ML20132C105
Text
DISTRIBUTION:
May 7,1984 Do'cket File 50-412 CSB Rdg. File JSGuo JShapaker WButler b
AD/RS Rdg. File MElORANDUM FOR:
T. M. Novak, Assistant Director for Licensing, DL FRON:
R. W. Houston, Assistant Director i.
for Reactor Safety DSI
SUBJECT:
DRAFT SAFITY EVALUATION REPORT FOR THE BEAVER VALLEY POWER STATION, UNIT 2 Plant Name: Beaver Valley, Unit 2 Docket No. : 50-412 Responsible Branch: LB No. 3, DL Project Managers:
M. Licitra, H. Ley Review Branch: CSB: DSI Review Status:
Incomplete Enclosed is the Draft Safety Evaluatio.n Report (SSER) (Enclosure 1), for the Be awr Valley Power Station, Unit 2 (DVPS-2) prepared by the Containment Systems Branch (CSB). Tnis 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
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that information will be provided later.
In many cases, this information has not been filed; the applicant should be requested to provide a schedule for filing suitable amendments to complete the FSAR.
In addition, the following unresolved items in the DSER also need to be addressed by the applicant:
1.
The methodology used by the applicant to compute the mass and energy release rates from postulated reactor coolant pipe breaks for the contai: ' ant analysis and the containment subcompartment analysis.
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 postulated main steam line breaks have not been documented in the FSAR.
Completion of our review of the applicant's main steam line break analysis is dependent on the receipt of this information.
3.
The applicant should discuss and justify the barometric pressure used in the containment depressurization analysis. This information is needed to verify the acceptability of the containment functional design.
OFFICt)
SUANAMEI DATE )
OFFiClAL RECORD COPY usara muus m Sac Fonu og 00 SQ NACW CN0
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T. M. Novak -
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4.
The subcompartraent design pressure differentials for the reactor cavity, steam generator and pressurizer compartments have not been documented in the FSAR.
In addition, the applicant has i
not provided an. analysis of the forces and moments on the teactor vessel due to the differential pressure caused by a RCS break within the reactor cavity. Completion of our review 3
of the applicant's subcompartment analysis is dependent on the receipt of this information.
5.
The applicant should address the potential for whipping pipes, high velocity jets of water or steam, or direct streams of water to adversely affect the integrity or performance of the sump's protective screen assembly.
Furthermore. the applicant should discuss and justify the acceptability of the anticipated water velocity at the fine mesh screen (based on one-half of the available free area to account for blockage).
6.
The applicant should further justify the acceptability of the 50 percent blockage assumption that is used to assess emergency sump performance, by specifying the types (and total quantity of each type) of insulation used in the Beavsr Valley 2 containment and discussing the susceptability of the insulation to become dislodged by virtue of its proximity to high energy line piping.
7.
The applicant should provide the results (test report) of the
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sump model testing conducted by the Alden Research Laboratory, r.hd discuss the significance of the results relative to the performance of the as-built. Deaver Valley 2 emergency sump.
C.
The applicant should discuss the performance characteristics of the post-accident hydrogen monitoring system to be installed.
j 9.
The applicant proposes to g clude certain valves from Type C testing (for example, the s afetr injection system penetrations t
and recirculation spray system penetrations). The applicant should provide justification for excluding these and any other penetrations from Type C testing to facilitate staff review of the plant's Technical r,pecifications.
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Please contact our staff if you have any questions regarding these open i tems. (s the SALP input for this DSER input in accordance with Office Letter No. 44.
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'.. :.'.; !.':3 53 R. W. Houston, Assistant Director for Reactor Safety, DSI
Enclosure:
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R. Mattson D. Eisenhut cc w/ encl.:
G. Knighton H. Licitra li. Ley R. Capra t
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CONTAINMENT SYSTEMS BRANCH
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INPUT FOR DRAFT SAFETY EVALUATION REPORT BEAVER VALLEY POWER STATION - UNIT NO. 2
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DOCKET NO.. 50-412 6.2 CONTAINMEtJT SYSTEMS Tne Beaver Valley Power Station, Unit 2 containment Systems j
include the c o n t ai nm e nt s t ruc t ure s and a s soc iat ed s ys t ems, ;9vCN 45
&n k fog containment heat removat systems, containment isolation system, and containment hydrogen control system, i
that function to prevent or control the release of radioactive fission products which might be released into the c o n t a i nm e n t atmosphere folLowing a postulated Loss of j
coolant accident (LOCA), secondary system pipe rupture, or fuel handling accident.
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/,vp sr* s4AA 7~ s on #Ce~l.4 r/ /J4 70 r//G The staff has r e v i e w e d t h e augsQ;sEssta d e s i g n, d e s i g n b a s e s
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and safety analyses for the containment and the containment systems 'provided in the F The acceptance criteria used as the basis for our evaluation are contained in Section 6.2.1, " Containment Functional Design," 6.2.2, " Cont ai nment H ea t R emova t Sy st ems," 6.2.4, " Cont a inme nt Isolation Sy s tem," 6.2.5, "C ombustible Ga s C ont rol in Containment,"
and 6.2.6, " Containment Leakage Testing," of the Standard Review PLen (SRP), NUREG-0800.
These acceptance criteria include the applicable General Design criteria (GDC) of Appendix A to 10CFR Part 50, Regulatory l
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Guides, Branch Technical Positions, and industry codes and, standards, as specified in the above cited sections of the SRP.
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6.2.1 Containment Functional Design 6.2.1.1 Containment Structure The containment structure for Beaver Valley, utilizes the subatmospheric containment concept, and houses the Nuclear Steam Supply System (NSSS), including o
the reactor coolant system (RCS), associated y
auxiliary systems and certain components of the plant engineered safety feature systems.
