ML20154D780
| ML20154D780 | |
| Person / Time | |
|---|---|
| Site: | 05000605 |
| Issue date: | 09/14/1988 |
| From: | Marriott P GENERAL ELECTRIC CO. |
| To: | Chris Miller NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation |
| References | |
| MFN-64-88, NUDOCS 8809160103 | |
| Download: ML20154D780 (114) | |
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l l GE Nuclear Energy l Pv v / w c Con f #v 1 $ Cu :m A.t a L,* Krt CA 9 1H September 14,1988 MFN No.64 88 Docket No STN50-605 i
Document Control Desk U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 i
Attention: CharlesI. Miller, Director Standardization and Non Power Reactor Project Directorate
Subject:
Submittal of Responses to Additional Information as Requested in NRC Letter from Dino C. Scaletti, Dated July 7,1988
Dear Mr. Miller:
Enclosed are thirty four (34) copies of the Responses to Request for Additional Information (RAI) on the Standard Safety Analysis Report (SSAR) for the Advanced Boiling Water Reactor (ABWR). These responses principally pertain to Chapters 4,5,6 and 15. Also included are other committed responses to RAls from Scaletti's letter, dated February 22,1988.
It is intended that GE will amend the SSAR with these responses in December 1988.
Sincerely, y
(', b P. W. $larriott, Manager 1.leensing and Consulting Services cc: D. R. Wilkins (GE)
F. A. Ross (DOE) k J. F. Quirk (GE) Q D. C, Scaletti (NRC) L y$k l
8809160103 680914 PDR ADOCK 05000605 -
A PDC l
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i s .a i ABWR m-n i' Sandard Plant nr n 4.6 FUNCTIONAL DESIGN OF (5) Each positioning device shall provide a !
REACTIVITY CONTROL SYSTEMS means to present or limit the rate of .
control rod ejection from the core due to a l The reactivity control systems consists of break in the drive mechanism pressure boun- l conteol rods and contral rod drives, dary. This is to prevent fuel damage tc. [
supplementary reactivity controlin the form of a sulting from rapid insertion of reacti ity. ;
gandolinia (Section 4.3), and the standby liquid (
control system (described in Subsection 9.3.5). 411.1.2 Power Generation Design kasis ,
Evaluations of the reacthity control sptems The control rod drhe system (CRDS) design l l
agalast the applicable General Design Criteria shall meet the following power generation design (GDC) are contained in the folowing subsections: bases. j i
QDC subsection (1) The design shall proside for controlling i changes in core reactivity by positioning I
23 3.1.23.4 neutron absorbing control rods within the !
25 3.123.6 core. !
26 3.123.7 !
L 27 3.12 3.S (2) The design shall proside for mosement and po.
25 3.1.23.9 sitioning of control rods in increments to j 29 3.123.10 enable optimited power control and core !
power shaping.
4.6.1 Iriformation for Control Rod Dri e 5,$ stem 4A1.2 Description f 411.1 Design kases The CRDS consists of fine motion con'rol rod drive (Th1CRD) mechanisms, and the CRD hydraulic 4.6.1.1.1 Safet) Design Bases system (including pumps, filters, hydraulic :
control units, interconnecting piping, instru. [
The control rod drive CRD mechanical sptem mentation and clectrical controls). The CRDS, ;
shall mect the following safety desi'gn bases: in conjunction with the rod control and infor. !
mation sptem (RC&lS) and reactor protection (1) The design shall proside for rapid control sptem (RPS) performs the following functions:
tod insertion (scram) so that no fuel damage !
results from any moderately frequent csent (1) Controls changes in core reactisit) by !
(see Chapter 15), positioning neutron absorbing control rods i within the core in response to control !
(2) The design shall include positioning signals from the RCAIS. I desices, each of which indisidually supports f and positions a control rod. (2) Prosides mosement and positioning of contiol l rods is increments to enable optimized power l (3) Each positioning desice shall be capabic of control and core power shape in response to l holding the control rod in position and control signals from the RCAIS.
preventing it from inadscrtently withdrawing I outward during any non accident, accident, (3) Prosides the ability to position large I post accident and seismic condition. groups of rods simultaneously in response to l control signals from the RCAIS. J (4) Each positioning desice shall be capable of (
detecting the separation of the control rod (4) Prosides rapid control rod insertion (scram) !
from the drise mechanism to present a rod in response to manual or automatic signals l
drop accident. from the RPS to that no fuel damage results a from any plant transient, f AnmJmen 3 4g
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ABWR mom.m Etandard Plant ut v A (5) Gathers rod status and rod position data for rod pattern control, performsnee monitoring, operator display and scram time testing.
(6) Presents undesirable rod pattern or rod mot.ons by imposing rod motion blocks in order to protect the fuel.
(7) Prevents and mitigates the conscquences of a rod drop accident by detecting rod separation and controlling rod pattern.
Avestment 3 4Me
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ABWR mun Standard Plant R TT H (8) Prosides alternate rod insertion (ARI), an inside each FhtCRD, high pressure water lifts the alternate means of actuating motor drisen hollow piston off the ball nut and drives the t ad insertion should an anticipated control rod into the core. A spring washer transient without scram (ATWS) occur, buffer assembly stops the hollow riston at the end of its stroke. Departure from the ball nut (9) Automatically drives in the drhe rnechanisms releases spring loaded latches in the hollow with the electric motors upon scram initia- piston that engage slots in the guide tube.
tion. This provides an additional, diserse These latches support the control rod in the means of fully inserting a control rod. Inserted position. The control rod cannot be withdrawn until the ball out is drisen up and
,, (10) Proddes selected control rod run in (SCRRI) engaged with the hollow piston. Stationuy
{ for reactor stability control, (See fingers on the ball nut then cam the latches out Subsection 7. 7.1. 2. 2. ( 2 ) ) . of the slots and hold them in the retracted position. A scram action is complete when every The design bases and further discussion of FhtCRD has reached their fully inserted position.
both the RC&ls and RPS, and their control inter.
faces with the CRDS, are presented in Chapter 7. The use of the rhtCRD rnechanisms in the CRD system prosides sescral features which enhace 4 1.1 Fine blotion Control Rod l) rise bot h t h e s y st e m r e h a bilit y a n d pl.i n t hlechanisms operations. Some of these features are listed and discussed briefly as follows:
The Fne motion control rod drive (FMCRD) used for positioning the control rod in the reactor (1) Diverse hicans of Rod insertion core is a rnechanical/ hydraulic actuated rnechanism (Figures 4.6 1, 4.6 2 and 4.6 3). An electric The FhtCRDs can be inserted either motor drisen ball out and spindle assembly is hydraulically or electrically, in response capable of positioning the drise at a minimum of to a scram signal, the ThtCRD is inserted 18.3mm increments. Hydraulic pressure is used hydraulically us the stored energ) in the for fast scrams. The FMCRD penetrates the bottom scram accumulators. A signal is aho gisen head of the reacter prenure scuel. The FMCRD simultaneously to insert the FMCRD does not interfere with refueling and is electrically sia its motor drive. This operatise esen when the head is remosed from the diversity provides a high deFree of reactor sessel. assurance of sod insertion or demand.
The fine motion capability is achiesed with a (2) Absence of fMCRD Picon Seah ball nut and spindle arrangernent drisen by an electric motor. The ball nut is keyed to ihe The FMCRD pistons base no seats and thus, do guide tube (roller key) to present its rotation not require maintenance, and traserses asially as the spindle rotates. A hollow riston rests on the ball nut and upward (3) FMCP.D Dhcharge motion of the ball nut drises this piston and the control rod into the core. The weight of the The water which scrams the control rod control rod keeps the hollow piston and ball nut discharges into the reactor sessel and does in contact during withdrawal. not require a scram discharge solume, thus climinating a potential for common mode A single hydraulic control unit (HCU) powers f ailu r e, the scram action of two FMCRDs. Upon scram sahe initiation, high prenure nitrogen from the HCU (4) Improved Plant Mancuserability raises the piston within the accumulator forcing water through the scram piping. This water is The fine motion capability of the FMCRD directed to each FMCRD connccted to the llCU. allows rod pattern optimitation in response A*t RJ'm4 *t 3 b
s i 23A6t00AI)
Standard Plant nrv n i operating buffer under criteria for normal 4.62JJ.2.2 Rupture of Hydraulic Line to and upset events and for an abnormally Drise Housing Flany operating buffer under criteria for upset !
, events. For the case of a scram insert line break, a partial or complete circumferential oacning is (5) The control rod is designed for lateral postulio d at or near the point whe re the line
- displacements due to the maximum fuel enters the housing flange. This fallac,if not ;
e channel deflection allowed within fuel mitigated by special design features, cosid ,
channel design criterig under upset (OBE) result in rod ejection et speeds exceeding maxi. ;
events and faulted ($5E) esents, mum allowable limits cf 4 in/sec (assuming rod ,
] pattern control) or 6 inches maximum travel j distance before full stop. Failure of the scram 44JJ.2 Centrol Red Driies i insert line would cause loss of pressure to the [
44JJ.2.1 Esalvation of Scram T1me underside of the hollow piston. The force l
) resulting from full reactor pressure acting on l i
The rod scram function of the CRD system the cross sectional area of the hollow piston, l provides the negative reactivity insertion litus the weights of the control rod and hollow ,
j required by safety design basis 4.C.1.1.1(1), pi; ton, is imposed on the ball out. The ball 4
The scram time shown in the description is nut in turn translates this resuliant force into i adequate as shown by the transient analyses of a torque acting on it.c spindle. When this
, Chapter 15. torque exceeds the motor residual torque and i seat friction, reverse rotation of the spindle L 44JJ.2.2 Analpis of hlalfunction Relating to will occur permialog iod withdrawal. Analyses a Rod %1thdrawat show that the fo, ,rs generated during this post.
! , ulated event can result in rod ejection speeds g There are no known single malfunctions that which exceed the masimum allowable limits. ,
cause the unplanned withdrawal cf even a single t control rod. Howeser,if multiple malfunctions The FhlCRD design presides two diserse means .
I are postulated, studies show that an unplanned of protection against the results of a (
. rod withdrawal can occur at withdrawal speeds postulated scram insert line failure. The first :
i that sary with the combination of malfunctions means of protection is a ball check valse f I postulated. located in the middle flange of the drise at the .
I scram port. Reserse flow during a line breal l 3
44JJJ.2.1 Drhe Housing rallure will cause the ball to move to the closed i i
position This will present loss of pressure to s The bottom head of the reactor sessel has a the underside of the hollow ristor'. which in :
penetration for each CRD location. A drise turn will present the generation of loads on the
! housing is raised into position inside cach drise which could cause rod ej:ction.
1 penetration and fastened by welding. The drhe ;
J is raised into the drise housing and bolted to a The second means of protection is the ;
l flange at the bottom of the housing. centrifugal brake described in Subsection ;
- 4.6.1.2.2.8. In the esent of the failure of the
l In the unlikely esent of a failure of the check vahe, the centrifugal brake will actuate !
j drive housing to sessel attachment weld and stop the ball spindis rotation before the !
j (including a failure through the housing or along maximum allowable ejection spced (approximately j the fusion line of the housing to stub tube weld) 5M rpm corresponding to 4 in/sec) is reached.
i or the flange bolting a:taching the drise to the i housing, ejection of the CRD and attached control 44JJJJJ Total Failurs:of All Drbe
]
1 mod is presented by the integral internal blorout Flange Bolts f
i support. The details of the this internal l blowout support structure are contained in The ThlCR1> design prosides an anti. rotation l Section 4.6.1.2.2.9. l i j i I i wm nu l 4
e- e AoM 234aooan Standard Plant _, ,
Rw c material adjacent to welds in Type 304 and Type manipulate the corrosioa potential in laboratory 316 stainless steel piping systems has occurred tests) (Reference 10).
in the past. Substantial research and deselop- 'n ment programs have been undertakea to understand As the corrosion potential is reduced below A the IGSCC phenomenon and develop remedial mea- the range typical of normal BWR power operation s sures. For the ABWR, IGSCC resistance has been (+ 50 to .50 mVSHE), a region of immunity to N achieved through the use of IGSCC resistant mate- IGSCC appears at 230 mVSiiE. It is apparent
- g rials such as Type 316 Nuclear Grade stainless that a combination of corrosion potential (which
' steel and stabilized nickel base Alloy 600M and can be achieved in a BWR by inlecting usually < l i 182M. 1 ppm hydrogen into the feedwater) plus tight Q conductivity control (0.2 pS/cm) should permit A Much of the early remedy development work fo- BWRs to operate in a regime where sensitized .
cused on alternative materials or local stress stainless steels are immune to IGSCC. In the "
reduction, but recently the effects of water che- reactor vessel, the excess hydrogen reacts with mistry parameters on the IGSCC process have rece- the radioloytic oxygen and reduces the electro- u ived increasing attention. Many important fea- chemical corrosion potential (Reference 10a and a tures oithe relationship between BWR water chem- 10b). The reactor water cleanup system, which ,'g istry and IGSCC of sensitized stainless steels processes reactor water at a rate of 2% of reted ,
have been identified, feedwater flow, removes both dissolved and <
- undissolved impurities that enter the reactor "I #
r N Laboratory studies (References 3 and 4) have water. The removal of dissolved impurities >
shown that although IGSCC can occur in simulated reduces the conductivity into the region of BWR startup endronments, most IGSCC damage pro- immunity to IGSCC.
l bably occurs during power operation. The normal BWR emironment during power operation is -2SO Since the ABWR has no sensitized stainless OC water containing dissolved oxygen, hydro- steel, IGSCC control by hydrogen injection is gen and small concentrations of ionic and non- not required, liowever, irradiation assisted !
r ionie impurities (conductivity generally below stress corrosion cracking (IASCC) can occur in '
a N 0.3 pS/cm at 250C), it has been well docu- highly inadiated annealed stainless steel and mented that some ionic impurities (notably sul- nickel base alloys. Preliminary irricactor and l fate and chloride) aggravate IGSCC, and a number laboratory studies (Reference 11) have indicated of studies have been made of the effects of inai- that liWC will be useful in raitigating lASCC. r vidual impurity species on IGSCC initiation and g growth rates (References 3 thru 7). This work In reactor and laboratory evidence also indi. l
- h clearly shows that IGSCC can occur in water at cates that carbon and low a!!oy stects also tend 2800C with 200 ppb of disselved oxygen, even at to show improved resistance to environmentally low conductivity (Iow impus" levels), but the assisted cracking with both increasing water pu-
- rate of cracking decreases wi h decreasing impu- rity and decreasing corrosion potential
- nty u ntent. Although BWl' neer chemistry (Refereace 12). Z guidelines for reactor water cannot prevent A i IGSCC, maintah.ing the lowest practically achiev. 5.2.3.2.2.1 Fuel Performance Considerations !
able impurity ?cvels will minimize its rate of D progression (References 5 and 9). Nuclear fuel is contained in Zircaloy tubes A
s that constitute the first boundary or primary Stress corrosion cracking of ductile materials containment for the highly radioactive species j in aqueous environments often is restricted to generated by the fission process; therefore, the
- specific ranges of corrosion potential *, so a integrity of the tubes must be ensured. Zirca.
1 number of studies of impurity effects on IGSCC loy interacts with the coolant water and some
] have been made as a function of either corrosion coolant impurities. This results in oxijation 2
potential or dissolved oxygen content (the by the water, increased hydrogen content in the dissobed oxygen content is the major chemical Zirealoy (hydriding), and, often, buildup of a variable in BWR type water that can be used to layer of crud on the outside of the tube. Ex.
& *Also called electrochemical corrosion potential or ECP, see Reference 9.
$28
- Amendment 3 4
MM , 33A61 NAB Standard Plant nry A cessive oxidation, hydriding, c crud deposition may lead to a breach of the cladding wall.
Metallic impurities can result in neutron losses and associated economic penalties which increase in proportion to the amount being introduced Ina the reactor and deposited on the fuel. With respect to iron oxide type crud depo- ,
sits,it can be concluded that cperat.%n f
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ABM 234enoan l Standard Plant nry c !
within the BWR water cheraistry guidelines (speci- suggests that these impurities change both the fically the limits on feedwater iron levels) ef- corrosion rate and the oxide film character-fcctively precludes the buildup of significant istics to adversely increase the cobalt 60 up-deposits on fuel elements. take. Tnus, controlling water purity should be beneficial in reducing radiation buildup.
5.23.2.2.2 Radiation Field Bulldup Prefilming of stainless steel in cobalt 60 The primary long term source of radiation free water, steam, or water / steam mixtures also fields in most BWRs is cobalt 60, which is formed appears to be a promising method to reduce by neutron activation of cobalt 59. Corrosion initial radiation buildup rates. As an example, poducts are released from corroding and wearing the radiation buildup rates are reduced surfaces as soluble, colloidal, and particulate significantly when samples are prefilmed in high species. The forination of cobalt 60 takes place temperature (2880C), oxygenated (200 ppb l after the corrosion products precipitate, adsorb, oxygen) water prior to exposure to cobalt 60 or deposit on the fuel rods. Subsequent re en- containing water. Mechanical polishing and trainment in the coolant and deposition on out- electropolishing of piping internal faces shot:Id of core stainless steel surfaces leads to buildup also be effective in reducing radiation buildup, of the activated corrosion products (such as co-balt 60) on the out of core surfaces. The depo. 5.23.2.2J Sources of Impurities sition may occur either in a loosely adherent -
layer created by particle deposition, or in a Various pathways exist for impurity ingress tightly adherent corrosion layer incorporating to the primary system. The most common sources radioisotopes during corrosion and subsequent ion of impurities that result in increases in reac- ,
exchange. Water chemistry influences all of tor water conductivity are condenser cooling '
these transport processes. The key variables are water inleakage, improper operation of ion ex-the concentration of soluble cobalt 60 in the re- change units, air ialeakage, and radwaste re. '
actor water and the characteristics of surface cycle, in addition to situations of relatively oxides. Thus, any reduction in the soluble co- continuous ingress, such as from low level con-balt 60 concentration will have positive denser cooling wster inleakage, transient events
, benefits. can also be si,cuificant. The major sources of impuritles during such events are resin intru.
As a means to reduce cobalt, GE has redt.ced sions, organic chemical intrusions, inorganic cobalt content in alloys to be used in high chemical intrusions, and improper rinse of re.
, , , fluence areas such as fu:1 assemblies and control sins. Chemistry transients resulting from intro- ;
Ia rods, in addition, cobalt base alloys used for duction of organic substances into the radwaste <
M3 pins and rollers in control rods have been system comprised a significant fraction of the ;
replaced with noncobalt alloys. transients which hase occurred. ,
~
The reator water cleanup system, which pro. The following factors are measured for cesses reactor water at a rate of 2% of rated control or diagnostic purposes to maintain feedwater flow, will remove both dissolved and proper water chemistry in the ABWR.
., undissolved impurities which can becor e radio.v 2 tive deposits. Reduction of these radioactoe (1) Conducthity M *g deposits will reduce occupational radiation exgv ,
sure during operation and maintenance of the Increasing levels of many ionic impurities '
plant compot.cnts, adversely influence both the stress corrosion cracking behavior of RCS Water quality parameters can have an influence materials, the rate of radiation field on radiation buildup rates, in laboratory tests, buildup and also can affect fuct the water conductisity and pH were varied syste. performance. Therefore, conductisity lesels ;
matically from a high purity base case, in each in the reactor water should be maintained at case, impurities increased the rate of cobalt.60 the lowest levcis practically achievable.
j uptake over that of the base case. The esidence
! Amendmtat 3
ABM 2346iootti Standard Plant arv ^
(2) Chloride Chlorides are among the most potent promo-ters of IGSCC of sensitized stainless steels
~
and are also capable of inducing transgran-I ular cracking of nonsensitized stainless steels. Chlorides also promote pitting and crevice attack of most RCS materials. Chlo- 0 l
rides normally are associated with cooling water inleakage, but inputs via radwaste
. processing systems have also occurred, j
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ABWR um*w nry c Standard Plant to have an important influence on IGSCC ini- tion monitor readings increase uith the hy.
tiation times for smooth stainir ss steel drogen addition rate. During initial plant specimens in laboratory tests, in addition, testing, the amount of hydrogen addition pli can serve as a useful diagnostic parame- required to reduce the electrochemical cor- ,
ter for interpreting severe water chensistry rosion potential to the desired range is g transients and pli measurements are ret s- determined at various power levels. Chan-mended for this purpose, ges in the main steam line radiation moni-tor readings at the same power level indi.
(10) Electrochemical Corrosion Potential cate an over addition (high readings) or under addition (low readings) of hydrogen.
The electrochemical corrosion potential (ECP) of a metalis the potentialit attains (11) Constant Extension Rate Tef,1 when immersed in a water environment. The ECP is controlled by various oxidizing Constant extension rate tests (CERTS) are agents including copper and radiolysis pro- accelerated tests that can be completed in ducts. At low reactor water conductivitics, a few days, for the determination of the
- the ECP of stainless steel should be below susceptibility to IGSCC. It is useful for 0.23 VS11E to suppress 1GSCC, verifying IGSCC suppression during initial
, implementation of hydrogen water chemistry l (11) Feedwater Hydrocen Addition Rate (llWC) er following plant outages that could have had an impact on system chemistry
- ' A direct measurement of the feedwater hydro- (e.g., condenser repairs during refueling). '
gen addition rate can be made using the hy.
drogen addition system flow measurement de. (15) Continuous Crack Gromh Monitorinc Test vice and is used to establish the plant-spe-cific hydrogen flow requirements required to This test employs a rescrsing DC potential satisfy the limit for the ECP of stainless drop technique to detect changes in crack steel (Paragrain 10). Subsequently, the ad- length in IGSCC test specimens. The crack dition rate m',asurements can be used to help growth test can be used for a variety of diagnose '.he origin of unexpected ECP pu poses, including the following:
changes (a) Initial verification of IGSCC suppres-
- (12) Enheulation System Water Dissolved sion following IlWC implementation.
liyd .oc e n (b) Quantitatise assessment of water che.
