ML20132C542
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'p November 15, 1984
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MEMORANDUM FOR:
T. M. Novak, Assistant Director for Licensing, DL FROM: ,
R. W. Houston, Assistant Director for Reactor Safety, DSI
SUBJECT:
SAFETY EVALUATION REPORT FOR BEAVER VALLEY, UNIT 2 Plant Name: Beaver Valley Power Station, Unit 2 Docket No.: 50-412 Responsible Branch: LB #3, DL '
Project Manager: M. Licitra, M. Ley Review Branch: CSB: DSI Review Status: Incomplete Enclosed is the Safety Evaluation Report (SER) (Enclosure 1), for the Beaver Valley Power Station, Unit 2 (BVPS-2) as prepared by the Containment Systems Branch (CSB). This report is based on the staff's review of the applicant's Final Safety Analysis Report (FSAR) as amended, and the applicant's response to staff requests for additional information. We have noted that the FSAR (
contains blank tables with statements to the effect that information has been I forwarded to the staff under separate cover. In many cases, this information has not been received; the applicant should be reouested to provide a schedule
( for filing suitable amendments to complete the FSAR. In addition, the following unresolved items in the SER need to be addressed by the applicant.
- 1. The methodology used by the applicant to cocoute the mass and energy release rates from postulat'ed reactor coolant pipe breaks for the containment analysis is currently under separate staff review. In this regard, the applicant's response to NRC Question 480.7 did not fully justify the use of the unapproved methodology.
- 2. The mass and energy release data for the postulated main steam line breaks have not been documented in the FSAR. Completion of our review of the applicant's main steam ;1ine break analysis is dependent on the receipt of this information.
- 3. The subcompartment design pressure differentials for the reactor cavity, steam generator and pressurizer compartments have not been documented in the FSAR. The applicant is required to complete Table 6.2-26 in the FSAR for completion of the ' staff's review.
CONTACT: J. S. Guo, CSB: DSI i -
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l T. Novak l i
4 In the event of a LOCA, coolant flow velocities on the floor of the i containm'ent, at the time of recirculation spray system actuation, when '
l the water level is low, are substantial and may transport debris to the '
sump; this could adversely affect pump performance. Therefore, the '
applicant should justify the acceptability of the 50 percent blockage I assumption that was used to au ess emergency sump performance, by :
assessing the susceptibility of insulation to become dislodged by virtue !
of its proximity to high energy line piping and be transported to the i sump.
j Please contact our staff if you have any questions regarding these items, i
Enclosure 2 is the SALP input for this SER input in accordance with Office l Letter No. 44 ,
C&!.';; TipM Cy I h.~v.ty .. Lr.ca R. W. Houston, Assistant Director for Reactor Safety, DSI 1
( Encicsures:
As statec cc w/c encl.:
D. Eisenhut i
cc w/ encl.:
G. Knighton M. Licitra M. Ley R. Capra DISTRIBUTION:
Docket File CSB R/F JGuo JShapaker WButler AD/RS/RF
- PREVIOUS CONCURRENCES ON FILE IN CSB n J-DS!:CSB DSI:CSB DSI:CSB/BC DSl/./.D ,
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I-CONTAINMENT SYSTEMS BRANCH INPUT FOR SAFETY EVALUATION REPORT BEAVER VALLEY POWER STATION - UNIT NO. 2
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DOCKET NO.: 50-412 6.2 Containment Systems The Containment Systems for Beaver Valley Power Station, Unit" 2 include the containment structures and associated systems, such as the containment heat removal systems, containment isolation system, and containment hydrogen control system. These systems function to prevent or control the release of radio-active fission products which might be released into the containment atmosphere following onset of a postulated loss of coolant accident (LOCA), or fuel handling accident, mitigate the accumulation of combustib!e gases that can potentially be generated and mitigate the effects of secondary system pipe ruptures.
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. The staf f has reviewed the information relating to the design, design bases and safety analyses for the containment and the containment systems provided
.in the FSAR. The acceptance criteria used as the basis for our evaluation are contained in Section 6.2.1, " Containment Functional Design," 6.2.2, "Contain- i ment Heat Removal Systems," 6.2.4, " Containment Isolation System," 6.2.5,
" Combustible Gas Control in Containment,"and 6.2.6, " Containment Leakage Testing," of the Standard Review Plan (SRP), NUREG-0800. These acceptance criteria include the applicable General Design Criteria (GDC) of Appendix A to 10 CrR Part 50, Regulatory Guides, Branch Technical Positions, and industry codes and standards, as specified in the above cited sections of the SRP.
6.2.1 Ccntainment Functional Design 6.2.~.I' Containment Structure
( The containment structure for Beaver Valley, utilizes the subatmospheric containment concept, and houses the Nuclear Steam Supply System (NSSS),
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6-1 . Beaver Valley 2 SER Input Sec 6.2 )
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.' including the reactor coolant system (RCS), associated auxiliary systems and certain components of the plant engineered safety feature systems. It is a steel-lined reinforced concrete structure with an internal free volume of about 1,800,000 cubic feet. The maximum and minimum internal design pressures of the containment structure are 45 psig, and 8 psia, respectively, and the design temperature is 280*F. (See also Section 3.8 of the SER). -
During normal operation, the containment structure is maintained at a subatmoe-pheric pressure (i.e., about 9 to 12 psia). In the event of a high energy line break accident, the containment would be depressurized and a subatmos-pheric condition reestablished within 60 minutes; this condition would be maintained for at least 30 days following onset of an accident.
Maximum Pressure / Temperature and Depressurization Analyses The applicant has performed containment response analyses for a spectrum of
- postulated reactor coolant system and secondary system pipe ruptures to verify the containment funtional design; i.e., the acceptability of the containment
-. k design pressure and containmen't depressuri:ation criterion, and establish the pressure and temperature conditions for environmental qualification of safety-related equipment located inside containment. Tne containment functional ar.2%.ses inciade the peak containment pressure analysis and the containment depressurization analysis.
With respect to the peak containment pressure analysis,~ the loss of coolant accid'ents (i.e. , RCS pipe breaks) analyzed by t5e applicant include a spectrum of het leg and cold leg (pump suction and pu= discharge) breaks, up to and including the double-ended rupture of the largest reactor coolant line. The spectrum of secondary system pipe breaks analyzed by the applicant include double-ended and split breaks of the main steam line at different reactor power levels (i.e., 102%, 70% and 30% of full power, and the hot shutdown condition). A single failure analysis is not necessary for the peak containment pressure evaluation since the peak pressure for each case analyzed occurs t.efore active engineered safety feature syste.s can influence the results.