It is a steet-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 normat operation, the containment structure is maintained at a subatmospheric 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 subatmospheric condition e-established within 60 minutes; this condition would k
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b be maintained for at least 30 days fotLowing an
. accident.
t 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 func'tional de' sign; i.e.,
the accept-ability of the containment design pressure and con-tainment de p re s su ri za ti on criterions and establish the pressure and temperature conditions for environ-mental qualification of safety-related equipment located inside containment.
The containment l
functional analyses include the peak containment pressure analysis and the containment depressurization analysis.
i With respect to the peak containment pressure analysisi l
the Loss of coolant accidents (i.e., RCS pipe breaks) analyzed by the applicant include a spectrum of hot i
leg and cold leg (pump suction and pump discharge) breaksi up to and including the double ended rupture of the largest reactor coolant Line.
The spectrum of
f secondary system pipe breaks analyzed by the applicant
- include double ended and split breaks of the main steam
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,line at di f ferent reactor power levels (i.e., 102%,
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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 before active engineered safety feature systems can influence the results.
The design basis accident for peak containment pressure' (contain' ment integrity DBA) was determined to 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 al.so performed a sensitivity study and found that the initial conditions which result in the highest peak calculated pressure are the maximum initial containment pressure (11.6 psia)r maximum initial containment temperature (105 ' F) and maximum initial containment dewpoint (105 ' F),
i.e.,
relative humidity.
These are the limiting valves that will be allowed by the Technical Specifications.
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- 40 The staff has performed a confirmatory analysis of
.this design basis accident using the CONTEMPT-LT/28A z
computer code.
The results of the staff's analysis g
are in godd agreement with the applicant's results.
For the secondary system pipe break analysis, the applicant analyzed a spectrum of main steam Line break accidents covering different double ended ruptures 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
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1.4 FT2 by a flow r e s t r i c t o r,.
Two different single active failures were considered, namety, the failure of a main steam isolation valve to close and the f ailure of an eme rgency bus to energize (causing the f ailure of 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 l
containment dry bulb and dewpoint temperatures of 1
j 105' F.
The highest containment temperature, 333' F, k
was calculated for a 0. 707 ft2 split break at 30%
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t power, assuming either a MSIV failure or emergency bus failure, and with an initial containment pressure of 9.11 psia, initial dry bulb temperature o f 105 ' F f
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and initial dewpoint temperature o f 55 ' F.
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With respect to the containment depressurization i
analysis, only pump suction ruptures were determined I
to be of concern since they produce the highest l
L energy flow rates during the post-blowdown period.
The design basis accident for. maximum depressurization i
time and subatmospheric peak pressure (containment
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depressuri zation BBA) was found to be the double-ended rupture of the pump suction line (PSDER),
l with miminum ESF (loss of offsite power and emergency
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diesel generator failure resulting in the loss of one engineered safety feature train, i.e.,
one charging pumps one safety inje ction pump, one quench I
spray pump and two containment recirculation pumps with associated coolers).
The applicant also performed a sensitivity study and found that the initial conditions which result in the maximum depressurization time are:
initial containment pressure of 9.85 psia, initial containment temperature of 85' F, L
initial containment dewpoint of 85' F, service water l
tempera ture of 86' F, and re f ue Ling water storage tank i
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i temperature of 50' F.
These are the limiting valves t
.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 substmospheric peak pressure -0.08 psig.
The staff is unable to conclude on the acceptabil'ity of the applicant's containment de p re ss u ri zat i on analysis at this time because'the applicant has not stated the a
barometric pressure used in the analysis.
The applicant will be required to discuss and justify the barometric I
pressure for the plant site.
This matter will remain an open item pending the receipt of additional information.
The staff's review of the applicant's containment response analysis has included the postulated reactor I
coolant system and secondary system pipe breaksi initial conditions, input parameters and assumptions.
However, the methodology umed to calculate the mass and i
energy release rate data for the LOCA and MSLB accident has not been reviewed due to a lack of information (see 8 e c t ion 6.2.1.3 and 6.2.1.4 o f t he SER).
Therefore, the staf f can not conclude on the acceptability of k
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the applicant's analysis at this time.
This wiLL be l
an open iten untiL further information is provided by i
the applicant regarding the calculation of the mass and energy release data.
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Protection Against Demane from External Pressure Thg containment structure is designed to withstand the external (differential) pressure load due to a post-u lat ed inadve rt ent actuation of the containment quench sp ray syst em during normal plant operation.,The maximum pressure differential is based on the dif f erence between the maximum barometric pressure and the minimum attainable internal containment pressure.
l The applicant catculated a minimum internal pressure of 8.0 psia for this po,stulated event.
The st af f has reviewed the applicant's analysis and W45 found that theappLicent'sassum/ptionsregarding initial containment conditions and containment quench
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spray system operation tend to minimize the
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containment pressure (e.g., minimum initial air partial pressure, maximum initial containment temp-erature and final containment temperature,wh+th equalg TU I
the minimum RWST temperature).
The applicante.?_7.Z?A assumed a barometric pressure of 14.36 psia, which is the maximum expected barometric pressure for the Beaver Valley 2 site.
Based on the conservative analysis performed by the applicant, the staf f conc Ludes that the containment external (differential) pressure i
design basis is acceptable.
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6.2.1.2 Subcompartment Analyses Subcompa rtment analyses a re required to det ermine the a
acceptability of the design dif ferential pressure i
Loadings on containment internal structures from high i
energy line ruptures.
The applicant has performed the necessary subcompartment analyses for the reactor c avi ty, )( s t eam gene rat or compa rtment$ and the i
pressurizer compartment, wh ere high energy Line
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ruptures are postulated to occur.