A direct measurement of the dissolved hydro- mistry transients. 6 g6n content in the reactor water serves as a ,
cross check against the hydrogen gas flow (c) Long term quantification of the success '
meter in the injection system to confirm the of the llWC program.
actual presence and magnitude of the hydrogen addition rate. The major impurities in various parts of a 1 BWR under certain operating conditions are
- (13) Main Steam Line Radinbn Level listed in Table 5.2 5. The plant systems have i been designed to achieve these limits at least The major activity in the main steam line is 90% of the time. The plant operators are nitrogen 16 produced by an (n, p) reaction encouraged to achieve better water quality by with oxygen 16 in the reactor water. Under using good operating practice. !
j , conditions of hydrogen water chemistry, the , l l g fraction of the nitrogen 16 that volatilizes Water quality specification; require that 4 i with the steam increases with increased dis. crosion corrosion resistant low alloy steels are 3a solved hydrogen. The main steam line radia. to be used in susceptible steam extraction and M2 ,
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I Amendment 3 S 2 11 t
~0 D ABM 3346 ooan Standard Plant RFV A drain lines. Stainless steels are considered for baffles, shields, or other areas of severe duty, e Provisions are made to add nitrogen gas to
,7 extraction steamlines, feedwater heater shells, Oj heater drain tanks, and drain piping to minimize N- corrosion during layup. Alternatively, the system may be designed to drain while bot so that dry layup can be achieved.
st " Condenser tubes ond tubesheet are required to Ej be made of titanium alloys.
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4 O ABM 23A6 00AD Standard Plant nry c ballon per minute, thus meeting Position C.2 and testing is prosided.
re guirements.
These satisfy Position C.8 requirements.
By monitoring (1) floor drain sump fillup and pumpout rate, (2) airborne particulates, and (3) Limiting unidentified leakage to the range of air coolers condensate flow rate. Position C.3 is 1 to 5 gpm and identified to 25 gpm satisfies satisfied. Position C.9.
Monitoring of the reactor building cooling 5,2.6 Interfaces water heat exchanger coolant return lines for radiation due to leaks within the RilR, R** and The remainder of plant will meet the water RWCS beat exchangers (and the fuel pool cooling chemistry requirements given in Table 5.2 5.
system beat exchangers) satisfies Position C.4.
For system detail, see Subsection 7.6.1.2. 5.2,7 References !
Tbc f1oor drain sump monitnring, air particu- 1. General Elec;ric Standard Application for lates monitoring and air cooler condensate moni- Reactor Fuel (NEDE 24011 P A, latest app-toring are designed to detect leakage rates of roved version),
one gpm within one hour, thus meeting Position C.5 requirements. 2. (Deleted)
The fission products monitoring subs > stem is 3. D.A. liale, The Effect of Bil'R Startup En.
l qualified for SSE. The containment floor drain vironments on Crack Growth in Structural sump monitor, air cooler, and condensate flow Alloys, Trans. of AShlE, vol 10S, January meter are qualified for OBE, thus meeting 1986.
Position C.6 requirements. l
- 4. F.P. Ford and bl. J. Povich, The Effect of i Leak detection indicators and alarms are Oxygen l Temperature Combinations on the provided in the main control room. This Stress Corrosion Susceptibility of Sensi-satisfies Position C.7 requirements. Procedures tised T-304 Stainless Steel in High Purity and graphs will be provided by the applicant to ll'aler, Paper 94 presented at Corrosion 79, plant operators for converting the various Atlanta, GA, blarch 1979. ;
indicators to a common leakage equivalent, when i necessary, thus satisfying the remainder of 5. Bil'R Normal IVater Chemistry Gu!: felines 2st 4 2 Position C.7. The leakage detection system '- 19S6 Revision, EPRI NP 4946 SR, July 198S. :si.7 j equipped with provisions to permit testing 1.s :stto l
- operability and calibration during the plant 6. B.hl. Gordon, The Effect of Chloride and tiem 4 & lt I operation using the following methods: Oxygen on the Stress Corrosion CracAing of i Stainless Steels: Review of Literature. l (1) simulation of signals into trip units; blaterial Performance, NACE, Vol.19, No. 4, April 19S0.
(2) comparing channel A to channel B of the same leak detection method (i.e., area tempera. 7. W.J. Shack, et al, Environmentally Assist-ture monitoring)t ed Cracking in Light it'aler Reactors: Annual Report, October 1983 September 1984, 4 (3) operability checked by cornparing one method NUREG/CR 4287, ANL 85 33, June 1985.
Setsus another (i.e., sump fillup rate ver-sus pumpout rate and particulate monitoring 8. D.A. Itale, et al, Bil'R Coolant Impurities l on air cooler condensate flow versus sump Program, EPRI, Palo Alto, CA, Final Report fillup rate); and on RP2293 2, to be published.
(4) continuous monitoring of floor drain sump lesel and a source of water for calibration
! n.:s j AmendmeH 3 4,
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- 9. K.S. Brown and G.ht. Gordon, Effects of Bil"R Coolant Chemistry on the Propensity ofIGSCC ,
Initiation and G wth in Crevleed Reactor internals Corrpo sents, paper presented at 3g the Third Ints. ational Symposium of Envi.
~
r amental Degradation of blaterials in Nucle-ut.7 ar Fower Systems, ANS NACE TMS/AIME, Muo 3,g , 3 3 Traverse City, Michigan, September 1987.
- 10. B.M. Gordon et al, EAC Resistance of Bil'R Materials in Hil'C, Proceeding of Second international Symposium Environmental Degration of Materials in Nuclear Power Systems, ANS, LaGrange Park, ILL 1986, 10a. BWR Hydrogen Water Chemistry Guidelines:
1987 Revision, EPRI NP 4947 SR LD (To be published).
N 10b. Guideline for Permanent BWR liydrogen Water Chemistry Installations: 19S7 Revision, EPRI NP 5203 SR.A.
- 11. B.M. Gordon, Corrosion and Corrosion 2
au Controlin BilRs, NEDE 30637, Deeember 1984.
281.7 a1.10 12. B.M. Gordon et al, Halogen li'ater Chemistry item t a ns for Bil'Rs Materials Behavior, EPRI .
NP 5080, Palo Alto, CA, March 1987. !
i l
i i-AmtMment 3 5 2.N l
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. o ABWR 234siooan Standard Plant nnv n Table 5.2 5 2st.10 HWR WATER CHEMISTRY liem s Electro-Chemical Corrosion Concentrations
- Conducthity Potential Parts Per Billion (ppb) ;
Iroit Cooner Chloride Sulfate Oxwen" .at 250C ,210C, Y at 250C condensatt < 20 <2 <4 <4 < 10 - 0.15 -7 -
! Condensate Treatment Efiluent and Feedwater :2.2 < 0.02 < 0.4 < 0.4 20-50 < 0.059 ~7 -
Reactor Water (a) Normal Operation <1 " ~7
< 20 < 20 < 20 < 0.3 < 0.23 l
(b) Shutdown < 20 <1 < 20 < 20 - < 1.2 ~7 --
5 (c) Hot Standby < 20 <1 < 20 <20 < 200 < 0.3 ~7 -
l t
(d) Depressurized < 20 <1 < 20 < 20 high (may < 1.2 5.68.6 -
1 be 1000 to 8000) 4 Control Rod Drive l J Cooline Water <2.2 < 0.1 < 0.4 < 0.4 20 50 .50.059 ~7 --
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l These limits should be met at least 9% of the time.
Some resision of o9 gen vahtes may be established ofte? hydrogen nattr chemistry has been established.
i Amendment 3 $ 2.M
ABM 33462 man Standard Plant RTs* B TAELES Table .T.!!.ls East 6.21 Containment Parameters 6.2 45 6.22 Containment Design Parameters 6.2 4 h
6.2 2a Enginected Safety Systems loformation for Containment Response Analyses 6.2 46a M 6.2 2b Net Positive Suetion Head (NPSH) Available
- 6 to RHR Pumps 6.2 46c 6.2 3 Compartment Nodal Description 6.2 47 6.2-4 Compartment Vent Path Description 6.2-43 6.25 Reactor Coolant Pressure Boundary (RCPB)
Influent Lines Penetrating Dr>well 6.2 49 6.24 Reactor Coolant Pressure Boundary (RCPB)
Effluent Lines Penetrating Drysell 6.2 50 ;
ILLUSTRATIONS Figure Title Eage 6.2 1 A Break in a Feedwater Line 6.2 11 6.22 Feedwater Line Break RPV Side Break Area 6.2 52 -
6.2 3 Feedwater Line Break Flow Feedwater System Side of Break 6.2 53 ,
6.24 Feedwater Line Break Flow Enthalpy -
Feedwater System Side of Break 6.2 54 6.25 Lower Dr>well Air Transfer Percentage for Model Assumption Versus Actual Case 6.2 55 ;
6.2-6 Pressure Response of the Primary j Containment 6.2 56 j 6.27 Temperature Response of the Primary j Containment 6.2 57 '
6.28 Temperature Time liistory After a Feedwater Line Break 6.2 58 6.29 ADWR Main Steamlines with a Brcak 6.2 59 6.2lx l
MM.
Standard Plant 33A6100AD RIN D i
6.2.1.1.3.1 Summary Evaluation 6.2.1.1.33.1 Feedwater Line Break The key design parameter and the maximum Immediately following a double ended rupture t calculated accident parameters for the pressure in one of the two main feedwater lines just out-suppression containment are shown in Table 6.21. side the vessel (Figure 6.21), the flow from both sides of the break will be limited to the The maximum drywell pressure would occur maximum allowed by critical flow considerations.
I during a feedwater line break. The maximum The effective flow area on the RPV side is given dt)well temperature condition would result from a in F;gure 6.2 2. During the inventory depletion main steam line break. All of the analyses pericd, subcooled blowdown occurs and the effec-
- assume that the primary system and containment tive flow area at saturated condition is much
- system are initially at the mr,ximum normal less than the actual break area. The detailed operating conditions, calculational method is provided in Reference
- 1. The RPV blowdown through the break is 6.2.1.1J.2 Containment Design Parameters prevented by the check valves.
l Table 6.2 2 provides a listing of the key The feedwater system side of the feedwater f design parameters of the primary containment line break (FWLB) was modeled by adding a time [
I system including the design characteristics of variant feedwater mass flow rate and enthalpy 4
the drywell, suppression pool and the pressure directly to the drywell airspace. The time
! suppression vent system. histories of the mass flow and enthalpy were -
determined from the operating characterFtics of Table 6.2 2a prevides the performance para- a typical feedwater system.
meters of the related engineered safety feature ,
systems which supplement the design conditions of The maximum possible feedwater flow rate was (
, Table 6.2 2 for containment cooling purposes dur- calculated to be 164% of nuclear boiler rated ,
j ing post blowdown long term accident operation. (NBR), based on the response of the feedwater Performance parameters given include those pumps to an instantaneous loss of discharge pres-applicable to full capacity operation and reduced sure. Since the feedwater control system will capacities assumed for containment analpes, respond to decreasing RPV water level by demand. ,
ing increased feedwater flow, and there is no .
6.2.1.1.3.3 Accident Response Analpis FWLB sensor in the design, this maximum feed.
water flow was conservatively assumed to contin- ,
The containment functional evaluation is ue for 120 seconds, as shown in Figure 6.2 3. l l based upon the consideration of several This is very conservatise because: 1) all feed-j postulated accident conditions which would result water system flow is assumed to go directly to
- in the release of reactor coolant to the the dr>well,2) flashing in the broken feedwater ,
- containment. These accidents include
- line was igno ed,3) initial feedwater flow was '
assumed to be 10$% NBR, and 4) the feedwater (1) an instantaneous guillotine rupture of a pump discharge flow will coastdown as the feed-feedwater linet water system pumps trip due to low suction pres.
i sure. During the inventory depletion period, (2) an instantaneous guillotine rupture of a the flow rate is less than 164% because of the main steam linet or highly subcooled blowdown. A feedwater line
- length of 100 M was assumed on the feedwater j (3) small break accidents. sptcm side.
- The containment design pressure and The enthalpy of the feedwater flow is 120% of j temperature were established based on enveloping a typical BWR/5 feedwater system inventory I
the results of this range of analyses plus enthalpy. The specific enthalpy time history, l providing NRC prescribed margins. assuming the break flow of Figure 6.2 3, is i
f Amtadrnent 3 6N l
1 MM 23A61ooAn Standard Plant _
nrv A shon in Figure 6.2-4.
6.2.1.1.3.3.1.1 Assumptions for Short. Term Response Analysis ,
The response of the reactor coolant system and the containment system during the short term 3 blowdown period of the accident has been analyzed using the following assumptions: ,
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ABM 2346woan Standard Plant REV D (10) Actuation of SRVs is modeled, pressure flooder feeding a broken feedwater line,in case of a FWLB). A single failure (11) Wetwell to-dtgell vacuum breakers are not of one RHR heat exchanger was assumed for modeled. conservatism.
(12) Drywell and wetwell sprays and RilR cooling (2) The ANS decay heat is used. Fission mode are not modeled. energy, fuel relaxation heat, and pump heat are included.
(13) The dynamic back pressure modelis used.
(3) The suppression poolis the only heat sink (14) Initial drywell conditions are 15.45 psia, available in the containment system.
1350F, and 20% relative humidity.
(4) After 10 minutes, the RHR heat exchangers (15) Initial wetwell airspace conditions are are activated to remose energy via 15.45 psia, 950F and 100% relative recirculation cooling of the suppression humidity. pool with the RCWS and ultimately to the RSWS. This is a conservative assumption (16) The drywell is modeled as a single node. since, the R11R design permits initiation of All break flow into the drywell is containment cooling well before a 10 minute 3
homogeneously mixed with the drywell inventory.
period. (See response to Question 430.26) f l (5) The maximum service water temperature is
- (17) Because of the unique containment geometry assumed to bc 950F. This is a of ABWR, the inert atmosphere in the lower conservative assumption that maximires the drywell would not transler to the wetwell suppression pool temperature, i
until the peak pressure in the drywell is achieved. Figure 6.2 5 shows the actual (6) The lower drywell flooding of 28,760 ft3 l case, and the model assumption. Because the was assumed to occur 70 seconds after t lower drywell is connected to the drywell scram. During blowdown phase, a portion of connecting vent, no gas can escape from tha break flow flows into the lower drywell, lower drywell until the peak pressure This is conservative since lowe drywell 4 a
occurs. This situation can be compared to a flooding will probably occur at j bottle whose opening is exposed to an approximately 110 to 120 second time period atmosphere with an increasing pressure. The . (See Figure 6.2 6).
contents of the lower drywell will start
, transferring to the wetwell as soon as the (7) At 70 seconds, the feedwater specific i pressure starts decreasing. A conservative enthalpy becomes 180 Blu/lb (2120F l 4
credit for transfer of 50% of the lower saturation fluid enthalpy),
drpell contents into the wetwell was taken.
6.2.1.lJJ.1J Short Term Accident Responses 6.2.1.1JJ.l.2 Assumptions for temg. Term Cooling Analpis The calculated containment pressure and temperature responses for feedwater line break Following the blowdown period, the ECCS are shown in Figures 6.2 6 and 6.2 7, I discussed in Section 6.3 provides water for core respectisely. The peak pressure (39 psig) and flooding, containment spray, and long term decay temperature (2840F) occur in the drywell. The heat removal. The containment pressure and containment design pressure of 45 psig is 116%
. temperature response during this period was of the peak pressure, analyzed using the following assumptions:
J The drywell pressurtration is drhen by the (1) The ECCS pumps are available at yecified in wetwell pressurization for stable peaks. The Subsection 6.1.1.3.3.1.1 (exce pt one low wetwell pressurization is a function of threc Amudent 3 624
ABWR 3346ioorn Standard Plant RIT B of channels, recording of parameters, instrument condensate running back into the suppression range and accuracy and post accident monitoring pool. All water that leaves the suppression equipment is discussed in Section 7.5. poolis cooled by the RiiR heat exchangers during the three operational modce indicated above.
6.2.2 Containment Heat Removal System For each of the three loops, water is drawn from 6.2.2.1 Design Bases the suppression pool, pumped through a R11R heat exchanger and injected into the reactor vessel l
for the LPFL mode. Also, for each of the three The containment heat removal system, loops for the suppression pool cooling mode, consisting of the suppression pool cooling mode water is drawn from the suppression pool, pumped and the wetwell and drywell spray features are through a R11R heat exchanger and delivered to integral parts of the RilR system. The purpose of the suppression pool. On two of the loops this system is to prevent excessive containment (BAC), a portion of the water returned to the temperatures and pressures, thus maintaining suppression pool may be passed through wetwell containment integrity following a LOCA. To spray headers. These two loops also have a
- fulfill this purpose, the containmeut cooling manual feature for providing drywell spray, system meets the following safety design bases
- Water from the RCWS is pumped through the heat exchanger shell side to exchange heat with the (1) The system limits the long term bulk processed water. Three cooling loops are pro-temperature of the suppression pool to vided, each being mechanically and c!cctrically 2070F when considering the energy separate from the other to achieve redundancy.
additions to the containment following a A piping and instrumentation diagram (P&lD) is LOCA. These energy additions, as a function provided in Section SA. The process diagram, of time, are provided in the previous including the process data, is provided for all ,
section. design operating modes and conditions.
4 (2) The single failure criterion applies to the All portions of the containment cooling g system. system mode are designed to withstand operating loads and loads resulting from natural (3) The system is desigacd to safety grade phenomena. All operating components can be requirements including the capability to tested during normal plant operation so that perform its function following a Safe reliability can be assured. Construction codes Shutdown Earthquake. and standards are cosered in Subsection 5A.7. .
(4) The system maintains operation during those The low pressure flooder (LPFL) mode is 4 environmental conditions imposed by the automatically initiated from ECCS signals. The s* t 4
LOCA. suppression pool cooling mode is started l manually or automaticitly. The RilR system must i (5) Each active component of the system is be realigt ed for suppression pool cooling by the testable during normal operation of the plant operator after the reactor vessel water
- nuclear power plant. lesel has been recovered (Subsection 6.2.1).
The RilR pumps are already operating. i 6.2.2.2 Containment Coollrg System Design Suppression pool cooling is initiated in any of 4
the three loops by manually closing the LPFL The containment cooling system encompasses injection valve and opening the pool return i several of the RilR operating modes, which are the valve. In the event that a single failure has ;
low pressue flooder (LPFL) mode, the suppression occurred, and the action which the plant [
g pool cooling mode, and the containment spray operator is taking does not result in system L
- s' modes (dry *cll and wetwell). Containment cooling initiation, then the operator will place the "
- starts as soon as the LPFL injection flow other totally redundant system into operation by begins. The suppression pool cooling mode cools following the same initiation procedure. If the
- the containment. The cor.tainment sprays cool the operator chooses to utilire the containment j dry *cll and wetwell by condensing steam and the sprays, he must close the LPFL injection valvc5 Amendmem 3 6216 i
i
- O ABM ursiooan Standard Plant nirv ^
and open and spray valves. The drywell spray mode may be initiated manually only after a high drywell pressure permissive occurs.
Preoperational tests are performed to verify individual component operation, individual logic element operation and system operation up to the containment spray spargers. A sample of the sparper nozzles is bench tested for flow rate versus pressure drop to evaluate the original hydraulic calculations. Finally, the spargers are tested by air and visually inspected to verify l-that all nozzles are clear. (See l
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ABM a>x61oorn nrv c Standard Plant Subsection 5.4.7.4 for further discussion of initial suppression pool temperature and the R11R preoperational testing.) service water temperature are at their maximum values. This assumption maximizes the heat sink 6.2.2J Design Esaluation of the Containment temperature to which the containment heat is Cooling Sptem rejected and thus maximizes the containment
! temperature. In addition, the RIIR heat 6.2.2J.1 Splem Operation and Sequence of exchanger is assumed to be in a fully fouled Esents condition at the time the accident occurs. This conservatively minimizes the heat exchanger heat In the event of the postulated LOCA, the removal capacity. Even with the degraded short term energy release from the reactor pri- conditions outlined above, the maximum mary systern will be dumped to the suppression temperature is maintained below the design limit pool. Subsequent to the accident, fission ,ao- specified in Subsection 6.2.2.1.
duct decay heat will resQ in a continuing en-crgy input to the pool. The RilR LPFL mode and It should be noted that, when evaluating this suppression pool cooling mode will remove this long term suppression pool transient, all heat energy which is released into the primary contain- sources in the containment are considered with ment system, thus resulting in acceptable sup- no credit taken for any heat losses other than 4 pression pool temperatures and containment through the RiiR beat exchanger. These heat pressures, sources are discussed to Subsection 6.2.1.3.
In order to evaluate the adequacy of the RilR lt can be concluded that the conservative system, the following is assumed: evaluation procedure described above clearly demonstrates that the R11R system in the (1) With the reactor initially operating at 102% suppression pool cooling mode limits the of rated power, a LOCA occurs. post LOCA containment temperature transient.
l (2) A single fa'ilure of a R11R heat exchanger is 6.2.2..I Test and Inspections ,
the most limiting single failure.
! The containment cooling system is required (3) The ECCS flows assumed asailable are 2 HPCF, to have scheduled maintenance. The system 1 RCIC, and 2 LPFL (R}{R). testing and inspection will be performed .
i periodically during the plant normal operation f 4 (4) Containment cooling is initiated after 10 and after each plant shutdown. Functional !
O minutes. (Sec Response to Ouestion 43026) testing will be performed on all active i components and controls. The sptem reference Analysis of the net positive suction head characteristics will be established during i
g (NPSil) available to the RilR pumps in accordance preoperational testing to be used as base points s* with the recommendations of Regulatory Guide 1.1 for checking measurements obtained from the '
is prosided in Table 6.2 2b. system tests during the plant operation.
