLB accidents psing the methodology described in Westinghouse Topical 3
Report WCAp-8860. Th'is topical report was found acceptable under the conditions delineated in the staff's evaluation, which was transmitted to Westinghouse by letter dated August 22, 1983 The staff evaluation of the topical report j
states that the mass and energy release data should be revised to reflect the I
model changes made by Westinghouse during the review. The mass and energy release data in FSAR Tables 6.2.1-61 through 6.2.1-63 do not reflect the
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changes in the methodology.
k Since the peak calculated pressure and temperature for the containment analyses are based on MSLB accidents, the MSLB analysis t i should be redone using mass and energy release data that reflects the topical report methodology approved by the staff.
The concerns of IE Bulletin 80-04, regarding the potential for main steam line breaks with continued feedwater addition, have been addressed. The main feedwater lines are equipped with redundant automatic isolation valves which provide acceptable assurance that main feedwater flow will be tenninated.
y Auxiliary feedwater flow was assumed to continue at maximum runout flow for 30 minutes until manually terminated by operator action. We find the MSLB L
analysis has adequately taken into account the design and perfonnance of the main feedwater and auxiliary feedwater systems relative to the concerns of IE Bulletin 80-04.
The staff has reviewed the spectrum of reactor coolant system and secondary system pipe breaks analyzed by the applicant and the applicant's choice of initial conditions, input parameters, and assumptions, and except for the following, finds the applicant's analyses acceptable. The applicant assumed
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Tectrical Specification limit (17.7 psia) during nonnal operation.
The applicant has proposed adding the difference (2.7 psig) to the peak calculatedpressure(41.9psig)todeterminethetestpressure(Pa)for g
containment leakage testing. However, since the MSLB analysis will have to be
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redone, for reasons stated above, the correct initial pressure can be used and the appropriate value of Pa recalculated. This will remain an open item pending,,
the receipt of the applicant's analysis.
c 6.2.1.2.2 Protection Against Damage from External Pressure To demonstrate the adequacy of the containment external (differential) pressure capability (3.0 psi),Y the applicant calculated the internal pressure that would result assuming inadvertent actuation of the containment spray system.
The applicant calculated a maximum external pressure load of 1.0 psi across the containment shell. However, the applicant reports that operation of the containment normal purge system may result in an external pressure load of 1.5 psi across the containment shell. An analysis of the likelihood and consequences of this event is not currently in the FSAR; the applicant has comitted to provide this analysis.
t We have reviewed the initial condition and assumptions used in the analysis of inadvertent spray actuation and find the analysis acceptable..Misoperation of the containment purge system is reported by the applicant to produce a 50%
greater external load. This is a confinnatory item pending the receipt of the applicant's analysis. Nevertheless, the containment design external pressure of l
3.0 asi is capable of accomodating with sufficient margin the maximum calculated external load.
6.2.1.3 Minimum Containment Pressure Analysis for Energency Core Cooling System Perfonnance Capability Studies Appendix K to 10 CFR 50 requires that the containment pressure used for evaluating cooling effectiveness :during reactor core reflood shall not exceed a pressure calculated conservatively for this prpose. The calculation must include the effect of operation of all installed containment pressure reducing y........-....-
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systems and processes. The corresponding reflood rate in the core will then be reduced because lessened containment pressure reduces the resistance to steam flow in the reactor coolant loops and increases the boiloff rate from t
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the core.
.j The app 1 Mat has performed the containment backpressure calculation using the 1
methods and assumptions described in Appendix A of WCAP-8339, " Westinghouse
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Emergency Core Cooling System Evaluation Model-Su' mary". A break spectrum analysis was perfomed that considered various break sizes, break locations, and Moody discharge coefficients for the double-ended cold leg guillotine.
Mass and energy release rates for this break were calculated using the methods described in Section 15.6.5 of the FSAR, and are evaluated separately in Section 6.3.5 of this'SER.
1 The mass and energy release data during the blowdown phase are missing from Tables 6.2.1-69 and 6.21-70 of the FSAR.
The applicant should provide the appropriate blowdown data.
The staff has reviewed the applicant's input parameters used in the minimum containment pressure analysis, including initial containment conditions, contain-ment net free volume, passive heat sinks, and containment active heat removal
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capability, and it has found them to be acceptably conservative and in conformance with BTP CSB 6-1, "Minic14 Containment Pressure Model for PWR ECCS Performance Evaluation," with one exception. This exception is the nonconservative assumption of an initial containment pressure of 14.7 psia which is greater than the minimum allowable contabment pressure under normal operating conditions p
(13.2 psia). The applicant should redo the analysis or discuss the impact of the lower initial pressure on the result. The applicant calculated heat transfer coefficients to passive heat sinks in accordance with Appendix A to WCAP-8339 and the containment response using the COCO code.
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L previously approved the use of WCAP-8339, and the C0CO code for ECCS performance analyses by a letter to Westinghouse dated May 30, 1975.
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We are nnable to complete our review of the analysis results.
The applicant should' provide the mass and energy data used in the ECCS back pressure analys'is and justify using an initial containment pressure of 14.7 l
psia. This is a confimatory item.
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6.2.2 Subcompartment Analyses Subcompartment analyses are required to determine the acceptability of the design differential pressure loadings on containment internal structures from high-energy line ruptures. The applicant has perfonned the necessary subcom-partment analyses for the reactor cavity, the steam generator compart-ments, and the pressurizer compartment where high-energy line ruptures are postulated to occur.
A spectrum of pipe breaks was analyzed by the applicant to determine the break sizes and locations that resulted in peak differential loads on each of the walls around each subcompartment.
The short term mass and energy release rate data used in the subcompartment analyses (VogtleFSA Tables 6.2.1-26through6.2.1-28)werecalculatedusing R
the analytical model described in WCAP-8312-A, which was approved by NRC in a letter dated March 12, 1975. We therefore conclude that the methodology for computing the mass and energy release rate data used in the subcompartment,-
analyses is acceptable.
The applicant used the COPDA computer program to analyze the pressure transients
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in the reactor cavity, all of the steam generator compartments, and the pressur-izer compartment. The COPDA code, described in BN-TOP-4, Rev. 1, was found acceptable by NRC in a letter dated February 23; 1979. We, therefore, conclude that the calculational method for the subcompartment analyses.is acceptable.
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A separate discussion and review of the analyses of the reactor cavity and the steam generator and pressurizer compartments are presented below.
Reactor Cavity Analysis The reactor cavity is a heavily reinforced concrete structure that performs the dual function of providing reactor vessel support and radiation shielding.
l The applicant's sub. compartment analysis postulated a 144 in2 cold leg break.
i The inherent stiffriess of the sys,tems, together with the pipe whip restraints, limits-the postulated pipe rupture and flow area to this break sizes.
The reactor cavity design basis break was found to be the 144 in2 cold leg rupture.
l The staff has reviewed the applicant's analysis and concudes in the selection '
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of the design basis pipe break contingent upon the acceptability of the mechanically constrained limit on the pipe break size (see SER Chapter 3.0).
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The reactor cavity nodalization model used by the applicant accounts for all physical restrictions in the vent flow paths. The upper reactor cavity seal ring was assumed to be completely plugged. Access ports to the annelar inspection cavity were assumed to be closed. For the purpose of nodal volume i
calculations, all themal insulation was assumed to remain in place.
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reactor cavity nozzle penetrations of the unaffected loops were assumed to be completely plugged. The nozzle penetration to the steam generator compartments y for the affected loop was assumed plugged. The applicant followed the recommendations of Section 3.2 of NUREG-0609, "Asymetric Blowdown Loads _on pWR Primary Systems in nodalizing the reactor cavity." We have examined the applicant's nodal model, shown in Figure 6.2.1-16 of the FSAR, and find it acceptable, peak differential pressure loads on the reactor cavity walls of 192.9 psid were obtained for the 144-square inch cold leg break. The applicant also analyzed the forces and moments acting on the reactor vessel.
The application of these loads in designing the reactor cavity is discussed in Section 3.8.3.3 of the FSAR. The design adequacy,of structures is discussed in Chapter 3 of this report.
Our review of the applicant's reactor cavity analysis included an examination
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- of the nodal model, initial conditions, and major assumptions. Based on our review of the infomation provided, we conclude that the applicant's reactor cavity analysis is acceptable.
Steam Generator Compartment Analysis The steam generator compartment encloses two steam generators and two reactor coolant pumps. The applicant analyzed the pressure response of the steam generator compartment to a spectrum of breaks of reactor coolant system pipe breaks including both break size and location (double-ended breaks at l
pump outlet nozzle, pump inlet nozzle, steam generator outlet nozzle, steam generator inlet nozzle, and loop closure weld, and a split break at the steam generator inlet elb.ow). The limiting pipe break size and location were identified to be a 436 square-inch double-ended break at the steam generator outlet' nozzle. The calculated maximum pressure differential load is 23.5 psid.
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We have reviewed the applicant's assumptions and input data used in the analysis.
We have found that the applicant's analysis is consistent with the guidance i
in Section 3.2 of NUREG-0609. Based on our review, we conclude that the applicant's steam generator compartment pressure response analysis is acceptable q
for the design evaluation of the steain generator supports as well as for the
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steam generator compartment structures.
.b Pressurizer Compartment Analysis The pressurizer compartment is a small, nearly square compartment that encloses the pressurizer vessel. The subcompartment is vented at the top to the upper containment. Two break locations, namely a double-ended surge line break and a double-ended spray line break, were analyzed by the applicant.
Reductions in flow area and subcompartment volume due to insulation around pipino and vessels were accounted for in the analysis. Flow restrictions and maor vent flow areas were accounted for in nodalizing. The results of the
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applicant's analysis showed a peak differential pressure of 21.08 psid across
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the walls of the pressurizer compartment.
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[t Based on our review of the applicant's nodal model, initial conditions, and i
assumptions, we find the applicant's pressurizer. compartment analysis a
acceptable.
i 6.2.3 Containment Heat Removal' Systems P
a The function of the containment heat removal system (CHRS) is to remove heat l
from the containment atmosphere to limit, reduce, and maintain at acceptable j~
low levels the containment pressure and temperature following a LOCA or secondary B-system pipe rupture. In addition to heat removal provided by passive means O
such as heat transfer to containment walls, structures, and equipment located j,
inside containment, the Vogtle design includes active containment heat removal j,
systems. The activ,e containment heat removal systems consist of the containmen't i
cooling system and the containment spray system. The containment cooling i
i-system 'also functions during norEa1 operation to maintain a suitable atmospher a
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for the equipment located within the containment. The containment spray system does not have a normal operating function.
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b The containment spray system consists of two redundant and independent trains I!
powered from separate sources independent of offsite power. Each of the two jj containment spray pumps has a design flow rate of 2600 gpm. All active compo-P nents of the containment spray system are capable of being tested during plant ij operation. The system is designed and fabricated to codes consistent with
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Regulatory Guide 1.26 and seismic Category I requirements.
In Table 3.2.2-1 of the FSAR, the containment spray additive tank and its downstream valves and piping are specified to Quality Group C which is contrary to the statement in Section 6.2.2.2.3 of the FSAR that the containment spray system is, designed to i
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Quality Group B requirements. Since the spray additive system is required during the operation of the containment spray system, it should meet the same design requirements.g This is an open item. The two containment spray
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recirculation intake' pipes take suction from two separated containment emergency sumps. The sump intakes are protected by trash guards and fine mesh screens from debris that could clog the spray nozzles.
I The containment spray system is automatically actuated by a coincident two-ou of-fourcontainment(high-3)pressuresignal. Operation of the spray system may also be manually initiated from the control room. The spray system will initially take suction from the refueling water storage tank. When the tank reaches a low-low level alann, a switchover from injection to recirculation will
' initiated manually. The applicant has also provided a failure mode i
and effects analysis and other information demonstrating the ability of the containment spray system to function follofwng postulated single active failures. The containment spray system has been designed to allow periodic inspection of the components and functional testing to assure the operability and performance of the system.
s Thestaffhasreviewedtheapplicant'snetpostivesuctionhead(NPSH)calcula-tions and finds the NPSH available in either the recirculation spray mode or the ECCS cold leg recirculation mode to be adequate. The applicant has complied with the provisions,of Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Contai,nment Heat Removal Systems," with one exception.