The design basis accident for peak containmert pressure' (containment integrity 11/14/84 6-2 . Beaver Valley 2 SER Input Sec 6.2 m
DBA) was determined 9,o be the double-ended guillotine break in the hot leg (HLDER).
The peak containment pressure calculated by the applicant (using the Stone and Webster LOCTIC computer code) was 44.7 psig, which is below the containment design pressure of 45 psig. The applicant also performed a sensi-tivity study- and found that the initial conditions which result in the highest -
peak calcula,ted pressure are the maximum initial containment pressure (11.6 psia-),
maximum initial containment temperature (105*F) and maximum initial containment dewpoint (105*F), i.e., relative humidity.
These are the limiting values that will be allowed by the Technical Specifications.
Tne staff has performed a confirmatory analysis of this design basis accident using the CONTEMPT-LT/28A computer code. The results of the staff's analysis are in good agreement with the applicant's results.
For the secondary system pipe break analysis, the applicant analyzed a spcc-trum of main steam line break accidents covering different double ended rup-tures and split breaks of the main steam line, and reactor operating power levels from hot shutdown to full power. For the DER, the forward flow area
( (effective break area) is limited to 1.4 ft2 by a flow restrictor in the main steam line. Two different single active f ailures were considered, namely, the fai'ure of a main steam isolation valve te close and the failure of an emer-ge cy sus to energize (causing the failure cf one ESF train which results in minimum containment heat removal capability). Redundant valves are provided for automatic isolation of the main feedwater lines. The highest containment pressure, 41.2 psig, was calculated for a full DER at 30% power, with a MSIV failure, and with an initial containment pressure of 11.6 psia and initial containment dry bulb and dewpoint temperatures of 105 F. The highest contain-ment temperature, 333*F, was calculated for a 0.707 ft2 split break at 30%
power, assuming either a MSIV failure or emergency bus failure, and with an initial containment pressure of 9.11 psia, initial dry bulb temperature of 105*F and initial despoint temperature of 55 F. The staff has not performed confirmatory analyses for the two MSLB cases due to a lack of information (see 5.2.1.4). Therefore, we are not in a position to conclece our evaluation at this time.
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With respect to the, containment depressurization analysis, only pump suction ruptures were determined to be of concern since they produce the highest energy flow rates during the post-blowdown period. The design basis accident for maximum ,depressurization time and subatmospheric peak pressure (contain-ment depressurization OBA) was found to be the double ended rupture of the '
pump suction line (PSDER), with miminum ESF (loss of offsite power and emer- ~
gency diesel generator failure resulting in the loss of one engineered safety feature train, i.e.,
one charging pump, one safety injection pump, one quench spray pump and two containment recirculation pumps with associated coolers).
The applicant also performed a sensitivity study and found t6at the initial conditions which result in the maximum depressurization time are: initial containment pressure of 9.85 psia, initial containment temperature of 85 F, initial containment dewpoint of 85 F, service water temperature of 86 F, and refueling water storage tank temperature of 50 F. These are the limiting values that will be allowed by the Technical Specifications. The applicant calculated a maximum containment depressurization time of 3480 seconds, which
, is within the design limit of 3600. seconds, and a subatmospheric peak pressure of -0.08 psig. A barometric pressure of 14.36 psia was used in the analysis,
.. ( which is based en climatological data for Pittsburgh (U.S. Dept. of Ccmmerce, Weather Bureau, 1963-1964 Local Climatolocical Data, Pittsburgh, Pa, Greater Pittsburgr Airport), ar.d adjusted to the plant grade.
The staff's review of the applicant's containment response analysis has included the postulated reactor coolant system and secondary system pipe breaks, initial conditions, input parameters and assumptions. However, the methodologies used to calculate the mass and energy release rate data for the LOCA and MSLB accident have not been completely reviewed due to a lack of certain information (see Section 6.2.1.3 and 6.2.1.4 of the SER). Therefore, the staff can not complete its review of the applicant's analysis at this time. This will remain a confirmatory item until further information is provided by the applicant regarding the calculation of the mass and energy release data.
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4 Protection Against Damage from External Pressure The containment structure is designed to withstand the external (differential) pressure load due to a postulated inadvertent actuation of the containment quench spray system during normal plant operation. T,he maximum pressure -
differential. is based on the difference between the barometric pressure at the -
plant site and the minimum attainable internal containment pressure. The applicant calculated a minimum internal pressure of 8.0 psia for this postu-lated event.
i The staff has reviewed the applicant's anaysis and has found that the appli-cant's assumptions regarding initial containment conditions and containment quench spray system operation tend to minimize the containment pressure (e.g.,
minimum initial air partial pressure, maximum initial containment temperature and final containment temperature equal to the minimum RWST temperature).
Based on the conservative analysis performed by the applicant, the staff con-cludes that the containment external (differential) pressure design basis is acceptable.
i 5.2.1.2 Subcompartment Analyses Subccr.partment analyses are requi ed te determine the acceptability of the design differential pressure loadings on containment internal structures from high energy line ruptures. The applicant has performed the necessary subcompart-ment analyses for the reactor cavity, steam generator compartments and the pressdrizer compartment, where high energy line ruptures are postulated to occur. The applicant has developed models for each subccrgartment, with a selected pipe break location, type and size, and initial conditions, that result in maximum differential; pressure loads on the subcompartment walls.
The applicant used the THREED c,omputer program to analy2e the pressure tran-sients in the reactor cavity, the steam generator compartment and the pres-surizer compartment. The staff's confirmatory analysis is based on the COMPARE-MDD 1A compuer code. ; 1
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The mass and energy, release rates used in the subcompartment analyses were
- calculated using the SATAN-VI computer program (WCAP-8312A). The methodology described in WCAP-8312A was previously approved by the staff.
Separate dis'c ussions and review of the analyses of the reactor cavity, steam '
generator and pressurizer compartments are presented below. -
Reactor Cavity Analysis The reactor cavity is a heavily reinforced concrete structurs that performs the dual function of providing reactor vessel support and radiation shielding.
For the reactor cavity analysis the applicant postulated a 150-in2 cold leg, limited displacement rupture (LDR) at the reactor vessel nozzle. The staff has reviewed the applicant's analysis and concurs in the selection of the design basis pipe break, contingent upon the acceptability of the mechanically con-strained limit on the pipe break size. (See Section 3.6 of the SER).