General compliance for Regulatory Guide 1.26 The preoperational test program of the l
.way be found in Subsection 3.2.2. containment cooling system is described in i Subsection 14.2.12. T . following functional l Failure modes and effects analpes for the RilR tests will be performed. The RilR pump will bc l and RCWS are provided in Appendix 158, icsted through the suppression pool cooling loop l operation by measuring flow and pressure. Each 6.2.23.2 Summary of Containment Cooling pump will be tested indisidually, i Analpis l Containment spray spargers will be tested When calculating the long. term, post LOCA pool during reactor shutdown by air, and by visual temperatwe transient, it is assumed that the inspection to verify that all the norrles are AmeMmem 3 1 6217 4
ABWR umman Standard Plant REV H pressurized with air at a reduced test pressure Pg, which will result in a measured leakage 1 4m Lmt rate, identified as Ltm. The second phase is L" t La for values of _ .s,0.7 then conducted at pressure P, resulting in a L am Lam measured leakage rate identified as Lam. The $
absolute method shall be employed for determining the leakage rate (see ANSI N45.4 Subsection 5.2.1 L" La (P h t Lmt
> 0.7 h
t I for values of _
and Section 7.9). Test duration of each phase (Pa / Lma shall be sufficient for pressure and temperature stabilization. To ensure uniform temperature (Pg and Pa are psig),
distribution, fans will be provided to circulate air in the containment during the test. Prior to The leakage Lam shall be less than 0.75 L a commencement of the tests, the test prerequisites and not greater than the design leakage rate described in Subsections 6.2.6.1.2.1 a n d (Ld )-
6.2.6.1.3 will be m e t.
6.2.6.1.2 Periodic leakage Rate Tests 6.2.6.1.1J Supplement Verification Test Leakage rate tests are conducted periodically The accuracy of the leakage rate tests is in conformance to Appendix J of 10CFR50 to ensure verified by using a supplemental method of that the integrity of the containment is leakage measurement. Verification is obtained by maintained and to determine if any leakage superimposing a controlled and measurable leak on increase has developed since the previous ILRT.
the normal containment leakage rate or other The tests are performed at regular intervals, methods of demonstrated equivalency. The after major repairs or upon indication of difference between the total leakage and the excessive leakage, as specified in the standard superimposed k sown leakage results in the actual technical specification for the ABWR.
leakage rate. This leakage rate is a check against its accuracy and is acceptable prosided 6.2.6.1.2.1 Integrated Leakage Rate Test (ILRT, the correlation between tbc supplemental test Type A) data and integrated leak test data demonstrates an agreement within 2,25%. Conduct of the Type A tests are conducted periodically, verification test is normally accomplished after following the initial preoperational tests, at completion of each test phase of the ILRT. test pressure Pg only. Except for the Complete descriptive details are found in climination of the P pressure test, all ILRTs Appendix C of ANSI N45.4. follow the same format as the initial ILRT, as outlined in Subsection 6.2.6.1.1.
6.2.6.1.1.4 Instrumentation Requirements in addition to the normal test prerequisites, Instrumei tation provided to monitor the the following requirements are mandatory prior to containmen' iakage rate testing is designed, all periodic Type A tests:
calibrated and tested to accurately ensure that the containment atmosphere parameters can be (1) A detailed visual examination of critical precisely measured, areas and general inspection of the accessible interior and exterior surfaces 6.2.6.1.1J Acceptance Criteria of the containment structure and components shall be performed to uncover any evidence The iuitial allowable leakage rate (Li m) at of structural deterioration which may
- test pressure P shall not exceed 75% of the affect either the structural integrity or maximum allowable test leakage rate (Lt ), where leaktightness of the containment. If there Ltis defined as follows: is evidence of significant structural Amm mw 1 y
ABM a346 ooan Standard Plant RM B deterioration, Type A tests shall not be per- until two consecutive Type A tests meet the formed until corrective action is taken in acceptance criteria, after which time the k accordance with approved repair procedures. previously established periodic retest $
If leak repairs of testable components are schedule may be resumed.
performed, the reduction in leakage shall be measured (at test pressure Pg) and added to 6.2.6.1J.3 Test Frequency the Type A test result. Except for inspec.
tions and actions taken above, no preliminary After initial ILRT, a set of three Type A leak detection surveys and repairs shall be tests shall be performed at approximately equal performed prior to the conduct of the Type A intervals during each 10 yr service period, with test. the third test of each set coinciding with the end of each 10 yr major inservice inspection (2) Closure of containment isolation valves shall shutdown, in addition, any major modification be accomplished by normal mode of actuation or replacement of components of the primary and without preliminary exercises or adjust- reactor containment performed after the initial ments. All malfunctions and subsequent cor. ILRT shall be followed by either a Type A or a rective actions shall be reported to the NRC. Type B test of the area affected by the modifi.
cation, with the affected area to meet the 6.2.6.1.2.2 Acceptance Criteria applicabic acceptance criteria. The basis for the frequency of testing is established in The measured leakage rate Ltm shall not accordance with 10CFR50, Appendix J.
exceed 0.75 Li as established by the initial ILRT. 6.2.6.1J Additional Criterla for Integrated Rate Test (1) If during a Type A test including the supple.
mental test, potentially excessive leakage (1) Those portions of fluids systems that are paths are identified which will interfere part of the reactor coolant pressure with satisfactory completion of the test, or boundary, that are open directly to the which result in the Type A test not meeting primary reactor containment atmosphere the acceptance criteria, the Type A test under post accident conditions and become shall be terminated and the leakage through an extension of the boundary of the primary g such paths shall be measured using local leak. reactor containment, shall be opened or y age testing methods. Repairs and/or adjust. vented to the containment atmosphere prior J
ments to equipment shall be made and a Type A to or during the Type A test. Portions of test performed. The corrective action taken closed systems inside containment that l ano the change in leakage rate determined penetrate primary containment and are not from the tests and overall integrated leakage relied upon for containment isolation determined from the localleak and Type A purposes following a LOCA shall be sented tests shall be included in the report to the containment atmosphere. ;
! submitted to the NRC. '
(2) All vented systems shall be drained of wa.
(2) If any Type A test f ails to meet the ter to the extent necessary to ensure expo.
4 acceptance criteria, prior to corrective sure of the system primary containment iso.
action, the test schedule applicable to lation valves to the containment air test ;
subsequent Type A test shall be subject to pressure, j 1 review and approval by the NRC. l (3) Those portions of fluid systems that pene. :
(3) If two consecutive periodic Type A tests fail trate primary containment, that are exter. l 4 to meet the applicable acceptance criteria, nal to containment and are not designed to l g prior to corrective acilon, notwithstanding provide a containment isolatio2 barrier,
- g the established periodic retest schedule, a shall be vented to the outside atmosphere
- Type A test shall be performed at each plant as applicable, to assure that full post ac.
j shutdown for major refueling, or approximate. cident differential pressure is maintained ly every 18 months, whichever occurs first, across the containment isolation barrier.
1 Amendment 3
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ABWR 234accan l
Standard Plant REV H (4) Systems that are required to maintain the quired to meet this limit, the results shall be plant la a safe condition during the Type A reported in a separate summary to the NRC. The test shall be operable in their normal mode summary shall include the structural conditions '
and are not vented. of the components which contributed to failure.
(5) Systems that are normally filled with water 6.2.6.2.3 Retest Frequency and operating under post LOCA conditions need not be vented, in compliance with the requirement of Section Ill.D.2(a) of Appendix J to 10CFR Part 6.2.6.2 Containment Penetration leakage Rate 50, type B tests (except for air locks) are f Test (Type B) performed during each reactor shutdown for major fuel reloading, or other convenient intervals, 6.2.6.2.1 General but in no case at intervals greater than two years.* Air locks opened when containment
- Containment penetrations whose designs integrity is required will be tested in manual incorporate resilient seals, bellows, gaskets, or mode within 3 days of being opened. If the air scalant compounds, airlocks and lock door seals, lock is to be opened more frequently than once ,
equipraent and access hatch seals, and electrical every 3 days, the air lock will be tested at 'I i
canisters, snd other such penetrations are leak least once every 3 days during the period of tested during preoperational testing and at frequent openings. Air locks will be tested at l
periodic intervals thereafter in conformance to initial fuel loading, and at least once every 6 Type B leakage rate tests defined in Appendix J months thereafter. Testing may be initiated of 10CFR50. The leak tests ensure the continuing automatically at the end of each interval by the .
s t r u ct u r al a r.d le a k in t e g rit y o f t he seal test instrumentation system, with manual I pcnetrations. override of the s.utomated sequence provided for
, in the associated logic. Testing involves the To f acilitate local leak testing, a injection of air under pressure (15 psig) into permanently installed system may be provided, the space between the two redundant seals in consisting of a pressurized gas source (nitrogen each door of the air lock. The leakdown rate is or air) and the manifolding and vahing necessary measured by sensing the pressure drop and/or i to subdivide the testable penetrations into flow rate necessary to maintain the pressure.
l groups of two to five. Each group is then Main control room readout of time to next test, l pressurized, and if any leakage is detected (by test completion and test results is provided, l pressure decay or flow meter), individual An alarm sounds if the specified interval passes 4
penetrations can be isolated and tested until the without a test being effected. No direct,
, source and nature of the leak is determined. All safety.related function is sersed by the seal
- Type B tests are performed at containment peak test instrumentation system. l accident pressure, Pa. The local leak detection tests of Type B and Type C (Subsection 6.2.6.3) 6.2.6.2A Design Proshions (c,r Periodic !
must be completed prior to the preoperational or Pressurization :
periodic Type A tests. !
In order to assure the capability of the i 6.2.62.2 Acceptance Criteria containment to withstand the application of peak accident pressure at any time during plant life
, The combined leakage rate of all components for the purpose of performing ILRTs, close i subject to Type B and Type C tests shall not attention is given to certain design and l
exceed 60% of La (cfm). If repairs are re- maintenance provisions. Speciiically, the i
- *In compliance with the requirement of Section l fil.D.2(b)(lii) of Appendix J to 10CFR Part 50 l l
l Amendment 3 6242
^
ABM as46waan Standard Plant RIV A ;
! effects of corrosion on the structural integrity '
-of the containment are compensated for by the l inclusion of a 60 yr service life corrosion allowance, where applicable. Other design features that have the potential to deteriorate with age, such as flexible seals, are carefully inspected and tested as outlined in Subsection i 6.2.6.2.2. In this manner, the structural and i
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ABM a34610048 Standard Plant nry si leakage integrity of the containment remains reverse direction is equivalent, or more essentially the same as originally accepted. conservative. The correct direction for this design is defined as flow from inside the 6.2.6.3 Containment isolation Vahe l.4akage containment to outside the containment.
Rate Test (Type C)
- 6.2.6.3.2 Acceptance Criteria 6.2.6.3.1 General The combined leakage rate of all components Type C tests are required on all isolation subject to Type B and Type C (Subsection valves. All testing is performed pneumatically, 6.2.6.3) tests shall not exceed 60% of aL . If except hydraulic testing may be performed on iso- repairs are required to meet this limit, the lation valve Type C tests using water as a scal- results shall be reported in a separate summary ant provided that the valves will be demonstrated to the NRC, to include the structural conditions
- to exhibit leakage rates that do not exceed those of the components which contributed to the l in the ABWR standard technical specifications. failure. {
Type C tests (like Type B test) are performed 6.2.6.4 Scheduling and Reporting of Periodic i by local pressurization using either pressure Tests l decay or flowmeter method. The test pressure is d
applied in the same direction as when the valve The periodic leakage rate test schedules for i is required to perform its safety function, un- Type A, B and C tests are described in Chapter less it can be shown that results from tests with 16. i pressure applied in a different direction are equivalent or conservative. For the pressure de- Type B and C tests may be conducted at any cay method, test volume is pressurized with air time during normal plant operations or during or nitrogen to at least Pa. The rate of decay shutdown period;, as long as the time interval i of pressure of the known test solume is monitored between tests for any individual Type B or C to calculate leakage rate. For the flowmeter tests does not exceed the maximum allowable j method, required pressure is maintained in the interval specified in the standard technical !
j test volume by making up air, nitrogen or uter specifications for the ABWR Each time a Type B [
i (if applicable) through a calibrated flowmeter, or C test is completed, the overall total j 3
The flowmeter fluid flow rate is the isolation leakage rate for all required Type B and C tests l valve (or Type B test volume) leakage rate. is updated to reflect the most recent test j results. In addition to the periodic tests, any Aliisolation valve seats which are exposed to major modification, replacement of component l containment atmosphere subsequent to a LOCA are which is part of the primary reactor containment !
! tested with air or nitrogen at containment peak boundary, or rescaling a seal wclded door, accident pressure, Pa . performed after the preopertionalleakage rate test will be followed by either a Type A, Type i l Those valves which are in lines designed to B, or Type C test as applicable for the area ,
be, or remain, filled with a liquid for at least effected by the modification. Type A, B and C - l 30 days subsequent to a loss of coolant accident test results shall be submitted to the NRC in 3" j are leakage rate tested with that liquid. The the summary report approximately three months i liquid leakage measured is not converted to after each test. ;
equivalent air leakage not added to the Type B ;
and C test total. Included in the leak rate test summary report will be, a report detailing the containment in.
For Type C testing of containment penetra- spection, a report detailing any repairs neces-l tions, all testing, with the exception of the sary to pass the tests, and the leak rate test j ECCS systems will be done in the correct direc- results.
tion unless it can be shown that testing in the 4
j Amtedmtet 3
MM 33ActooAn
. Standard Plant nry A :
4 6.2.6J Special Testing Requirements The maximum allowable leakage rate into the secondary containment and the means to verify
.l that the inleakage rate has not been exceeded, as ,
] well as the containment leakage rate to the environinent, are discussed in Subsections 6.2.3 ,
ii and 6.5.1.3.
i
. i d
I i I
1 i !
i l
1 I i i 6
,} i l l 1
I l t
I !
i i i i i
I i
i l
4 I
i !
3 l
! i
) !
I 1
4 1 !
i a
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1
! Amndant 3 6.24h H
ABM 234siman .
Standard Plant ,
na c Table 6.2 2 CONTAINMEST DESIGN PARAMETERS A. Drmell and Wetwell(I) h.cIl Wetwell
- 1. Internal Design Pressure (psig) 45 45
- 2. Negative Design Pressure (psid) 2.0 2.0
- 3. Design Temperature (OF) 340 219
- 4. Net Free Volume (ft3 ; 239,563 210,475
. j 5. Maximum allowable leak rate (2) ;
' 0.5 l (%/ day) 0.5
! 6. Minimum Suppression Pool Water ,
Volume (ft3 ) - 126,427 i 7. Suppression pool depth (ft)
Low 1.evel - 22.97 High Level - 23.29 l B. Vent S> stem I
- 1. Number of Vents 30 l
- 2. Nominal Vent Diameter (ft) 2.3
- 3. TotalVent Area (ft2) 125 1 4. Vent Centerline Submergence ,
j Low Level),(ft) i j I
- Top Row 11.48 Middle Row 15.98 Bottom Row 20.48 e
i
- 5. Vent Loss Coefficient (Varies with number of vents open) 2.53.5 :
I (I) Item A.1, A.2, A.3 and A.5 apply to related structures including lower dr) sell access tunnels, drywell equipment hatches, dr>well personnel locks and drywell head.
j (2) Corresponds to calculated peak containment pressure related to the design basis accident conditions.
a J
Amenoment 3 gg i
ABM n46imio Standard Plant arv A TABLE 6.2 2a ENGINEERED SAFETY SYSTEMS INFORMATION FOR CONTAINMENT RESPONSE ANALYSES :
Full Capacitt Containment Analuis Value ;
A. Containment Snrav
- 1. Number of RHR Pumps 1(1) 1(1)
- 2. Number of Lines 1(I) 1(I)
- 3. Number of Heat Exchangers 1(2) g(2)
- 4. Drpell Flow Ratc (Ib/hr) 1.81 x 106 1.8) x 106 i 5. Wetwell Row Rate (Ib/ht) 2.46 x 105 2.46 x 105 ,
B. Containment Cooline Systtm 8
]
- 1. Number of RHR Pumps 3 2 i 2. Pump Capacity (gpm/ pump) 4200 4200
- 3. RilR licat Exchangers
- a. Type U. tube,
! b. Number 3 2 1
- c. Ileat Transfer Area (3) (3) 2 '
- (ft / unit) i i
1 d. Overall lleat Transfer (3) (3) ;
3 Coefficent (Btu /hr- !
2 ft .0F/ unit) {
- e. Service Water Rowrate (lb/hr) 2.63 x 106 2.63 x 106
(
i f. Maximum Senice Water l Temperature ('F) 95 95 i l 1 !
l I l
._m ., , .,. ,
23A6100All Standard Plant ,
REV A TABLE 6.2 2a ENGINEERED SAFETY SYSTES1S INFORMATION FOR CONTAINS1ENT RESPONSE ANALYSES (Continued)
NOTES 1, Two redundant loops available with one pump each.
- 2. One header each for drywell and wetwell.
- 3. Tbc RHR beat exchanger characteristic has bsen defined by an overall K coeficient h based on a temperature differecce and the heat rate. The defining equation is:
Q = (K)(AT) 0, Blu = j (K, __ Btu ji ST, F
)f 8
( /( )
The K value is 195 Bru/sec'F.
The applicable temperature difference occurs from the RilR heat exchanger's reactor side inlet to the service water temperature. Thus, K is a characteristic of the combined RilR
- and reactor senice water sptem's heat exchangers. i t
l i
1
)
i >
)
f i
1.
t i
(
i l'
Armadrunt 3 6MA i i
ABWR Standard Plant amman arv A TABLE 6.2 2b NET POSITIVE SUCTION HEAD (NPSH) AVAILABLE TO RHR PUMPS 1
A. Suppression pool is at its minimum depth, El. 3740mm ( 12.27 Ft).
B. Centerline of pump suction is at El. 7200mm (23.62 Ft).
C. Suppression pool water is at its maximum temperatue for the given operating ;
mode; 97'C (207*F).
D. Pressure is atmospherie above the suppression pool.
E. hiaximum suction strainer losses are 2.0 psi. .
NPSH = HAThi + HA HyAp Hp i where: .
1 [
) HAThi = atmospheric head 1 >
.i ils = static head g R
llyAp = vapor pressure head i ,
j Hp = Frictional head including strainer i
4 hfinimum Etnected NPSH ..
i j RiiR Pump Runout is 1100 m3/h (4S43 gprn). -
1 l 51aximum suppression pool temperature is 97'C (207'F) l
=
+
llAThi 10.73m (35.20 ft) -
=
j 113 3.46m (11.35 ft) l i
HyAp =
9.74m (31.95 ft) !
Hp =
1.82m (5.97 ft) l l
Strainer head loss = 2.0 psi = 1.46m = 4.80 Ft ;
NPSH availabic = 10.73 + 3.46 9.74 1.82 = 2.63m (8.63 Ft) :
4
! NPSH required = 2.4m (7.87 ft) {
! 'i 1 i l
l l
i l
1 Am w...o sw I
l 1
I ABM usuaoar Standard Plant avn two channel operability. The operator has scram time performance based on the scram the capability to invoke bypass conditions timing data received from the RC&lS.
within the following system or subsystems:
7.7.1.2.2 Other Systems Interfaces (a) Symchro A or B position bypass (b) Rod server module channel A or B bypass (1) Alternate Rod Insertion (ATWS) (Anticipated (c) Uncoupled condition bypass Transient Without Scram)
(d) File control module channel A or B bypass The RC&lS logic, during an anticipated tran-(e) AF8ht chanuel A or B bypass sient without scram (on receipt of signals (f) hfRBM channel A or B bypass as a result of high reactor dome pressure or (g) RPC chancel A cr B bypass low reactor water level) initiates ARI sig-(b) RACS chaanel A or B bypass n,.ls which controls the fine motion control (i) DAh! channel A or B bypass rod drive motors such that all control rods are driven to their full in position automa-(11) Scram Time Test Data Recording tically. The four divisions of tue nuclear boiler system provide cach of the two The logic of the RC&lS prosides the capabi- channels of the RC&lS logic with the reactor lity to automatically record indisidual fine high dome pressure and reactor low water motion control rod drive (Fh!CRD) scram level signals for generation of the ARI timing data based upon scram timing reed signal based on two out of four logic.
switches. When a particular Fh1CRD scram timing switch is activated, the time of The operator at the RC&lS dedicated display actuation is recorded by DAhls for time can take action and initiate the ARI func-tagging of stored scram time test data for tion. Two manual actions are required to l that particular fine motion cont'ol rod manually initiate ARI. The RC&lS logic has drive. The time tagged da;a is stored in been designed to complete the ARI functions memory until the next actuation of that in the worst case non accident environment, particular reed switch is detected again, completely independent of reactor pressure transient conditions. This capability is The RC&!S also time tgs the receipt of a accomplished with control logic for inser-reactor scram condition being actisated tion of all control rods by an alternate and based upon the scram fo. lowing function diserse method, based on receiving reactor input signals from the reactor protection high dome prenure and low water lesel(Le-system which is received via the essential vel 2) signals for generating its own ATWS multiplexing system. (anticipated transient without scram) sig-nal. The logic of the RC&!S has been de-The resolution of this time tagging feature signed such that no single failure results is less than 10 milliseconds. Contact it' failure to insert more than one operab!c bounce of the reed switch inputs and the DAh! control rod when the ARI function is inputs are properly masked to support this activated, function. The reference real time clock for time tagging is the real time clock of the (2) Recirculation now Control system RC&lS.
The recirculation flow control cystem (RFCS)
When RC&ls detects a reactor scram provides each of the two channels ei fla condition, the current position of all RC&lS with two separate isolated trip control rods in the core are recorded, tims signals indicating the need for automatic ,
tagged, and stored in memory. RC&lS logic selected con rol rod run in (SCRRI). Tbc $
stores this data in memory until a request signals are treated as nonsafety related is receised from the performance monitor ng signals within the logic of the RC&lS.
and control system. The tran.mitted data i.
used by the PhfCS to calculate and summarire Tbc RFCS prosides signals to both channels AN"*3 u.o
O
- ABWR mer Standard Plant nry si of the RC&lS that represent validated total The presclected control roo, for an SCRRI core flow. These signals are used for part function are selected at the RC&lS dedicated of the validity checks when performing an operators control panel and the CRT displays ARBh1 operating limit setpoint update. The of the performance monitor control system in RC&lS can obtain these signals from the RFCS the main control room. The preselected via the multiplexing system of direct communi- SCRRI rod data is stored in memory in the cation links to the RC&iS channels. These rod action and position information subsys-signals are also completely independent from tem of the RC&lS. The total control rod the process computer system. worth for the preselected control rods is designed to bring down the reactor power le-The RFCS receises refstence power level sig- set from the 100% tod line to the 80% line.
cals from the neutron monitoring system and m compares the reference power level signals The RC&ls dedicated operat ers control panel $
with the nominal power lesel setpoint, also provides control switches that requires two manual operator actions for the operator Selected control rod run in (SCFRI) is auto- to manually initiate the SCRRI function.
matically initiated when a trip of two or more reactor internal pumps (RIPS) occur. For the manual or the automatic initiation This function is part of Ihe stability of the SCRRI function the RC&lS dedicated op-control and protection logic. erators panel provides status indications and alarm annunciators in the control room.