Regula' tory Guide 1.1 states that containment heat removal systems should be designed so that adequate NPSH is provided to system pumps assuming maximum g.-..........,
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i-expected temperatures of pumped fluids and no increase in containment pressure i
from that present before postulated LOCAs.
Instead, the applicant has calcu-
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lated NPSH available using a saturated sump model (that is, the containment atmospheric pressure is conservatively assumed to be equal to the vapor pressure j
of the liquid in the recirculation sumps, ensuring that credit is not taken i
forcontainmentpressurizationduringthetransient). The staff has previously j
found the saturated sump model to be conservative (SRP Section 6.2.2, Acceptanc d
Criterion 2)and,therefore, acceptable.
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Regulatory Guide 1.82 " Sumps for Emergency Core Cooling and Containment Spray Systems" provides guidelines to be met by the containment sumps that are designed to be sources of water for ECCS and the containment spray systems following a LOCA. Yhe guideline addresses redundancy, location and arrange ment of sumps as well as provisions to screen out debris and to ensure adequate pump performance. The staff has reviewed the applicants sump design and finds it to be in conformance with the guidelines of Regulatory Guide 1.82 Revision O.
We, therefore,. find the design acceptable.
The staff has reviewed the infonnation in the FSAR and the responses to staff N
requests for additional information concerning the containment spray systems i,
to ensure confonnance to all of the acceptance criteria in SRP Section 6.2.2.
The staff concludes that the containment heat r'emoval systems satisfy the requirements of GDC 38, 39, and 40 and the provisions of Regulatory Guides 1.1 and 1.82 and are, therefore, acceptable with one exception. That is the spray additive tank and the associated downstream valves, and piping should be designed to Quality Group B instead of Group C.
The containment cooling system consists of eight 25% capacity fan cooler units.
Each fan cooler unit consists of an axial fan, a fan motor, cooper-nickel cooling coils with cooper fins, a carbon steel housing, round metal ducting, and a concrete discharge duct.
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y The system has two modes of operation which are:
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Nomal operation - during which four fan coolers ere operating at a rated r
flow of 97,000 CFM each.
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Accident operation - during which at least two fan coolers are assumed to be operating at a reduced flow of 43,000 CFM each.
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During normal operation four of the eight fan coolers are operating at the
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higher capcity. Upon receipt of SIAS, the four fans are automatically placed
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into low speed operation. Only two fan coolers (one of two trains) are assumed i
in operation in accident analyses. The system design will pemit remote operation from the c"ontrol room and from the shutdown panels.
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the system has been designed to allow periodic inspection of the components y
and functional testing to assure the operability and performance of the system.
The containment cooling units are designed to seismic Category I, Quality Group B.
Based on our review of the containment cooling system, we conclude that the p
system design is in accordance with the requirements of GDC 38, 39 and 40 and, therefore, is acceptable, 6.2.4 Containment Isolation System The function of the containment isolation system is to allow the normal or emergency passage of fluids through the containment boundary while preserving the ability of the boundary to prevent or limit the escape of fission products that may result from postulated accidents. The containment isolation provi-(
sions are safety grade design (ASME Section III, Class 2 and Seismic Category 1).
The containment isolation for Vogtle is accomplished in two phases.
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containment isolati,on phase A (CIA) signal, wi:ich shuts all non-essential system lines penetrating the containment, is initiated by any of the following:
h-(1) safety injection signals (including (a) high conuinment pressure (high-1);
(b) low steam line pressure; (c) low pressurizer pressure; (d) manual safety injection actuation); (2) containment high radiation; or (3) manual containment g
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isolation actuation.
The containment isolation phase B (CIB) signal is i
initiated by containment high-3 pressure or by manual actuation.
Although the Phase B containment isolation signal is not actuated by diverse parameters, i
this is acceptable because the only affected lines are considered important to safe shutdown of the plant, and thes.e'11nes can be remote-manually isolated.The.
i staff concludes that adequate diversity has been provided with regard to the
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different monitored parameters which actuate containment isolation.
The containment isolation provisions for the Ifnes penetrating containment p
must conform to the requirements of GDC 54, 55, 56 or 57, as appropriate. As permitted by GDC 55 and 56, there may be penetrations whose isolation provi-sions do not satisfy the explicit requirements of the GDC but which are accep-l tableonsomeother'ditfinedbasis. These penetrations are discussed below.
The applicant states in the response to Q480.23 that the containment pressure instrument lines and the reactor vessel water level instrument lines are designed in accordance with Regulatory Guide 1.11 and, therefore, are not subject to the requirements of GDC 54, 55, 56, or 57. The applicant, however, f
has not adequately justified the applicability of R.G. 1.11 to the containment
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pressure instrument lines, or the confonnance of the vessel water level
. instrument lines to R.G. 1.11. This is a confinnatory item.
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Penetrations associated with the secondary (ste m/feedwater) system were designed to comply with GDC 57.
In Amendment 5 to the FSAR, the applicant revised its position and stated that these penetrations are not subject to the requirements of GDC 57. We find this approach unacceptable; penetrations E
associated with the secondary side should comply with GDC 57. This is an open item.-
Penetrations 59 and 60 are the residual heat removal (RHR) suction lines, I
from the reactor coolant system (RCS) hot legs. Each line has two normally I
closed, motor-operated gate valves in series inside containment but no. valve
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outside containment; this is a deviation of GDC 55. The valves are interlocked i
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to prevent them from being inadvertently opened and are designed to remain closed following an accident. These lines connect to the ECCS outside the
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containment. In view of the above design, the staff has concluded that the normally closed system isolation valve closest to the containment and the
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closed safety grade system outside estainment constitute an acceptable alternative to the explicit isolation requirements of GDC 55 as pennitted by the "other defined basis" provisions of that criterion.
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4 Penetration 87, the containment leak test line, is provided with a blind flange inside containment and a locked closed isolation valve outside contain-This represents a deviation from the explicit requirements _of GDC 56 ment.
in that a blind flange is used in liue of a valve. The staff has concluded that a blind flange is an appropriate substitute for an isolation valve, provided it is leak testable. Staff acceptance of the isolation provisions for penetration 87 is contingent on the applicant's confirmation that the blind flange is leak testable.
The containment sump recirculation lines (penetrations 36 through 39), which supply suction to the RHR pumps and the containment spray pumps, are each provided with a single isolation valve outside containment; this is a deviation L
from GDC 56 requirements since there is no isolation valve inside containment.
The ' isolation valve outside containment is enclosed in a valve isolation tank, and the piping from the sump to the valve is enclosed in a guard pipe.
The guard pipe and valve isolation tank serve to confine leakage that may occur between the valve and the containment.
In view of the design, the staff finds
. that the isolation provisions for the sump recirculation lines are an acceptable alternative to the explicit requirements of GDC 56, contingent on the applicant's confirmation that the valve isolation tank and guard pipe are leak tested.
We have reviewed the closure times for the containment isolation valves.
Most valves close in 15 seconds or less.
In particular, the 14-inch minipurge lines, which may bg,open during nonnal plant operation, are designed to close in less than 5 seconds. We conclude that the containment isolation valve closure times are scceptable.
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The applicant has designed the containment isolation system in accordance with the provisions of NUREG-0737, Item I.E.4.2, " Containment Isolation Dependability".
As previously described, the applicant has complied with the provisions regarding diversity in parameters sensed for initiation of containment isolation.
All nonessential lines, which are identified in Table 6.2.4-1 of the FSAR, are automatically isolated by the containment isolation signals. The 24-inch preaccess purge lines are sealed closed during operational modes 1, 2, 3, and 4 and are verified to be closed at least every 31 days. This requirement will be included in the plant's Technical Specifications. The containment isolation provision for the purge lines are designed to ASME Section III Class 2, and seismic Category 1 requirements. The inboard and outboard isolation valves are supplied with Category 1E power from bus A and bus B, respectively.
The isolation valves of the minipurge lines are designed to fail closed upon loss of air pressure or electric power, and are testable from the control The containment purge system is automatically isolated by any signal room.
resulting from safety injection actuation, high containment pressure, or containment high radiation level. Debris screens are provided on both the inlet and outlet of the mini-purge ducting inside the contair. ment.
The debris screens are designed to withstand a differential pressure of 60 psi and are supported to Seismic Category I requirements. The piping between the isolation valve and the debris screen was dynamically analyzed to seismic Category 1 requirements. The staff has reviewed the applicants containment isolation system design for conformance to the provisions of BTP CSB 6-4, Containment Purging During Normal Plant Operation, and find it to be accept -
able.
Based ori our review, we conclude that the design of the containment system conforms to GDC 54, 55, 56 and 57, NUREG-0737 Item II.E.4.2, and SRP 6.2.4 and CSB BTP 6-4, with the exception of the issues discussed in the above evalua-tion.
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't 6.2.5 Combustible Gas Control System y
Following a LOCA, hydrogen may accumulate within containment as a result of d
(1) metal-water reaction between the zirconium fuel cladding and the reactor coolant, (2) radiolytic decomposition'of the water in the reactor core and the
{
containment sump, and (3) corrosion of metals by emergency core cooling and I
I containment spray solutions. To monitor and control the buildup of hydrogen n
within containment, the applicant has provided a hydrogen recombiner system, a
[
hydrogen monitoring system, a post-LOCA purge system, a post-LOCA cavity-hydrogen purge system, and containment hydrogen mixing system.
The hydrogen recombi.ner system consists of two redundant 100 percent-capacity, electric hydrogen re' combiners and associated control units located in the containment.
Each recombiner train has a capacity of 100 SCFM and is designed to seismic Category 1 design criteria. The recombiner system.is supplied from the Class IE emergency buses, and is manually started and operated from a control panel located in the control building.
A redundant containment hydrogen monitoring system is provided in the Vogtle design. Each train contains a hydrogen analyzer and two sample lines.,The analyzers are located in accessible areas outside the containment.
The hydrogen monitors are normally isolated from the containment atmosphere by the closed containment isolation valves which are operated remote manually from the control room.
- h The piping and valves of the hydrogen monitoring system outside containment are designed to Quality Group C.
This is not consistent with the l
provisions of Regulatory Guide 1.7, in which it is stated that the containment combustible gas control system be designed, fabricated, erected, and tested to the Group B quality standards of Regulatory Guide 1.26.
This is an open item.
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A purge system is provided for post-accident containment atmosphere cleanup.
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in accordance with,Section 50.44 of 10 CFR Part 50. The system consists of an j.,
exhaust penetration line and a filter system. The containment isolation valves y
and interconnec' ting piping are seismic Category 1.
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The system is actuated manually. The outside isolation valve is locked closed i[
and manually controlled by the operator. The nonna11y closed remote-manual, j
parallel isolation valves inside the containment have position indication in
[
the control room.
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L The post-LOCA reactor cavity hydrogen purge system is designed to preclude hydrogen pocketing following a LOCA by supplying air to the reactor cavity for
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dilution of the hydrogen that may be released in the reactor cavity area.
The
[
system consists of two redundant 100% capacity fans which automatically start on a safety injection signal. The fans take suction from the atmosphere within
{
the steam generator compartments and discharge the. air into the reactor cavity.
The air flows out ofghe cavity along the reactor vessel nozzles or through the J
ventilation openings surrounding the seal ring.
Containment hydrogen mixing is facilitated by the containment fan coolers, g
which take suction from above the operating deck and discharge to the lower levels of the containment. They are, therefore, able to mix the containment
- f atmosphere and prevent hydrogen pocketing in the containment.
The applicant has analyzed the production and accumulation of hydrogen within f,
the containment using the guidelines provided in Regulatory Guide 1.7 and SRP 6.2.5.