The reactor cavity subcompartment model employed by the applicant was
.! developed to account for all important obstructions to flow. This is con-sistent with the recommendations concerning nodalization that are presented in NUREG/CR-1199, "Subccmpartment Analysis Procedures Report." The staff has exa 'nec tne applicant's necal model and finds it to be in accordance with current NRC guideline as specified in NUREG/CR-1199 and, therefore, is accept-able.
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Selection of the break size, location and use of restraints to limit the break area are discussed in Section 3.6. The assumed initial conditions were chosen to maximize the differential pressure response. The applicant calculated a peak differential pressure load on the reactor cavity wall of 115.9 psid, for the design basis 150-in 2 LDR.
The staff performed a confirmatory analysis using the COMPARE-MOD 1A computer code, which confirmed that the applicant's result is conservative. However, the cesign basis for the reactor cavity wall is not documented in the FSAR; therefore, the staff can not confirm that the reactor c~avity wall design is t.'
acceptable. The applicant is requested to complete Table 6.2-26 in the FSAR.
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The applicant also performed dynamic force and moment calculations on the
- reactor pressure ves'sel (RPV) from postulated pipe ruptures in the reactor cavity (see FSAR Section 5.4.14.3.1.1). The staff has reviewed the appli-cant's analytical approach, including methods and modeling for calculating asymmetric loads, and finds that it conforms withthe guidelines of NUREG-0609,
" Asymmetrical Blowdown Loads on PWR Primary Systems". The staff's review of -
the structural aspects of the applicant's calculation of forces and moments on the RPV is discussed in Section 3.6.
Steam Generator Subcomoartment Analyses i Steam generator cubicle 2 was selected as the representative steam generator cubicle since all three steam generator cubicles are similar in design. The applicant analyzed three RCS breaks in the steam generator compartment to evaluate loads on the subcompartment walls and component supports. Main steam lines are not routed through the steam generator cubicles and are, therefore, not considered in the analysis. The three pipe ruptures analyzed include a 360-in2 LDR at the steam generator outlet nozzle, a 180-in2 LDR at the reactor
( cociant pump (RCP) outlet nozzle, and a 707-in2 longitudinal intrados split break at the steam generator inlet elbow. These breaks were chosen from the nine treaks in the applicant's sensitivity study as being limiting cases which erc.e:cp ccr.ditions resulting froc all nine breaks. The staff has reviewed the spectrum of postulated breaks analyzed by the applicant and finds them accept-able.
The applicant's nodalization scheme of the steam generator subcompartment was developea to take into account all significant physical obstructions to flow.
The staff has reviewed the applicant's model and finds it acceptable. The results of the applicant's analyses predict a peak differential pressure of 12.9 psid for the design basis 707-in2 longitudinal intrados split break.
However, the design basis for the steam generator wall.is not documented in the FSAR.
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11/14/84 6-7 Beaver Valley 2 SER Input Sec 6.2
Pressurizer Subcompartment Analyses The applicant. considered three breaks for the pressurizer cubicle, and the pressurizer , relief tank cubicle; namely a spray line double ended rupture (DER) in the upper pressurizer cubicle, a surge line DER at the pressurizer '
nozzle and a. surge line DER in the pressurizer relief tank cubicle. The applicant's nodalization models of the pressurizer subcompartment were developed to take into account all critical restrictions to flow. The staff has reviewed the applicant's models and the spectrum of postulated breaks and finds them appropriately conseivative and acceptable. i The results of the applicant's analysis of the spray line DER in the upper pressurizer cubicle gave a peak differential pressure of 18.07 psid across the pressurizer nodal boundary surface. However, the design basis for the pres-surizer cubicle walls is not documented in the FSAR.
6.2.1.3 Mass and Energy Release Analyses for Postulated LOCA i
. The applicant calculated the mass and energy release rate data for reacter cociant system pipe breaks at three break locations including the hot leg piping between the reactor vessel and steam generator, the cold leg piping at the pur; section, and the CCld leg piping at the pump discharge. The results indicate the pump suction break is the worst case for long term containment depressurization, and the hot leg break is the worst case for containment peak pressure. (See Table 6.2-4 in the FSAR). The applicant assumed minimum safeguards in determining the mass and energy releases, i.e. , the loss of one emergency diesel resulting in minimum safety injection. The staff has reviewed the applicant's spectrum of breaks, the description of the LOCA transient models and the single failure considerations, and finds them acceptable.
The applicant's mass and energy release analysis is considered in four phases:
blowdown, refill, reflood and post reflood. The blowdown phase is the phase of the accident during which most of the energy contained in the reactor coolant system is released to the containment. The SATAN VI computer code was used by the applicant to determine the mass and energy accition rates to the
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containnment during the blowdown phase of the accident. The model used for e
the blowdown transient is described in a Westinghouse letter (NS-TMA-2075) that is currently under staff review. At this time, we are not in a position to complete our review of the blowdown methodology. This will be a confirma-tory item pending the completion of the staff's review. -
The time delay due to lower plenum refill has been neglected by the applicant for containment analysis. Instead, the applicant has conservatively assumed that the bottom of the core is covered immediately af ter the end of blowdown.
i Tne analysis of the reflood phase of the accident is important to pipe ruptures in the reactor coolant system cold legs since the steam and entrained liquid carried out of the core for these break locations, pass through the steam generators which constitute an additional energy source. The steam and entrained water leaving the core and passing through the steam generators will be evaporated and/or superheated to the temperature of the steam generator secondary fluid. The rate of enegy release to.the containment during the reflood phase is proportional to the core flooding rate. The rupture of the
( ccic leg at the pump suction results in the highest mass flow through the ccre, and thus through the steam generators.
2; #ng the ccre reficod phase of the accident, when the ccre is filling with water, mass and energy release rates were calculated by the applicant using the modified WREFLOOD code. This model is described in a Westinghouse letter that is also under staff review in conjunction with the review of the SATAN VI code. ' Staff acceptance of this model will be a confirmatcry item pending the receipt of outstanding infctmation and completion of the staff's review.