When two or more RIPS are tripped, the trip signal is *ANDED' with the power lesel 'AND' The RC&l5 prosides the capability for manual flow rate signals and RFCS automatically or autonetic initiation of the SCRRI sends . request for control rod blocks to the function arm 3e total delay time to start RC&lS. When the power lesel signal with two of control rod motion for the preselected or moee RIPS tripped k "ANDED' with the flow control rods is less thaa 350 milliseconds, e rate the RFCS automatically sends four sig-5 nals to the RC&lS to initiate the SCRRI func. (3) Feedwater ControlSptem tien.
The feedwater control system provides The SCRRI function is bypassed when power signals to both channels of the isgic of the level is below the specified setroir.t. or RC&lS that represents validated total when the core flow is abme the specified feedwater flow to the vessel, validated setpoint. narrow range vessel dome pressure, and validated feedwater temperature. Thece The SCRRI function if designed as a non- signals are used as part of the validit) safety related sptem. The function i; de- checks when performing an ARB51 operating signed to meet the reliability requirement limit setroint update.
that no single failure shall cause a loss of the function. The RC&ls can obtain these signals from the feedwater control sptem sia the multiples.
The RFCS automati: initiation signal for the ing sptem of direct communication links to SCRRl function is sent as two independent the RC&lS channels. These signals are also sets of signels, one set to each channel of completely independent from the process com-the RC&lS, each channel of the RC&ls uses the puter sptem.
input in two out of two logic to control the fine motion control rod drise (FhtCRD) motors (4) Neutron hionitoring 5ptem of presclected control rou. The preselected control rods are driven to their full in posi. Each of the four disisions of the neutron tion on receipt of the automatic initiation monitoring splem prosides independent signah. Either channel of an RC&ls is ca- signals to both channels c.* the RC&ls that pable of initiating the SCRR1 function on re- indicate when the following conditions are ceipt of the automatic signal from the RFCS actise:
AmeeJment 3 7716
O e ABWR 234sicoar Standard Plant nrV A (a) Startup range neutron monitor (SRNM) pc- (c) Status of hydraulic control unit (HCU) riod alarm scram test switch.
(b) SRNM downscale alarm (c) SRNM upscale alarm Tbc essential multiplexing system provides (d) Average power range monitor (APRM) up- the above signals to the RC&IS with complete scale alarm isolation between the safety related system (e) SRNM inoperative and the nonsafety related system equipment.
(f) APRM downscalc (g) Flow biased APRM rod block (6) Nuclear Boiler System (h) APRM inoperative The four divisions of the nuclear boiler sys-Whether or not some of the signals result in tem (NBS) provide each of the two channels
- a rod block depends on whether or not the re- of the RC&lS with the reactor high dome pres-j ector is in the RUN mode. The reactor mode sree e'd acactor water level signals for gen-status is providc
- 1 to the RC&lS from the reac- . v. of N RC&lS alternate rod insertion tor pro;ection system via the essential multi. .,'. ) fuucslon.
plexing system.
, Each o' the four divisions of NMS rJgnal, pro-sides APRM, LPRM and core flow signals to the two channels of logic in the rod action and
), positioa information system for determining whether reactor power is above or below the i Iow power setpoint, j The four divisions of NMS signals to the RC&lS two channel system are isolated sig-
, nals between the Class 1E NMS and the l
nonsafety related equipment of the RC&lS.
(5) Reactor Protection Spterr 1
i Each of the four divisions of the reactor pro-
) tection system provides the RC&lS two channel l system with separate isolated signals for indication of the reactor mode switch positions: SHUTDOWN, REFUEL, STARTUP, HOT STANDPY and RUN.
Each of the four divisions of the reactor pro-tection system (RPS) provides to the RC&ls
! two separate isolated signals for each of the 2
following conditions:
l
- (a) Reactor scram condition. This signal re-i mains active if initiated until the
1 (b) Low charging water header prest,ure trip j
switches in bypass position.
Amendet 3 II' 4
o__
ABWR 23A6100AT Sinndard Plant arv. n NRC* Reslew Question SSAR Response RAl** ,
i Branch Area Number Subsection Subsection letter t 1
Materials 252.1 4.5.1.1(1) 20 3.1 1 Application 252.2 4.5.1.1(2) 20 3.1 1 :
2523 4.5.2.2 20 3.1 1 252.4 4.5.2.3 20 3.1 1 252.5 4.5.2.4 20 3.1 1 1 252.6 4.5.2.5 203.1 1 252.7 $.23.2.2 20 3.1 1 252.8 5.2.3.23 20 3.1 1 l 252.9 5.233.1 20 3.1 1 252.8 5.23.23 20 3.1 1 l 252.10 5.2.3.4.1.1 20 3.1 1
! 252.11 5.23.4.2.3 20 3.1 1
(
I ECEB Chemical 2S1.1 5.1 20.3.1 1 Techr. ology 251.2 5.23.2.2 20.',.1 1
' l 2813 5.23.2.2 20 3.1 1 :
281.4 5.23.2.2 20 3.1 1 l 231.5 5.23.2.2 20 3.1 1 281.6 5.23.2.2.2 20 3.1 1 l
4 231.7 5.23.2.23(4) 203.1 1
[
2S1.8 5.23.2.23(13) 20 3.1 1 r i 281.9 6.4.9.2 203.1 1 I 281.10 Chap. 5 20 3.1 1 l
i, l SPLB Plant 430.1 4.6 20 3.2 2 [
Sptems 430.2 5.2.5 203.2 2 L 1 430 3 5.2.5 20 3.2 2 i
! 430.4 5.2.5.4.1 20 3.2 2 j
430.5 5.2.5 20 3.2 2
, 430.6 5.2.5 20 3.2 2
- 4 430.7 6.2 20 3.2 2 ;
1 430.8 0.2 20 3.2 2 i i
430.9 6.2 20 3.2 2 l 430.10 6.2 20 3.2 2 l 430.11 6.2 20 3.2 2 (
430.12 6.2 20 3.2 2 i i
430.13 6.2.1.13 20 3.2 2 !
430.14 6.2 20 3.2 2
[
i 430.15 6.2 20.3.2 2 ,
430.16 6.2 20 3.2 2 i n 430 17 6.2.1.23 20 3.2 2 j 430.1S 6.2 20 3.2 2 l l 430.19 6.2 203.2 2 i a
430.20 6.2 203.2 2 l
j Amen.iment 3 2012 i 8
ABWR u^amn Standard Plant nrv. A NRC* Re$lew Question SSAR Response RAl" Branch Area Number Suberction Subsection latter 430.21 6.2 20 3.2 2 430.22 6.2 20 3.2 2 430.23 6.2 20 3.2 2 -
430.24 6.2 203.2 2 420.25 6.2 20 3.2 2 430.26 6.2 20 3.2 2 430.27 6.2 20 3.2 2 4M.28 6.2 20 3.2 2 430.X 6.23 20 3.2 2 r 43030 6.2 20 3.2 2 43r31 6.2 20 3.2 2 ,
- 43032 6.2 20 3.2 2 4M33 6.2 20 3.2 2 43034 6.2 20 3.2 2 43035 6.2 20 3.2 2 43036 6.2 20 3.2 2 !
i 43037 6.2 20 3.2 2 l 43035 6.2 20.3.2 2 l 43039 6.2.4 20 3.2 2 l 430.40 6.2 20 3.2 2 430.41 6.2 20 3.2 2 430.42 6.2 20.3.2 2 .
l l 430.43 6.2 20 3.2 2
, 430.44 6.2 20 3.2 2 i 430.45 6.2 20 3.2 2 i 430.46 6.2 20 3.2 2 j 430.47 6.2.53 20 3.2 2 !
430.45 6.2.6 20 3.2 2 430.49 6.2.6 20 3.2 2 i 430.50 6.2.6 20 3.2 2 430.51 6.26 20 3.2 2 430.52 6.2.6 20 3.2 2
- 430.53 6.2.6 20 3.2 2
- 430.54 6.4 20 3.2 2 430.55 6.5.1 20 3.2 2 430.56 6.53 203.2 2 430.57 6.7 20 3.2 2
. 430.58 15.7 3 20 3.2 2 ,
i l SRXB Reactor 440.1 4.6 203.2 2 {
Sptcm5 440.2 4.6.23.2.2 20 3.2 2 [
l 440 3 4.6.1.2 20 3.2 2 1 440.4 4.6 20 3.2 2 1 1
440.5 4.6 20 3.2 2 l
- 440.6 4.6 20 3.2 2 1 440.7 4.6 20 3.2 2 l
U
.i l Amendment 3 , ,
ABMR 234610a41 Standard Plant MY A NRC* Review Question SSAR Response RAl**
Branch Area Number Subsection Subsectica letter 440.8 4.6 20 3.2 2 440.9 4.6 20 3.2 2 440.10 4.6.23.1 20 3.2 2 440.11 4.6 20 3.2 2 440.12 4.6 20 3.2 2 PRPB Radiological 470.1 15.5.2 203.1 1 Report 470.2 15.6.2 20 3.1 1 4703 15.6.4.5.1.1 203.1 1 470.4 15.6.5.5 20 3.1 1 470.5 15.6.5 20 3.1 1 470.6 15.7.5 20 3.1 1 470.7 15.7 20 3.1 1 470.8 15.7 20.3.1 1 470.9 15.7 20 3.1 1 470.10 15.7 20 3.1 1 Amendment 3 20,1.h
ABWR 2mmr Standard Plant Rrv H SECTION 20.2 CONTENTS SectIon Ihlt East 20.2.1 Chanter 1 Ouestion, 20.2 2 20.2.2 Chanter 2 Ouestions 20.2 3 20.23 Chanter 3 Ouestions 20.2 4 20.2.4 Chanter 4 Ouestions 20.2 5 20.2.5 Chapter ! Ouestions 20.2 6 l i
20.2.6 CAapter 6 Ouestions 20.2 10 1 20.2.7 Chapter 7 Ouestions 20.2 13
+ 20.2.8 Chanter R Ouestinns 20.2 14 4
20.2.9 Cnanter 9 Ouestions 20.2 15 1
4 20.2.10 Chapter in Ouestiong 20.2 15 20.2.11 Chapter 11 Ouestions 20.2 17 20.2.12 Chapter 12 Ouestions 20.2 18 20.2.13 Chanter 13 Ouestion.t 20.2 19 20.2.14 Chanter 24 Ouestions 20.2 20 l
1 20.2.15 Chanter 15 Ouestions 20.2 21 l 20.2.16 Chanter 16 Ouestions 20.2 23 20.2.17 Chapter 17 Ouestions 20.2 24 r 20.2.18 Chapter la Ouestions 20.2 25 20.2.19 Chanter 19 Ouestions 20.2 26
, 20.2 ii I
Amerkiment 3
. 1 Mkb 23A61 COAT Standard Plant nry n '
20,2A Chapter 4 Questions 252.1 Subsection 4.5.1.1 (1) should state: *The properties of ;be materials selected for the control rod drive mechanism must be equivalent to those given in Appendix 1 to Section til of the ASME Code, ,
or parts A and B of Section 11 of the ASME Code, or are included in Regulatory Guide 1.85, except that cold worked austenitic stainless steels should have a 0.2% offset >ictd strength no greater than j 90,000 ps!.*
l 252J J
l '
Subsection 4.1.1.1 (2) should state: 'All materials for use in this system must be selected for ,
thei' compatibility with the reactor coo!=nt as described in Articles NB 2160 and NB.3120 of the ASME l j Code.'
i 2!23 i
. Subsection 4.5.2.2: The first sentence should read: Core support structures are fabricated in
- accordance with the requirements of ASME Code, Section 111, Subsection NG 4000, and the examination and acceptance criteria shown in NG 900.*
l j 252.4 ;
j Subsection 4.5.2.3: The following statement should be added to the last sentence of the first .
- paragraph
- 'The examination will satisfy the requirements of NG 5300.* '
1 i 252J .
i .
Subsection 4.5.2.4 should state:
- Furnace sensitired material should not be allowsd.'
i 252.6 i Subsection 4.5.2.5 should state: 'All materials used for reactor internals will be selected for
- their compatibility with the reactor coolant as shown in ASME Code Section Ill, NG 2160 and NG 3120. .
i The fabrication and cleaning controls will preclude contamination of nickel based alloys by chloride .
l ions, fluoride ions, or lead.'
i l
l 430.1 I e
Proside a failure modes and effects analysis of the control rod drise system (CRDS) in tabular ;
i form with supporting discussion to delineate the logic employed. The failure analysis should j l demonstrate that the CRDS can perform the intended functions with the loss of any active single i
component. These evaluations and assessments should establish that all essential elements of the i CRDS are identified and prositions are made for isolation from nonessential CRDS elements. It should i
) be established that all essential equipment is protected from common mode failures such as failure of J
moderate and high energy lines. The failure mode and effects analysis of the control rod drives !
I should include water, air and electrical failures to CRDs and how the CRD system operation is l
- affected due to air contamination or water contamination. Before finalizing the scope of the I analysis, refer to ACRS subcommittee meeting pru
- cedings on the ABWR dated June 1,19SS. It is noted I that the above information is to be included in Appendix 15B of the SSAR which will be submitted at a ;
) later date. However, the evaluation of the functional design of the reactivity centrol systems ;
j cannot be completed until this information is provided. (4.6) ;
i f l !
l Amendment 3 2025 l i
l a
- o e kMN 2.M6100AT t
Standard Plant RFV A i
440.1 1
SRP 4.6 identifies the following GDCs 23,25,26,27,28 and 29 in the acceptance criteria.
Confirm that the reactisity system, described in Section 4.6 of the SSAR, meet the requirements of the above GDCs. ,
i
! 440.2 In Section 4.6.2.3.2.2 Analysis of malfunction relating to rod withdrawal, it is stated 'There are known single malfunctions that cause the unplanned withdrawal of even a single control rod.' Confirm that this is a editorial mistake and correct it if so. Otherwise, explain in detail the basis for this statement and why this is acceptable.
1
] 440.3 i
! In Section 4.6.1.2 it is stated that CRD system in conjunction with CRC &lS and RPS systems provides selected control rod run in (SCRRI) for reactor stability control. Describe in detail how l SCRRI works.
a 4M.4 i t
In Figure 4.6 8a, CRD system p&lD, sheet 1, piping quality classes AA.D. FC D, FD D, FD.B. etc. ;
! are shown. Submit the document which explains these classes and relates them to ASME code classes.
i -
440.5 l
In Figure 4.6-Sb, the leak receiver tank is shown. What is the function of this tank? How big'it [
this tank? Will a high level in the tank impact the operation of the control rod drive? l
{
440.6 i 1 ;
identify the essential portions of the CRD system which are safety related. Confirm that the i I safety related portions are isolable from non.cssential portions. (4.6) I 440.7 [
t in the old CRD sptem, the major function of the cooling water was to cool the drise mecbanism and l i its seals to preclude damage resulting from long term exposure to reactor temperatures. What is the i j fur tion of purge water flow to the drises? (4.6) j 1
1 440.5 !
0 We understand that the LaSalle Unit 2 fine motion control rod drive demonstration test is still in i j progress. Submit the test results as soon is it is available. (4.6) [
] 440.9
(
l j in the present CRD system design, the ball check valve ensures rod insertion in the event the i accumulator is not charged or the inlet scram value fails to open if the reactor pressure is abose l
- 600 psig. Confirm that this capability still exists in the ABWR design. (4.6) i 9
i 1,
l Amendment 3 i j 2024a d
5
. f ABWR mmo.u Standard Plant nrv A 440.10 in section 4.6.2.3.1, it is stated the scram time is adequate as shown by the transient analyses of Chapter 15. Specify the scram time.
440.11 For both the low (*2ero') power and operating power region describe the patterns of the control rod groups that are expected to be withdrawn simultaneously with the new rod system, and estimate the maximum for the total and differential reactivity worth of these groups. What sort of margin to period scram will exist in the low power range. (4.6) 440.12 Describe the relative core location of control rods sharing a scram accumulator. Can a failure of the scram accumulator fail to insert adjacent rods? If so, discuss the consequences of that failure.
(4.6)
Amendment 3 ;o ;.3d
o o MkN 33A6t00AT arv n S tandard Plant 20.2.5 Chapter 5 Questions 210.1 In Subsection 5.2.1.2. the statement is made that Section 50.55a of 10CFR$0 requires NRC staff i approval of ASME code cases only for Class 1 components. Revise this statement to be consistent with i I
the current (1987) edition of 10CFR50.55a, which requires staff approval of code cases for ASME Class 1,2, and 3 components, i 210.2 l
Revist Table 5.21 or provide additional tables in Subsection 5.2.1.2 which identify all ASME code
~l cases that will be used in the construction and in plant operation of all ASME Class 1,2, and 3 -
components in the ABWR, All code cases in these tables should be idtattfied by code case number, l revision, and title. These tables should include those applicable code cases that are listed either l i
as acceptabic or conditionally acceptable in Regulatory Guides 1.84,1.85, and 1.147. For those code l 3 '
cases listed as conditionally acceptable, verify that the construction of all applicable components j will be in compliance with the additional Regulatory Guide conditions. {
250.1 i Subsection 5.2.4.1 should state that the system boundary includes all pressure vessels, piping, pumps, and valves which are part of the reactor coolant system, or connected to the reactor sys' ems, ,
up to and including. l l
2 (1) The outermost containw ent isolation valve in system piping the penetrates the primary reactor ;
- containment. l i (2) Tbc second of two valves normally closed during normal reactor operation in system piping that {
i does not penetrate primary reactor containment, j i
i (3) The reactor coolant system and relief vahes. !
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250.2
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J Subsection 5.2.4.2 rhould satisfy the requirements in ASME Code IWA 1500.
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1 251.1 l
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Subsection 5.3.1.1 should state that the materials will comply with the provisions of the ASME !
j Code, Section 111, Appendix 1, and meet the specification requirements of 10CFR50, Appendix G. l 4
251.2 l Subsection 5.3.1.2 should state the specific subsection NB of ASME Code to which the moufacturing l j and fabrication specifications were alluded. !
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AmeridmeM 3 20.2 4 i
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= f ABWR mamu Standard Plant nry A 251J Subsections 53.1.4.4 and 53.1.4.5 should be restitten; the cross reference is unacceptable.
Subsections 5.3.1.4.7, 5.3.1.5.2, 5.3.1.6.3, and 5.3.2.1.5: Revision 2 of Regulatnry Guide 1.99 should be added in these subsections.
251.4 Subsection 5.3.1.6.1: the third capsule of the vessel surveillance program is designated as a standby; however, according to ASTM 155 82, the capsule should be withdrawn at the end of life.
Provide justification for this desiation.
Amendment 3 33 ;y
ABM 234siooar Standard Plant arV A 430.2 Regarding Reactor Coolant Pressure Boundary (RCPB) leakage detection systems, proside information on the following: (5.2.5) 4 (a) Describe how the leakage through both the inner and outer reactor vessel head flange seals '
will be detected and quantified.
(b) List the sources that may contribute to the identified leakage collected la the Reactor Building Equipment Drain Sumps.
(c) Describe how potentialintersystem leakages will be monitored for the (1) Low pressure !
Coolant Injection System,(2) High Pressure Core Spray System,(3) Reactor Core Isolation Cooling System (RCIC) . Water side and (4) Residual Heat Removal System Inlet and discharge sides. Your response should include all the applicable (for the ABWR Jesiga) systems and components connected to the Reactor Coolant System that are listed in Table 1 of SRP Section L 5.2.5 and other systems that are unique to ABWR (except those that you have already ;
discussed in SSAR Subsection 5.2.5.2.2, item 11). ,
r 430.3 ,
Discuss compliance of reactor coolant leak detection systems with Regulatory Guide (RG) 1.45, f
' Reactor Coolant Pressure Boundary Leakage Detection Sptems,' Positions C4, C5, C6, C8 and C9 with l l
respect to the following items: (5.2.5)
(a) Indicators for abnormal water levels or flows in all the affected areas in the event of f intersystem leakages.
l (b) Sensitivity and response time of leak detection systems used for unidentified leakages outside the dr)well. [
(c) Qualification relating to seismic events for dr> Sell equipment drain samp monitoring system l and leak detection systems outside the dr>well. t i'
(d) Teting Procedures . Monitoring sump levels and comparing them with applicable flow rates of fluids in the sumps. j (c) Inclusion u! reactor building and other areas floor and equipment drain sumps in ABWR f Technical Specifications for leak detection systems. [
l Note that a few of the questions above arise because in Subsection 5.2.5.4.1 you state that the ;
total leakage rate includes leakages collected in drywell, reactor building and other area floor drain and equipment drain sumps.
(
r 430.4 l Clarify whether the RCIC makeup capacity is sufficient to provide also for main stem line !
leakage through to tbc main turbine stop valves. Also, clarify whether this leakage is included in the total leakage mentioned in Subsection 5.2.5.4.1. i i
l, Amendmtal 3 20.211a l
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I- 23A6100AT Re--d-ed Plant R.EV A
-i WJ i
! Clarify bow Position C.2 of RG 1.29,'Selsmic Design Classification
- is met for all applicabic i leak detection systems (also include the leak detection systems outside the drywell). (5.2.5) !