- l The applicant's analysis shows that a single recombiner, started on the second day following the onset of a LOCA, at a containment hydrogen concentration of 3.5 volume percent, is c'apable of limiting the hydrogen i
concentration in containroent to below the Regulatory Guide 1.7 lower t
flamability limit of 4.0 volume percent.
p e
Based on our review of the Vogtle combustible gas control system we conclude l
that the system, with one exception, satisfies the design and performance requirements of 10 CFR 50.44, the provisions of Regulatory Guide 1.7 the
[
requirements of GDC 41, 42, and 43, and the requirements of NUREG-0737 Item l,
II.E.4.1 and 11.F.1 Attachment 6.
The one exception deals with the open item
{
on use of Quality G'roup C instead of Quality Group B criteria for design of the j
piping and valves of the hydrogen' monitoring system. We will need to discuss i
this matter further with the applicant. Moreover, the applicant will I
be required to confirm that procedures have been developed to ensure that the h
recombiner system is actuated in a timely manner. This is a confirmatory item.
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6.2.6 Containment Leakage Testing Program The containment design includes the provisions and features necessary to satisfy the testing requirements of Appendix J to 10 CFR Part 50. The design of the containment penetration and isolation valves permits preoperational and periodic leakage rate testing at the pressure specified in Appendix J to 10 j
CFR Part 50.
The staff has reviewed the containment leakage testing program in the FSAR and in response to NRC questions, and finds that the proposed reactor contair.-
ment leakage testing program complies with the requirements of Appendix J to 10 CFR Part 50 with the exception discussed below:
1 The applicant indicated in the FSAR and the response to question, Q480.33, that lines which penetrate containment and are required to perform a safeguard function following an accident need not be Type C tested. The lines which fall into this category include:
(1) the safety injection pump
~
dischargelines(penetrations 30,31,and33);(2)theresidualheatremoval pump discharge lines (penetrations 56, 57, and 58); (3) the centrifugal charging
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pump cold leg injection path (penetraticn 32); (4) the containment spray pump dischargelines(penetration 34and35);(5)thecontainmentemergencysump lines to the residual heat removal and containment spray pump suction lines (penetration 36 through 39); ar,ei (6) the residual heat removal pump hot leg suction lines (penetrations 59 and 60).
i l
The valves in the penetrations, discussed in Q480.33 and listed above, are j
nmually closed.
If the isolation valves in the above penetrations fail te open when called upon to do so in the event of an accident, the valves will j
remain closed and provide a containment isolation function. Therefore.
Type C testing of these valves is required, unless a water seal can be
}
justified.
For water-sealed penetrations it is necessary to demonstrate (as prescribed by Appertdix J) that the fluid inventory is sufficient to assure the l
sealing function, t.e., preclude, containment atmosphere leakage for at least 30 days. -
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i The applicant has stated that these lines will always be water filled; during ECCS operation water will be flowing into containre.ent, or if the valves are
]
closed the pumps will maintain an inward acting head of water on the valves.
However, it is not clear that the valve stems will not become containment atmosphere leak paths when the valves'are closed. Therefore, additional
[
h information is needed to justify excluding these valves from~ the Type C testing program. This is a confirmatory item.
4 j.h With regard to the reactor coolant pump (RCP) seal water injection lines
[
(penetrations 51through54),theapplicantstatesintheresponsetoQ480.33 that seal injection is desirable at' all ti.nes. Therefore, these lines are not automatically isolated in the event of an accident, and are remote-manually controlled by the operator. The valves in these penetration lines are designed to be open during normal operation as well as post-accident condition.
In addition, the system is designed as a safety graded closed system both inside i
and outside containment. These lines can not become containment atmosphere leak paths; therefore, an exemption cf Type C Testing for penetration 52 is p
)
. acceptable.
However, the infonnation about post-accident valve position for,
[
penetrations 51, 53, and 54 in Table 6.2.4-1 of the FSAR is not correct. A revision of the table is needed.
I[
Except as noted, containment leakage testing program c1mplies with the requirements of Appendix J to 10 CFR Part 50. Such compliance provides adequate assurance that containment leaktight integrity can be verified throughout service lifetime and that the leakage rates will be periodically checked during service on a timely basis to maintain such leakages within the specified limits of the Technical Specifications. The Plant's Technical Specifications will contain appropriate surve111ance requirements for containment leak testing, including test frequencies. Based on the above discussion, the staff concludes that the applicant's leakage testing program is acceptable and complies with the requirements of GDC 52, 53, and 54; L
Appendix J to 10 CFJi 50; and 10 CFR 100.
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1l ENCLOSURE 2 DRAFT SAFETY EVALUATION REPORT OUTSTANDING ISSUES V0GTLE,t! NITS 1 AND 2
)
CONTAINMENT' SYSTEMS BRANCH I.
Confirmatory Items
- 4 1.
A more detailed discussion of long tem containment heat removal capability is needed to confirm the acceptability of the long-term containment pressure transient following a LOCA.
.\\
f 2.
The applicant should provide an analysis of containment external load resulting from misoperation of the containment purge system.
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3.
The applicant should provide a complete spt of the mass and energy h
data used in the ECCS back pressure analysis, and justify using an initial containment pre:sure of 14.7 psia.
~
4.
The applicant should provide addit:onal infomation to justify the applicability of R.G.1.11 to the containnant pressure instrument lines, and to demonstrate the confomance of vessel water level instrument lines to R.G. 1.11.
r 5.
The staff's acceptance of the isolation provisions for penetration' l
-87 (containment leak test line) is contingent on the applicant's confirmation that the blind flange is leak testable.
6.
The staff's acceptance of the isolation provisions for penetrations 36 through 39 is contingent on the applicant's confimation that the valve isolation tank and guard pipe are leak testable.
s 7.
The appi cant should confim that procedures have been developed to ensurt: that the recombiner system will be actuated in a timely manner following the onset of an accident.
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. '. 8.
The applicant should justify excluding the isolation valves in the following penetrations from the Type C testing program: penetration numbers 30 through 39, a d' 56 through 60
,9.
The applicant should revise Table 6.2.4-1 of the FSAR on post-accident valve positions for penetrations 51, 53, and 54 to j
reflect the operation of these isolation valves.
Open Items
' II.
t 11 o
1.
The applicant should provide revised mass and energy release data using the approved versions of the MARVEL /TRANFLO codes and to assess the impact of these changes on the containment responses following a MSLB. 'The applicant,should assume an initial
,f containment pressure that is conservative.for the purpose of calculating the peak containment pressure and temperature.
2.
The spray additive system should meet Quality Group B req ~uirements.
/
Lr 3.
Containment penetrations associated with the secondary side should l
comply with'GDC 57.
l E 1 ;
y 4.
The piping and valves of the hydrogen monitoring system outside containment should bt jesigned to Quality G or up B standards.
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Response from" Reviewers Survey of OLs Docketed After December 1980
.,'.i Structural Analfsis Considering Flexibility of Tanks _2 General Instructions:
1.
Prepare this form for each major tank (e.g., RWST, CST &
AFWST for PWRs).
2.
BWRs may not have any large Category I tanks.
I$4 neb'-
A.
Plant Description Reviewer (s) -- '--
- 1. Plant: b /
- 2-u
- 2. Docket No.:
Po 4 / /'78-r' 4..
- 3. NSSS System: h) b [WR.
s
- 4. A/E:
M dL Cl.. A -)
r B.
Tank Description
- 1. Type of Tank (Function): bi (KM WST-) ~ f Ta~4.
M Sdrv
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- 2. Height:
- 3. Diameter:
33'
- 4. Wall, Thickness
- 2. '
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- 5. Anchoring Syster. (Brief Description):
- 6. Foundation (Brief Description): #lbf/'X 3' M /.vus cn c.o r
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Response from" Reviewers Survey of OLs Docketed After December 1980 Structural Analfsis Considering Flexibility of Tanks _;. _ -
General Instructions:
1.
Prepare this form for each major tank (e.g., RWST, CST &
AFWST for PWRs).
4 2.
BWRs may'not have any large Category I tanks.
NI?C-Nb'-
A.
Plant Description Reviewer (s) -- '--
- 1. Plant: kb / * : -
V
- 2. Docket No.:
Vo dt V/'r's-r*
..e
- 3. NSSS System: D)Mbw OWR.
- 4. A/E:
bt-cAh t-(L. A.) _-
B.
Tank Description
- 1. Type of Tank (Function):
%/db f A~4.-(EwSd I
d i /
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- 2. Height:
- 3. Diameter:
N'
- 4. Wall, Thickness 3,
g
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- 6. Foundation. (Brief Description):
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Response from Reviewers
~
Survey of OLs Docketed After December 1980
_ n.i t'
Structural Analfsis Considering Flexibility of Tanks _2 General Instructions:
1.
Prepare this form for each major tank (e.g., RWST, CST &
AFWST for PWRs).
S 2.
BWRs may not have any large Category I tanks.
K e sk A.
Plant Description Reviewer (s)
O'-
NLb'-
- 1. Plant: kb / * : -
V -
y
- 2. Docket No.:
fD d2M/'r's-r' f
- 3. NSSS System: $.
@b [WR.
- 4. A/E:
b2-clk L (L. A.)
B.
Tank Description
- 1. Type of Tank (Function):
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- 2. Height:
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- 3. Diameter:
M
- 4. Wall. Thickness
- 2. i Mlmd A
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- 5. Anchoring System (Brief Description):
- 6. Foundation (Brief Description):
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Review Findings 1
1.
Was the Seismic Analysis Proced0re for Category. I tanks reviewed bb and/or audited?-
W By Whom? d I-g 2.
Did the tank analysis procedure include the tank wall flexibility consideration:
e 3.
Briefly describe the tank analysis procedure and give pertinent references (if known) T/3-704 [h M p ZA A 9'A. A)
D.