The applicant has developed modeis for each subcompartment, with a f
selected pipe break location, type and size, and initial conditions, that result in maximum
'c differential pressure loads on the subcompartment walis.
i The mass and energy release rate data used in the sub-compartment analyses were calculated using the SATAN-VI computer program (WCAP-8306).
The acceptability of using SATAN-VI for this purpose is currently under separate staf f review.
This matter wilL remain an open l
item until such time that pending staf f information i
needs under the We stinghous e Top ic al R eport Review I
are satisfied.
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T he applicant used the THREED comput er program to anaJyze the pressure transients in the reactor cavity, t he steam generat or compartment and the pressurizer compartment.
The staff's confirmatory analysis is based on the COMPARE-MOD 1 A computer code.
A separate discussion and review of the analyses of the reactor cavity, steam generator and pressurizer compartments are presented below.
I Reactor Cavity Analysis a The reactor cavity is a heavily reinforced concrete structure that performs the dual function of providing reactor vessel support and radiation shielding.
For the reactor cavity analysis the applicant postulated a 150 in cold Leg, limited displacement rupture (LDR) at the reactor vessel nozzle. The staff has reviewed the applicant's ana Lysis and concurs in the selection of the design basis pipe break, contingent upon the acceptability of the mechanically constrained Limit on the pipe break size.
(S ee Section 3.6 of the SER).
The reactor cavity subcompartment modeL employed by the applicant was developed to account for atL important model obstructions to flow.
This4 s consistent with the 1
recommendations concernin nodalization that are
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p presented in'NUREG/CR-1199, "Subcompartment Analysis Procedures Report."
We have examined the applicant's is eh in accordoorc e; b coem ue c.
nodat model and find it acceptable.
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g in *J.'EM /0C itec +4xrefece, is 4j A%,
The applicant calcuated a peak differential pressure Load on the reactor cavity wa LL of 115.9 psid, for the r,M sele e - of +t br-ok ne. !ce von o.s.J 2
c design basis 150 in LDR.
L ', ::: c;;;;;t i:n:-wt4.l-4e:d bj 4e. u se oc mttnin45 += Irmit the. brek. oreo. ore 45 cuss 5J in Sec. won 3,c.,
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.nnl4enn+ 4n the reartne emu 4+y annfye4e kaum kann
""T~h t o g$ u m e.l i nt A'e l Ce nc!ib'o ris o re. c h en e ri +rs nmssim f 3t. M e* viewed and fanad
+a h-ete=atebte.
In addition, the cit!.(* r e nk'o ( pras Wf e r'es ye.%e.5, staff performed a confirmatory analysis using the COMPARE-MOD 1A computer, code,wh confirmed that the applicant's result is conservative.
However, the design basis value of the differential pressure load on the reactor cavity wall is not documented in the FSAR; therefore, the staff can not confirm that the reactor cavity wall design basis is satisfied.
This wiLL be an open it em pending the receipt of additional information from the applicant.
The applicant has not provided in the FSAR an analysis of the forces and moments on the reactor vessel due to the differential pressure across the vessel caused by a reactor cootant system pipe break within the reactor cavity.
This matter wilL be an open item pending the receipt of additional information from the applicant.
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e Steam Genera' tor Subcompartment Analyses g.
S t eam generator cubicle 2 was selected as the rep-I retentative steam generator cubicle since atL three steam gene rat or cubicle s are simila r in design.
The applicant analyzed thrce RCS breaks in the steam generator compartment to evaluate loads on the sub-c omp a rt me nt walts 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 rupture s analyzed include thes5eamgeneratoroutlet a 360-in t.D R at nozzle, a
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180-in LDR at the reactor coolant pump (RCP) outlet nozzle, and a 7
-in Longitudinal intrados split Q_ _.. _.
break at the steam generator inlet elbow.
These I
breaks were chosen from the nine breaks in the applicant's sens itivity study as being Limiting cases which envelop conditions resulting from atL nine I
breaks.
The staff has reviewed the spectrum of postut at ed breaks analyzed by the applicant and finds them acceptable.
The applicant's nodalization scheme of the steam generator subcompartment was developed to take into account atL significant physical obstructions to flow.
The staff has reviewed the applicant's modeL and finds it acceptable.
The results of the applicant's analyses
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predict ap ak differential pressure of 12.9 psid for the design basis 707-in Longitudinal intrados split
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break.
However, the design basis value of the differ-Lomb
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ential pressure on the steam generator wall' is not 3
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documented in the FSAR.
This wilL be an open item pending the receipt of additional information from the applicant.
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Pressurizer Subcompartment Analyses i
i The applicant considered three breaks for the pressurizer cubicle, and the pr,essurizer relief tank cub i c le; namely a sp ray line DER in the upper pressurizer cubicle, a surge line DER at the
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pressurizer nozzt e and a surge line DER in the p
Il pressurizer relief tank cubicle.
The applicant's 1
nodalization models of the pressurizer subcompartment i
vere developed to take into account atL criticat l
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restrictions to flow.
The staff has reviewed the applicant's models and the spectrum of postulated breaks and finds them appropriately conservative and acceptable.
The results of the applicants analysis of the spray line DER in the upper pressurizer cubiete gave a peak dif f erentia L pressure of 18.07 psid ac ross the l
pressurizer nodel boundary surf ace.
However, the k
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design basis' value of the differential pressure load on the pressurizer cubie te walls is not documented in the FSAR.
This wi LL be an open item $$ pending the receipt
- er5, of ' additional information from the applicant..