- m.6 'i 4 i identify all the laterface requirerr.ents relating to RCPB leakage detection systems. (5.2.5) {
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20211b ,
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9
2M6t00AT [
Saadard Plant arv n j 20.2.6 Chapter 6 Questions -
t 230 3 i
l Subsection 6.6.8 should discuss the augmented insersice inspection for those portions of high I l
caergy piping enclosed in guard pipes.
l 282.12 Subsection 6.1.1.1 should discuss ferritic steel welding in detail. It should also discuss the I control of ferrite content in stainless steel weld metal similar to that of Regulatory Guide 1.31. ,
b 4 252.13
$ Subsections 6.1.1.1.3.1, 6.1.1.1.3.2, a nd 6.1.1.1.3.5 should be rewritten because the !
cross reference is unacceptable. l
. 281.9 Subsection 6.4.4.2 (page 6.4 6) discusses personnel respirator use in the event of toxic gas i intrusion into the control room, lloweser, the chlorine detection system is not discussed. Also, any j control functions that are automatically triggered by a chlorine detector alarm (closing intake !
dampers, energiring control room IIVAC system recirculation) should be identified. j 430,7 f i !
- In the SSAR section desoted to containment functional design, identify clearly those areas that j are not paet of the ABWR scope and proside relevent interface requirements. (6.2) l 430JL j l i
, With respect to the design bases for the containment:(6.2) l 3
(a) Discuss the bases for establishing the margin between the maximum calculated accident [
t pressure or pressure difference and the corresponding design pressure or pressure i l difference. This includes the design esternal pressure, internal pressure, and pressure i between subcompartment walls.
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(b) Discuss the capability for energy remosal from the containmeut under various singic failure ,
i conditions. State and justify the design basis single failure that affects containment heat I removal, f
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Amendmtat 3 202.t 3
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23A6100AT Standard Plant arY A I !
' 40.9 :
The Standard Safety Analysis Report (SSAR) states that the analytical models used to evaluate the containment and drywell responses to postulated accidents and transients are included in General
- Electric Co report NEDO 20$33 and its supplement 1, entitled The G.E. Mark 111 fressurc Suppression i 1
Containment Analytical Nodel. Provide justification that these references are appropriate to use t for the ABWR Containment design which is not specified as Mark 111. Discuss the similarities and I differences of the ABWR design to previously approved Mark II and Mark Ill designs as they relate to i
! the containment and drywc!) responses to the postulated accidents and the analytical model used for i
] the analyses. Include in the discussion the conservatism used in the model and assumptions, the
- applicable test data that support the analpical models, and the sensitivity of the analyses to key '
parameters (6.2) l ,
1 i 430.10 !
! l i With regard to the design features of the containment. (6.2) !
l (a) Provide general arrangement drawings for the containment structure.
l (b) Provide appropriate references to Stction 3 of the SSAR which mcludes the information on [
i the codes, standards, and guides applied in the design of the containment and containment !
j internal structures.
[
, I l (c) Discuss the possibilities of water entrapment inside containment and its effect on the !
accident analysis. .
T l (d) Proside information on qualification tests that are intended to demonstrate the functional !
capability of the containment structures, systems and components. Discuss the status of any i 3
developmental tests that may not base been completed, j l
l 4 0.11 i 4 ;
j Provide a detailed discussion of the likelihood and sensitisity to steam bypass of the suppression !
- pool for a spectrum of accidents. Include in your discussion the following information
- (6.2)
)
j (a) A comparison of the ABWR pool bypass capability with that for Mark 11 and Mark lit designs. ;
1 i (b) The measures for minimiting the potential for steam bypass and the systems presided to ,
mitigate the consequences of pool bypass. Discuss and demonstrate the conservatism of j assumptions made in the analysis of steam bypass.
[ [
(c) Identify alllines from which leakage (or rupture) could contribute to pool bypass and l wetwell air space pressurization. l l
(d) Identify all fluid lines which traverse the wetwell air space and identify those lines which !
i are protected by guard pipe. l l
} (c) Discuss the rationale and basis for the wetwell spray flow capacity.
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- AmeMment 3
- 20112. ;
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MM 11A6ttCAT taandard Plant aty a :
W.12
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With regard to containment response to external pressure: (6.2)
(a) Describe the wetwell.to dr3well vacuum breaker system and show the ex snt to which the requirements of subsection NE of section til of the ASME B&PV Code anc satisfied. Discuss ,
the functional capability of the system. Provide the design and performance parameters for l the vacuum relief devices. ;
, 1 (b) Discuss the basis for selecting a low design capability for external pressure acting across the drywell to wetwell boundary. It is not apparent that the drywell negative design ,t pressure of 2.0 psid is desirabic or sufficient, (c) The margin between the calculated wetwell.to. reactor building negative differential pressure t
( 1.8 psid) and the sign differential pressure ( 2.0 psid) is not conaldered adequate. A higher margin of 15% should be provided at this stage of the design. Further, design on containment venting to control pressure, discuss the basis for not providing wetwell-to renetor building vacuum breakers.
1 (d) In the analysis of wetwell to reactor building negative differential pressure calculation, a ,
' 500 gpm wetwell spray flow rate was used. Provide the basis for the assumption and the design basis for the wetwell spray capacity. !
i 430.13 ,
1 1
- Section 6.2.1.1.3 of the SSAR states that the containment functional evaluation i based upon the i i consideration of several postulated accident conditions including small break accidents. Proside the l assumptions, analysis and results of the small break accidents considered, ad demonstrate that the identified (in the SSAR) feedwater line and steam line breaks are the limiting accidents.
W.14 l
Provide analyses of the suppression pool temperature for transients involving the actuation of l safety / relief vahes. Proside the assumptions and conservatism employed in the analyses so that an
, assessment could be made for conformance to the acceptance criteria set forth in NUREG 07S3, Suppression Pool Temperature Limits for BWR Containments. (6.2) I j i
! 430.15 ,
j !
Prodde the pressure at which the maximum allowable leak rate of 0.5%/ day is quoted. (6.2) l l
430.16 !
B L
Provide enginected safety systems information for containment response analysis (full capacity i operation and capability used in the containment analysis), as indicated in Table 6 7 of Regulatory i Guide 1.70, Revision 3. (6.2) l 430.17 la the design evaluation section for containment subcompartments (Section 6.2.1.2.3), proside the
- information necessary to substantiate )our assessment that the peak differentiat pressures do not exceed the design differential pressures. Guidance for the informetion required is provided in l Regulatory Guide 1.70, Resision 3. Section 6.2.1.2, ' Containment Subwmpartments'. Design Euluation.
1 Amendment 3 4
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MkN 83A6100AT 3 Standard Plant RTV A m.ls Describe the manner in which suppression pool dynamic loads resulting from postulated loss of.
coolant accidents, transients (e.g., relief valve actuation), and scismic esents have been integrated into the affected containment structures. Provide plan and section drawings illustrating all equip- ;
~;
ment and structural surfaces that could be subjected to pool dynamic loads. For each structure or j group of structures, specify the dynamic loads as a function of time, and specify the relative magni.
- tude of the pool dynamic load compared to the design basis load for each structure. Provide justifi- '
r
- cation for each of the dynamic load histories by the use of appropriate experimental data and/or j analyses.
l Describe the manner by which potential as>mmetric loads were considered in the containment l
design. Characterire the type and magnitude of possible asymmetric loads and the capabilities of the j affected structures to withstand such a loading profile. (6.2)
M 19
- Provide information to demonstrate that the ABWR design is not vulnerable to a safety relief vahe i discharge line break within the air space of the wetwell, coupled with a stuck open relief valve I after its actuation as a result of the transient. (6.2) i G 20 j j Discuss suppression pool water makeup under normal and accident conditions. (6.2) ,
l
@.21 l With respect to mass and energy release analyses for postulated loss of coolant accidents identify !
- the sources of generated and stored energy in reactor coolant system that are considered in the !
! analyses of loss of coolant accidents. Describe the methods used and assumptions made in !
! calculations of the energy available for release from these sources. Address the conscrsatism in the !
, calculation of the available energy from each source. Tabulate the stored energy sources and the !
amounts of stored energy. For the sources of generated energy, provide curses showing the energy [
] release rates and integrated energy release. (6.2) {
G.22 l In the SSAR sections desoted to containment heat removal syst:ms, identify clearly those areas !
t that may not be part o! the GE scope and proside relevant interface requirements. (6.2) !
! f i G .23 i i
I The SSAR states that the containment heat removal system is designed to limit the long term i
, tempegature of the suppression pool to 207 F. The calculated peak pool temperature is l t 206.46 F for the feedwater line break. With respect to this analysis proside the following i j information: (6.2)
! (a) The justification that this is the limiting accident with respect to the maximum temperature !
in the suppression pool, j (b) The bases for the design margin between the design and calculated temperatures. {
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Amendment 3 20 21 e j i I
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ABM 2miwAT uv. A man =dard Plant (c) All assumptions ased in the analysis and conservatism associated with each, include the I effects of potential temperature stratification in the suppression pool end its effects on heat removal capability of the system. (
l (d) The identification of the decay heat curve used in the analpis. [
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430.24
)
> Provide the design bases for the spray features of the containment heat removal system. Provide [
the safety classification of the components associated with the spray feature of the system. (6.2) j
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! 430 25 1
Discuss the rationale for continued reliance on sprays as the sole active engineered safety fea.
ture for drpell atmosphers pressure and temperature. Discuss the recrits of upgrading the design of r drywell fan coolers to provide some capacity for pressure, temperature, and humidity control 3
following an accident. (6.2) i i
430.26 t The time period assumed for initiation of the containment heat removal system after a 1.OCA is 10 !
I minutes requiring operator action. It is the staff's position that this time period is too restric- l j tise. In fact presious BWR designs (Grand Gulfs hlark Ill) use 30 minutes actuation time. Proside !
A the reasons why the ABWR does not provide more flexibility with respect to the time required for !
actuation. (6.2) [
- 430.27 Describe the design features of the suppression pool suction strainers Specify the mesh sire of r e the screens and the maximum particic sire that could tie drawn into the piping. Of the systems that i j receive water through the suppression pcul suction strainers under post accident conditions identify [
the sptem component that places the limiting requirements on the masimum site of debris that may be allowed to pass through the strainers and specify the limiting particle size that the component can s circulate without impairing system performance. Discuss the potential for the strainers to become i clogged with debris. Identify and discuss the kinds of debris that might be developed following a j loss of. coolant accident. Discuss the types of insulation used in the containment and describe the [
, behasior of the insulation during and after a 1.0CA. Include in your discussion information regarding i
] compliance with the acceptance criteria associated with USI A 43 as documented in NUREG-OS97. (6.2) l i !
430.2s ,
I Provide analpes of the net positise suction head 'NPSil) availab!c to the RilR pumps in accordance ,
with the recommendations of Regulatory Guide 1.1. Compare the calculated values of available NPSil to !
the required NPSit of the pumps. (6.2) l i
> 430.29 t l
j in SSAR Section 6.2.3, identify clearly those areas that may not be part of the ABWR scope and
- proude relevant interface requirements. (6.2) ,
1 i l
l 430.30 1
Provide a tabulation of the design and performance data for the secondary containment structure. ,
- Proside the types of information indicated in Table 617 of Regulatory Guide 1.70, Resision 3. (6.2) i mo 20212J l
ABM msimar RTV A Standard Plant 43031 Describe the valve isolation features used in support of the secondary containment. Specif) the plant protection system signals that isolate the secondary containment and actisate the standby gas treatment system. (6.2) 43032 Identify and tabtlate by sire, piping which is not provided with isolation features. Proside an analysis to demonstrate the capability of the standby gas treatment system to insintain the design negatise pressure following a design basis accident with all non isolated lines open and the esent of the worst single failure of a secondary containment isolation valve to close. (6.2) 43033 Discuss the design prosisions that prevent primary containment leekage from bypassing the secondary containment standby gas treatment system and escaping directly to ibe environment. Include a tabulation of potential bypass leakage paths, including the types of inforraation indicated in Table 618 of Regulatory Guide 1.70, Resision 3. Proside an esaluation of potential bypass leakage paths considering equipment design limitations and test sensitivities. Specify and justify the maximum allowable fraction of primary containment leakage that may bypass the secondary containment structure. The guidelines of BTP 6 3 should be addressed in considering potential bypass leakage paths. (6.2) 43034 Proside a list of the secondary containment openings and the instrumentation means by which each is assured to be closed during a postulated design basis accident. (6.2) 43035 Proside a table of design information regarding the containment isolation prosisions for fluid system tir.cs and fluid instrument lines penetrating the containment which are within the GE scope of the ABWR design. Include as a minimum the following information:
(1) General design criteria or regulatory guide recommendations that have been met or other defined bases for acceptabilit);
(2) Sptem name; (3) Fluid contained; (4) Line size; (5) E.SF sptem (>cs or no);
(6) Through.line leakage classification (7) Reference to figure in SSAR showing atraepement of containment isolation barriers; (S) location of salve (inside/outside containment);
(9) Type C leakage test (>cs or no);
AmeWmeet )
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l 2M6100AT I fl!amedard Plant arY A I
(10) Valve type and operstor; ,
l (11) Primary mode of vahs actuation; (12) Secondary mode of vahe actuation; (13) Nwaal valve poshion; l (14) Shutdown suht poskion 1 1
1 (15) Poetaccident vaht position; l (16) Power failure vaht position; (17) Containment isolation signals; (18) Vaht closure time; and (19) Power source.(6.2)
M 34 For isolation valve design in systems not within the ABWR scope, identify the systems and the relevant interface requirements. Include a discussion on essential and non essential systems per Regulatory Guide 1.141 and the means or criteria provided to automatically isolate the nonessential systems by a containment isolation signal. Also, include a discussion on the requirement that the ;
setpoint pressure which initiates containment isolation for nonessential penetrations be reduced to the minimum value compatible with normal operations. (6.2)
W.37 Specify all plant protection signets that initiate closure of the contalement isolation valves.
(6.2)
W 38 Describe the leakage detection means provided to identify leakage for the outside. containment remote manualisolation vabes on the following influent lines: Feedwater, RHR injection, HPCs, i standby liquid control, RWCU connecting to feedsater line, RWCU reactor vessel head spray. (6.2)
(
W39 The containment isolation design prosisions for the recirculation pump seal water purge line do l l not meet the explicit rtquirements of GDC 55 nor does the design satisfy the GDC on some other i defined basis as outlined in SRP Section 6.2.4. It is our position that the isolation design in the lastance is inadequate and 7hoeald be modified to satisfy GDC 55 cither esplicitly or on some other l defined basis, with the appregwte justification. (6.2) l l
l Ancoment 3 202.tM l l
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MM 23At,100AT t
i lEsamA=rd Plant arv A i
m.4o l i
i With tespect to Figure 6.2 384 (a) loclude the isolation valve arrangem<nt of the atandby liquid control system line.
i (b) Identify the line lebeied in the figure as 'WDCS A* (it joins the RWCU line prior to its connection to the feedwater line), and discuss the isolation provisions for that line. ;
m.41 I '
Provide a diagram or reference to figtre(s) showing tbc isolation valve arrangement for the lines
. identified below. For the isolation vahe design of each of these lines, proside justification for t
, not meeting the explicit requirements of GDC 56, and demonstrate that the guidelines for acceptable ;
j alternate containment isolation prosisions contained in SRP 6.2.4 are satisfied. The lines in question are 4
l j
o HPCS and RHR test and pump mini!)ow bypass lines j o RCIC pump miniflow bypass line
- r i o RCIC turbine exhaust and pump minillow bypass hnes
, o SPCU suction and discharge lines j t
I I @ .42
! i i
i Describe the isolation provhions for the containment purge supply and enhaust lines and discuss l
! design conformance with Branch Technical Position CSB 6-4,' Containment Purge During Normal Opera-tions.* (6.2)
G.43 [
i Discuss the closure times of isolation sahes in sptem lines that can proside an open path from the primary containment to toe ensironment (e.g., containment purge sptem). Also discuss prosisions q
of radiation monitors in these lines having tbc capability of actuating containment isolation. (6.2)
G ,44 j identify the sptem lines whose containment isolation requirements are covered by GDC 57 and
- discuss conformance of the design to the GDC tcquirements. (6.2) a j m ,45 I
For de combustible gas control $3 stems dedge,identif) <!ca5ly those areas that may not be part of the ABWR scope ans pruside relwant inti;rtue cquirements. (6.2)
]
l j Accordin sific accepta,ce :riteria re!ated to the concentration of hydrogen er oxygen in 4 e shcis ami,3r abers ue the following:
j (a) The w . i egen t roductios should be based on the parameters listed in j Table 1 m -., ,6h 1.7 for the purpose of establishing the design basis for l
combustibis , rol Wicms, i
j Amentment 3 20 212g
_ _ _ _ . _ _ _ _ _ _ _ _ _ , - , , _ . . . . . . . _ o-
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ABM imico41 Standard Plant _
mrv A (b) The fission product decay caergy used in the calculation of hydrogen and oxygen production from radiolysis should be equal to or more conservative than decay energy model given in Branch Technical Position AS89 2 in SRP 9.2.$.
Provide justification that the assumptions used in the ABWR in establishing the design basis for ;
the combustible gas control systems are conservative with respect to the criteria a. and b. above .
(6.2) l 4M 47 Provide an analysis of the production and accumulation of combustible gases within the containment following a postulated loss of coolant accident including all applicable information specified in l Section 6.2.5.3 of Regulatory Guide 1.70, Resisjon 3.
430.48 !
I Regarding Containment Type A leakage testing (6.2.6) ,
(a) Proside the values for Pa and Pg.
(b) include the acceptance criterion for L during preoperational leakage rate tests, i.e., !
Lt*L a(Lm/ t Lam), for the case when La (Lim / Lam) = 0.7. j (c) Your acceptance criterion for L m i (SSAR Subsection 6.2.6.1.2.2, item 1) is at variance i with th: staff's current practice for acceptance of Lt m. Also,it does not comply with [
the 10 CFR Part 50, Appendix J. Section 111, item A.I.(a) requirement. Therefore, either j preside sufficient supporting justification for the above requirement or correct the i criterion as appropriate to comply with the requirement. Also, correct the stated i acceptance criterion (SSAR Subsection 6.2.6.1.2.2, licm 3) as appropriate to comply with ;
Appendix J, Section 111, Item A 6.(b) requirement. [
(d) Regarding ILRT, identify the systons that will not be vented or drained and provide reasons for the same.
(c) Provide PilDs and process flow drawini;s for systems that will be sented or drained.
{
430.49 f l
Regarding Tspe B tests (6.2.6) !
(a) Clarify how air locks opened during periods when containment integrity is required by l plant's Technical Specifications will be tested to comply with Appendix J, Section Ill, item (
D.2.(b).(iii). ,
(b) Proside the frequency for periodic tests of air locks and associated inflatable seals.
(c) Provide the acceptance criteria for air lock testing and the associated inflatable seal !
testing, j l
(d) List all containment penetrations subject to Type B tests.
(e) List all those penetrations to be excluded from Type B testing and the rationale for excluding them.
AmMment 3 20213
MM 23A6tooAT manadard Plant arv A -
430Je Regarding T)pe C tests (6.2.6)
(a) Correct the statement (Subsection 6.2.6.3.1 Paragraph 1) as appropriate to ensure that the hydraulic Type C tests are performed only on those isolation valves that are qualified for .
such tests per Appendix J. The current statement implies that these tests are not
[
necessarily restricted to the valves that qualify for such tests. t (b) List all the primary containment isolation valves subject to type C tests and provide the ;
necessary P&lDs.
l (c) Provide the list of valves that you propose to test in the reverse direction and justification for such testing for each of these valves.
l (d) Identify the valves that you propose to test hydrostatically based on their ability to I
maintain a 30 day water leg scal. Also, identify other valves which >ou propose to test l hydrostatically and provide the basis for such tests. Proside the test pressure for all the f I
valves mentioned above. l L
(c) Indicate test pressures for htSIVs (with justification if it is less than Pa) and isolation i valves scaled from a sealing system. -
f (f) Indicate how you will perform Type C leak tests for ECCS systems and RCIC system isolation i vahes. !
(g) Confirm that the interval between two consecut!ve periodic Type C tests will not exceed 2 [
) cars as required by Appendix J. :
(h) State what testing procedures you will follow regarding the vahes that are not coscred by i Appendiz J requirements. !,
430J1 Identify the reporting requirements for the tests. Note that )our response should address compliance with the requirements in this regard as stated in Appendiu J Sections lil.A.(a),IV.A and V. (For example, regarding follow up tests after containment modification,)ou hase not included l Type C testing for affected areas). (6.2.6)
I l 430J2 Regarding Secondary Containment (6.2.6) !
l (a) Identify the special testing procedures you will follow to assure a maximum allowable in leakage of 50 percent of the secondary containment free column per day at a differential pressure of .0.25' water gauge with respect to the outdoor atmosphere (See Section 6.5.1.3.2 ) .
(b) Identify all potential leak paths which bypass the secondary containment. (For such IJentification, see (DTP) CSB 6 3, ' Determination of bypass Leakage Paths in Dual Containment Plants *)
(c) Identify the total rate of secondar) containment bypass leakage to the emironment.
Amenderwas 3 20 2.th
ABWR mum Standard Plant nrv A 430.53 Identify all the interface requirennnts relating to containment leak testing. (6.2.6) 430.54 Regarding Control Room liabitability sptems, (6.4)
(a) Provide the minimum p, ,i;ise pressure at which the control building envelope (which includes the mechanical equipment room) will be maintained with respect to the surrounding air spaces when makeup air is supplied to the design basis rate (295 CFM).
(b) Provide the periodicity for serification of control room pressurization with design flow rate of makeup air.
(c) Clarify whether all the potentialleak paths (to be prosided in Section 9.4.1) include dampers or vahes upstream of recirculation fans.
(d) Identify the action to be taken when there is no flow of the equipment room return fan and consequently the equipment room is oser pressuriied (Table 6.41 contains no information on the abose).
(c) Provide the actual minimum distances (lateral and vertical) of the control room ventilation inlets from major potential plant release points that hase been,used in your control room l dose analpis. Also, preside a schematic of the location of control room intake sents.
(f) Proside Figure 6.4 5 (plan siew) which you state shows the release points (SGTS sent).