Any Other Pertinent Comments T J A t y -
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O dlNb WN nob bk'3 v, h bk hit $ \\wdWha.ni k12p'M Further support for this conclusion comes from'Prowell's later (1983) report ( en Cretaceous and younger faults of eastern North America. The Millett and Statesboro Faults are not included as documented faults. It is concluded, therefore, that no surface fab 1 ting capable of localizing c2rthquakes is present at the plant site on in th.e vicinity of the site. l I4 2.5.4 Stability of Subsurface Materials an'd Foundations l S:ctions 2.5.4.1 through 2.5.4.7 summarize the staff's geotechnical engineering review of the Vogtle Electric Generating Plant, Units 1 and 2, as pglented Tri_ thn Final Safety Analysis Report (FSAR) through Amendment ' dated Su h b k4$ S
- and the applicant's response to staff questions Q241 1 through Q241 24 The st:bility of subsurface materials and foundations (FSAR Section 2.5.4) has been evaluated in accordance with the applicable criteria outline in 10 CFR 50; 10 CFR.100; Appendix A of 10 CFR 100; RG 1.70, ' Standard Format and Content of Snfety Analysis Reports for Nuclear Power Plants" (Rev. 3); RG 1.132, " Site Invcstigations for Four! ations of Nuclear Power Plants;" RG 1.138, " Laboratory d
Invastigations of Soils for Engineering Analysis and Design of Nuclear Power Plants"; and the Standard Review Plan (SRP), NUREG-0800, July 1981. 2.5.4.1 Site Conditions 9tmw.W The site conditions which exist at the Vogtle site do not involve stability of x slopes nor embankment and dams, and therefore, FSAR Sections 2.5.5 and 2.5.6 are not addressed in this eport. 2.5.4.1.1 General Th] Vogtle Electric' Generating Plant. Units 1 and 2 f r located on the southwest sid3 of the Savannah River in Burke County, Georgia, approximately 26 mi s:utheast of Augusta, Georgia. The topography of the site is one of rolling hills wnh original ground surface elevations in the immediate plant area (excluding th,e river. intake canal and structure) generally ranging from c1 255 ft ab:ve mean sea level to el 280 ft. Firlal plant grade at el 219.5 ft required t ' I)b. 10/26/84 2-42 V0GTLE DSER SEC 2 ~
a the' removal f the upper natural soils. The Savannah River at its closest point to the site is approximately 3000 ft northeast from the main plant area and has a normal water elevation of 80 ft. The maximum water level in the Savannah River has been estimated at el 1,65 ft under assumed probable maximum flood (PMF) conditions that include allowance for wave runup. As described in Section 2.4 of this SER, groundwater movement has been observed in an upper water table aquifer system and a confined aquifer system that is located below approximately el 70 ft. The foundation designs of seismic Category I struc-tures have been based on a maximum groundwater elevation of 165 ft and this maximum level would be, located in the upper wcter table aquifer system. ,q 2.5.4.1.2 Site Foundation Description W The subsurface conditions as revealed by explorations and foundation excava-tions in the plant site area may be divided into three principal strata. The top stratum consists of sands (SP), silty sands (SM), and clayey sands (SC) and in the FSAR this top layer is identified as the upper sand stratum. The upper sand stratum is about 85 ft in thickness below plant grade and has a bottom elevation at approximately el 135 ft. A sh~elly liinestone (Utley Lime-stone), which subsurface explorations showed to be subjected to extensive leaching and to solution cavities, is located at the base of the upper sand - stratum and ranged up to 12 ft in thickness. The stratum below the Utley Limestene is the major foundation-supportir.g layer and is identified as the clay marl bearing stratum. The clay marl stratum is approximately 65 ft in thickness in the main plant area and ranges in elevation between 135 ft and 70 ft. The clay marl stratum is a gray to greenish gray, calcareous silty clay with shell fragments and interbedded with limestone and sand lenses. Drilling recoveries show the earl to be predominantly a hard to very hard, weakly cemented material with some zones of softer marl. Seismic explorations indicated a velocity interface about 15 ft beloithe top of the clay marl stratum which is a reflection of weathering in the upper 15 ft of the marl A thick, dense, coarse-to-fine sand zone with minor interbedding of zone. sil,tyclayandclayeysilt'iayersislocatedbeneaththeclaymarlstratum. This lower sand stratum is estim'ated to be in excess'of 750 ft in thickness; 10/26/84 2-43 V0GTLi OSER SEC 2
x, ?, ,w, recorded blow counts per foot of penetration in' the standard penetration test ~ (SPT) are generally in excess of 100 blews @ M d M n,q.r The applicant decided to excavate the upper natural soils and extend this excavation into the clay marl stratum in order to avoid foundation difficulties with the shelly limestone layei and to eliminate any potential for itquefaction ~ in the upper sand stratum. Liquefaction had been indicated to be a possibility in the upper sands when evaluated, allowing' for a seismic event equivalent to the safe shutdown earthquake (SSE). This extensive foundation excavation operation required the removal of approximately 5 million cubic yards of soil tr el 130 ft and measured approximately 1000 ft along each side at the bottom-of the excavation which was roughly square in shape. A deeper exca,vation to el 108.5 ft in the c' lay marl stratum was made over a rectangular area measuring 120 ft x 440 ft to accommodate the basemat for the deeper portion of'the auxiliary building. Description of the geologic mapping, dewatering at.rivities, rebound monitoring, surface cleanup and protection measures, and foundation inspection and approval procedures are provided in Appendix 28 of the FSAR. The. foundations of seismic Category I structures that are evaluated in this ~ report include the reactor co,ntainment buildings, nuclear service cooling water (NSCW) towers and pumphouses, auxiliary building, fuel-handling buil/.ing, control building, diesel generator butidings, diesel fuel oil storage tanks and buildings, condensate storage tanks, auxiliary feedwater pumphouses, refueling water 4torage tanks, reactor makeup water stor' age tanks, and Category I piping, conduits and tunnels. Reinforced concrete mat foundations were used for Category I structures with the exception of wall footings for certain tanks, and box culverts for piping and tunnels. FSAR Figure 241.2-1 provides _ a plan view of the main plant layout and ' identifies the outline of seismic Category I structures. resc N m h % H9 Excavating the natural soils to the clay marl stratum p'- --d the foundations of most seismic Category I structures on compacted backfill. Only the more d:eply. founded auxiliary building, NSCW towers, and instrumentation cavity of the containment building are founded on the clay marl. All other foundations of the power block structures are supported on Category 1 backfill and have 10/26/84 2-44 V0GTLE OSER SEC 2
s. ~o ~ foun'dation elevations ranging from el 158 ft ("eactor building) to el 218 ft 1 (reactor makeup water storage tanks). Category I backf'ill was selectively excavated from nearby borrow sources and cAton consisted ofjmedium to fine sands (SP) and sands with some silt (SP-SM). Although permitted by PSAR and'FSAR documentation to contain up to 25% by weight. passing the No. 200 sieve, the percent of fines actually contained in the Category I backfill that was placed and compacted was limited in the field to about 12%. All Category I backfill in the power block area was to be compacted to an average of 97% o,f the maximum dry density determined by American Society of Testing Materials (ASTM) D1557, with no tests below 93% and not more than 10% of the tests between 95 and 93%. On the basis of'the results of test fill studies, the applicant indicated his intent was to control the placement moisture content of the Category 1 backfill to within* 2% of the optimum moisture content determined by ASTM D1557 (FSAR Section 2.5.4.5.2.7). The staff's evaluation on the adequacy of the compacted backfill is subsequently discussedinthis$5ERinSection2.5.4.3. 2.5.4.1.3 ' Site Investigations Fiald investigations at the site were initially started in January 1971 'and were continued during construction; 38 borings were drilled from the bottom of the foundation excavation on the top of the clay marl, bearing stratum in 1977' in the po v block area.. The field investigations have included drilling, geophysical seismic surveys, groundwater studies, and geologic mapping of foundation excavations. A total of 474 holes have been drilled, of which 111 holes were completed subsequent to the PSAR investigations. Table 2B-1 of the, FSAR provides a list of borings with summary information for foundation investi-gations completed for the PSAR. The, site investigations were completed to define the various subsurface materials and stratifica, tion, to obtain soil samples for laboratory testing and the es,tablishment of engineecing properties, to identify sources of suitable borrow, to perrait measurement of shear and compression wave velocities, to determine in situ foundation material permeabilities and groundwater movement -Q 10/26/84 2-45 V0GTLE DSER SEC 2 i
..a end for geologic mapping and inspection (e.g. for faulting, cavities, soft I nnes) of foundation excavations for approval before concrete was placed. \\ The investigation's completed at the Vogtle site did not extend to firm bedrock 1 which is estimated to be approximately 750 ft.below the bottom of the clay marl stratum. 3 On the basis of its review of the information presented in the FSAR, the staff concludes that the site investigations comp'leted by the applicant are accept-cble and adequate to identify the important subsurface features and foundation conditions, with the exception of the "CS" series holes drilled in 1977 from on top of the clay marl stratum. A In Question 241.3 of its review of the FSAR, the staff attempted to understand the reasons for the poor core recovery in the clay marl stratum that was lb) SERT [ indicated in 9 of the 36 borings which were drilled in 19778S'::.aa the
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Q241.5 whiuh,iere .=2pvu== ww p-a"4 M 4a 8 =txt 6 (u.y 30'.), tM: renc:m-rema4mwn-epen eeview4 tem. Jh,hasis-for tM: 'a H' t no sitiq1.inpJ.udae +h: '+11wirg. - hi ) The Applicant's response to Q241.3 indicates that the 107 ari. I hig ' pr ram was not an exploration program and was not designed to obtai core r overy but, rather, was intended to obtain selected samp of the clay marl for oratory testing. The staff has great difficult n understanding this response. The staff finds that'the borings of jth S" series, some of 'which were drille ithin the foundation limits o 4he NSCW towers, auxiliary building, containment 11 dings, control bu ng, and fuel-handling building, were important to assessi the founda n competency of the clay marl stratum. -and should have been drilled ac dance with good engineering practice and the guidelines of RG l'.132 "5 . estigations for Foundations of Nuclear Power' Plants." Good eng,ijueri,ng prac e would require a full and complete description of the m rit:ls encountered the entire depth and an explanation. en the boring lo for 0% recoveries in order t reperly assess this condition en the a5eq y of foundation design and future but ng peformance. Supplementary explora ns specifically intended to determine the feat s of the zones of th oor core recovery would normally be completed. The a &,t'1rygam l 10/26/84 2-46 V0GTLE DSER SEC 2 C..'
c. m. -. INSERT 1 The applicant in meetings with the Staff and in its response to Q241.3 has indicated that the poor core recovery.resulted from the difficulties experienced when drilling with a 4-inch diameter single tube core barrel. When the harder limestene lenses were encountered in the marl during drilling, turning and grinding of the core would occur. In the borings of poor recovery, when the core was eventually withdrawn from the borehole after a drill run had been completed, the core, because of the past grinding difficulties, would drop out of the core barrel and be lost for meaningful recovery. The applicant has noted that no drilling fluid was lost during the 1977 exploration program and there were no instances of rapid advancement during drilling which would indicate the presence of voids. In the staff's review of the many borings drilled in and through the clay marl layer during the various exploration phases, only bor.ing 107 had indicated the loss of drilling fluid in the clay marl. Losses of drilling fluid had been reported in boring 107 during.1971 explorations in the clay marl at the top and bottom of this layer. The applicant has attributed the upper fluid loss to be due to leakage into the Utley Limestone formation from around the hole casing which had just been seated into the top of the clay marl and the lower fluid loss to be due to seepage into the more pervious lower sands. Because of the depth intervals where the fluid losses were observed, the. staff finds the appli, cant's explanations to be reasonable. ( e e e
www =,w w = . - _ -._= E that this program was intended to )btain selected representative samples of [ the earl stratum needs to be further explained, if it implies only good in 't' k core specimens were to be laboratory tested. (2) The s ff is unclear as to the significance in the appl.icant's response of indicating t six of the nine borings with zero recove~ry are located outside of the lim of seismic Category I structure / s. Certainly the borings are close enough to sa -related structures,.to/ reasonably permit extrapolation of the subsurface informati to these str0ctures. The staff recognizes that the applicant has made,similar ex apo tion of subsurface data in its assess- . ment of thi clay marl stfatum w ar/e it s relied on information /l rom caisson excavations and outcrop locations that exis considerably grea~ter distances from seismic Categoryf' structures. The staff als ecognizes that widely spaced borings maymot, in many instances, allow detect adverse anomalies, discontinuities #or lenses or pockets of unsuitable material and hat there is / an important need to respond to these indications, such as 0% reco e when it doe y /ecur, particularly at locations where leaching and solution cavit MSf-F54
- h he staff agrees with the applicant that the preponderance of subsurface information indicates that no open cavities exist below the top of theclaymarlstratum,Ihestaffislesscertainthatzonesofsoftermaterial I
do not exist in the clay marl. h c-h so.eer zones couia oe a rauwur....3.o.. ring p.we.1 i., th:.. wwwid h: ;;3aii;u..J.b h u-Lhou..iues usou .a ivuuu u un i edu! E. The staff attempted to gain confidence in the foundation adequacy of 3 the marl layerJ ap ;. vig. J:TT;uwl;:n f th th "CF ogongs, by reviewing the recorded settleme structures founded on "Ws.evoivateedsbTeemr+e 4 ov$od condvuom dyiM th. - th .t 'M c ay marl stratum &s l . $... 1'.'.13 Y. W..., h.... hjI : 'h.5 _O A'. H.... % W?0 6 ^'- ' ' ~ , mm..u n, ..m ...w .. uun..um.ut 4uildingara 4 ~""
- M +c ti? ret-a+ e ' r;;;. au.u u u. p..d Tw, so,y. rs of ni et :peratier.
f th-apprex4mately C7%-cf-the total oi.itu iording airesur 1 9 ead 50tthnr.t..wwr us iv.