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1 6.2.1.3 Mass and Energy Release Analyses for Postulated LOCA i
The applicant calculated the mass and energy release rate data for reactor coolant system pipe breaks at three
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break locations including the hot Leg piping between f
the reactor vessel and steam generator, the cold Leg pi pi ng at the pump suction, and the cold leg piping at the pump discharge.
The results indicate the pump suction break is the worst case for Long term con-(
tainment depressurization, and the hot Leg break is PcAE the worst case for containment pressure.
The staff cS- -
g has reviewed the applicant's spectrum of breaks, the desc ription of the LOC A t ransient models and the s i ng le failure considerations, and finds them acceptable.
1 The met hod used by the applicant to compute the mass l
i and energy release rates from reactor coolant pipe j
l breaks for the containment functional analyses is i
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described in a reference Westinhouse lett er that is j
currently under staff review.
At this time, we are not in a position to conclude on the acceptability of the blowdown methodology.
This matt er wilL remain an
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open item pending the completion of the staff's review.
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6.2.1. 4 M a s s and Ene'rgy Release Analyses for Postulated 4
Secondary System Pipe Ruptures The applicant has computed the mass and energy release t
rates for postulated main steam Line breaks using the M ARVEL Ccaput er Code (W CA P-8843, 1977).
H o w'e ve r, the I
mass and energy release data for the MSLB analysis were not documented in the FSAR.
The staff has requested this information for review, and to facilitate the
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1 staff's confirmatory analysis.
This matter wilL remain an open it em pending the receipt of additional infor-1 mation.
6.2.1.5 Minimum Containment Pressure Analysis for Emergency I
Core Cooling System Performance Capability Studies Appendix K to 10CFR Part 50 requires that the con-t ai nm e nt pressure used for evaluating core cooling effectiveness 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 resistance to i
st eam flow in the reactor coolant loops and increases the boitoff rate from the core.
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The applica'nt
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has performed the required containment os.JJrm-J in swir n 4. 2, i. 5 c4 M F5 A R,
back pressure calculationsusing the methods and ass-A umptions described in " Westinghouse Emergency Core Cooling System Eva luation Mode-Summa ry," WC AP-833 9, Appendix A, for the Limiting case LOCA, the double-ended cold leg guitLotine break (Cp = 0,4). (i.e, the h
fieb break found to produce the highest peak clad temp-erature).
Mass and energy release rates for this break were calculatedg using the method described in Se ction 15.6.5 of t he FS AR.
This method is evaluated separately in Chapter 15 of this SER.
The staff has reviewed the applicant's input parameters
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used in the minimum containment pressure analysis including initial containment conditions, containment net free volume, containment active heat removal, passive heat s i nk s, heat transfer to passive heat sinks, and found them to be acceptablpr conservative, and in conformance with BTP CSB 6-1.
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-6.2.1.6 Summary and Conclusions The staff has evaluated the Beaver Valley, Unit 2 con-tainment functional design with respect to the I
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 k
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General Design criteria 13,16, 38 and 50 have been me,t with the following exceptions:
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The method used by the applicant to compute the mass and energy release rates f rom postulated AIL 6 f
y's reactor coolant system pipe breaks for the
$gF gP g) containment analyses and for the subcompartment
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analysis h:: x t h approved by the staff.
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JJ g4 2.9 The mass and energy release data for postulated g// A"Mf k' gI' ' A ' stJ fplC,
.O 1 j main steam Line breaks have not been documented M
ak in t he FSAR.
S t af'f a c c ep ta'nce of t he appli c ant 's T
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V6 6a main steam line break analysis is continge(upon 6 jV the receipt of this information. h f
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There are two open items concerning the staff's review of the applicant's subcompartment analysis.
i First, subcompartment design pressure dif ferentials for the reactor cavity andg-steam generator and j
pressurizer compartments have not been documented in the FSAR.
Second, the applicant has not provided an analysis of the forces and moments on the reactor vesset due to the dif ferential I
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pressure caused by a RCS break within the reactor cavity.
Staff acceptance of the applicant's subcompartment analysis is contingent upon the
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receipt of this information.
6.2.2 Containment. Heat Removat Systems The function of the containment heat remova L syst ems is to remove heat from the containment atmoshpere to i
limit, reduce and maintain at a c c ep ta b ly low levels, the I
containment temperature and pressure fotLowing a loss of coo tant ac ci dent ormai$ steam Line break.
In addition to heat removat provided by passive means such as heat transfer to containment structures and components, the
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Beaver Valley 2 design includes active containment heat remova L systems (CHRS).
The active CHRS includes two spray systems; namely, the quench sprray system (GSS) and
-41 the re circulation spray system (RSS); the containment air coolers are not included in the CHRS.
The CHRS is designed to depressurize the containment to a sub-atmospheric condition within one hour.
For a discussion of the fission product removal fun ction of the CHRS, see SER Section 6.5.
The QSS is composed of two redundant 10D percent capacity trains each containing a quench spray pump, a chemical j
injection system and riserpipe leading to two spray headers.
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The two trains, connect to the two common 360-degree
[s spray headers in paratLet with risers 180 degrees apart.
There,areatotalof159;fpracomodel1713Anozzleson 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 h e a'd e r s.
Both spray pumps operating together can supply approximately 4500 spa to the spray headers.
The QSS is designed to spray c.old borated 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 (NaOH) solution from the chemical additive tank (CAT) is added
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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 chemical injection pump is automatically diverted to the containment sump.
The RSS is designed to provide additional depressur-ization of the containment and to maintain the con-tainment at a subatmospheric condition in the Long ters l
fotLowing the accident.
The RSS consists of two 360 l'
I degree spray ring headers and four pumps and heat exchangers.