(g) Section 6 4.2 4 and Figure 6.41 indicate only one air intet for suppl >ing makeup air to the emergency zone. Ilowescr. Tables 6.4 2 and 15.6 8 and Section 15.6.5.5.2 indicate that there l are Lws autonutie air inlets for the emergency rene. Correct the abose discrepancy as appropriate. Also describe the characteristics of these inlets with respect to their relatise locations and automatic selection control features. State how both flow and isolation in each inlet assuming single actise components failure will be ensured.
(b) Describe the design features for protecting against confined area releases (e.g., raultiple barriers, air flow patterns in sentilation rones adjacent to the emergency zone).
l (i) Describe the specific features for protecting the control room operator from airborne i
radioactisity outside the control room and direct shine from all radiation sources (e g.,
shielding thickness for control room structure boundary, two door sestibules).
(j) Clarify what you mean by *suitained occupancy * (See SSAR Section 6.4.1.1, item 3) for 12 persons.
(k) Proside justification for not specifying any unfiltered infiltration of contaminated air !
into the control room in SSAR Table 15.6 8.
i (1) Proside Subsection 6.3.1.1.6 which you state (SSAR Section 6.4.6) contains a complete I description of the required instrumentation for ensuring control room habitability at all I times.
l Am4mu3 NNh l
/ WUi1 2wt00AT 5.w prd Plant arv A (m)Give schematics for control room emergency mode of operation during a postulated 1.OCA (this is required for calculating control room LOCA doses).
(n) The source terms and control room atmospheric dispersion factors (X/O values) used in the control room dose analysis (See SSAR Tables 15.6-8 and 15.612) to demonstrate ABWR control -
room compliance with GDC 19 are non. consecutive. Therefore, reevaluate control room doses during a postulated 1.OCA using RG 1.3 source terms and assumptions and the methodology given in Reference 4 of SSAR Section 15.6.7. Include possible dose contribution. from containment shine, ESF filters and altborne radioactivity outside the control room. Also check and correct as appropriate the recirculation rate in the control room (22.4 m3 /sec) given in Table 15.6 8.
I l
(o) Section 6.4.7.1,' External Temperature,' provides design maximum external temperatures of ,
100'F and 10'F. How are these values used in the design and assessments related to the ABWR7 What factors, such as lasulation, heat generation from control room personnel and [
a equipment and heat losses, are taken into account? D, these values represent l
' instantaneous
- values or are they temporal and/or spatial averages?
- l (p) Clarify your position on potential hazardous or toxic gas sources onsite of an ABWR, If I applicable, indicate the special features provided in the ABWR design in this regard, to J casure control room habitability. ;
1 !
(q) Identify all the interface requirements for control room habitability systems (e.g.,
instrumertation for protection against toxic gases in general and chlorine in particular; j potential toxic gas release points in the ensirons). l i
430.!5 ,
Regarding ESF Atmosphere Cleanup System (6.5.1) *
(a) Proside a table listing the compliance status of the standby gas treatment s) stem (SGTS) [
- with gzh of the regulatory positions specified under C of RG 1.52. Proside justifications ?
I for each of those items that do not fully comply with the corresponding requirements. In i this context, you may note that the lack of redundancy of the SGTS filter train (the staff
! considers that filter trains are also actise components See SRP 6.4, Acceptance Criterion II.2.B) is not acceptabic. Further, the described sizing of the charcoal adsorbers based on ,
assumed decontamination factors for sarious chemical forms of iodine in the suppression pool [
, is not acceptable (RG 1.3 assumes a decontamination factor of 1 for all forms of iodine and l RG 1.52 requires compliance with the above guide for the design of the adsorber section). !
Therefore, resise charcoal weight and charcoal iodine loading given in SSAR Table 6.51 as
) appropriate.
1
- (b) Specify the laboratory test criteria for methyl iodine penetration that will be identified I as an interface requirement to be qualified for the adsorber efficiencies for iodine gisen l J
in SSAR Table 15.6 8. Also, provide the depth of the charcoal beds for the control room '
emergency system.
l (c) Provide a table listing the compliance status of the instrumentation prodded for the SGTS i for read out, recording and alarm provisions in the control room with ush of the
. instrumentation items identified in Table 6.5.1 1 of SRP 6.5.1. For partial or non. compliance items, provide justif aations.
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t l Amendmcat 3 i
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ABM 234sioarr Standard Plant nry A 1
i (d) Clarify whether primary containment purging during normal plant operation when required to I limit the discharge of contaminants to the environment will always be through the SGTS (See l SSAR Section 6.5.1.2.3.3) . Clarify whether such a release prior to the purge system isolation has been considered in the LOCA dose analysis.
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l (c) Provide the compliance status tables referred to in Items (a) and (c) above for the control room ESF filter trains. (The staff notes that you have comenitted to discuss control room l
ESF filter system under SSAR Section 9.4.1. However, since evaluation of the control room habitability sptem cannot be completed until the information identified above is provided, the above information is requested now.)
(f) Identify the applicable interface requirements for the SGTS and the control room ESF atmosphere cleanup splem.
430J6 ,
Regarding Fission Product Control Sprems and Structures (6.5.3) i (a) Provide the drawdown time for achieving a negative pressure of 0.25 inch water gauge for the l secondary containment with respect to the ensirons during SGTS operation. Clarify whether j the unfiltered release of radioactisity to the environs during this time for a postulated
not be considered prosided the required negative pressure differential is achiesed within 60 ;
seconds from the time of the accident.) i (b) Provide justification (See SRP Section 6.5.3, 11,4) for the decontamination factor assumed ;
in SSAR Table 6.5 2 and 15.6 8 for iodine in the suppression pool, correct the elemental, ;
particulate and organic iodine fractions gigen in the tables to be consistent with RG 1.3, j and 'acorporate the correction in the LOCA analpis tables. Alternatively, taking no credit for iodine retention in the suppression pool, resise the LOCA analysis tables. Note that i the revision of the LOCA analpis tables (this also includes the control room doses) mentioned above is strictly in relation to the iodine retention factor in the suppression f pool (also, there may be need for resision of other parameter (s) given in the tables and 1 these will be identified under the relevant SRP Sections questions).
(c) Identify the applicable interface requirements. l 430J7 i
Regarding SSAR Section 6.7, the staff notes that the Nitrogen Supply Sptem has been discussed (
under this section,instead of the Main Steam isolation Valse Leakage Control Sptem (MSIV LCS) as !
I required by the Standard Format for SARs. The staff will review the material presented in SSAR l l Section 6.7 along with the material that will be presented in SSAR Section 9.3.1.
Regarding MSIV LCS, the staff notes that you are committed to provide a son safety related MSIV l leakage processing pathway tonsistent with those evaluated in NUREG.1169,
- Resolution of Generic l Issue C.8,* August 1956. Since the staff has not finalized its position so far on the acceptability :
of the NUREG findings with regard to the design of the MSIV.LCS, proside pertinent information on the i
l sptem design including interface requirements to esaluate the to.bc. proposed design against the
(
i acceptance criteria of SRP 6.7. (6.7) !
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AmMast 3 20213 l
ABM 2. m .wi Standard Plant ni v n 20.2.15 Chapter 15 Questions 430.!8 The accident analyicd under this section considers only the airborne radioactivity that may be released due to potential failure of a concentrated waste tank in the radwaste enclosure. The SRP acceptance criteria, howeser, requires demonstration that the liquid radwaste concentration at the nearest potable water supply in an unrestricted area resulting from transport of the liquid radwaste to the unrestricted area does not exceed the radionuclide concentration limits specified in 10 CFR Part 20, Appendix B Table 11. Colun,n 2. Such a demonstration will require information on possible dilution and/or decay during transit which, in turn, will depend upon site specific data such as surface and ground water hydrology and the parameters goserning liquid waste mosement through the soil. Additionally, special design features (e.g., steel liners or walls in the radwaste enclosure) may be provided as past of the liquid radwaste treatment systems at certain sites. The staff will, therefore, resiew the site specific characteristics mentioned above indisidually for each plant referencing the ABWR and confine its resiew of ABWR, only to the choice of the liquid radwaste tank.
Therefore, preside information on the following. (15.7.3)
(a) Basis for determining the concentrated waste tank as the worst tank (this may scry well be the case, but in the absence of information on the capacities of major tanks, particularly the waste holdup tanks it is hard to conclude that the abose tank both in terms of radionuclide concentrations and insentories will turn out to be the worst tank).
(b) Radionuclide source terms, particularly for the long lised radionuclides such as Cs.137 and St 90 (these may be the critical isotopes for sites that can claim only decay credit during transit) in the major liquid radwaste tanks.
470.1 Subsection 15 6.2 of the ABWR TSAR prmides your analysis for the radiological consequences of a failure of smalllines carrying g rimary coolant outside of containment. This analysis only considers the failure of an instrument line with a 1/4. inch flow restricting orifice. Show that this failure scenario presides the most sescre radioactiu releases of any postulated failure of a srnali line.
Your esaluation should dcluJc lincs that moct GUC 55 as well as small lines exempt from GDC 55.
470.2 Proside a justificsiion for your assumption that the plant continues to operate (and therefore no iodine peaking is experienced) during a small line break outside containment (Subsection 15.6.2) accident scenario. Also preside the basis for the assumption that the relcase duration in onl) two hours.
470.3 Subsection 154 4.5.1.1 of the FSAR pises the iodine source term (concentration and isotopic mis) used to anal >ic the steam line break outside of containment accident. The noble gas source term, howeser,is not addressed. Proside the noble gas source term used. Also, the table in Subsection 15.6.4.5.1.1 seems heasily weighted to the shortrr lived actisities (i.e., (3134). Proside the bases for the isotopic mix used in your analysis (iodine and noble gas).
Amcem ) M:1
ABWR : m ia m Standard Plant Rt v A 470.4 Subsection 15.6.5.5 states that the analysis is based on assumptions prosided in Regulatory Guide 1.3 except where noted. For all assumptions (e.g., release assumtd to occur one hour after accident initiation, the chemical species fractions for iodine, the temporal decrease in primary containment leakage rates, credit for condenser leakage rates, and dose consertion factors) which desiste from NRC guidance such as regulatory guides and ICRP2, proside a detailed descript!on of the justification for the desistion or a reference to another section of the SSAR where the deviations are discussed in detail. Pros'de a comparison of the dose estimates using these assumptions sersus those which would result from using the NRC guidance.
470J Provide a discussion of, or reference to, the analysis of the radiological consequences of leakage from engineered safety feature components af'er a design basis LOCA.
470.6 For the spent fuel cask drop accident, what is the assumed period for deca) from the stated power condition? What is the justification fc,r that assumption?
470.7 The tables in Chapter 15 should be checked and resised as appropriate. In sescral cases the footnotes contain typographical errors related to defining the scientific notatioe. Table 15.712 also appears to contain inappropriate references to Table 15.716, rather that Table 15.713.
470Ji it is stated that Regulatory Guides 1.3 and 1.45 were used in the calculations of N/O salues k',
Based on the salues presented, it appears as though a Pasquill stability Class F and one meter per second wind speed were sssumed, with adjuatment for meander per Figure 3 of Regulatory Guide 1.145.
If this is not the case, describe the assumptions and justification used in calculuing the N/O salues which are used in Chapter 15 dose assessments.
470.9 -
The SGT5 filter efficiencies of Wi for inorganic and organic iodine are higher than the vi and 70ri salues, respectisel), assumed in Regulatt.r> Guide 1.25 if it can be shown that the building atmo.pbere is eshausted through adsorbers designed to remose iodine. Proside a justification for the ,
use of the higher salues.
e 470.10 Dose related factors such as breathing rates, iodine conscrsion facton and finite scrsus infinite p:
cloud assumptions for calculating the whole body dose are not stated esplicitl), although reference W is made to Regulatory Guide 1.25 and another document. State these assumptions esplicitly and :1 justify use of any salues which desiate from Regulater) Guide 1.25.
A*pJmt ) E . la Qj:
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23A610;AT Standard,Elant Rr v 11 SECTION 20.3 CONTENTS Section Ihlt has i
203.1 Response to Fint RAI Refemmet i 20 } 1 203.2 Renposse to hecend Eti Reference 2 20 } 22 l
TABLES t
Table Ihlt East 4
i 1 20 3-1 s ensithity Stt;dy of Parameters for LOCA Analy 4 203 21 Response to Question 470.4)
I 20 3-2 Core Decay lleat Following LOCA . Short Term Analyses (Response to Question 430.21) 20}$6 1
20 } 3 Integrated Core Decay liest Values . Shert Term I
Anal)vs (Response to Question 430.21) 20}$7 i
20}4 Ilot Stattup Criticality Rod Sequenec (Response to Question 440.11) 20 } 55 i
j ILLUSTRATluNS i Eisne Iais na 20 3-1 Rcd Groups 14, Sequen:e A (Response to Question 440.11) NJ $9
) 20 3-2 Rod Groups 510, Sequenee A j (Response to Question 44111) 20 } 60 i
i 203 3 llot Recovery Criticality Control Rod Pattern (Response to Question 440.11) 203 61 203-li Amendment 3
-_.a_.
D O ABM 23461004r Standard Plant _ arv_.a SECTION 20.3--
, CONTENTS (Continued)
ILLUSTRATIONS Figure Iltle East 203 4 Src Power Control Rod Pattern (Response to Question 440.11) 203-61 20 3-5 10% Power Control Rod Pattern (Response to Question 440.11) 203 62 203-6 257 Power Control Rod Pattern (Response to Question 440.11) 203-62 20 3-7 40To Power Control Rod Pattern
- (Response to Question 440.11) 203-63 20 3-8 53rc Power Control Roa Pattern (Response to Question 440.11) 20 3-63 203 9 1007c Power Control Rod Pattern <
(Response to Question 440.11) 20 3-64 203 10 Grouped 11CU to Control Rod Drive Assignments (Response to Question 440.12) 20 3-65 l
d 203.iii Amendment 3 1
- e ,- --+ ,-,e---r,- e._-+-m,, m---- - --- -%yg-p.-- **----=--ep- - = - - - -v*- - - - - *
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. o MN 23AM00 *T Stiendard Plant arv n 20.3 QUESTIONS / RESPONSES This subsection provides the responses for each of the NRC questions identified in Sections 20.1 and 20.2. For convenience, each question is repeated here before its corresponding response. These questions / responses are provided in grouga corresponding to the NRC Requests for Additional Informa-tion (RAl) referrriced in Section 20.4. Within each group, the questions / responses are presented in the numerical order of the question numbers. Tables and Figures are prosided at the end of each RAI group.
20.3.1 Response to First RAI. Reference 1 ,
QUESTION 210.1 in Subsection 5.2.1.2, the statement is made that Section 50.55a of 10CFR50 requires NRC staff approval of AShfE Code Cases only for Class 1 components. Revise this statement to be consistent with the current (1987) edition of 10CFR50.55a which reouires staff approval of Code Cases for ASME Class 1,2, and 3 components.
RESPONSE 210.1 Response to this question is prosided in revised Subsection 5.2.1.2.
QUESTION 210.2 Revise Table 5.21 or provide additional tables in Subsection 5.2.1.2 which identifies all ASME Code Cases that will be used in the construction and in plant operations of all ASME Class 1,2, and 3 components in the ABWR. All Code Cases in these tables should be identified by Code Case number, revision and title. These tables should include those applicable Code Cases that are listed either as acceptable or conditionally acceptable in Regulatory Guides 1.84,1.85 and 1.147. For those Code Cases listed as conditionally acceptable, verify that the construction of all applicable components will be in compliance with the additional Regulatory Guide conditions.
RESPONSE 210.2 i c Response to this question is provided in reshed Subsection 5.2.1.2 and Table 5.21.
QUESTION 250.1 Subsection 5.2.4.1 should state that the system boundary includes all pressure vessels, piping, pumps, and valves which are part of the reactor coolant system, or connected to the reactor coolant systems, up to and including (A) The outermost containment isolation valve in system piping that penetrates the primary reactor containment.
(B) The second of two valves normally closed during normal reactor operation in system piping that does not penctrate priiary reactor containment.
(C) The reactor coolant system and relief valves.
RESPONSE 250.1 Response to this question is provided in resised Subsection 5.2.4.1.
AmeMment 3 20 5 )
ABM 23A6100AT Standard Plant nrw n QUESTION 252.11 Subsection 5.23.4.23 states that the ABWR design meets the intent of this Regulatory Guide (1.71) by utilizing the alternate approach given in Section 1.8. We cannot review this subsection because we have not received Section 1.8. In additioa, this subsection should be rewritten because it lacks detailed discussion about welder qualification.
RESPONSE 252.11 Response to this question is prosided in revised Subsection 5.23.4.23.
QUESTION 281.1 In Section 5.1 (pas 3.12) the function of the reactor cleanup system filter demineralizer should include the removal of radioactive corrosion and fission products in addition to particulate and dissolved impurities.
RESPONSE 281.1 Response to this question is prosided in revised Section 5.1.
QUESTION 281.2 In Subsection 5.2.3.2.2 (page 5.2 7) irradiation assisted stress corrosion cracking (IASCC) of reactor internal components and its mitigation are not discussed. Present laboratory data and plant experience has shown that IASCC can be initiated even at low conductivity (< 0.3yS/cm) after long ,
exposure to radiation. l RESPONSE 281.2 Response to this question is provided in the new Subsection 5.23.2.4,IGSCC Considerations.
QUESTl0N 281.3 t in Subsection 5.2.3.2.2 (pages 5.2 7 and 8) the ABWR standard plant design does not clearly incorporate hydrogen water chemistry to mitigate IGSCC. Since the plant design life is 60 years, hydrogen water chemistry may be of greater importance in reducing reactor coolant electrochemical l corrosion potential to prevent IGSCC as well as IASCC. If hydrogen water chemistry is the referenced ABWR standard design, the following documents should be cited:
EPRI NP 5283.SR.A, Guidelincs for Pennanent BifR llydrogen li'ater Chemistr i lnstallations 1981
! Resision.
EPRI NP 4947 SR LD, BilR lipirogen li'atcr Chemistry Guidelines 1987 Resision (to be published).
RESPONSE 281.3 Response to this question is prosided in resised Subsection 5.23.2.2.
Amendment 3 20.3-9
Mk 23A6t00AT Standard Plant RTN 11 (2) Criteria for Selecting Stellite Materials:
- 1. Wear resistance
- 2. Weldability
- 3. Experience and senice history
- 4. Radiation levelin area of application (3) Esaluation of Noncobalt containing Material to Replace Stellite:
The major source of cobalt fro n the reactor core has been 11aynes 25 and Stellite 3 (cobalt based alloys) for pins and rollers, respectively, in BWR control rods. Replacement of the cobalt alloy pins and rollers with noncobalt alloys has been extensively investigated under a joint GE EPRI program (Project 13311). The results of this investigation are documented in the report, EPRI NP.2329, Project 1331-1, final Report, March 1982. Tbc current design noncobalt materials are alloy X 750 for control rod rollers and 13 8 Pli for the pins.
QUESTION 281.7 Subsection 5.2.3.2.2.3(4) (page 5.210) states that control of reactor water oxygen during startup/ hot standby may be accomplished by utilizing the de aeration capabilitier, of the condenscr.
In addition, this section states that independent control of control rod drive (CRD) cooling water oxygen concentrations of < 50 ppb during power operation is desirable to protect against IGSCC of CRD materials. Are either one or both of the above dissolved oxygen controls incorporated in the ABWR stai.dard plant design?
l RESPONSE 281.7 in Subsection 5.2.3.2.2.3, control of reactor water orygen by using the condenser and control of control rod drive water were mentioned as dissolved oggen control methods. These two plant features j are not in the Nuclear Island scope, llowever, an interface requirement has been added (see new j Subsection 5.2.5) that requires the remainder of the plant to meet the water quality requirements of !
Table 5.2 5.
QUESTION 281.8 in Subsection 5.2.3.2.7.3(13) (page 5.211) it states that the main steam line radiation monitor indicates an excessive amount of hydrogen being injected. An explanation of this occurrence should be discussed. j l
RESPONSE 281.8 Reponse to this question is provided in revised Subsection 5.2.3.2.2.3(13).
QUESTION 281.9 Subsection 6.4.4.2 (page 6.4 6) discusses personnel respirator use in the event of toxic gas intrusion into the control room, liowever, the chlorine detection system is not discussed. Also, any control functions that are automatically triggered by a chlorine detector alarm (closing intake dampers, energiring control room }{VAC system recirculation) should be identified.
Amendment 3 201t1
ABM 234aooar arv n -
Standard Plant RESPONSE 281.10 item 1 i
Response to item 1 of this questio,is prosided in resised Subsection 5.2.3.2.2.2.
Item 2 Response to item 2 of this question is prosided in resised Subsection 5.2.3.2.2.
Item 3 Information is being obtained and evaluated from operating plants with GEZIP. llowever, this feature is not in the Nuclear Island scope.
Item 4 This feeture is not in the Nuclear Island scope, liowever, an interface requirement has been added (see new Subsection 5.2.6) that requires the remainder of the plant to meet the watcr quality requirements in Table 5.2 5.
7 4
Item 5 New and improved water quality monitoring instrumentation is being constantly developed and l
introduced for use in BWR p! ants. Several useful instruments have been developed and introduced within the past few years. GE will evaluate the state of the art when a BWR is undergoing detailed design and willincorporate such instruments that are necessary to assure proper water quality.
Item 6 d Response to item 6 of this question is prosided in revised Subsection 5.2.3.2.2.3.
l Item 7 l Pesponse to item 7 of this question is provided in revised Subsection 5.2.3.2.2.3.
Item 8 i
Response to item 8 of this question is provided in revised Subsection 5.2.3.2.2.2 and Table l I
5.2 5.
Item 9 l
Response to item 9 of this question is prosided in resised Subsection 5.2.3.2.2.3.
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Amendment 3 20114
ABWR mano.u Standard Plant nrv n Item 10 Response to item 10 of this questbn is provided in revised Subsection 5.23.2.23.
Item 11 Response to item 11 of this question is provided in resised Subsection 5.23.2.23.
Item 12 This design feature is not is the Nuclear Island scope. However, an interface requirement has been added (see new Subsection 5.2.6) that requires the remainder of the plant to meet the water quality requirements in Table 5.2 5.