Tne applicant has included consideration of a possible additional energy release to the containment during the post reflood phase. The post reflood phase begins after the core has been recovered with water. During this phase, decay heat generation will produce boiling in the core and a two phase mixture cf steam and water will exist. In calculations performec by the FROTH code, ine applicant assumed that this two phase mixture of steam and water rises atove the core and enters the steam generators. By this process the remainder
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of the available steam generator energy is removed by boiling of the water a
entrained in the two phase mixture and is carried into the containment as steam. In calculating the rate of energy removed from the steam generators, the applican,t has used the maximum steam flow based on the hydraulic resistance of the system and the maximum steam generator heat transfer. A portion of the steam that flows through the unbroken loops through the ECCS '
injection points is assumed to be quenched before exiting to the containment.
The mass and energy rates calculated by FROTH are used in the containment analysis to the time of containment depressurization.
i 6.2.1.4 Mass and Energy Release Analyses for Postulated Secondary System Pipe Ruptures The applicant has computed the mass and energy release rates for postulated main steam line breaks using the MARVEL Computer Code (WCAP-8843, 1977) which was previously approved by the staff. The MARVEL code describes the primary
, and secondary systems of a pressurized water reactor including the power excur-sion which may occur in the core following a MSLB. The code calculates heat I
. ' low frcm the core and intact ' steam generators into the primary system, and f
heat 'lo. from the primary system into the af fected steam generator. The reima y system heat flow produces additional steam which is added to the centa'9,ert.
It is assumed that the flow frem the break contains no entraineo liquid so that the break flow is pure steam. This assumption maximizes the energy release to the containment. The analysis includes the blowdown of steam from the intact steam generators before closure of the isolation valves and from the unisolated steam lines and turbine plant piping. Feedwater flow is acted to the affected steam generator based on the reduction in the dis-cnarge pressure calculated by the KARVEL code. No credit is taken for any feec=ater flow reduction during the valve closure period. The unisolated feedwater mass is added to the steam generator inventory during the blowdown.
In the applicant's mass and energy release analysis, the unisolated feedwater line .>olume between the steam generator and the isolation valve was included as a source fcr additional high energy fluid to be dischargec through the pipe t eat. Accition of auxiliary feedwater to the affected' steam generator was i
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assumed to start at time zero and continue until manually stopped by the plant operator. The applicant has considered the long term blowdown of the water i
supplied by the auxiliary feedwater system. Auxiliary feedwater flow to the affected ste,am generator is limited to 43 lbm/sec by passive flow control devices installed in the line to each steam generator, and, for the analysis, -
is assumed to be manually terminated 30 minutes after the break. The blowdown -
rate is also limited by the rate at which water is added to the steam genera-tor from the auxiliary feedwater system (if the main feedwater isolation valve in the broken loop fails to close, main feedwater will be terminated by the feedwater control valve). The staff has found the applicant':s analysis has adequately considered the long term blowdown of water supplied by the auxi-liary feedwater system, in accordance with the guidelines of IE Bulletin 80.04.
The mass and energy release data for the worst case MSLB's are not provided in the FSAR. The staff requests that this information be provided for review and to support the staff's confirmatory analysis. This matter will be a confirma-tory item pending the receipt of the information.
( 6. 2.1. 5 Minimum Containment Pressure Analysis for Emergency Cere Cooling System Performance Capability Studies Appe. dix K : 10 CFR Part 50 requires that the containment pressure used for evaluating core cooling ef fectiveness during reactor vessel reflood shall not exceed a pressure calculated conservatively for this purpose. The calculation must include the effect of operation of all installed containment pressure reducing systems and processes. The corresponding reflood rate in the core will then be reduced because lessened containment pressure reduces the resis-tance to steam flow in the reactor coolant loops and increases the boiloff rate from the core.
The applicant has performed the required containment back pressure calcula-tions, (see Section 6.2.'1.5 of the FSAR) using the methods and assumptions described in " Westinghouse Emergency Core Cocling System Evaluation Mode-Summary," WCAP-8339, Appendix A, for the limiting case LOCA, the double-ended colc leg guillotine break (CD = 0.4) (i.e. , the break found to produce the 11/14/84 6-11 Beaver Valley 2 SER Input Sec 6.2
highest peak claddi,ng temperature). Mass and energy release rates for this
, break-were calculated using the method described in Section 15.6.5 of the FSAR. This method is evaluated separately in Section 6.3.5 of this SER.
The staff has reviewed the applicant's input parameters used in the minimum -
containment pressure analysis including initial containment conditions, con- '
tainment net free volume, containment active heat removal, passive heat sinks, heat transfer to passive heat sinks, and found them to be acceptably conserva-tive, and in conformance with BTP CSS 6-1.
i 6.2.1.6 Summary and Conclusions The staff has evaluated the Beaver Valley, Unit 2 containment functional design with respect to the acceptance criteria in SRP Section 6.2.1.1. A, 6.2.1.2, 6.2.1.3, 6.2.1.4, and 6.2.1.5 and concluded that General Design Criteria 13, 16, 38 and 50 have been met with the following exceptions:
- 1. The method used by the applicant to compute the mass and energy release
,( rates from postulated reacter coelant system pipe breaks for the contain-ment analysis is currently under separate staff review. In this regard, the applicant's respor.se to NRC :.estion 480.7 dic not fully justify the ese of the unapproved methodclog .
- 2. The mass and energy release data for postulated main steam line breaks have not been documented in the FSAR. Staff acceptance of the appli-cant's main steam line break analysis is contingent upon the receipt of this information.
- 3. For the subcompartment analyis, the design bases for the reactor cavity, steam generator and pressurizer compartments have not been documented in the FSAR.
6.2.2 Containment Heat Removal Systers T e 'onction of the containment heat removal systems is'to remove heat from
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ine containment atmosphere to limit, reduce and maintain at acceptably low 11/14/84 6-12 . Beaver Valley 2 SER Input Sec 6.2
levels, the containment temperature and pressure following a loss of coolant accicent or main st6am line break. In addition to heat removal provided by passive means such as heat transfer to containment structures and components, the Beaver Val' ley 2 design includes active containment heat removal systems (CHRS). The- active CHRS includes two spray systems; namely, the quench spray -
system (QSS),and the recirculation spray system (RSS); the containment air -
coolers are not included in the CHRS. The CHRS is designed to depressurize the containment to a subatmospheric condition within one hour. For a discus-sion of the fission product removal function of the CHRS, see SER Section 6.5.
i The QSS is composed of two redundant 100 percent capacity trains each con-taining a quench spray pump, chemical injection system and riserpipe leading to two spray headers. The two trains connect to the two common 360-degree spray headers in parallel with risers 180 degrees apart. There are a total of 159 SPRACO model 1713A nozzles on the two quench spray ring headers; 120 nozzles on the lower header and 39 nozzles on the upper header. Each quench spray pump is rated at 3000 gpm of spray flow to the spray headers. Both spray pumps operating together can supply approximately 4500 gpm to the spray
( headers. The QSS is designed to spray ccid berated water into the containment from the refueling water storage tank (RWST) no later than 83 seconds after receipt of a containment isolation Phase B signal (CIB). Sodium hydroxide (Na0M) soluticn from the chemical adcitive tank (CAT) is added to the quench spray by means of the chemical injection system upon receiving a CIB signal.