- 3C';'
,,.. hov. usi,, i4eer, y.r;'id. un. 2.5 4.2 Engineering Properties 'of Foundation Materials 10/26/84 2-47 V0GTLE DSER SEC 2
__. = _==g .-~~-__ = j._. ge The types of foundation materials have been described in Section 2.5.4.1.2 of- / this report. The engineering properties of these materials were established \\- ,by laboratory testing and field testing and are summarized in the following FSAR tables and figures: Range of Engineering Static Properties for Site (Natural) soils (FSAR o Table 2.5.4-1) Engineering Static Properties Adopted in Design-Site (Natural) Sofis (FSAR Table 2.5.4.2) Engineer,ingDynamicPropertiesAdoptedinDesign-SiteSoils(FjA Table 214.12-1 and FSAR Figures 3.7.B.1-9, 3.7.8.1-10, 3.7.B.2-6, 3.7.B.2-7) On the basis of its review of the information provided in the FSAR, the staff concludes that the engineering properties determined for foundation materials are acceptable and conform with the applicable portions of the Commission's - regulations, the Standard Review Plan, and RG 1.138, " Laboratory. Investigations of Soils for. Engineering Analysis and Design of Nuclear Power Plants." The staff also notes, however, the extremely variable properties of undrained shear strength and soil modulus of elasticity for the clay marl that have been es' ablished in laboratory testing. It is N extreme variability that t the staff concern for the appropriateness of the adopted design values for undrained shear 4trength and soil modulus of elasticity (FSAR Table 2.5.4-2) and for the strain-dependent dynamic soil shear moduli and damping curves g2 d (Figures crns are... ?_.g and 3.7.B.1-9) for the clay marl stratum. The staff's con- - M r = d in this report in Sections 2.5.4.1.3, 2.5.4.4.3, and 2.5.4.4.6. i 2.5.4.3 Engineering Properties of Backfill. Materials' L A description of the, materials placed and compacted as Category 1 backfill soils has beert provided in Section 2.5.4.1.2. In localized areas that restricted compaction because of space lim'itations, lean concrete was used in place of backfill. The engineering properties of Category I backfill were established \\ 10/26/84 2-48 V0GTLE DSER SEC 2 .m.. m. ..._._,,_.,..,,.,,,.__.r..my. y .,y_, .,y,, _,,,__,_.,,,._,,y,_,7. 7 ,,.,_y
mm w cm l' E. ' r by la'boratory testing and are summarized in the following FSAR tables and figures: Engineering Static Properties Adopted in Design (FSA.R' Table 2.5.4-8)' Engineering Dynamic Properties Adopted in Design (FSAR Figures 241.12-1, 3.7.B.1-8,3.7.B.2-5) Test fills on Category 1 backfill were constructed to determine the appropriate lift thickness and the number of passes, and to evaluate the performance of different compactors in ' order to achieve the required maximum dergitiesr-The FSAR, as originally submitted, did not provide information on the actual ~ results from field control testing on compacted Category 1 backfill. In response to staff Q241.4 and to discussions at a March 1984 site visit, the applicant provided compaction control records for backfill material placed during the first 6 months of 1983. Following its review and evaluation of the field records, the staff expressed the following difficultg.s with the subitted .information. gsq NsTm ) $57(hMMbh5h$, (1) Many of the laboratory-determined maximum dry densitkt9dths4GER).- appeared unusually low. These low densities, when used to establish the percent compaction,,would result in the reporting of values in excess of y] og nd 01% de o)Wafdd den % bb monwm Ichoto. hey 100%. (2) The field procedures used by the applicant to demonstrate that fill placement moisture contents met FSAR commitments also gave a problem to the staff. The field procedures followed would consist of running a fill' moisture conterit immediately before compaction to verify that the fill moisture was within the specified, range of an average optimum moisture content that had been predetermined on stockpiled fill material. The 4 staff's problem resulted from the moisture testing of the fill before compaction and using.this result to decide on moisture acceptability 3 l rather than the more normal practice of testing the fill after compac-i tion. The normal practice of testing after compaction has the advantage ~ 10/26/84 2-49 V0GTLE DSER SEC 2 m--- ,,...--_._-r,_,_____ --.,.__r_--. _m,._-_,_w-.-__._-_,____,,,
. aw_ (
- a..
of " verifying the uniform mixing of' water t'hroughout the entire lift thickness, which is required by the compaction control specification. (- Also it is the compacted condition pf the fill (density and molding water content) which will govern the resulting engineering properties. The applicant's procedure of using an averagetoptimum moisture content also presented a problem to the staff because it differs from normal procedures where the optimum moisture is directly estabitshed in the lab on the same type of material that is field tested for density. In order to address the staff's concern with compaction of Category 1 backfill,
- a. confirmatory laboratory testing program was agreed upon with the applicant--
and testing was initiated in June 1984. The major objectives of th.e confirma-tory testing program consisted of the following: (1) Evaluate the acceptability of the quality control test procedures and test results for the compacted Category 1 ~ backfill by determining whether FSAR commitments (Section 2.5.4.5.2) had been met in obtaining the required maximum dry densities. The check on acceptability was to be made by ( requiring both the field laboratory and an independent te' sting laboratory i to perform control tests (gradation, moisture-density relationships, rela-tive density and permeability) on the same Category 1 backfill material. 3dentical samples of fill material were selected from existing stockpiles.,he pe4 Mc<tsbrses. -(2) Reexamine the FSAR commitments on compaction control (maximum dry density and placement moisture contents) after evaluation of the results from the confirmatory testing program and determine if modifications of FSAR commitments are warranted for the future control of Category I backfill that remains to be placed. Tha laboratory results of the confirmatory testing program were,p%h n rovided to the NRC in an August 10, 1984 submittal. Theapplicantimealsosubmittedalbe j report tgtgNRC bated.Septemb r 27,7m%w%%ch evaluates the testing
- program t
1984 hi m .e4 )ae-
- rse i@m H --' 'xs W cq) p 3 it:s+si.hPreliminary observations of the d
i pr:;;r:- on the results provided in the August 10,1984,submittalindicatelthefollowing: ( 10/26/84 2-50 V0GTLE DSER SEC 2
m2 z._ 7-mm
- s.
(1) A comparison of the maximum dry densities dete ined by the field laboratory 1 and the independent testing laboratory indicate that the independent '\\ laboratory results show higher valueSof maximum densities in all of the 12 tests performed using ASTM D1557. The increase in densities ranged from 0.8 lb/ft3 up to 3.5 lb/ft. The maximum difference in dry density 3 from the loosest state to the densest state for the N 9fum to fine sand (SP) is about 20 lb/ft. 3 The differences in results between the testing laboratories for optimum moisture content determinations were more widely scattered--differences ranged from 7.6% moisture below optimum to 2.5% above for the tests on the same type of material. g, g .4 (2) Thetestresultsalsoindicatethatthebackfillhanswhich'haveasmall p amount of fines'(less than 6% pgng 2 highest densities when testedri-t J=......gg a ghe,ig '.;nfAT".MGQ y in six of the seven tests performed. The increase in maximum dry densities between W i;d f ar: m-gASTM D1557 71.. m = = r. $ ASTM D4253{ testing 3 range $ f$o}m Y 4 1$ $tgh 3 up to 4.5 lb/ft.3 Recognition of these results would encourage a modift, cation . to current control procedures that recuires the running of both en ATTm M'253 T Jc % u... ?.., - and +"Trn D 555 f' O. Ns m"- 2-w in crder + gb sg f the maximum dry densities and percent compaction for this type # ='"f O e .;hid Sn '" <-=" _.aatoi eines. The opportunity for the staff to obs'erve a portion of the actual testing by the independent laboratory has helped the staff understand why higher densities are not more consistently obtained in the ASTM D1557 test. During this labora m PCTM Dis test a large part of the heavy compaction effort that is specified 1s actual y 3 lost during compaction, because of the large shear displacements which repeatedly Cccur in the test sample mold under the impact of the hammer weight. These displacements and resulting' loss in compactive effort appear to be greatest for soils being compacted at moistures on the wet side of optimum moisture The staff believes the large displacements and resulting loss in content. compactive effort are a maj6r reason for the differences in test results between the field laboratory which was indicated to use a mechanical tanper, and the independent testing laboratory, which manually compacted the test 4 10/26/84 2-51 V0GTLE DSER SEC 2
. m_.. a x. sp:cimens. Undei manual compaction conditions'there is a natural tendency to-I locate the next hammer blow where the displacements are occurring, whereas in mechanical tamping, a set pattern and sequence in compaction e'ffort is followed. Th] differences in results between the two testing laboratorie3ptherefore, are d more the result of the particular soil behavior under the specified compactor and allowable test procedures of ASTM D1557, rather than the result of errors cr unacceptable test procedures between th laboratories. The staff also believes the densities obtained in t..:fqDn D, n rc...i= &n ity tests are higher because AsTv.3 % % the displacement.s do not occur and that thehe!4va damity test is better [T 1 suited for Vogtle backfill meNutr with less than 6% fines. Tha -*=" ***cfp:te;-u..i :n-th: :pp!< m +'= futura repor>_: which addresses-ahe nhiectivae nf the--confi.maLurf-teste-program 3-the higher-maximumMry-4 :n:tti:: Obteined, fvr.he in ci-t-yp:: O f-bac-k-f-i-1-1-ma teri a 1 s-teste d,' w M 1-b e -d *e ::t-ablish inc gc. won; ww..@actien=fcr-a4-i--Gategory-1--backf14 Frempacted -te J. Lc. Pi ci iminary ooservasiuna, when saing-the-h+gher-denstt-ies--for-the field rece,d2. rum 1.ne first six months-of4 983, indfeate-that-fSAR require-ment: -h ve essefrthi4y-been-met-but--at-lowen-percent-compaction:valwa Lhan' ~ deiticaHy reported. 2.5.4.4 Foundation Stability With the exception of the CW towers, the instrumentation cavity of the i:entainment buiking and the auxiliary building, all seismic Category I struc-(tures are founded on compacted Category I backfill. 'T%.l. c..M r.-fe ~ E e~2teff c e=L.wc. 02'1.17 indie.Lua u..; i; Liic t u. Li ca.e n s v. nw 4... pre tly available for the auxiliary feedwater pumphou Tesel generator ,- buildings, 1 (uel oil storage pumphouses and Ca cry I tanks since these structures are either initial stages o onstruction or construction has not begun. Also in response . ques on Q241.17, the applicant indicates that settlement records for the CdoDsN Category I tunnels are to be submitted to the NRC. Un this construction is com eted and described and th] settlement dat eprovidedtotheNRCforevaluation) taff would n t be in sition to complete its final report on foundation stab tys rTl~tnis is an operi"Treia. t 10/26/84, 2-52 V0GTLE DSER SEC 2
- p..
b ~* i INSERT 2 In1the September 27, l'984 and March 11, 1985'submittals, the applicant provided the following observations and conclusions on the results of the confirmatory test program:
- 1..A relative density of 80 percent as established by ASTM D2049 (Relative. Density of..Cohesionless Soils) was the basis in the PSAR design stage for establishing dynamic soil properties to be used in the evaluation of resistance to liquefaction. A relative density of 80 percent for Category 1 backfill materials resulted in a margin of safety against liquefaction of 1.9.