QUESTION 470.1 Subsection 15.6.2 of the ABWR FSAR provides you analysis for the radiological consequences of a failure of small lines carrying primary coolant outside of containment. This analysis only considers the failure of an instrument line with a 1/4 inch flow restricting orifice. Show that this failure scenario prosides the most severe radioactive releases of any postulated failure of a small line.
Your evaluation should include lines that meet GDC 55 as well as smalllines exempt from GDC 55.
RESPONSE 470.1 The analysis for failure of a small line carrying primary coolant was conservatively analyzed as a failure of an instrument line with full flow for a period of two hours. This analysis is deemed conservative for the reason given below.
Of all the lines carrying coolant penetrating the primary containment wall, only the instrument lines are exempt from GDC 55. All other lines use some form of check valve / motor operated valve combination to stop the flow of primary coolant in the event of a line break. Typically, the motor operated valves close at the rate of two inches per ten seconds. Considering a two inch line and assuming that a flow of 175 pounds per second would result in operator action within 60 seconds, the total mass released over the 70 second period would be approximately 12,000 pounds or about one half of the assumed reles.se over two hours from the instrument line. Using this logic and these simplified calculations, it is found that a two hour instrument line break bounds releases for small lines.
QUESTION 470.2 Provide a justification for your assumption that the plant continues to operate (and therefore no iodine peaking is experienced) during a small line break outside containment (Subsection 15.6.2) accident scenario. Also provide the basis for the assumption that the release duration is only two hours.
I Amendment 3 20 M$
e . o MVR ' unaooar Standard Plant arv ti TABLE 20.31 SENSITIVITY STUDY OF PARAMETERS FOR LOCA ANALYSIS (RESPONSE TO QUESTION 470.4)
Site Boundary 24 Hr. LPZ Dose for 30 Days Dose at 300 m (REM) at 800 m (REM)
"Jh3reid whole Body Tinrold %% ole Hody
- 1. LOCA Results 1.5 0.62 22. 12.
- 2. No Initial I Hr. Hold up 1.5 0.90 22 13
- 3. No Pressure Reduction NC NC 22 13
@ 24 Hrs 4 Iodine Species Consistent 10.0 0.64 1700 13 with Regulatory Guide 1.3 l S. No Suppression Pool 140. 0.92 930 13 Scrubbing
- 6. No Steaml:ne Platcout 1.5 0.62 23 12
- 7. No Steamline Plateout 1.5 0.64 23 12 or Hold up
- 8. No Condenser Plateout 2.3 0.62 340 12
- 9. No Condenser Plateout 280 41 1300 70 or Hold up l
NOTE:
l All evaluations are nuade independently of cach other. l l
Amendment 3 20 M I
\ 23A6100AT Standard Plant arv A 20.3.2 Response to Second RAl Reference 2 QUESTlON 430.1 Provide a failure modes and effects analysis of the control rod diive system (CRDS) in tabular form with supporting discussion to delineate the logic employed. The failure analysis should demonstrate that the CRDS can perform the intended functions with the loss of any active single component. These evaluations and assessments should establish that all essential elements of the CROS are identified and provisions are made for isolatios from nonessential CRDS clements. it should be established that all esseeial equipment is protected from common mode failures such as failure of moderate and high. energy lines. The failure mode and effects analysis of the control rod drives should include water, air and electrical failures to CRDs and how the CRD system operation is affected due to air conta:nination or water contamination. Before finalizing the scope of the analysis, refer to ACRS subcommittee meeting proceedings on the ABWR dated June 1,1988. It is noted that the above information is to be included in Appendix 15B of the SSAR which will be submitted at a later date. However, the evaluation of the functional design of the reactivity control systems cannot be completed until this information is provided. (4.6)
RESPONSE 430.1 FMEAs for the CRDS and other selected systems will be submitted by December 31,1988. The scope of CRDS FMEA willinclude appropriate consideration of the June 1,19S8 ACRS subcommittee meeting proccedings.
QUESTlON .130.2 Regarding Reactor Coolant Pressure Boundary (RCPB) leakage detection systems provide information on the following: (5.2.5)
(a) Describe how the leakage through both the inner and outer reactor vessel head flange seals will be detected and quantified.
(b) List the sou*ces that may contribute to the identified leakage collected in the Reactor Building Equipment Drain Sumps.
(c) Describe how potential intersystem leakages will be monitored for the (1) Low Pressure Coolant injection System, (2) liigh Pressure Core Spray System, (3) Reactor Core Isolation l Cooling System (RCIC) Water side and (4) Residual lleat Removal System Inlet and '
discharge sides. Your response should include all the applicable (for the ABWR design) systems and components connected to the Reactor Coolant System that are listed in Table 1 of SRP Section 5.2.5 and other systems that are unique to ABWR (except those that you have already discussed in SSAR Subsection 5.2.5.2.2, item 11).
RESPONSE 430.2 Response to this question will be provided by November 11,1988. ,
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Amendment 3 20 M2 j
a v 33A6iOOAT Standard Plant REV A QUESTION 430.3 Discuss cotapliance of reactor coolant leak detection systems with Regulatory Guide (RG) 1.45,
' Reactor Coolant Pressure Boundary Leakage Detection Systems", Positions C4, C5, C6, C8, and C9 with respect to the following items: (5.2.5)
(a) Indicators for abnormal water levels or flows in all the affected areas in the event of intersystem leakages.
(b) Sensitivity and response time of leak detection systems used for unidentified leakages outside the drywell.
(c) Qualification relat'ng to seismic events for drywell eqwipment drain sump monitoring system and leak detection systems outside the drpell.
(d) Testing Procedures Monitoring sump levels and comparing thern with applicable Dow rates of fluids in the sumps.
(c) Inclusion of reactor building and other areas floor and equipment drain sumps in ABWR Technical Specifications for !cak detection systems.
Note that a few of the questions above arise because in Subsection 5.2.5.4.1 you state that the total leakage rate includes leakages collected in drywell, reactor building and other area floor drain and equipment drain sumps.
RESPONSE 430.3 Response to this question will be provided by November 11,19SS.
QUESTION 430.4 Clarify whether the RCIC makeup capacity is sufficient to provide also for main turbine stop valves. Also, clarify ..acther this leakage is included in the total leakage mentioned in Subsection 5.2.5.4-1.
RESPONSE 430.4 The RCIC system has sufficient capacity to account for this leakage. The totalleakage mentioned in Subsection 5.2.5.4.1 does not account for this leakage.
QUESTION 430.5 Clarify how Position C.2 of RG 1.29, "Scismic Design Classification" is met for all applicable leak detection systems (also include the leak detection systems outside the drywell). (5.2.5)
RESPONSE 430J All elements of the lear detection and isolation system (LDS) and supporting systems that must accomplish a safety fucction or w' n ose failure could prevent accomplishment of a safety function will be designed to accomrrodm a SSE and remain functional. All such equipment will be designated as Seismic Category I equipment.
l Amendment 3 20.k21
s t 23A6100AT Standarri Plant REV A All LDS equipment related to isolating functions and all equipment of interfacing systems, either providing input signals to the LDS, or which receive LDS isolation signals and accomplish the safety functions related to isolating the reactor coolant pressure boundary (RCPB) or the primary contain-ment vessel (PCV) will thus conforrn to Position C.2 of RG 1.29. Such conformance shall be applied to the LDS itself and to all systems which support the LDS in monitoring for leaks from the RCPB, internal to the drywell or external to the drywell, e.g., the nuclear boiler system and the process radiation moni:oring system provide such support.
The LDS and associated safety systems will also conform to the RG 1.100 position related to satisfying requirements of IEEE 344. Note that RG 1.100 effects interfacing mechanical systems (e.g., the isolation valves and motor control centers, etc., of these systems) to a greater degree than it effects the LDS.
The airborne particulate radioactivity monitoring system of the LDS will also meet the guidelines of RG 1.45, Position C 6 and will be designed to remain functional when subjected to a SSE.
QUESTION 430.6 Identify all the interface requirements relating to RCPB leakage detection systems. (5.2.5)
RESPONSE 430.6 There are no RCPB leakage detsetion system safety related interfaces for the ABWR Standard Plant.
This will be reflected in Section 1.9.
QUESTION 430.7 In the SSAR section devoted to containment functional design, identify clearly those areas that are not part of th: ABWR scope and provide relevant interface requirements. (6.2)
RESPONSE 430.7 There are no containment safety related interfaces for the ABWR Standard Plant. This will be reflected in Section 1.9.
QUESTION 430.8 l l
With respect to the design bases for the containment: (6.2) l QUESTION 430.8a Discuss the bases for establishing the margin between the maximum calculated accident pressure or pressure difference and the corresponding design pressure or pressure difference. This includes the design external pressure, internal pressure, and pressure between subcompartment walls.
RESPONSE 430.8a The containment pressure response to a postulated accident is divided into three different time periods: vent clearing; short term; and long term. The most dynamic processes occur during vent I clearing and result in a very rapid rise in containment pressure and the maximum differential l pressure across the diaphragm floor. Because these processes are so dynamic, a margin of 30% between ;
the maximum calculated pressure and the design pressure is specified for design purposes. The peak l containment pressures are reached during the short term period. For this time period a margin of 15% l Amendment 3 20.3 24 l
l
i r ABM 2346woar Standard Plant nrv A to the maximum calculated pressure a specified. This 15% margin is judged to be adequate, since the blowdown and containment response are relatively stable and predictable. The short term maximum calculated pressure will bound the long term pressure response.
The 30 and 15% margins described above are the same as those recommended by the Standard Review Plan. i QUESTION 430.8b Discuss the capability for energy removal from the containment under various single failure conditions. State and justify the design basis single failure that affects containment heat removal. I RESPONSE 430.8b ,
l The containment heat removal system, which comprises of three independent loops, has energy I l removal capability to keep the suppression pool temperature within the acceptable limits and other i l guidelines. The design basis of the heat removal system assumes a single failure of a RHR heat ;
exchanger which is the most limiting single failure. l 1
l l QUESTION 430.9 l The Standard Safety Analysis Report (SSAR) states that the analytical models used to evaluate the containment and drywell response to postulated accidents and transients are included in the General Electric Co. report NEDO 20533 and its supplement 1, entitled "The G.E. hlark 111 Pressure Suppression Containment Analytical hiodel". Provide justification that these references are appropriate to use
- for the ABWR Containment design which is not spccified as hf ark 111. Discuss the similarities and i differences of the ABWR design to previously approved hlark Il and blark Ill designs as they relate to the containment and drywell responses to the postulated accidents and the analytical model used for l
the. analyses. Include in the discussion the conservatism used in the model and assumptions, the ,
l applicable test data that support the analytical models, and the sensitivity of the analyses to key l l
parameters. (6.2) l RESPONSE 430.9 1
The analytical models described in the NEDO 20533 are appropriate to calculate the ABWR (containment and dr>well) short term responses to postulated accidents. Though originally written I for prediction of hlark 111 transients, these models, which simulate from first principles the transient conditions in the containment, can be adapted for the ABWR containment configuration.
These models have the capacity to model the reactor pressure vessel, drywell, vent systems, and wetwell (suppression pool and airspace). They are, therefore, adaptable to other containment configuration having the same basic components. Comparison of these analytical models with test data is described and contained in NEDO 20533. In calculating the ABWR containment responses to postulated accidents, these models are used with conservative modeling assumptions. These assumptions are described in Subsection 6.2.1.1.3.3.
The ABWR design, basically, utilizes combined features of htark 11 and blark 111 design, with the exception of a unique feature of two drywell volumes (upper and lower). The veut system is a combination of vertical (blark 11 design) and horizontal (blark III design) vent system, and the wetwell(suppression pool and airspace) is similar to hlark II. Tbc above models have capabilitics to predict the containment and drywell responses to the postulated accidents. The vent system (combined vertical and horizontal vents) can be modeled by employing appropriate vent loss coefficient values.
The unique lower drywell feature of ABWR can be modeled by taking credit for transfer of a Amendment 3 20 W
e f ABM 2346ioaar St.indard Plant REY A conservative fraction of the lower drywell contents into the wetwell airspace. Because the lower drywell is connected to the drywell connecting vents, the inert atmosphere in the lower drywell would not transfer to the wetwell until the peak pressure in the drywell is achieved.
QUESTION 430.10 With regard to the design features of the containment:(6.2)
QUESTION 430.10m Proside general arrangement drawings for the containment structure.
RESPONSE 430.10a The general configuration and the major dimensions of the containment are shown in Figure 3.818.
The nomenclature for various part of the containment and the internal structures are shown in Figure 3.817. The horizontal cross sections of the reactor building and the containment are shown in Figures 3.81 through 3.8 7; the vertical cross sections are shown in Figures 3.810 and 3.811. The code jurisdictional boundary for various codes is shown in Figure 3.812.
QUESTION 430.10b Provide appropriate references to Section 3 of the SSAR which includes the information on the codes, standards, and guides applied in the design of the containment and containment internal structures.
RESPONSE 430.10b )
The applicable codes, standards, and specifications applied in the design of the containment and internal structures are provided in the following subsections of Chapter 3:
htm Subsection Concrete Containment 3.8.1.2 SteelComponents of the 3.8.2.2 Reinforced Concrete Containment Concrete and SteelInternal 3.8.3.2 Structures of the Concrete Containment QUESTION 430.10c Discuss the possibilities of water entrapment inside containment and its effect on the accident analysis.
Amendment 3 20126
. r MN 33A6t00AT Standard Plant RFV A RESPONSE 430.10c The ABWR containment unique design feature lower and upper dr)well volumes has potential for some water entrapment inside containment. Water could be trapped in the lower drywell cavity sod the wetwell equipment and personnel tunnel from two possible sources: (1) from the suppression pool draw-down through the suppression pool return path (see Figure 3.818) or (2) directly from the reactor pressure vessel (RPV). Effect of this possible water entrapment was considered as described below.
For the short term response analysis which determines sizing of the suppression pool, water entrapment was not considered in the analysis, it was found that the short term blowdown is practically over before the spill over from the suppression pool through the return path starts. Any drawdown directly from the RPV to the lower drywell cavity will result in reduced pool heatup which, in turn, will require a smaller pool volume. Therefore, for conservatism, no weler entrapment was considered in determining the minimum suppression pool volume required. For the long term response analysis which determines maximum pool temperature rise, water entrapment was considered in the pool temperature response analysis. This is conservative since water entrapment reduces the suppression pool heat sink capacity and therefore maximizes the ,nool temperature rise. ,
QUESTION 430.10d Provide information on qualification tests that are intended to demonstrate the functional '
capability of the containment structures, systems and components. Discuss the status of any developmental tests that may not have been completed.
RESPOSSE 430.10d The structural integrity pressure test is discussed in Subsection 3.8.1.7.1. The preoperational and inservice integrated leak rate test is discussed in Subsection 6.2.1.6. The shop tests related to reinforced concrete containment vessel which were performed in Japan between 1981 and 1987 are ,
listed below; 1
l
, (Il Fundamental Test
- 1. Transverse Shear
, 2. Openings in RCCV
- 3. Rebar Joints I
(II) Partial Test
- 1. Top slab
- 2. 1.iner and liner anchors
- 3. Diaphragm floor slabjoint
- 4. Penetrations I
(Ill) TotalTest l l
- 1. Large scale (1/6) model All of the developmental tests are complete.
QUESTION 430.11 Proside a detailed discussion of the likelihood and sens. lvity to steam bypass of the suppression pool for a spectrum of accidents, include in you discussion the following information: (6.2)
Amnhent 3 20M
. . r M tb 33A6100AT Standard Plant nrv A RESPONSE 430.11 The ABWR design uses a pressure suppression type containment which is similar to that used in the Mark I,11, and til containment designs. In a pressure suppression type containment, any steam released from the primary system following a postulated LOCA will be condensed by the suppression pool. However, the potential exists for steam to bypass the suppression pool through leakage paths between the drywell and the wetwell airspace. The steam from the drywell leaking directly into the wetwell airspace would produce pressurization of the ABWk costainment as is the case for the other (Mark 11 and 111) containment designs.
Large primary system ruptures generate high pressure differentials across the assumed leakage path which, in turn, give proportionally higher leakage flow rates. However, large breaks also rapidly depressurize the reactor and terminate the blowdown. As the size of the assumed primary system rupture decreases, the magnitude of the differential pressure across any leakage path also decreases. Small breaks, however, result in an increasingly longer reactor blowdown period, which in turn, results in longer durations of the leakage flow. The limiting case is a very small reactor system break which will not automatically result in reactor depressurization. For larger breaks the maximum allowable area of the leakage path is larger, since leakage into the wetwell airspace is of limited duration.
QUESTION 430.lla A comparison of the ABWR pool bypass capability with that for hiark 11 and Mark 111 designs.
RESPONSE 430.lla Tbc ABWR containment design has a steam bypass capability for small breaks of the order of 0.05 ft' (A/[E), same as for the Mark 11 plants.
QUESTION 430.11b The measures for minimizing the potential for steam bypass and the systems provided to mitigate a
the consequences of pool bypass. Discuss and demonstrate the conservatism of assumptions made in the analysis of steam bypass.
RESPONSE 430.11b The potential leakage paths for steam bypass incorporate design features which help in minimizing the potential for steam bypass. The ABWR design includes a wetwell spray system to mitigate the consequences (wetwell airspace pressurization) of suppression pool bypass. Detailed analysis results, discussing the analysis assumptions will be provided by December 31, 1988. (
QUESTION 430.1Ie ,
Identify alllines from which leakage (or rupture) could contribute to pool bypass and wetwell air space pressuritation.
RESPONSE 430.lle !
ll Response to the question will be prosided by December 31,19S8.
Amendment 3 20 M s
. o 23A6t00AT Standard Plant RN A QUESTION 430.11d Identify all fluid lines which traverse the 'vetwell air space and identify those lines which are protected by guard pipe.
RESPONSE 430.11d Response to the question will be prosided by December 31,1988.
QUESTION 430.11e Discuss the rationale and basis for the werwell spray flow capacity.
RESPONSE 430.11e The primary purpose of the wetwell spray system (manually operated) is to provide mitigation for the adverse consequences of the steam bypass. The basis for the wetwell spray flow capacity (500 gpm)is to assure that the maximum containment pressure due to pool bypass does not exceed the containment design pressure QUESTION 430.12 With regard to containment response to external pressure: (6.2)
QUESTION 430.12a Describe the wetwell to dtpell vacuum breaker system and show the extent to which the require-ments of subsection NE of section til of the ASME B&PV Code are satisfied. Discuss the functional capability of the system. Provide the design and performance parameters for the vacuum relief devices.
RESPONSE 430.12a The wetwell to-drywell vacuum breaker system (WDVBS) is safety related consisting of eight (8) 20-inch vacuum breaker valves. Seven valves are required to open to provide an effective flow arca ade-quate to keep the differential pressure between the drywell and wetwell within the negative design value of 2 ;,.i Juring all operating and accident transients. Therefore, the system design accounts for the single failure case in which one valve falls to open. Each vacuum breaker valve shall cpen fully within 1.0 second (start to open at a pressure differential of 0.2 psi and fully open at 0.5 psi).
The vacuum breaker valves shall be installed on the RPV pedestalin separate penetrations from the lower drywell to the suppression chamber airspace, with one valve per penetration. The vacuum breaker valves shall be swinging disk valves which will be actuated by the differential pressure across the valve ports. No external power shall be utilized to open the valves. Valves shall be capable of being manually operated and remotely operated with air operated piston to verify the movement of valve disk. The valve shall be supplied with a position indicator switch in the control roorn that will permit remote indication of valve position in control room.
I QUESTION 430.12b Discuss the basis for selecting a low design capability for external pressure acting across the drywell to wetwell boundary, it is not apparent that the drywell negative design pressure of 2.0 psid is desirable or sufficient.
Amendment 3 20S29
ABM 334acoar Standard Plant RIN. A RESPONSE 430.12b The drywell negative design pressure of 2.0 psid is specified mainly for designing the steel liner. The ABWR primary containment vessel (PCV)is a steellined reinforced concrete containment vessel (RCCV). The main purpose of steel liner is to provide the leaktightness required. This desigs value of 2.0 psid, which has also been specified for the Mark 11 design, is judged to be adequate based on the experience for the Mark 11 plants.
Engineering analyses were performed (with no vacuum breakers) to calculate the negative differential pressure between the wetwell and the reactor building. All possible wetwell ,
depressurization events which may result in the negative differential pressure were considered, and analyses were conducted for the limiting transient event. The negative differential pressure was !
determined to be 1.8 psid, which is below (by 109c) the negative design pressure of 2.0 psid.
l QUESTION 430.12c i The margin between the calculated wetwell to reactor building negative differential pressure ( 1.8 psid) and the design disferential pressure ( 2.0 psid) is not considered adequate. A higher margin j of 159c should be provided at this stage of the design. Further, given the reliance of the BWR pressure suppression design on containment venting to control pressure, discuss the basis for not i prosiding wetwell to reactor building vacuum breakers.
RESPONSE 430.12c Experience indicates that a margin of 10To between the calculated and the design differential pressure should be adequate. As noted in response to Part b of this question this design pressure is for the steelliner and this is not a load carrying component to provide structural integrity of the primary containment boundary. The reinforced concrete walls (about 6 ft. thick) are the main load carrying components whose design is controlled by the internal design pressure 45 psig which is <
aarried by rebar. The concrete walls (are not vulnerable) are subjected to compression under the 2.0 l psid negative design pressure. Therefore,it is not necessary to provide wetwell to reactor building !
vacuum breakers. )
1 QUESTION 430.12d in the analysis of wetwell to reactor building negative differential pressure calculation, a 500 gpm wetwell spray now rate was used. Proside the basis for the assumption and the design basis for the wetwell spray capacity.
RESPONSE 430.12d See response to Question 430.11e.
QUESTION 430.13 Section 6.2.1.1.3 of the SSAR states that the containment functional evaluation is based upon the consideration of several postulated accident conditions including small break accidents. Provide the assutaptions, analysis and results of the small break accidents considered, and demonstrate that the identified (in the SSAR) feedwater line and steam line breaks are the limiting accidents.