Once the quench spray discharge has ended, flow from the checmical injection pump is automatically diverted to the containment sump.
The RSS is designed to provide additional depressurization of the containment anc to maintain the containment at a subatmospheric condition in the long term following the accident. The RSS consists of two 360 degree spray ring headers and four pumps and heat exchangers. Each spray ring header contains 292 SPRACO model 1713A nozzles, and is fed by two risers, with each riser origi-nating from one of the recirculation coclers.
Tne two redundant recirculation spray pumps that feed each header are each supplied with emergency power from separate ciesel generators. Each RSS pump
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takes suction from the containment sump at approximately 3480 gom (50% heat removal capacity). The RSS is capable of operating in the post-accident environment to maintain a subatmospheric pressure for 30 days following a high energy line preak.
The RSS pumps are started automatically about 628 seconds after receipt of a -
CIB signal, and the spray becomes effective about 714 seconds after the CIB signal. When the water in the RWST reaches a predetermined low level, the flow from two of the RSS pumps is automatically diverted to the cold leg recircu-lation mode of ECCS. -
The CHR5 satisfies the provisions of Regulatory Guide 1.26, " Quality Group Classifications for Water, Steam, and Radioactive-Waste Containing Components of Nuclear Power Plants," and 1.29, " Seismic Design Classification," for engineered safety features. The applicant has provided information (FSAR Section 14.2, " Initial Test Program") in accordance with the guidelines of r Regulatory Guide 1.68, " Initial Test Program for Water Cooled Nuclear Power Plants") which will ensure the ability of the quench spray system and recircu-
,, ( 1ation spray system to function fellovring a postulated single active failure.
Regulatory Guide 1.82, " Sumps for Emergency Core Cooling and Centainment 5; ray Systems," provides cesign guidelines for cor.tainment sum;s th'at are to serve as sources of water for the ECCS and containment spray system following a LOCA.
The guidelines address redundancy, location and arrangement criteria, as well as debris screen provisions to ensure adequate pump performance. The staff'has reviewea tha Beaver Valley 2 sump design against this guidance.
A single containment sump has been provided, and is enciesed by a protective screen assembly that has a total screen area of about 150-ft2 . Furthermore, the containment sump is divided at the center line by screening and vertical bars so that a failure of either half would not adversely affect the other half. The redundant recirculation pump suctions are located in separate halves of the sump. Therefore, even though the single su D design is not in acccreance with Regulatory Guide 1.82 recommendations, the staff has concluded
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' that adequate measures have been taken to assure that the RSS function will not be lost.
- The protective screen assembly provides three stages of screening, namely, vertical trash bars, a coarse mesh screen (3/4" opening) and a fine mesh ;
screen (3/32" opening). The fine mesh screen opening is smaller than the -
smallest coolant passage gap in the reactor core and smaller than a spray nozzle orifice. The screen assembly rises vertically approximately 5 feet above the containment floor, and is arranged so that no single failure could result in the clogging of all suction points of the recirculat. ion spray system.
Following a LOCA, the top of the screen assembly would be under about 10 feet of water.
System design allows for 50 percent blockage of the sump screening without loss of function.
The applicant has conducted sump model testing at the Alden Research Labora-tory. using a 1/3 scale hydraulic model of the sump. The objective of the testing was to evaluate the sump performance characteristics with a view towards eliminating conditions that may be conducive to the formation of air
( entrainment vortices and lowering the threshold containment water level at which vortex formation would be expected to be completely suppressed. The test progra- included two and four pump operation, up to 50 percent blockage cf t'is su .p screens er trash racks, and varicus farfield flo.c distributions and water levels. By lowering the pump suction line inlets several inches and installing horizontal gratings above the inlets, the containment water level for vortex free operation was reduced from EL 697 ft (with no blockage) to EL 693.8 "f t (with up to 50 percent blockage of the screens or trash racks). (The maximum water level in the containment following a LOCA would be EL 708.5 ft, anc :ne lowest sump elevation is EL 691 f t).
The applicant states that in the long term following onset of a LOCA, when the sump structure is fully submerged, the average flow velocity at the sump screen would be 0.31 fps (assuming maximum safeguards equipment conditions and 50 percent blockage of the screen area); Regulatory Guide 1.82 recommends a cesigr. velocity for the coolant at the inner screen of about 0.2 fps. Further-
, ccre, initially, as the RWST water is being discharged i.o the containment via
(
11/14/84 6-15 Beaver Valley 2 SER Input Sec 6.2
+ .
the quench spray system and emergency core cooling system, flow velocities on o
tne floor of the containment, as reported in the Aleden Research Laboratory test report submitted by the applicant, would be on the order of 1.7 fps. The applicant further states that the velocity distribution would be such that reflective ' metallic type insulation (used on most of the piping in the contain -
ment) would not be transported to the sump. However, the Alden Research Laboratory reported in NUREG/CR-3616, " Transportation and Screen Blockage Characteristics of Reflective Metallic Insulation Materials", that flow velo-cities well below 1.7 fps are capable of transporting the various component parts of this type of insulation to the sump structures. Inflight of this, the staff recommends that the applicant provide a debris generation and transport analysis, which describes the transient behavior of the sump as the water ?
level in the containment is rising, to justify the acceptability of the 50 percent sump blockage assumption throughout the accident. The staff considers this to be a confirmatory item and will report on its resolution in a future supplement to this SER. !
i The staff has reviewed the applicant's net positive suction head (NPSH) cal- 1
[
,( culation submitted by letter dated February 21, 1984, and the updated results f
reported in amendment 5 (February 198c) to the FSAR, which reflect the head Icss across the m0dified sump structure. The analysis shows the NPSH avail-aMe to t'.e recirculation pumps during noth the recirculation spray mode and the combined recirculation spray and low head safety injection mode is always greater than the required NPSH. At the beginning of the recirculation spray mode (when the containment water level is low, and conservatively calculated f
{
to be'at El 694 ft), the NPSH margin is calculated to be 0.9 ft (assuming minirum ESF operation to achieve a higher flow rate, and 50 percent blockage f of the sump. The NPSH margin will continue to increase as the containment water level rises to its maximum level. The applicant has complied with the provisions of Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal Systems", with one exception.