2. While ASTM D2049 was an appropriate laboratory test for determining the density of the clean sands, which sands proved to be the controlling backfill material in the liquefaction analysis, it was also acknowledged to be a less appropriate field test at Vogtle 'because ASTM D2049 type test apply reliably only to clean sands andtheVogtlefilliscomposedofcleansands(approximately 23 percent of the fill placed to date) and silty sands. In the applicant's opinion, ASTM D1557 is clearly the appropriate test to l control compaction of both the clean sands and silty sands. 4 3. To compare the maximum densities which were established by ASTM D
- 1557 test procedures with maximum densities established using
. vibratory methods in the labs, the applicant' initially made a 1 + ,--w..-wne.- --.w.n~_e,e,,.-,,_,-,.,.,me.,._,,-..-_nn_,.._.---m., ,,.n. -.,,--,,,.,m.n., m,,,-,m
m-,.+-- .., ~ ~ .-.._ ~ _ _ _ - - n - - - -.. = 2 reduction of 2.2 pounds per cubic foot from the maximum densities established in ASTM D4253 tests.'.The ASTM D2049 test method had been discontinueo in 1983 and.had been replaced with ASTM Test Methods D4253 and D4254~(Minimum Index Density of Soils and Calculation of Relative Density)'. The major difference between D4253 and D2049 is that ASTM D4253 allows variations in the double amplitude of vibration whereas D2049 required a single maximum double amplitude of 0.025 inch to be used for determining maximum density. The reason for allowing the variations in double amplitude in D4253 procedures is the recognition that the range of vertical vibrations had been shown to. have significant effects on the values of maximum density obtained. The basis for the 2.2 pounds per cubic foot reduction.,results from the applicant's testing of one of the clean sand samples from the confirma-tory program under both ASTM D2049 and D4253 procedures. The applicant's conclusion, after reducing all 04253 test.results by 2.2 pounds per cubic foot to produce the equivalent maximum densities of D2049, is to indicate that maximum densities determined by ASTM D1557 are approxim-ately the same as ASTM D2049 densities, therefore, ASTM D1557 testing is appropriate for field compaction control of all Category I backfill. 4. For backfill sands with less than 6 percent ines (material passing the200me,s.hsieve),thedifferencesbetweentheindependentlaboratory
- and the field laboratory in optimum moisture content results exceeded the precision limits specified for ASTM D 1557.
A
_a_ x w- + 2 . -. s . (Noexplanationforthisexceedancewasevidentfromtheresultsof the confinnatory test program). '. 5. In its evaluation of test results from the confirmatory test program, the applicant has concluded that Vogtle's Category I backfill soils (both clean sands and silty sands) are relatively insensitive to moisture content in achieving the required compaction, as evidenced by the flatness of the compaction curves. As a follow-up to this conclusion the applicant has proposed tne following revision to FSAR paragraph 2.5.4.5.2: "In accordance with the earthwork specification, Category I backfill is sand and silty sand with not more than 25 percent passing the U.S. No. 200 (0.074 mm) sieve size. The sand and silty sand materials actually used to data as Category 1 backfill consist of less than 15 percent passing the U.S. No. 200 (0.74 mm) sieve size. The laboratory compaction curves for these materials used in the backfill are relatively flat and indicate that 97 percent compaction can be achieved over a wide range of moisture contents. Therefore, because of the insensitivity of these sands to variations in moisture content, a broad range of moisture' content is acceptable for reaching the specified density. However, a target of 3 percent below to 2 ~ percent above optimum is specified as a construction aid to facilitate compaction with the understanding that a broader range is acceptable provided the required compaction is met."
.a_ = ,. u -. 't _4 6. In response to the staff's concern expressed on page 2-49 of this report that many of the maximum dry densities established in the field lab using ASTM D1557 procedures appeared unusually low, the applicant recomputed the percent compaction of fill soils tested between May 1980 and December 1984. Recalculations were completed after initially increasing the field laboratory densities by the largest differences obtained in the confirmatory test program between the ASTM D1557 results of the independent laboratory and the field laboratory for material with fines more than 6 percent and between ASTM D1557 and ASTM D4253 results for material with less than 6 percent fines. The increases to field lab densities were respectively 3.5 pounds per cubic foot (pcf) and 4.5 pcf. The reported results of the applicant's recalculations are: a. An average percent conpaction of 100 percent was obtained based on the results from 10,262 field density tests. An FSAR commitment requires an average of 97 percent compaction. b. 86.3 percent of the tests exceeded 97 percent compaction. c. 9.4 percent of the tests were between 95 and 97 percent compac. tion. 9
... - _ = _ - _~ ~ d. 3.8 percent of the tests were between 95 and 93 percent co.npaction. The FSAR requires that not more than 10% of the tests be between these limits. e. 0.5 percent of the tests (a total of 52 field density tests) were less than 93 percent compaction. The FSAR requires no tests be below 93 percent compaction. The applicant has concluded that FSAR commitments on compaction control have been exceeded, even when the very conservative addition of the largest differences established in the confirmatory test program have been made to field laboratory results. Based on the staff's review and evaluation of the results of the confirmatory test program provided by the applicant, we make the following observations and conclusions: 1. The confirmatory test program completed by both the independent laboratory and the field laboratory was well organized and properly conducted. The results have resolved most of the staff's concerns and have provided the basis for th'e positions which are subsequently given. w
- w. -
a w. . p -- c.------._. ~. 2. The staff wou!d agree with the applicant that FSAR commitments on compaction bntrol of Category 1 backfill have been met or exceeded based on the results of the recalculations given in Item 6 above. The staff does plan to audit the recalculations at a future date to verify the accuracy of the results and we consider this verification to be a confirmatory item.' A major reason for this verification effort is the recognition of the sensitivity in the dynam,1c engineering properties of the clean sands with only small (e.g. 2 pcf) variations in densities. 3. The staff does not agree with the applicant that ASTM D1557 is clearly the appropriate test to control compaction of the clean sands and this issue is an open item. The staff acknowledges that it is easier and simpler for the field to control all Category I backfill with one test standard, but the test results and observed behavior for the clean sand under ASTM D1557 test procedures indicate that maximum dry density is best determined using D4253 test procedures. It should be recognized that applicants at many of the nuclear plant sites where plant fill has been compacted have controlled compaction of cohesionless soils using the maximum density obtained in either I the relative density test or ASTM D1557 test. Both types of tests have been required in field control of compaction because of the uncertainty in establishing which test would produce the maximum ' densities for soils having.up to 12 percent fines. The basis for the i s I r ~ we e --e -e ,mme----- v-s----+ewv--
a_m 7 p.. ~ .. staff position which is subsequently given in paragraph d. fer requiring colnpaction control of the clean sands using ASTM D 4253 procedures includes the following: a. Figure 2. through 2. of this report present lab compaction test'results by the field lab using D1557 procedures which were submitted for the confirmatory test program in the applicant's September 27, 1984 submittal. The percent fines for the tested materials presented on these figures ranged from 2.6 percent to 5.5 percent. An examination of the plotted compaction curves reveals that the impact compaction procedures of of ASTM D1557 did not produce a well-defined moisture-density relationship curve' which would normally be anticipated and the densities 3 produced by the tamping method of ASTM D1557 varies in a very narrow range and are below the maximum density established by the vibratory method of D4253. It is this resulting curve and the observed displacement behavior during compaction which would lead one to conclude that vibratory compaction using D4253 is the more appropriate test to establish maximum density for Vogtle's clean sands. b. It is the staff's opinion that the final density obtained under ASTM D1557 test procedures is not a uniform density throughout the 4
_=a. w 1 '\\. ~.. ~ compaction test specimen but rather a random value that is significantly influenced by the irregular pattern of displace-ments which were still occurring at the completion of the test. c. In past meetings with the applicant and its consultants an argument had been offered which stated that what is important in compaction control of soils is to produce a fill with densities and engineering properties that were considered in design and which were shown to have an acceptable margin of safety. The staff agrees with this argument but believes the applicant's insistence on using ASTM D1557 to control all Category 1 backfill is a significant inconsistency in this argument. It should be recognized that dynamic engineering properties of plant fill were established on test specimens that were compacted by vibratory methods at optimum moisture content. It should be further noted that actual compaction of fill in the field is being performed with heavy vibratory rollers. It is the staff's opinion that the tight requirements established by the applicant on fill placement (maximum 6-inch uncompacted lift) and moisture conditioning of the fill (the required addition of water to bring moisture within 2 percent of' optimum moisture content) are th,e fortunate reasons why the applicant has been able to demonstrate that FSAR commitments have been met in spite of the problem of using the tamping method of ASTM D1557 as the control
, _:i;w .e . ~. _ , =,, x_. = '.o. ~.. - ~ test standard for the clean sands. d. In ASTM D4253 test. standard, the maximum index density is defined as the reference dry density of a soil in the densett state of compactness that can be attained using a standard laboratory compaction procedure that minimizes particle segregation and breakdown. This same definition is applicable to maximum dry densities established by ASTM D2049 test procedures. It is the staff's opinion that the improvements in guidance in the more recent ASTM D4253 test procedures over ASTM D2049 procedures have evolved from the learning experiences with ASTM D2049. The effects of variations in the double amplitude of vibrations en maximum density had been recognized and was being required long before the formal publishing of ASTM 04253 in 1983. The Corps of Engineers Laboratory Soils Testing Manual (EM 1110-2-1906, dated 30 November 1970) recognized the need for varying the displacement amplitude and required variations in amplitude be studied to produce the maximum density in the relative density test. For these reasons the staff does not agree with the applicant on the reduction of 2.2 pef in order to to produce equivalent ASTM D2049 test densities. 9 8
--m - _ - ~ ~ , = ow _ - ~ 10 For contro1 of the remaining backfill to be placed at Vogtle the staff will require determination of the maximum dry density using both ASTM D4253 and ASTM D1557 test procedures for fill materials with 8 percent or less fines passing the 200 mesh sieve. The variations in ASTM D4253 (wet and dry methods, range in double amplitude of vibrations) should be run and evaluated until.such time that initial test results demonstrate the procedure which consistently results in the highest maximum density. The applicant may request a reduction from running both tests and determining the effects of the variations of ASTM D4253 - procedures, after sufficient initial testing has been completed, by submitting a brief evaluation report with supporting data for staff approval that clearly demonstrates maximum dry densities are being consistently established by a single standard test procedure for the Vogtle clean sand backfill materials. The staff is aware of work by Geotechnical Engineers Inc. (GEI) which has lead to the conclusion that ASTM D1557 is appropriate for compaction of clean sandy soils. The staff rfotes, however, that in their study, GE! recomends modi,fication of the normal test procedures of ASTM D 1557'by requiring the addition of water for each layer being compacted in the test mold to bring the sample to near saturation and also places a requirement for compaction of a full soil layer in the collar of the compaction mold. The e