Each spray ring header contains 292 SPRACO j
I model 1713A nozzles, and is fed by two risers, with
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each riser originating from one of the recirculation l
coolers.
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The two redundant recirculation spray pumps that feed each beader are each supplied with emergency power from separate diesel generators.
Each RSS pump takes suction from the containment sump at approximately 3480 gpm (50%
heat removal capacity).
Th e RSS is capable of operating in the post-accident environment to maintin a sub-atmospheric pressure for 30 days folLoving a high energy Line break.
i The RSS pumps are started automatically about 628 seconds after receipt of a CIB si'gnal, and'the spray becomes E M FC. M E about 714 seconds after the CIB signal.
When the water in the ST reaches a predetermined low level, the flow from two of the RSS pumps is automatically diverted to
$F the cold leg recirculation mode)ry ECCS.
The CHRS satisfies the provisions of Regulatory Guide 1.26, " Quality Group Classifications for Water, Steam and Radioactive-Waste Containing Components of Nuclear Power
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Plants," and 1.29, " Seismic Design Classifications,"
i for engineered safety features.
The applicant has provided testing information ( F S A R S e c t i on 14. 2, 'LJ f>0T* "' '"
4:+evw..
.3o p w,+y, demonstrating the ability of the quench spray system and 14 ? ' TaihW h r Pa q w recirculation spray system to function following a Amy.pe postulated single active failure.
c.,ch.Jt GJur p-
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h I f
i 18 -
j Regulatory Guide 1.82, " Sumps for Emergency Core Cooling O
and Containment Spray Systems," provideds design
~
guide. Lines for containment sumps that are to serve as sources of water for ECCS and the containment sp ray system fotLowing a LOCA.
The guidelines address redundancy, location and arrangement criteria,'as welL as debris screen provisions to ensure adequat e SW pump performance.
The staff has review the Beaver t
Valley.2 sump design against this guidance.
A single containment sump,,has been provided, and is enclosed by a protective sc reen assembly that has a r
[
2
~
total screen area of about 150 ft Furthermore, the containment sump is divided at the center Line by screening and vertical bars so that a failure of either half would not a dve rs ely affect the other half.
The redundant recirculation pump suctions are Located in seperate halves of the sump.
Therefore, even though the single sump design is not in accordance with Regulatory Guide 1.82 recommendations, the staff has concluded that adequate measures have been taken to assure that the RSS function wiLL not be Lost.
t 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 mesch screen opening is 1
1
[
smalter than the smaltest coolant passage gap in the reactor core and smalter than a spray nozzle orifice.
The screen assembly rises vertically approximately 5 feet above the containment floor, and is arranged so that no single f ailure could result in the clogging'of aLL suction points of the recirculation spray system.
Fo l L owi ng a LOCA, the top of the screen assembly would t
be under about 10 feet of water.
System design attows for 50 percent b L ock age of the sump screening without loss of function.
HowSver, the" applicant should further justify the acceptability of 50 percent blockage assumption by specifying the types (and quantity of each ype) of insulation used within the Beaver Valley 2 containment, and discussing the susceptibility of the insulation of become distodged by virtue of its proximity j
i to high energy Line piping.
i The applicant has conducted containment sump model testing at the Alden Reserach Laboratory, but has not reported the results to the staff.
The staff has learned, however, that the sump modeL used dif fers from the sump design shown in the FSAR.
The staf f has requested the applicant ji!
to provide the results of the Alden sump tests and discuss the significance of the results relative to k
- 19 a
/
L the performance of the as-built, Beaver Valley 2 sump.
Tkis information has not been received.
This matter wi.it remain an cpen item pending the receipt of the Alden test report and an accompanying discussion of the applicability of the results to the as-built Beaver Valley 2 sump.
The staff has reviewed the net positive suction head (NPSH) calculations submitted by the applicant.
The analysis shows the NPSH a v a i la b'L e to the reciculation pumps during both the spray mode and the low head safety injection mode is always greater than the required NPSH.
The applicant has complied with the provisions of Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core
4.
[
Cooling and Containment Heat Removal Systems", with one exception.
Regulatory Guide 1.1 states that containment heat'removat systems should be designed so that adequate NPSH is provided to system pumps assuming maximum expected temperatures of pumd Luids and no increase in
-4 containment pressure f rom that present before the po stulated LOC A.
Instead, the applicant calculated the 65 Waa.! :n ihm A R, $ ce.H cm G.T. 2. S. ~4 NPSH available 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 th e sump, e nsur i ng t h at credit is not taken for containment pressurization during the transient).
The staf f has p reviously found the saturat ed sump model to be cons ervative and, the refore, accept able.
The staff has reviewed the information in the applicant's FSAR and in responses to staff reques ts for addtional information concerning the containment heat removat systems to assure conformance to the acceptance criteria contained in SRP Section 6.2.2.
The staff finds that the containment heat removat systems satisfy the requi rement of General Design Criteria 38, 39, and 40, and the provisions of Regulatory Guide 1.1 on an acceptable alternative basis as defined above.
- However, there are :everal issues in Regulatory Guide 1.82 which
(
the applicant has not adequately addressed, and for
, i
[
which additiona information is needed before the staff t
ca n conclude on t he a cceptability of the sump design.
In considering the Location of the sump within the containment, the applicant should discuss the potential f or whipping pipes, high velocity jets of water or steam, or direct st reams of wate r (which may contain entrained debris) to adversely af f ect the integrity or performance of the sump protective screen assembly.
The applicant shoutd.also address the acceptability of th e wa ter velo city at th e,.f ine mes h s creen, based on one-half of t he available free area to account for blockage.