Regulatory Guide 1.1 states that containment heat removal systems should be designed so that adequate NPSH is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure
(
11/14/84 6-16 , Beaver Valley 2 SER Input Sec 6.2 l
from that present before the postulated LOCA. Instead, the applicant calcu-lated the NPSH available (see FSAR, Section 6.2.2.3.2) using a saturated sump
('
model (i.e. , the containment atmospheric pressure is conservatively assumed to be equal to ,the vapor pressure of the liquid in the sump, e suring that credit is not taken- for containment pressurization during the transient). The staff -
has previous.ly found the saturated sump model to be conservative and, there- .
fore, acceptable.
The staff has reviewed the information in the applicant's FSAR and in res-ponses to staff requests for additional information concerning the containment heat removal systems to assure conformance to the acceptance criteria con-tained in SRP Section 6.2.2. The staff finds that the containment heat re-moval systems satisfy the requirements of General Design Criteria 38, 39, and 40, the provisions of Regulatory Guide 1.1 on an acceptable alternative basis as defined above, and the provisions of Regulatory Guide '.82, except as noted above. -
6.2.3 Secondary Containment Functional Design "
(
The Eeaver Valley 2 design does not include a secondary containment.
6.2.4 Cc.tainment Isolation System The function of the containment isolation system (CIS) is to allow the normal or emergency passage of fluids through the containment boundary while pre-servin'g the ability of the bou dary to prevent or limit the escape of fission products that may result from postulated accidents. In general, for each fluic system penetration at least two barriers are required between the con-tainment atmosphere or the reactor coolant system and the outside atmosphere, so that failure of a single barrier will not prevent isolation of the contain-ment.
i Centainment isolation for Beave'r Valley 2 is accomplished in two phases. The contai.. ment isolation Phase A (CIA) signal isolates all non-essential system
'.ines penetrating the centainment, and is initiated by any of the following:
k (1) high containment pressure (Hi-1 setpoint); (2) low compensated steam line 11/14/84 6-17 Beaver Valley 2 SER Input Sec 6.2
pressure; (3) press,urizer low pressure; or (4) manual actuation. The contain-g' ment isolation Phase B (CIB) signal isolates the component cooling water supply and return lines for the reactor coolant pumps (RCPs) and control rod drive mechan. ism (CRDM) shroud coolers, and the service water lines to the containment recirculation air coolers. The CIB signal is initiated by high '
containment pressure (Hi-3 setpoint) or by manual actuation. The containment '
isolation signals which initiate containment isolation functions are summarized in Tsble 6.2.4-1. The applicant has documented that each system line having automatic containment isolation valves, which must b'e imnediately isolated following an accident, is isolated by one of the signals in Tiible 6.2.4-1.
Although the Phase B isolation signal is not actuated by diverse parameters, it is acceptable because the affected lines are considered important to the safe shutcown of the plant and are capable of remote manual isolation. The staff concludes that adequate diversity has been provided with regard to the different monitored parameters which actuate containment isolation.
TABLE 6.2.4-1 CONTAINMENT ISOLATION SIGNALS
,, (
AND ACTUATION PARAMETERS Cc,tainment Isclation Phase A signal
- a. High Cer.tainment Pressure (Hi-1)
- b. Low Compensated Steam Line Pressure
- c. Pressurizer Low Pressure
- d. Manual Actuation Centainment Isolation Phase B signal
- a. High Containment Pressure (Hi-3)
- b. Manual Actuation Safety Injection Signal
- a. High Containment Pressure (Hi-1)
- t. Low Compensated Steam Line Pressure
- c. Pressurizer Low Pressure
, c. Manual actuation l.
11/14/84 6-18 . Beaver Valley 2 SER Input Sec 6.2
i Main Steam Isolation Signal
- a. High Steamline' Pressure Rate
- b. High Containment Pressure (Hi-2)
- c. Low Steamline Pressure
- d. Manual Actuation ~
l l ,
Feedwater Isolation Signal
- a. Steam Generator Hi-Hi Water Level
- b. Safety Injection Signal
- c. Low TAVG and Reactor Trip i Cotainment Vacuum System Isolation Signal i
- a. Containment Isolation Phase A Signal (Hi-1)
- b. Manual Actuation The staff has reviewed the applicant's containment isolation system design l
bases and containment isolation provisions as documented in Table 6.2-60 of I the FSAR, for conformance to General Design Criteria (GDC) 54, 55, 56 and 57
( and :.egulatory Guide 1.11, " Instrument Lines Penetrating Primary Reactor Con-l tainments." The applicant's containment isolation system design is summarized as follows:
l (1) There are at least two barriers between the atmosphere outside con-
{
tainment and the atmosphere inside containment (or the RCS) on each system line penetrating the containment.
k i
(2) The two barriers consist of one of the following arrangements:
a.
two normally closed manual valves with administrative control, one l
inside containment and the other outside containment; f
- b. two automatic isolation valves, cne inside containment and the other outside containment, a simple check valve may not be used as the automatic isolation valve outside cc,tainment;
(, !
11/14/84 6-19 Beaver Valley 2 SER Input Sec 6.2 m !
c.
one autorr.atic isolation valve inside containment and one normally 0 closed manual valve under acministrative control outside containment (or.the reversed arrangement);
d.
a sealed system (closed system) inside containment and one isolation ~
valve outside containment, which is either automatic, remote manual, or manual under administrative control.
(3) Isolation valves of the ESF related systems, which are essential to mitigate the effects of an accident, remain open or move- to their open position post-accident. These valves are remote manually controlled and operated from the control room.
(4) Motor operated valves (MOV) are used for system lines which are part of an ESF related system, and fail "as is" on loss of power supply. Sole-noid operated valves are used when greater reliability post accident and
, a safe-failure position are reovired. All power operated valves are designed to fail in the position that provides greater safety upon loss
,( ef power or contrcl air.