__,:2' 1.f,. '.. _ 1 _11 staff would censider acceptance of the proposed GEI modified version of ASTM D1557 to control Vogtle't clean sand till, if test results using this modified method resulted in maximum dry densities comparable to those established using D4253 test procedures and were shown to be reproducible using the GEI modified test procedures. 4. The staff is not in full agreement with the revisions to FSAR paragraph 2.5.4.5.2 on moisture control that are proposed by the applicant. Because the earthwork spe'cification for Category 1 backfill permits up to 25 percent fines, there continues to be a need for a specified moisture content range about optimum for Category 1 silty sand backfill with percent fines between 12 and 25 percent. The staff believes based on telephone discussions with Vogtle field personnel that compaction of both the clean sands and silty sands with up to 12 percent fines has been facilitated by the addition of large quantities of water to produce near saturation of the' lifts imediately before compaction with the vibratory roller. The staff could agree to eliminating the FSAR specified moisture content range for soils with less than 12 percent fines if a written comitment in the FSAR were given to continue the addition of water to p,roduce near saturation of the lifts imediately before compai: tion by the vibratory roller. e 0 e *
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zw 2 w .n+ amw-g', ya.. ~ INSERT 3 In response to staff question Q241.1, the applicant has provided information on the foundation design of the radwaste transfer building, which is located adjacent to the Auxiliary Building, and the radwaste transfer tunnel. The radwaste' transfer building and tunnel are founded on Category I backfill with the bottom foundation elevations ranging from el. 216.5 ft. to el. 197.2 ft., respectively. These structures are not seismic Category I, but have been conservatively designed to withstand the safe shutdown earthquake (SSE) because of their proximity and potential to impact seismic Category I structures. The staff's assessment on foundation stability of plant structures and conduits, which is given in the following paragraphs, is intended to cover the foundations of the radwaste transfer building and tunnel. e G e 8 e
w c_ _~ m n =v +. ). TS @. M=t '- e 7---- +a 0241.1 indicates that the vnd= rte tr c,.~.. T,g radwaste transfer tunnel, although not seismic Category
- ctures, I
could p tially adversely affect seism.ic Catego eructures. On the basis of guida in RG 1.29,'the foundati sign a~nd construction of t'hese structures would also ngrequi to meet the equivalent of seismic Category I requirements. Th un,ation design and constrdction information ~ for th'ese structure's d, therefore, rovided for staff review. The applicant is asked to provide the reas'ons sthe radwaste solidification buildi hich has been founded on drilled caissons fer-o staff ques- / Q2M.20),- wouid iot"ftTT 1'n'trthis same<ategory : TML Wmm an-a_- a ^ j 2.5.4.4.1 Construction Notes detdopej or During the years of plantgenstguction, ggn g ag g a g g ent that couldaffectthelong-termstabilityofSt....:=d;;.Re% These conditions 3 included (1) detecting the solution cavities in the Utley Limestone above the clay marl stratum M ... and (2) detecting the E m erosion of already placed Category I backfill (November 197,9) that resulted .from th.e heavy rainfall and surface runoff that entered the foundation excavation area. Sgybn aegne} 30in Q$56.M.M 3 noch k4ssuerlcavitiespe# ::.c' q .a were exposed on theaslopeT of the foundation exca'vation where the slopes intersected the limestone shell bed. The largest cavitywasQocatedonthe.northwestcornerofthepowerblockareaandmeasured 10 ft x 10 ft at the opening and extended approximately 30 ft back into the slope where it narrowed to a small size. Wh-rm " ~ ' tin... u n..tu d uv. i.;. is [:.. p;.ecf bhd f=t M;; e varvina 4-+e~gh;;;.ivus$nc w en.. :... r T:e cavities were cleaned of looss debris and then backfilled j with crushed rock. 'In the larger cavity the crushed rock was forced into the s opening for at least a 25-f,t le,ngth beyond the o'penin by m attached to wt CY YOCC the blade of a bulldozer. The cavities were filled to prov e a buttress on 3 the foundation excavation slope against which structural bailfM1 could be placed and compacted. 3 \\ 10/26/84 2-53 V0GTLE DSER SEC 2
y _.a. w :. 2- - - c_~ c =. w. ....j, i L x Th3 areas affected by the soil erosion problem that became evident in / hvember 1979 included zones (1) between the electrical shafts of the control building for. Units 1 and 2 and the turbine building, (2) between the containment buildings for Units 1 and 2 and the electrical punnels, (3) ahng the perimeter of the Unit I containment building, and (4) under the mud slab of the Unit 2 tendon gallery. The applicant provided detailed descrip' ion of the areas t affceted and the remedial measures completed in the August 15, 1980 report, " Final-Report on Dewatering and Repair of Erosion in Category 1 Backfill in Power Block Area." The limits of the areas disturbed by erosion were determined by inspection, field explorations, and testing, using proving ring and dynamic coh penetrometers and sand cone density tests. The remedial measures domplete~d b2 tween January and August, 1980, included (1) reshaping foundation e_xcavation slopes and protecting them with gunite, (2) improving surface water controls, (3) installing additional dewatering measures and piezometers to ensure that .the level of the water table was deep enough in the Category I backfill to allow mud slabs and disturbed soils to be replaced under dry conditions, and (4) pumping grout into voids in the backfill in1 space-restricted areas. -The staff concludes that the applicant's investigations and remedial measures are acceptable, but also recognizes that the success of these actions in cddressing the erosion problem and filling the cavities can best be judged by ' continued visual inspections of the structures' performance and long-term settlement behavior. T:.. s w. T T..b unu . / k=. lece m oil udlvrvmr 6.i. n *th: :...pprug.;;t " #h"~ -d i ;; rib: th;; :. .no ur sneir axten+ me 4 a#a-et4">nt>1d m aus e Ni ahcr. '-tar; ::tt4ement., cuoras tJ i # . ' is u aeVi2^4ed. Iii i a~~ii a w wis 7aasu. w.,, it;&, l L -2.5.4.4.2 Bearing Capacity Th2 applicant's responses to staff questions.Q241.5 and 241.15 indicate that ~ .the results of bearing capacity analysis under static and dynamic loading will ' be submitted to the staff by December 1, 1984. The staff will complete its ( 'e safety ey'11uation on the acceptability of the resulting margins of safety .against bearing-capacity-type fa'ilure after it reviews the' December 1984 submittal. This item is open. L. 10/26/84 2-54 V0GTLE DSER SEC 2 a
t c_ _ - a s - l~.. s m m 2,.. J',. ',:. l? ~ q 2'.5.4.4.'3 Settlement The ap cant r onded to staff que.stions Q241.17 and 241.18 by providing [y_ ytt:
- f ^.h. settlement records for seismic Category I structures.( The v,i on ve
.;.13_....-- _u. i, m +.--g t; u ;2,,,m.d ,; acat,,ue ta-: mica 22.u o f w i...yvi o. L ntth,.ai ruert previded in.e wuse w Q241.17... irr r ~ a ':r; i. hat .k. ..,i w ud avslu.Liwu di TTisult-- Titt WpTTcVNt7veds--te ee n..... +4.. e,21. far nin++4a-6t: an ;;;;h,,,.n, ~ y,rou,,;,,,.ug ro, g;o L L ;,,p %Am+= nn annlication af la"dina Ytre'. : tic:. Only-thea e.+t1.===+ 5ehwiec - -=n una., <+-n-+n. leadjpg be reasonatfly.voi..ted. S'xh e provemeni is important he f"+" : -tt: : t itt; h.ad....noi.; L.fe.. th ;t;TT won d gid wu i wn3 6.. -:-tti.....n 6-moni to r i ns..q ui......a t: foe the--Techn4 cal-Og i'ic 3th. " he epplicon 's reply to staff question Q241 18 is not acceptable for the/ fo wing reasons: (1) Cont to the staff's request there is no discussion o comparison between tota nd differential settlements allowed for in esign and actual settlement recor at specific locations of structures The applicant's statement that, bec se all major seismic Category structures are separated from each other by se mic gaps they are unaffe d by differential settlements, fails to recognize that l essive settlement can cause (a) unacceptable e cracking in structures (b) h h stresses d unacceptable tipping of the 3 structure. The staff's review the imited settlement records that the applicant has provided indicates t t values of total settlement lar'ger than the upper predicted; values have een -orded for certain settlement markers (Nos. 128, 133, 134, 234, 2, 323, 324,
- 5) at the auxiliary building and Unit I containment build g.
These larger s tiements have been recorded with approximately 87% of e total static load appl in comparison to the pre-dicted values whic were estimated for. the 40 yr p1 t life. The discussion requested in st f question Q241.18 asked that the spe fic maximum recorded 4 settlements , identified for each structure and be compa d to design estimates. The recor d settlements may have potentially significant an dverse impact to fu re structural performance (e.g., cracking, high stresses). The larger ~ 's Jements being observed may be tne resuit or tne clay mart stratum 10/26/84 2-55 V0GTLE DSER SEC 2
~ .a. _ = _:-,x a u. e .~.e s.
- ~
a.. .n ,qw. INSERT The information and data provided by the applicant included: 1. The locations of all settlement markers (FSAR Figure 241.19-1 and 4 drawing no. AX 2055V001). '2. Graphical plots of recorded settlements versus time for seismic Category I structures up to the start of 1985. 3. Tables and graphical plots of load histories for major Category I structures. Not all Category I structures were presented because some structures are in the initial stages of construction and little / x . settlement or load history datadre presently available. Ib 4. Graphical plots of foundation soil heave data developed prior to and during the major foundation excavation work. 5. The location of representative piping systems at interfaces between structures for Unit 1. Unit 2 is a mirror image of Unit 1 but differential settlements pipinginterfaceshnotaddressedas-s construction at Unit 2 is not as advanced. 6. Typical calculation for determining differential settlements at the interface of structures having piping connections. 7. A table listing the calculated val es of the maximum recorded slope of a . settlement (deflection) curve across a given structure foundation. A figure illustrating a typical calculation was also provided. t f -- s h m00 eismic Category I structures whose construction is near completion and f whose full design load has essentially been imposed include Unit I reactor containment building, the fuel-handling, auxiliary and control buildings. The other structures are in various stages of construction with the reactor containment building for Unit 2 appearing the next
z e- = q mx. 2 pw_w2nc gu. Y.* p ~ clasest to reaching full design load intensity. The largest total settlements recorded up to early 1985 have been measured on the nuclear s rvice cooling water (NSC ) tower IA (3.4 inches), reactor containment building for Unit 1 (3.2 inches) and the auxiliary building (3.1 inches). O The NSCW towers and the auxiliary building are founded directly on chay A 3 marl layer. The reactor containment building foundations are primarily .fcunded on 28.5 feet of Category 1 backfill placed above the top of the ' clay marl layer with the exception of the reactor cavity which extends to the top of the clay marl. The reason for the larger total settlement at NSCW tower IA is not readily apparent when it is recognized that the loodmq actual imposedpinten'sity to date for this tower is approximately one-half the loading of the auxiliary building and only one-half of the tower s 4 final design loading pressure of 10.25 kips per square foot (KSF). s The total settlement recorded to date for three of the NSCW towers, for 4 both Unit 1 and Unit 2 reactor containnent buildings and for the auxiliary building have all exceeded the or yinal FSAR maximum predicted U settlement estimates by val)(es ranging from 0.35 inches to 1 inch. In x FSAR Amendment 15 and in a December 28, 1984 submittal on settlements, mducci4 abbdes ll,c tha applicant has revised previous FSAR settlement estimates andAtt:r 1arger than anticipated total settlements to elastic settlements in the lower sand stratum below the clay marl. The staff does not share this ~ opinion and believes the original FSAR position by the applicant, that ssttlements in this very dense lower sand stratum would be negligible. is th0 more realistic and correct position.