The a cceptability of the materials used
(
in the construction of the sump screen assembly, and the inservice ins p e c t.i on requirements for the sump I
components, as we LL as the provisions made to r
facilitate such inspections, should also be addressed.
6.2.3 Secondary Containment Functional Design The Beaver Valley 2 design does not include a secondary containment.
6.2.4 Containment Isolation System The function of the containment isolati on syst em (CIS) i s to allow the normat or emergency passage of fluids through the containment bound a ry while preserving the ability of the boundary to prevent or Limit the escape
[
of fission pro ucts that may result from postulated j
accidents.
In general, for each fluid system penetration at least two barriers are required
~
between the containment atmosphere or the reactor coolant system and the outside atmosphere, so that f ailure of a single brrier wilL not prevent isolation of the containment.
Containment isolation for Beaver Valley 2 is accomplished in two phases.
The containment isolation Phase A (CIA) e signal isolates atL non essential system Lines i
penet rating the containment, and is initiated by any of I
the fotLowing:
(1) high containment pressure (Hi-1 l
setpoint); (2) Low compensated steam Line pressure; I
/
(3) pressurizer low pressure; or (4) manual actuation.
The containment isolation Phase B (CIB) signal isolates l
i the component cooling water supply and return Lines for i
the reactor coolant pumps (RCPs) and control rod drive mechanism (CRDM) shroud cooters, 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 Table 6.2.4-1.
The applicant has documented that each system Line having automatic k.
containment isolation valves, which must be immediately i
)
o
[
isolated following an accident, is isolated by one of the signals in, Table 6.2.4-1.
Although the Phase B isolation
\\
signal is not actuated by divers e pa rameters, it is acceptable because the affected Lines are considered important to the safe shutdown of the plant and are capable of remote manual isota-tion.
The staff concludes that adequate diversity has been provided with regard to the dif f erent monitored parameters which actuate containment isolation.
TABLE 6.2.4-1 CONTAINMENT I S O L A.T.I O N SIGNALS AND ACTUATION PARAMETERS Containment Isolation Phase A signal
(
a.
High Containment Pressure (Hi-1) b.
Low Compensat ed Steam Li ne Pressure
/
c.
Pressurizer Low Pressure b h.h'er fb m 'd i
'.s a
5 "
I A. N "4'm Co.do?., m e ist hre%
d.
Manual Actuation
. + -
b' A ' " ' " # "
-C ent ei n;; nt Is;L:t';n "5::: 9 Si;r.:t Qf e.t Injec tio n Sh r a (
- a. ' Hi h Containment Pressure (Hi-1) b.
Low Compensated Steam Line Pressure c.
Pressurizer Low Pressure d.
Manual actuation
^ '
24 -
O Ma in Steam I s o l'a't i o,n S i gn a l
[
a.
High Steamline Pressure Rate b.
High C,ontainment Pressure (H i-2) c.
Low Steamline Pressure d.
Manual Act uation F eedwate r Isolation Signal a.
Steam Generator Hi-Hi Water Level b.
Safety Injection Signal c.
Low TAVG and. Reactor Trip O
Containment Vacuum Sy st em Iso [ation Signal a.
Containment Isolation Phase A Signal {Ni-l) b.
Manual Actuation I
The staf f has reviewed the applicant's containment iso Lat ion system design bases and containment isolation provisions as documented in Table 6.2-60 of the FSAR, for conformance to Genera L Design C riteria (GDC) 54, 55, 56 and 57 and R e gu l at o ry Guide 1.11, " Ins t rume nt Lines Penetrating Prima ry i
Reactor Containments".
The applicant's containment isolation system design is summarized as folLows:
(1)
There are at least two barriers between the atmosphere I
outside containment and the atmosphere inside cont ainment (or the RCS) on each system Line penetrating the containment.
(
(2)
The two barriers consist of one of the fot Lowing a r r a ng e men ts :
[
~ 25 -
a.
two normalLy closed manual valves with administrative control, one inside containment and the other outside con t ai nm e n t; b.
two automatic isolation valves, one inside containment and the other outside containment, a
simple check valve may not be used as the automatic isolation valve outside containment; c.
one automatic isolation valve inside containment and one normaLLy cloied manual' valve under administ rative cont rol outside containment (or th e revers ed arrangement);
d.
a sealed system (closed system) i nside 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 t *, their open position post accident.
These valves are remote manually controlled and operated from the control room.
1
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,,,,7,.,
(4)
Motor operated valvos (MOV) are used for systec Linos wh i ch a re pa rt,, of a n ES F re la t ed sy s t em, and f ail "as
[
is" on loss of power supply.
Solenoid operated valves are used when greater reliability post accident and a safe-failure position are required.
AlL power operated
~
valves are designed to fail in the position that provides greater safety upon loss of power or control air.
(5)
Mechanical and electrical redundancy are provided by designing two isolation barriers between the RCS or atmosphere inside c o nt a i nm ent and the a tm osp h e re outs ide containment with two separated IE power sources.
15 Lecer pl?s h e d tviN, (6)
Containment purge system isolationgazes two 42-in.
g w h ic. h a r 4'.,
but t e rf ly valves,3only open during plant cold shutdown andclose/ automatically within 10 seconds upon receipt
(
of a high radiation signal.
(7 )
The containment isolation system is designed to meet J
the singte failure c rit erion.
(8)
The closure time for each containment isolation valve is less than 60 seconds.
Sy s t em Lines which have no post-accident f unction a re provided with ai r-operated valves 4
(A0V) with closure time of 10 seconds.
The appticant's c on t a i nm e nt isolation provisions are reviewed against the requirements of GDC 54, 55, 56, and 57 (Appendix A t o 10 C FR Part 50) and the supplementary guidance of SRP 6.2.4, where applicabte.