(5) ",ectanical End electrical redunda .:y are previded by desicnine two isola-tien barriers between the RCS or a:mesphere inside centainment and the atmosphere outside containment with two separated IE power sources.
(6) Containment purge system isolation is accomplished with two 42-in. butter-fly valves, which are only open during plant cold shutdown and close automatically within 10 seconds u:on receipt of a high radiation sigr.al.
During normal operation, the containment is not purged. The containment airborne radioactive contaminants are removed by the containment atmosphere filt ation system.
(7) The containment isolation system is designed to meet the single failure criterion.
- l. l l
11/14/84 6-20 Beaver Valley 2 SER Input Sec 6.2
I (8) The closure time for each containment isolation valve is less than 60 i g seconds. System lines which have no post-accident function are provided with air-operated valves (A0V) with a closure time of 10 seconds.
The applicant's containment isolation provisions are reviewed against the '
requirements of GDC 54, 55, 56, and 57 (Appendix A to 10 CFR Part 50) and the supplementary guidance of SRP 6.2.4, where applicable. Staff review has confirmed that the containment isolation system meets the explicit require-9 ments of GDC 54, 55, 56, and 57 with the following exceptions: 0 4
(1) The containment vacuum pump and hydrogen recombiner suction lines are provided with two remotely-controlled solenoid-operated isolation valves i in series outside containment. Therefore, the containment isolation 41
'l provisions differ from the explicit requirements of GDC 56. However, the isolation valves are located as close as possible to the containment, and l the associated system piping is designed in accordance with the break / il o
crack exclusion criteria of Branch Technical Position MEB 3-1. Further-more, the valves are hermetically sealed, precluding the need to encap- .,
( sulate the valves. Since the lines are used post-accident, for contain-1 ment atmosphere sampling and hydrogen control, locating the valves out-side containment improves the functional reliability of the valves.
f Therefcre, the staff finds the isolation provisions for these lines to be acceptable alternatives to the explicit requirements of GDC 56. i (3) The emergency core cooling system safety injection lines and reactor coolant pump (RCP) seal injection lines are equipped with weight-loaded fi check valves inside containment and motor-operated valves (MOV), outside i containment which do not receive a containment isolation signal to close.
The safety injection lines discharging to the hot and cold legs of the reactor coolant system and the RCP seal injection lines are important to !
safe shutdown or are part of an engineered safety. feature system. Provisions have been made to detect possible leakage from these lines outside ,
containment, thereby allowing remote manual instead of automatic isolation valves. The staff, therefore, finds that the containment isolation i
provisions for these lines are acceptable alternatives to the explicit requirements of GDC 55.
11/14/84 6-21 Beaver Valley 2 SER Input Sec 6.2 !
i (4) The quench spr,ay pump discharge and recirculation spray pump discharge
{ lines are provided with a normally open, remotely-controlled, motor operated valve outside containment and a weight-loaded check valve inside containment. The isolation valves in the containment depressurization (quench'and recirculation spray) systems open upon receipt of a CIB ~
signal,'if not already open, with the exception of the caustic addition ~
line to the containment sump which automatically opens after the quench spray discharge has stopped. The recirculation spray pump suction lines l are provided with a single, normally open, remotely-controlled, motor operated valve outside containment since it is not prac6ical to locate a second valve inside containment where it would be submerged following a LOCA; these valves do not receive an automatic isolation signal for l closure. Therefore, the containment isolation provisions for these lines
! differ from the explicit requirements of GDC 56 regarding their actua-tion and number.
l l
These lines are part of ESF systems, and are required to be open to perform their post-accident safety function. The ESF systems are closed
.( outside containment, and'are safety grade. Therefore, the staff finds the use of remote-manual instead of automatic isolation valves accept-able. In addition, the single isolation valve outside containment in the 1
recirculation spray pump suction lines is acceptable because system L
reliability is improved with a single valve and the piping between the l outside of the containment wall and the isolation valve, as well as the valve, are contained within a leak-tight encapsulation.
l The staff has also reviewed information provided by the applicant to demon-strate compliance with the provisions of NUREG-0737 Item II.E.4.2, "Contain-ment Isolation Dependability." As previously described, the applicant has complied with the provisions regarding diversity in parameters sensed for initiation of containment isolation, and has considered the functional require-ments of all systems penetrating containment and has made acceptable provi-sions for isolation of systems not recuired for mitigation of the consequences cf an accident or safe shutdown of the plant. The applicant has also made
(
11/14/84 6-22 . Beaver Valley 2 SER Input Sec 6.2
t provision to assure that resetting of a containment isolation signal will not t result in the automatic reopening of containment isolation valves.
(
In addition, the applicant has designated all system lines penetrating the containment as essential or non-essential. Therefore, the staff concludes -
that the app,licant has complied with the provisions of NUREG-0737 Item II.E.4.2.'
The applicant has stated that all containment isolation barriers as well as electrical and control components required for initiation are protected from missiles and the effects of natural phenomena to ensure thein performance under all anticipated environmental conditions. The staff, therefore, finds that the containment isolation system meets the requirements of GDC 1, 2, and 4.
The containment isolation system also meets the provisions of Regulatory Guide 1.29, " Seismic Design Classification," and 1.26, " Quality Group Classi-fications and Standards for Water , Steam , and Radioactive-Waste-Containing Components of Nuclear Power Plants."
In summary, the staff has reviewed the information in the applicant's FSAR and
( in response to NRC Questions concerning the containment isolation system to assure conformance to all of the acceptance criteria contained in SRP Section 6.2.4 and the provisions of BTP CSB 6-4.
The staff concludes that the Beaver Valley 2 containment isolation system reets the requirements of General Design Criteria 1, 2, 4,16, 54, 55, 56, and 57, and is, therefore, acceptable.
6.2.5 Combustible Gas Control System Following a loss of coolant accident, hydrogen may accumulate within contain-ment as a result of (1) metal-water reaction between the zirconium fuel clad-ding and the reactor coolant, (2) radiolytic decomposition of the water in'the reactor core, (3) radiolytic decomposition of the water collected on the sump floor, (4) hydrogen released from the pressurizer gas space and reactor cool-ant, (5) corrosion of metals by:the alkaline solution used for containment spray. The function of the combustible gas control system (CGCS) is to moni-tor anc control the potential hyrogen accumulation within the containment atmospnere below 4-volume percent following a design bas ~is accident.