w x ~. ~ p S, .- Th] staff concludes that the larger measured settlements are due to M. ^ whn.h b compressibility of the clay marl layerfgreater than originally adopted in g d; sign. Rigardless of the cause for the larger than anticipated settlements, th? applicant has concluded in response to Q 241.18, that the settlements have ,br.n relatively uniform across a given structure foundation. The basis for this position is the applicant's calculations which compare the calculated maximum net slope of the settlement (deflection) curves across a given 0 structure foundation with published criteria for tolerable valies on this same slope ratio. The slope of the settlement curve is determined by dividing the maximum differential settlement between two measured points on a structure by the distance between the points. Published engineerina criteria for the naximum slope of a settlement curve for multi-story, heavy, concrete rigid frfre,- structures on mat foundations, similar to the structures at the Vogtle site, indicate a ratio of 1 inch differential settlement over a distance of approximately 670 inches would be a tolerable slope for settlement before i s".rviceability and functioning of the structures would be impared. At Vogtle, s th] applicant has calculated maximum slope ratios (differential settlement / distance) of 1/1490,1/1605 and 1/2680 for the control, fual handling and 9. auxiliary buildings, respectively. Based on thes, calculated ratios, the s applica.nt has concluded that differential settlements are presently well within industry accepted tolerable limits. The applicant plans to reexanine s:ttlements of Category I structures when construction is complete or when future settlement data would indicate the need for reexamin)dation. The x applicant also concluded that unless differential settlements across a
- = m,w a- , q... ' legcompetent than orginally anticipated or mav nacen'y M - '"- ~ ^- soi1*e, ton problem. i-oundation design modifications may arranted. The engineer g basis for the response to th,e above consi ations should be clearly des bed in the applicant's response. (2) The applicant's response to 8 provides general information on differential settlements whic re typica addressed in the design of seismic Category I piping but t esponse needs to be leted by providing information sp;cific for Vogt ocations, sectional views where repriate) showing where total differential settlements have been recorded discussing the signif nce of these settlements on the piping system's capabilit dafely-- g...t=L th;.T. ~ '2.5.4.4.4 Lateral Pressures The walls of seismic Category I structures below plant grade el 219.5 ft were designed for static loading to resist at rest lateral earth pressures using the equivalent fluid pressure concept. The adopted design pressyre diagrams e I are presented in response to Q241.21 and are discussed in FSAR Section 2.5.4.10.5. i. An at-rest lateral earth pressure coefficient of 0.7 was used in design for the backfill materials. A water level at el 165 ft was conservatively used to establish the hydrostatic pressure contribution to lateral pressures. For dynamic load'4 conditions,.the Seed simplified version of the Mononobe-Okabe method was used for active earth pressures. Dynamic passive pressures were calculated using a method by Kapila that is based on the Mononobe-Okabe method. A peak horizontal ground surface acceleration of 0.20 g was used for SSE ~ . condition to estimate inertial forces. Ytth the exception of the approach for dynamic passive' pressures, the staff concludes that the methods used to estimate lateral earth pressures are conserva-tive and acceptable and are in accordance with current state-of-the-art engi-nearing practice. The staff needs to complete its evaluation of the method us:d to establish dynamic passive pressures. This is a confirmatory item. y e 9 10/26/84 2-56 V0GTLE DSER SEC 2
m_mme~ x- =_ _ h= 2.5.43.5 Liquefaction Potential - The applicant's decision to remove all the upper sand stratum materials and excavate the power block area to the top of the clay marl stratum has the' significant advantage of eliminating the_ potential for liquefaction' that was indicated for the upper sand material. The staff agrees with the applicant that neither the clay marl stratum nor the deeper, dense lower sand stratum is susceptible to liquefaction under SSE conditons assuming a peak horizontal ground surface acceleration of 0.20 g. TodemonstYate'thatanacceptablemarginofsafetyagainstliquefgtion.i.s. available for structures and piping founded in the Category 1 backfill, the applicant conducted cyclic shear strength tests on a representative range of backfill materials which were compacted to 97% of maximum dry density determined by ASTM D1557. The lowest factors of safety against liquefaction type failure were on the order of 1.9 to 2.0 using the Seed and Idriss (1971) simplified method and the cyclic test results. The staff concurs with the applicant's findings that an act ptable margin of safety against liquefaction potential does exist for Category I backfill that j is compacted to 97% of maxium dry density. The staff plans to reexamine this conclusion on liquefaction potential following resolution of the concern discussed in SER Section 2.5.4.3 on compaction control procedures. This is'a portion of Q at open item. ,-Qe Sm,y,qe %2-%W'?'"'C Y'P"W 5 nar\\ 1 L 2.5.4.4.6 Dynamic Loading In staff questions Q241.11 and 241.12 the applicant was asked to provide the soil properties (shear modulus and damping values) for the soil springs used in the finite-element and lumped parameter dynamic studies and to compare these properties with the results from field geophysical surveys and laboratory cyclic triaxial testing completed for Vogtle. The staff has reviewed the applicant's responses and.co'ncludes that: 10/26/84 2-57 V0GTLE DSER SEC 2 - -. -. ~
. ~ %ym. ww %. - -_ m u. = ~ (1) The strain-dependent soil damping curves for the compacted sand backfill ~(FSAR Figure 3.7.8.1-8) and the lower sand stratum (FSAR Figure 3.7.B.1-10) [- ara reasonable best estimates and are acceptable. The staff does not under-stand the basis for the' change made by the applicant for the damping curve (FSAR Figure 3.7.B.1-9) for the clay marl stratom between the time of FSAR d:cketing and Amendment 6 (May'1984). The staff requires that the applicant provide the basis for this change and include a comparison that permits evaluation of the effects on structure behavior (e.g., reponse spectra) when bath soil damping curves are used in design. This is a part of the open item discussed previously in Secti.ons 2.5.4.1.3, 2.5.4.2, and 2.5.4.4.3 (item 1 in Tablo 2.2). ft (2) The staff finds' the strain-dependent shear moduli curves (FSAR Figures 3.7.B.2-5 through 3.7.B.2-7) to be reasonable best estimates. The staff r; quires, however, that the results of the study which varied the soil shear c:duli values (discussed in FSAR Section 3.7.B.2.4.J); by a factor of i 1.5 be provided and discussed, and that the results permit a comparison of the resulting response spectra with final design spectra for the range of shear / I moduli valyes considered. The acceptability of variations in soil dynamic \\. properties is an open item. 2.5.4.5 Instrumentation and Monitoring Because of the primary importance of the groundwater regime in the solution process and the resulting potential for ground subsidence, the staff will require adequate monitoring of both groundwater levels and settlement during tha life of the Vogtle project. Tha observation wells which will be active as indicated in the applicant's resp:nse to Q241.10 need to be supplemented with additional wells closer to tha cain plant complex and be located in both the upper water table aquifer and in the clay marl stratum at representative depths. The staff requires that the applicant provide a plari which locates, as requested, the additional wells and the pertinent information on well installation and monitoring that is. requested in Q241.10. i 10/26/84 2-58 V0GTLE DSER SEC 2
=q w -m VA. c- _ AMSF)M' t %.*en r< g.e -1: rift::ti f th; ;;;1'-f,-+8 h e,a wense to 0241.19 as-t$ her it is intended that all settlement markers shown on FS d 241.19-1 w monitored for the entire life,of the Voggt1he staff feels it is are to initially important to resolve' the issues id,pt1 Tied in Section 2.5.4.4.3 of this report, partic hav_e t d understanding of the significance of the settlements which h a rea y oc cred. Following resolution of the concerns expressed action 2.5.4.4.3, the st'affswould be in a better posi-tion to ev e the applicant's proposal fo' r long-termlettlement monitoring. N .o-awem. . 7, n wean 2.5.4.6 5emaining Issues ft ~ The remaining operating license safety review items which have been identified and discussed in the preceding SER sections are listed in Table 2.2. 2.5.4.7 Conclusions On the basis of the staff's review of the information provided by the applicant in the FSAR, the staff has concluded that the following features of foundation stability are acceptable, except as impacted by items in Table 2.2. (1) site investigations (2) engineering static properties of foundation materials (3) foundation preparation measures including treatment of cavities (4) methods for estimating lateral earth pressures (5) margin of safety against liquefaction potential (6) engineering dy'namic soil properties Final c:nclusions on plant foundation stability requires resoltition of the . remaining issues identified in Table 2.2. 2.55/g56 The site conditions which exist at Vogtle do not involve stability 1 pes nor embankment ane dams; therefore, these are not addressed. ~ / i 10/26/84 2-59 V0GTLE DSER SEC 2
m_- f.4k-msan W d In response to staff question Q 241.19, the applicant provide, the following w information on the long-term settlement monitoring program. 1. Settlement monitoring of the markers shown on FSAR Figure 241.19-1 will be completed at its present 60-day interval rate through completion of 5 construction and the first year of startup of both Unit,1 and 2. Some x additionN, subtraction and relocation of settlement markers may occur during this period. 2. Settlement monitoring beyond the initial year of operation will have a reduction in the number of markers to be surveyed and in the frequency of readings, as justified by a predicted decrease in the rate of settlement. As a minimum, a representative number of markers will be read on a yearly basis for the life of the plant. 3. Settlement monitoring will be made immediately following extreme environmental events such as a safe shutdown earthquake, tornado or major
- flood, s
4 Piping installation,will be completed as late as practical to allow most 4 of the structure settlements to take place before the piping is installed. After permanent installation of piping is made, differential settlements will be monitored at each building interface. If stresses induced by differential settlement approach unacceptable limits, appropriate steps will be taken in the field to reduce stresses. 5. The settlement monitoring program will be a separate program controlled by plant procedures.
u. c: -we 94 1-The staff finds the applicant's proposed plan for long-tern settlement monitoring of safety-related structures and piping to be acceptable except for the following comments and additions. 1. Important settlements markers should be replaced as near as possible to the original location and not subtracted from the program. Settlement observations should be continued without interruption. Important settlement markers would be at those locations where total settlements greater than 1 inch have been recorded and/or are needed to continue the determination of differential settlements across a structure or to establish differential settlements at piping interfaces. 2~. The staff requires that settlement monitoring be completed for important markers through the third year of plant operation at a frequency interval not to exceed 90 days. At the end of the third year of operations, the licensee should provide a brief technical report with supporting settlement data and graphical plots)and an evaluation that justifies any e-reduction in the frequency and number of markers to be monitored. Future plants for the settlement monitoring program would need to be ihdentified. A 3. Settlement monitoring should be performed immediately after any earthquake event and whenever there is a fluctuation in excess of ten feet in previously stabilized groundwater levels within the main power block complex. 4. The reexamination of settlement behavior to be completed at the end of construction,that is discussed in Section 2.5.4.4.3 of this SER, should A also address other details of the long-tern monitoring program which would include the level of survey accuracy to be reautred, igdentification x
w -. _.n n --_n _....----a.._ b / JT
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- 3.- O of the actual valyes of allowable total and differential settlement limits x 0 b; tween specific markers with the design basis for those valyes, and actions to be taken by the applicant in the event that the allowable limits are r: ached. 4 =*aww=, W- -.. men. e e..
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~ ps.i - by the Office of Research of the NRC should reduce the' uncertainty N ter identi ing (1) the causal mechanism of the Charleston earthquake /and (2) the / potential or the occurrence of large earthquakes throughout the eastern seaboard. probabilistic studies, primarily that being/ conducted for NRC by Lawrence Liv ore National Laboratory (LLNL), wil,1<ake into a' count c / existing uncertain tes. They will have as their aim to determine differences, if any, between the p abilitiesofseismicyroundmotionexceedingdesign levels in the eastern seaboard (i.e., as af}ected by the USGS clarified position on the Charleston earthqua and the pr;o[ abilities of seismic ground motion exceeding design levels elsew reint[ecentralandeasternUnitedStates. g Any plants 'for which the 'probabt iXies of exceeding design level ground.. motions are significantly higher than t ose calculated for other plants in the central and eastern United States wi be ide +1fied and evaluated for "possible further engineering analysis. Given the speculative nature of the hypothese with respect to the recurrence of large Charlestog/ type earthquakes as a result f present limited scientific knowledge and t generalized low probability assoc ted with such events, the staff does no see a need for any action for specific tes at this time. It is the staf s position, as it has been in the past, that acilities should be designed o withstand the recurrence of an earthquake the si e of the 1886 earthqu e in the vicinity of Charleston, S.C. Attheconclus\\cnofthe shorter rmprobabilisticprogramandduringthelongertermdekerministic studies,thefstaffwillbe,assessingtheneedforamodifiedpositionwith respect to specif1c sites. ~ 10/26/84 2-61 V0GTLE DSER SEC 2 L___
fWi' Table 2.1 Resident population vs. distance from reactor building ( \\. Year 0-1 mi 0-2 mi 0-3 mi 0-4 mi 0-5 mi 0-10 mi 495 773' ,885 1085 2560 1980 1987 517 806 '923 1133 2669 1990 27 74 121 262 1830 2030 66 153 235 499 2551 Table 2.2 Remaining safety review items son OErdi*a 2*I* V _A R; view item SER Sections Status (1) Foundation competency of clay marl 2.5.4.1.3, 2.5.4.2, stratum 2.5.4.4.3, 2.5.4.4.6 Open (2) Verificatiot,of FSAR commitments on compaction of Category 1 backfill 2.5.4.3 Open 1 (3) Submittal and evaluation of settle-2.5.4.1.3, ment records 2.5.4.4, 2.5.~4.4.3 Open (4) Foundation design and construction information'on radwaste buildings and tunnels 2.5.4.4 Open (5) Locations and description of observed cavitics 2.5.4.4.1 Confirmatory (8) Bearing capacity stability 2.3.4.4.2 Open (7) long-term groundwater and settle-ment monitoring requirements 2.5.4.4.3, 2.5.4.5 Open (8) Acceptability of variations in soil (ynamic properties 2.5.4.4.6 Open i (9) Method used to establish dynamic passive pressures - 2. 5. 4. 4. 4 Confirmatory e 0 10/26/84 2-66 V0GTLE DSER SEC 2 _ _ _ _ _ _ _ _. _ _ _ _ _. _ _ _ _ _}}