Staff *1ireview has confirmed that the k
containment isolation system meets the explicit requirements
.8.
of GDC 54, 55, 56, and 57 with fotLowing exceptions:
- 2'? -
(1)
The containment vacuum' pump and hydrogen recombiner suction Lines are provided with two solenoid operated
{
isolation valves in series outside containment.
T he r e'f o r e, the containment isolation provisions dif f er f r om 'the e xp lic it requirements of GDC 56. However, the isolation va Lves are located as close as possible to the containment, 'and the associated syst em piping is designed in accordance with the break / crack exclusion criteria of B ranch Technical Position MEB 3-1.
Furthermore, the valvesare hermetically sealed, precluding the need to encapsulat e the valves.
Since the lines are used post-accident, for containment atmosphere sampling and hydro-gen control, locating the valves outside containment im-1 proves the functional reliability of the valves.
There-fore, the staff finds the isolation provisions for these Lines to be acceptable alternatives to the explicit re-quirements of GDC 56.
(3)
The emergency core cooling system safety injection lines and reactor coolant pump (R CP) seal injection Lines are equipped with weight-loaded check valves inside con-tainment and motor operated valves (MOV), outside contain-ment which do not re ceive a containment isolation signal to close.
The safety injection Lines discharing to the hot and cold legs of the reactor coolant system and the RCP seat injection lines are important to safe shutdown or are pa rt of an engineered safety feature system.
..~.
f Provisions have been made to detect possible Leakage f rom these Lines outside containment, thereby atLowing z
remot~e manual instead of automatic isolation valves.
The st a ff, the ref ore, finds that the containment isolation provisions for these lines are acceptable alternatives to the explicit requirements of GDC 55.
(4)
The quench spray pump discharge and recirculation sp ray pump discharge Lines are provided with a normalLy open, remotely cont rol L ed, mot or operated valve outside c on t a i nm ent and a weight-Loaded check valve inside c o nt a i nm e nt.
The isolation valves in the
{
c ont ai nm ent depressurization (quench and recirculation I
spray) syst ems open upon re ceipt of a CIB signal, if not already open, with the exception of the caustic addition line to the containment sump which automatically opens af ter the quench spray discharge has stopped.
Th e r e c i r-culation spray pump suction Lines are provided with a s ing L e, no rma l L y open, remote Ly cont roll ed, motor oper-ated va Lve outside containment sinc e it is not practical to lo cate a second valve inside containment where it would be submerged folLowing a LOCA; these valves do not l
k.
i l
receive an au omatic isolation signal for closure.
The refore, the containment isolation provisions for these Lines differ from the explicit requirements of GDC 56 reg a rd i ng their actuation and number.
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 sa fe ty grade.
Therefore, the staff finds the use of re-note manual instead of automatic isolation valves accep-table.
In addition, the ssing Le isolation valve outside su We n c on t ai nm ent in the recirculation spray pump Lines is ac-3 ceptabte because system reliability is improved with a
(
singte valve and the piping between the outside of the containment watL and the isolation valve, as welL as the valve, are contained within a leak-tight encapsulation.
T,he st aff has a tso revi ewed inf ormation provided by the appli-cant to demonstrate compliance with the provisions of NUREG-0737 Item II.E.4.2, " Containment I solation Dependability".
As
'pr eviou sly desc ri bed, the applic ant has complied with the pro-visions regarding diversity in pa ramete rs sensed f or initiation of containment isolation, and has considered the functional k
- the functional reg irements of atL systems penetrating containment and has made acceptable provisions for isolation $of systems not required for aitigation of the consequences of an accident or safe shutdown of the plant.
The applicant also made provions that resetting of a containment isolation si gnal wilL not result i n the automatic reopening of containment isolation valves.
In addition, the applicant has designated atL system Lines penetrating the containment as essential or non essential systems by appropriate signals.
Therefore, the staff concludes that the pplicant has complied with the provisions of NUREG-0737 It em 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 t he eff ects of natural phenomena to ensure their p erf orman ce under a t L anticipat ed 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, I
i
._--,-.--_m---._
---__.m-
y r
"S eis mi c De si gn Classif ication", and 1.26, " Quality Group
(
C Lassifications and Standa rs ds for Water, steam, and R a di oa c ti ve-Wa s t e-cont ai n ing Components of Nuclear Power P lant s.".
In summa ry, the staff has reviewed the information in the applicant's FSAR and in response to 'NRC Questions conc erning the containment isolation system to assure conformance to all of the acceptance criteria contained i
in SRP Section 6.2.4 The staff conclude
- that the Beaver Valley 2 containment is,olation system meets the requi reme nt s o f Gene ral Design c rit e ria 1, 2, 4,16, 54, 55, 56, and 57, and is, therefore, a cceptable.
6.2.5 Combustible Gas Control System Following a loss of coolant accident, hy#rogen may accumulate as a result of (1) metal water reaction between the zirconium fuel cladding and the reactor coo tant, (2) radiolytic decomposition of the water in the reactor core, (3) radio-Lyt ic decomposit ion of the water cotLected on the sump floor, (4) hydrogen released from the pressurizer gas space and re-ac tor coolant, (5) corrosion of metals by the alkaline solu-tion used for containment sp ray.
The function of the com-bustib Le gas control system (C GC S ) is to monitor and control the potential hydrogen accumulation within the containment at-mosphere below 4 volume percent fotLowing a design basis
(
accident.
In the event of a LOCA, two redundant, independent, full capa-i city electric hydrogen recombiners wilL be available outside 1
l containment to control the containment hydrogen concentration.
Each recombiner has a capacity of 50 SCFM and is designed to i
Seismic Category I criteria.