( l 11/14/84 6-23 . Beaver Valley 2 SER Input Sec 6.2
In the event of a L0CA, two redundant, independent, full capacity electric
( hydrogen recombiners will be available outside containment to control the containment hydrogen concentration.
Each recombiner has a capacity of 50 SCFM and is designed to Seismic Category I criteria. One hydrogen recombiner is permanently installed in the safeguards area; the other recombiner will be '
transferred from Beaver Valley, Unit I and installed in the safeguard area '
following onset of an accident.
In addition to the two safety related hydro-gen recombiners provided, a non-safety grade backup containment purge system is available to purge the containment atmosphere as an aide to cleanup. Each hydrogen recombiner system includes flow control capability, is blower, a temperature-controlled electric preheater, a thermal recombiner, and an air blast heat exchanger. The safeguards area is a Seismic Category I concrete structure located adjacent to the containment. The penetrations, and compo-nents within the safeguard area are protected against tornados and missiles.
The hydrogen recombiners and all associated valves are remote manually con-trolled from panels located in the safeguards area, outside of the recombiner
. cubicles, to allow access and minimize exposure of personnel. The staff has reviewed the hydrogen recombiner system desigr. concept and finds it accept-( able.
Twc redundant, independent hydrogen analyzers are installed in the cable vault area ic monitor the hydrogen concentration in the containment atmosphere. The analyzers are also used to check the efficiency of recombiner operation. The hydrogen analyzer is classified as Class IE, Seismic Category I and func-tionally tested with a calibrated gas sample. Indicators are provided in the main control room to monitor hydrogen concentration. Annunciation is also provided in the main control room for hydrogen analyzer /recombiner local panel trouble.
Based on the staff's review, the post accident hydrogen monitoring system meets the requirements of NUREG-0737 Item II.F.1, Attachment 6, "Contain-ment Hydrogen Monitor," and the single failure criterion.
The applicant has analyzed the potential hydregen generation within the con-tainment using the guidelines provided in Regulatory Guide 1.7, and calculated the hydrogen concentration for both one and two recombiner cperation. The analysis shows that a single recombiner, initiated when 'the containment hydro-( gen concentration reaches 3.1 volume percent (i.e. , approximately 4 days 11/14/84 6-24 , Beaver Valley 2 SER Input Sec 6.2
- e. .
post-accident), is sufficient to maintain the hydrogen concentration in the containment atmosphere below the lower flammability limit of 4 volume percent.
I The design of the Beaver Valley, Unit 2 containment is similar to the Beaver Valley, Unit,1 and Surry containments, which use recombiners. The staff has previously confirmed, using the COGAP computer code, that there is sufficient time before ,the containment hydrogen concentration reaches 3.1 volume percent -
to manually initiate the post-accident hydrogen recombiners, and that a single recombiner can acceptably control the hydrogen concentration in containment below 4.0 volume percent.
i The applicant has stated in the FSAR that the containment design allows air to circulate freely. Furthermore, all cubicles and compartments within the containment are provided with openings near the top as well as openings in the floor to allow air circulation. The applicant has also performed an analysis to demonstrate that adequate mixing of the hydrogen in the containment atmo-sphere will be ensured by the turbulence created by the containment spray system and thermal convecti.on. Therefore, sufficient mixing of hydrogen in containment will occur to prevent stratification and to eliminate areas of
( potential stagnation. The staff finds that adequate passive and/or active design measures have been incorporatec into the containment design to ensure adecuate hydrogen mixing within containment 'end, therefc e. the applicant's hydrogen mixing provisions are accaptable.
In summary, the staff has reviewed the information in the applicant's FSAR and in response to our questions concerning the combustible gases control system to assure conformance to all of the acceptance criteria contained in SRP Section 6.2.5. The staff concludes that the applicant's combustible gas control system meets the requirements of GDC 41, -42 and 43, satisfies the design and performance requirements of 10 CFR 50.44, the provisions of Regula-tory Guide 1.7 and the requirements of NUREG-0737 Item II.F.1, Attachment 6, and therefore, is acceptable.
6.2.6 Containment Leakage Testing Program Tne centainment design includes the provisions and feat 0res required to satisfy
( the testing requirements of Appendix J to 10 CFR Part 50. The design of the 11/14/84 6-25 . Beaver Valley 2 SER Input Sec 6.2
. a .
.. l
- l containment penetrations and isolation valves permit preoperational and perio-dic leakage rate testing at the pressure specified in Appendix J to 10 CFR 50. .
b The staff haf reviewed the containment leakage testing program contained in .
the FSAR and-in the responses to NRC Questions, and finds it acceptable with -
the following exceptions. The applicant proposes to exclude certain valves -
from Type C testing (including the safety injection system penetrations and recirculation spray system penetrations). The applicant states that the ;
justification for excluding these penetrations from Type C testing is based on l the rationale presented in Technical Specification Amendment Wo. 65 to the !
operating license for BVPS, Unit 1. Excluding these valves from the Type c testing program was approved by a NRC letter dated March 22, 1983. The staff has examined the subject issue and the bases for approving Amendment No. 65. I Since both plants are identical in design, the staff finds the applicant's proposal acceptable.
Based on above discussion the staff concludes that the proposed reactor con-tainment leakage testing program complies with the requirements of Appendix
,( J to 10 CFR Part 50. Such compliance provides adequate assurance that con-tainment leak-tight integrity can be verified periodically throughout service lifetime on a tirely basis to maintain such leakage within the limits of the Tech-ical Specifications.
Maintaining containment leakage rates within such limits prcvides reasonable assurance that, in the event of any radioactivity releases within the contain-ment, the 1 css of the containment atmosphere through the leak paths will not be in excess of acceptable limits specified for the site. Compliance with the requirements of Appendix J constitutes an acceptable basis for satisfying the requirements of General Design Criterie 52, 53 and 54.
k 11/14/84 6-26 . Beaver Valley 2 SER Input Sec 6.2
g m g G . ,'
f i Dgr
[NClOSi!RT I ."
?.Al p prepared by the Cont ainment Systems Branch Regar: ling '
Evaluation
_ Criteria Reaver Vaticy 2 (Docket No.: 50-412)
- 1. Management Involvement N/A '
l
- 2. Approach to Resolution Technically sound and thorough approach in most cases.
. r of Technical Issues ,
Category I
- 3. Responsiveness frequently requires extensions of time Category 3
- 4. Enforcement llistory N/A
- 5. Reportable Events N/A -
- 6. Staffing N/A r
- 7. Training N/A
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