One hydrogen recombiner is per-l manently installed in the safeguards area; the other recombiner I
wilL be transf e rred from Beaver Valley, Unit 1 and installed in the safegua rd area fot Lowing an accident.
(In addition to the f
i two safety relat ed hydrogen recombiners provided, a non safety r
grade containment purge syst em is available to purge the con-
~
tainment atmosp he re as an aide to (Leanup.)
Each hydrogen re-combiner sy st em includes flow control capability, a blowe r, a temperature-controlled electric preheat er, a thermal recombiner,
(
and an air blast heat exchanger.
The safeguards a rea is a Sei-smic Category I concrete structure locIt ed adjacent to the con-l tainment.
The penet rations, and components within the safe-t l
gua rd area are protected against tornados and missiles.
The hydrogen recombiners and alL associated valves are remote man-f ualLy cont rolled f rom penets locat ed in the safeguards area,
[
i out s ide of the recombiner cubicles, to allow access and mini-l t
size exposure of personnel.
The staff has reviewed the hydro-l gen recombiner sys tem design con cept and finds it acceptable.
I i
P Two re dun da nt, i ndepend ent hyd rogen ana Lyzers are installed in the cable vault area to monitor the hydrogen concentration in the containment atmosphere.
The analyzers are also used to 1
check the efficiency of recombiner operation.
The hydrogen I
analyzer is classified as Class IE, Seismic Category I and func-
~
.tional tested with a calibrated gas sample.
Indicators are provided in the main control room to monitor hydrogen concen-tration.
Annunciation is also provided in the main control room for hydrogen analyzer /recombiner local panet troubL'e.
Based 'on the staff's review, the post accident hydrogen monitoring system meets the requirements of NUREG-0737 It em II. F.1, At t a chment 6, " Containment Hydrogen Monitor"j and the single failure cri-terion.
However, the applicant has not required a sufficient-Ly complete description of th operating characteristics of the hydrogen analyzer to be installed.
[
The applicant has analyzed the potential hydrogen generation within the containment using the guidelines provided in Regula-e tory Guide 1.7, and catculated the hydrogen concentration for both one and two recombiner operation.
The analysis shows that a single recombiner, initiated when the containment hydrogen concentration reaches 3.1 volume percent (i.e., approximately 4 days post accident), is sufficient to eaintain the hydrogen 9
concentration in the containment atmosphere below the Lower flammability Limit of 4 volume percent.
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 c on f i rme d, using the COGAP computer k
e.
code, that there is sufficient time before the containment hy-drogen concentration reaches 3.1 volume percent to manually initiate the post accident hydrogen recombiners, and that a single recombiner can acceptably control the hydrogen concen-tration in containment below 4.0 volume percent.
The applicant has stated in the FSAR that the containment de-sign attows air to circulate f reely.
Furthermore, atL cubi-cles and compartments within the containment are provided with openings near the top as weLL as openings in the floor to atlow air circulation.
The applicant has also performed an analysis to demonstrate that adequate mixing of the hydrogen in the con-tainment atmosphere wilL be ensured by the turbulence created by
(
the containment spray system and thermal convection.
There-fore, sufficient mixing of hydrogen in containment vill occur e
to prevent stratification and to eliminate areas of potential stagnation.
The staff finds that adecuate passive and/or ac-tive design measures have been incorporated into the contain-ment design to ensure adequate hydrogen mixing within contain-ment and, therefore, the applicant's hydrogen mixing provisions are acceptable.
In summary, the staff has reviewed the information in the appli-cant's FSAR and in response to our questions concerning the
)
combustible gases control system to assure conformance to atL k.
)
l-r of the acceptance briteria contained in SRP Section 6.2.5.
The
{
staff concludes that the applicant's combustible gas control sys-tem meets.the requirements of GDC 41, 42 and 43, satisfies the design and perf ormance requirements of 10 C FR 50.44, the provi-sions of Regulatory Guide 1.7 and the requirements of NUREG-0737 Item II.F.1, Attachment 6, except for the followind item.
The applicant has not discussed in sufficient detail the perfor-mance characteristics of the actual post-accident hydrogen non-itoring system to be installed.
Therefore, this wilL remain an open item pending the receipt of additional information.
.as 6.2.6 Containment Leakage Testing Program The containment design includes the provisions and features
(
required to satisfy the testing requirements of Appendix J to 10 CFR Part 50.
The design of the containment penetrations and isolation valves permit preoperational and periodic Leakage rate testing at the pressure specified in Appendix J to
The staff has reviewed the containment Leakage testing program contained in the FSAR and in the response to NRC Questions, and finds them acceptable with the fotLowing exception.
The appli-cant proposes to exclude certain valves f rom Type C testing (in-
%44 y irjee We'n my4RM cLuding the penetrations and recirculation spray system penetra-3 tions).
The justification for excluding penetrations from
. e 3
- 1 Type C testing wil'L be evaluated in conjunction with the staf f review of the facility Technical specifications.
Other than the exception mentioned above, the proposed reac-tor containment Leakage testing program complies with the re-quirements of Appendix J to 10 CFR Part 50.
Such compliance f
provides adequate assurance that containment Leak-tight inte-grity can be v'erified periodically throughout service Lifetime I
l on a timely basis to maintain such Leakage within the limits of the Technical Speci cations.
Maintaining containment Leak ge rates within such limits pro-vides reasonable assurance that, in the event of any radio-activity releases within the containment, the loss of the
(
containment atmosphere through the Leak paths wilL not be in excess of acceptable limits specified for the site.
Compli-ance with the requirements of Appendix J constitutes an ac-ceptable basis for satisfying the requirements of General De-sign criteria 52, 53 and 54.
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