RESPONSE 430.13 Response to this question will be provided by December 31,1988.
Amendm<nt 3 20 W
MN 23AnooAT Standard Plant RFV A ,
i QUESTION 430.14 Provide analyses of the suppression pool temperature for transients involving the actuation of safety / relief valves. Provide the assumptions and conservatism employed in the analyses so that an assessment could be made for conformance to the acceptance criteria set forth in NUREG 0783,
' Suppression Pool Temperature Limits for BWR Containments.* (6.2) t i
RESPONSE 430.14 Suppression pool temperature analyses, for transients involving the actuation of safety relief valves (SRVs) to show conformance to NUREG 0783 are not required. Recent studies conclude that the pool temperature limit for SRV discharge is not necessary cod may be climinated. Results of these studies are documented in the GE Report NEDO 30S32, Class I, December 19S4. This report has been prosided to the NRC Staff sia BWR Oncts Group letter, BWROG 8513, of March 21,1985.
A temperature limit for BWR suppression pools during SRV discharge was specified in NUREG 0783.
This limit was established because of concerns about unstable condensation and associated high loads
- on the containment structure at high suppression pool temperatures. The concern was raised because of experience in BWRs with prolonged SRV discharge without quencher devices. The NRC established the temperature limits in NUREG 0783 based on data available at the time it was issued in 19S1. At that time sufficient data was not available to confirm that quenchers were effective in eliminating the unstable condensation loads.
Since NUREG 0783 was issued, scaling laws have been developed and confirmed for the discharge and condensation of steam in a suppression pool. Also, the subscale data base has been expanded over a range of pool temperatures up to saturation temperature with both straight pipe geometries and with quencher devices. The confirmation of the scaling laws and the expanded data base now proside strong support for the climination of the pool temperature limit for SRV discharge with quenchers (T. and X-quenchers). For details, please refer to the GE Report NEDO 30832, Class I, December 1984, Elimination of Limit on BWR Suppression Pool Temperature for SRV Discharge with Quenchers.
The ABWR design utilizes X quencher discharge devices which are the same as that used for the Mark 11 and Mark 111 designs and evaluated in the recent study noted above. This study determined that the dynamic pressures (loads) due to SRV discharge decrease as the pool temperature approaches ,
, saturation temperature, and concluded that the pool temperature limit, specified in NUREG 0783, for 1 SRV discharge is not necessary and may be eliminated.
Therefore, the acceptance criteria set forth in NUREG 0783 are not necessary and, hence, suppression pool transient analyses involving the actuation of SRVs are not needed.
QUESTION 430.15
, Proside the pressure at which the maximum allowable leak rate of 0.5%/ day is quoted. (6.2)
RESPONSE 430.15 Response to this question is provided in revised Table 6.2 2.
QUESTION 430.16 Provide engineered safety systems information for containment response analysis (full capacity operation and capability used in the containment analysis), as indicated in Table 6 7 of Regulatory j Guide 1.70, Revision 3. (6.2)
AmenJment 3 20 M t
- - _ _ _ _ _ _ _ _ _ _ , _ . _ _ _ - , _ .._ , _ ~_.. _ __
O
- ABM 3346mur :
Standard Plant ,
nry A RESPONSE 430.16 i
Response to this question is prosided in resised Subsection 6.2.1.1.3.2 and new Table 6.2 2a.
QUESTION 430.17 in the design evaluation section for containment subcompartments (Section 6.2.1.2.3), provide the information necessary to substantiate your assessment that the peak differential pressures do not ex.
I cced the design differential pressure. Guidance for the information required is provided in Regula.
l tory Guide 1.70, Resision 3, Section 6.2.1.2., ' Containment Subcompartments', Design Evaluation.
l l RESPONSE 430.17 l
Response to this question will be prosided by December 31,198S. ;
l l
QUESTION 430.18 Describe the manner in which suppression pool dynamic loads resulting from postulated loss of coolant accidents, iransients (e.g., relief valse actuation), and seismic events have been integrated into the affected containment structures. Provide plan and section drawings of the containment illustrating all equipment and structural surfaces that could be subjected to pool dynamic loads. For each structure or group of structures, specify the dynamic loads as a function of time, and specify the relative magnitude of the pool dynamic load compared to the design basis load for each structure. Provide justification for each of the dynamic load histories by the use of appropriate experimental data and/or analyses.
Describe the manner by which potential asymmetric loads were considered in the containment l design. Characterize the type and magnitude of possible as)cimetric loads and the capabilities of the affected structures to withstand such a loading profile. (6.2) 1 RESPONSE 430.18
( Response to this question will be prosided by November 11,19sS.
QUESTION 430.19 i
Provide information to demonstrate that the ABWR design is not vulnerable to a safety relief valve j discharge line break within the air space of the wetwell, coupled with a stuck open relief valve l after its actuation as a result of the transient. (6.2)
RESPONSE 430.19 Response to this question will be prosided by November 11,1988.
QUESTION 430.20 Discuss suppression pool water makeup under normal and accident condition. (6.2)
RESPONSE 430.20 Under normal conditions, make up water to the suppression pool can be added by the suppression pool clean up (SPCU) system. Suction is taken from the condensate s'orage pool (CSP) through a line that primarily supplies the high pressure core flooder (itPCF) system and the reactor core isolation cooling (RCIC) system. The SPCU pump outlet is piped to the suppression pool.
Amendm<ni 3 20.3 32 1
ABM 31xeicoar Standard Plant arv A Under loss of coolant accident conditions the ECCS systems (HPCF and RCIC) take primary suction from the CSP and secondary suction from the suppression pool. Suction from the CSP is the preferred source of water. The containment accident response (pressure and temperature) analyses neglect this source of make up water for conservatism.
For post accident suppression pool makeup or containment flooding, the HPCF system can take suction from the CSP and pump water through the HPCF suppression pool return line. This will provide makeup to the suppression pool or fill the containment to a water level consistent with containment design pressure. For the extreme situation where containment flooding is desired, additional water can be added to the CSP using fire hoses or another alternate source of water. For containment flooding the suppression poolis completely filled and the drywell flooded to the desired level.
QUESTION 430.21 With respect to mass and energy release analyses for postulated loss of coolant accidents identify the sources of generated and stored energy in the reactor coolant system that are considered in the i analyses ofloss of coolant accidents. Describe the methods used and assumptions made in calcula-tions of the energy available for release from these sources. Address the conservatism in the calcu.
lation of the available energy from each source. Tabulate the stored energy sources and the amounts of stored energy. For the sources of generated energy, provide curves showing the energy release rates and integrated energy release. (6.2)
RESPONSE 430.21 The energy released for postulated loss-of coolant accidents is comprised of (1) the energy gene-rated by fission product decay, and (2) stored energy in the reactor system. For short term re-sponse analyses, ANS 5 decay heat curve plus 20% margin is used for added conservatism. The rate of release of core decay heat is provided ir, Table 20.3 2 as a function of time after accident initia-tion, and Table 20.3 3 provides integrated decay heat release rate. For long term analyses ANS 5 decay heat curve with to added margin is used.
The sensible stored energy in the reactor coolant system is made available to the reactor coolant by modeling the heat sources as heat capacity modes in the analyses. Following each postulated accident event, the total stored c.iergy is made available for transfer to the reactor coolant. An estimated total amount of available storr nergy is about 200 x 106 Blu.
QUESTION 430.22 in the SSAR sections devoted to containment heat removal systems, identify clearly those areas that may not be part of the GE scope and provide relevant interface requirements. (6.2)
RESPONSE 430.22 There are no containment heat removal system safety related interfaces for the ABWR Standard Plant. This will be reflected in Section 1.9.
QUESTION 430.23 i The SSAR states that the containment heat removal system is designed to limit the long term temperature of the suppression pool to 207'F. The calculated peak pool temperature is l 206.46*F for the feedwater line break. With respect to this analysis provide the following ;
information: (6.2) l Amendment 3 20333 1
MM 23A6100AT Standard Plant REV A QUESTION 430.23a The justification that this is tbc limiting accident with respect to the maximum temperature in the suppression pool.
RESPONSE 430.23a In determining energy removal capability of the containment heat removal system, various potential bounding transient and accident event were analyzed. The events analyzed are:
l (1) Potential Bounding Transients on Suppression Pool Temperature ;
- - Inadvertent Open Relief Valve (IORV)
- Loss-of Coolant Accident (LOCA)(whole spectrum of LOCAs)
(2) Normal Shutdon Cooling (3) Emergency Shutdon Cooling l (4) Anticipated Transients Without Scram (ATWS)
A feedwater line break (FWLB), which is the largest liquid break, was determined to be the most bounding event for pool temperature r:sponse. Liquid breaks are expected to be more bounding than steam breaks, since liquid breaks are expected to result in pool drawdown. Pool drawdown will substantially reduce the heat sink capacity of the suppression pool.
QUESTION 430.23b The bases for the design margin between the design and calculated temperatures.
RESPONSE 430.23b The wetwell design temperature is 219'F (see Table 6.21). The long term pool temperature of 207'F is to assure sufficient net positive suction head (NPSil) for the pumps. The calculated peak pool temperature of 206.46 F demonstrates that the containment heat removal system has ad.
equate energy removal capability.
QUESTION 430.23c All assumptions used in the analysis and conservatism associated with each. Include the effects of potential temperature stratification in the suppression pool and its effects on heat removal capa.
bility of the system.
RESPONSE 430.23c Analysis assumptions are listed in subsection 6.2.1.1.3.3.1.2.
During the LOCA blowdown, there exists a potential for temperature stratification in the suppres-sion pool. During this period most of the mass and energy is release to the pool through the top
, horizontal vents. As a result, the top portion of the pool will be heated more than the lower por-tion. The temperature in the lower part of the pool where the RilR suetion is located can be expected to be lower than the bulk pool temperature thus, the heat removal through the RilR heat exchanger may be less than that expected if a uniformly mixed pool temperature at the RilR suction is assumed.
Amendment 3 20.344
O 8 ABM a346iocar Standard Plant REV A The long. term pool temperature analyses assume a well mixed uniform suppression pool temperature.
It is believed that the location of the RHR suction and return lines in the suppression pool, and other conservatisms in the analyses will more than offset the effect of potential pool stratifica-tion. The RHR suction and return line con 0guration will be designed (similar to Mark 111 design) to provide adequate pool nuixing and reduce the pool thermal stratification. The long. term analyses con-servatively model and use a lower than expected suppression pool volume; no credit for heat sinks in the drywell and wetwell; and no credit foe the ECCS suction from the condensate storage pool. Fur-thermore, based on design practices, the RHR heat exchanger thermal performance is considerably bet-ter than the design minimum.
QUESTION 430.23d The identification of the decay heat cutve used in the analysis.
RESPONSE 430.23d i ANS $ decay heat curve.
QUESTION 430.24 Provide the design bases for the spray features of the containment heat removal system. Provide the safety classification of the components associated with the spray feature of the system. (6.2)
RESPONSE 430.24 The drywell spray performs iodine removal which is not a NRC requirement. The dr>well spray de-sign is based on Japan Atomic Energy Research Institute (JAERI) testing. JAERI ha$ tested the iodine capability of PCV spray with 0.1 < F/V < 0.4 and have determined that, as a minimum, this range is acceptable, where F = spray flow rate, m*/hr V = free air volume (drpell), m*
For ABWR E A = 0.11, which is within the acceptable range.
V 7350 (Note: 840 m*/hr - 1.81 x 10' lb/hr is from Table 6.2 2a).
n l
The design bases for the wetwell spray is provided in the Response to 0uestion 430.11e.
Both wetwell and dr>well spray headers are located inside the primary containment vessel and are classified as Safety Class 3.
QUESTION 430.25 Discuss the rationale for continued reliance on sprays as the sole active engineered safety fea.
ture for drpell atmosphere pressure and temperature. Discuss the merits of upgrading the design of ,
drywell fan coolers to provide some capacity for pressure, temperature, and humidity control follow. '
lug an accident. (6.2)
Arnendment 3 20 3-33 i
AMR 334610041 Standard Plant RFV.A RESPONSE 430.25 The ABWR containment design does not require nor does it rely upon sprays for controlling drywell l pressure and temperature below their design values following design basis loss of coolant accident (LOCA) conditions. The primary design objective of the drywell sprays (initiated by operator action) i is to provide removal of the fission products released in the drywell during LOCA. As an option, !
drywell sprays can be utilized in controlling equipment environmental conditions in the drywell.
I The ABWR drywell cooling system design is non. safety grade. Upgrading the design to safety te pro-vide some capacity for controlling drywell thermodynamic conditions following an accident is not re-garded as cost effective. Control of drywell conditions through the suppression pool cooling (RiiR heat exchangers)is an order of magnitude more effective in overall containment heat removal than the i
drywell cooling system. It is not necessary to have the drywell cooling system available for control.
ling conditions in the drywell following an accident. The RHR heat exchangers have adequate heat re-moval capability.
In order to upgrade the drywell cooling system to safety grade, extensive design modification will be required. The entire cooling system (cooling units, pipings, ducts, source of cooling water, etc) design will be required to withstand seismic loads and other loads due to a high energy pipe break.
In addition, this upgrading will require an increase in the emergency diesel generator capacity.
Cousidering that it is not necessary to have the drywell cooling system available following an
! accident and the upgrading requires extensive design modifications,it is concluded that there is no technical merit in upgrading the drywell fan coolers.
QUESTION 430.26 The time period assumed for initiation of the containment heat removal system after a LOCA is 10 minutes requiring operator action. It is the staff's position that this time period is too restric.
tive. In fact previous BWR designs ( Grand Gulf's Mark 111) use 30 minutes actuation time. Provide the reasons why the ABWR does not provide more flexibility with respect to the time required for ac.
tuation. (6.2)
RESPONSE 430.26 Response to this question is provided in revised Subsections 6.2.1.1.3.3.1.2, 6.2.2.2 a n d 6.2.2.3.1. In addition, the following clarification is provided.
For the RilR response to a LOCA,10 minutes was assumed as the time fo' lowing the LOCA initiation I when containment cooling is initiated. The ABWR RilR is designed with its heat exchanger always in se. I ries with the pump. As soon as RilR injection flow initiates after depressurization the RilR heat ex. i changer is in the flow path and cooling the water. For a larg: break depressurization can occur in 3 to 5 minutes, at which time containment cooling begins as RilR injection sta:ts. For the large break analysis,10 minutes was conservatively assumed as the start of containment cooling.
The question mentioned the presious Grand Gulf design. Unlike the ABWR, the Grand Gulf design re-quired operator act'on to perform valve alignment to bring the RilR heat exchanger into the flow path to initiate containment cooling, j t
The ABWR design requirement for core cooling is that the ECCS shall be completely automatic in op- ;
eration (i.e., no operator action required) for at least 30 minutes following a LOCA.
I n
Amendment 3 NW l
i
ABM 334amar Standard Plant REV A QUESTION 430.27 Describe the design features of the suppression pool suction strainers. Specify the mesh size of the screens and the maximum particle size that could be drawn into the piping. Of the systems that receive water through the suppression pool suction strainers under post accident conditions, identify the system component that places the limiting requirements on the maximum size of debris that may be allowed to pass through the stainers and specify the limiting particle size that the component can circulate without impairing system performance. Discuss the potential for the strainers to become clogged with debris. Identify and discuss the kinds of debris that might be developed following a loss of coolant accident. Discuss the types of insulation used in the containment and describe the behasior of the insulation during and after a LOCA. Include in your discussion information regarding compliance with the acceptance criteria associated with USI A-43 as documented in NUREG-0897. (6.2)
RESPONSE 430.17
, Response to this question will be provided by December 31,1988.
QUESTION 430.28 t Proside analyses of the net positive suction head (NPSil) available to the RilR pumps in accordance with tbc reccmmendations of Regulatory Guide 1.1. Compare the calculated values of available NPSil to the required NPSli of the pumps. (6.2)
RESPONSE 430.28
! Response to this question is prosided in resised Subsection 6.2.2.3.1 and new Table 6.2 2b.
QUESTION 430.29 i
i in SSAR Section 6.2.3, identify clearly those areas that may not be part of the ABWR scope and pro-side relevant interface requirements.
RESPONSE 430.29
) There are no secondary containment safety.related int:rfaces for the ABWR Standard Plant. This i will be reflected in Section 1.9.
i QUESTION 430J0 Provide a tabulation of the design and performance data for the secondary containment structure.
Provide the types of information indicated in Table 617 of Regulatory Guide 1.70, Revision 3. (6.2)
RESPONSE 430J0
) Response to this question will be provided by December 31,1988.
l QUESTION 430J1 4
Describe the valve isolation features used in support of the secondary containment. Specify the plant protection system signals that isolate the secondary containment and activate the standby gas treatment system. (6.2)
Amendment 3 20.337 i
M\ 33A6100AT Standard Plant arv A RESPONSE 43031 l l
Response to this question will be prosided by November 11,1988.
QUESTION 430.32 ,
l Identify and tabulate by size, piping which is not provided with isolation features. Provide an t analysis to demonstrate the capability of the Standby Gas Treatment System to maintain the design ne. ,
gative pressure following a design basis accident with all non isolated lines open and the event of i the worst single failure of a secondary containment isolation valve to close. (6.2) i RESPONSE 430J2 Response to this question will be prosided by December 31,19S8.
QUESTION 43033
- Discuss the design provisions that prevent primary containment leakage from bypassing the secondary containment standby gas treatment system and escaping directly to the environment, include 3
a tabulation of potential bypass leakage paths, including the types of infstmation indicated in Table l 618 of Regulatory Guide 1.70, Revision 3. Provide an evaluation of potential bypass leakage paths ;
considering equipment design limitations and test sensitivities. Specify and justify the maximum allowable fraction of primary containment leakage that may bypass the secondary containment 2
structure. The guidelines of BTP 6 3 should be addressed in considering potential bypass leakage t paths. (6.2)
RESPONSE 43033 Response to this question will be prosided by November 11,1988.
i QUESTION 43034 l
Provide a list of the secondary containment openings and the instrumentation means by which each (
is assured to be closed during a postulated design basis accident. (6.2) ;
i RESPONSE 430J4 l Response to this question will be prosided by November 11,1988.
l QUESTION 430J5 Provide a table of design information regarding the containment isolation provisions for fluid j system lines and fluid instrument lines penetrating the containment which are within the GE scope of the ABWR design. Include as a minimum the following information:
(1) General design criteria or regulatory guide recommendations that have been met or other defined i
bases for acceptability; (2) System name; (3) Fluid contained; (4) Line size; Amendment 3 20.3 38
ABM 2346ioorr nry A Etandard Plant (5) ESF system (yes or no);
(6) Through line leakage classification; (7) Reference to figure in SSAR showing arrangement of containment isolation barricts; (8) Location of vaht (inside/outside containment);
(9) Type C leakage test (yes or no);
(10) Valve type and operator; ,
(11) Primary mode of valve actuation; (12) Secondary mode of valve actuation; (13) Normalvahr position; (14) Shutoon vahr position; (15) Post accident valve position; (16) Power failure vahr position; (17) Containment isolation signals; (18) Valve closure time; and (19) Power source. (6.2) f RESPONSE 430J5 ,
Response to this question will be prosided by November 11,19SS.
QUESTION 430J6 1 For isolation valve design in systems not within the ABWR scope, identify the systems and the rel.
evant interface requirements. loclude a discussion on essential and non essential systems per Regula-tory Guide 1.131 and the means or criteria provided to automatically isolate the nonessential systems by a containment isolation signal. Also, include a discussion on the requirement that the setpoint pressure which initiates containment isolation for nonessential penetrations be reduced to the minimum value compatible with normal operations. (6.2)
RESPONSE 43036 Response to this question will be prosided by November 11,1988. I QUESTION 43037 Specify all plant protection signals that initiate closure of the containment isolation valves.
(6.2) 4 Amend:ntet 3 20S3) 1
ABM a3462004T Standard Plant REV A RESPONSE 430J7 Response to this question will be prosided by November 11,1988.
QUESTION 43038 Describe the leakage detection means provided to identify leakage for the outside contalsment remote manual isolation valves on the following influent lines: Feedwater, RHR injection, HPCS, q
standby liquid control, RWCU connecting to feedwater line, RWCU reactor vessel head spray. (6.2)
RESPONSE 430J8 r .
Response to this question will be prosided by November 11,1988, i QUESTION 43039 ,
The containment isolation design provisions for the recirculation pump seal water purge line do
, not meet the explicit requirements of GDC 55 nor does the design satisfy the GDC on some other de-fined basis as outlined in SRP Section 6.2.4. It is our position that the isolation design in the in-stance is inadequate and should be modified to satisfy GDC 55 cither explicitly or on some other de-fined basis, with the appropriate justification. (6.2)
RESPONSE 430J9 1
The ABWR RIP purge lines penetrating the primary containment are currently equipped with one check valve each inside and outside containment and are currently 15 n'm (1/2 inch) size. This size is less
' than the ABWR instrument line size of 20 mm (3/4 inch). Therfore, the same design criteria (GDC 55, Reg. Guide 1.11, and SRP 6.2.4) apply to the RIP purge line design.
I Paragraph C.lb(2) of Regulatory Guide 1.11 coolant loss must be within the reactor coolant makeup system capability. The ABWR RCIC system provides normal reactor coolant makeup and is capable of
' making up' coolant to the reactor with a nominal l' diameter broken pipe discharging reactor cool-ant. Therefore, due to the small 1/2' size of the RIP purge lines, the curret.t containment isolation valve configuration is in accordance with current NRC requirements.
QUESTION 430.40 4 With respect to Figure 6.2 3ea e
(a) loclude the isolation valve arrangement of the standbyliquid control sptem line. {
I (b) Identify the line labeled in the figure as 'WDCS A'(it joins the RWCU line prior to its con-nection to the feedwater line), and discuss the isolation pro $isions for that line.
RESPONSE 430.40 ,
L i
Response to this question will be presided by November 11,1988. j i
i ,
1
+
1
-n ABM ux6iotsr l Standard Plant arv.A QUESTION 430.41 Provide a diagram or reference to figure (s) showing the isolation valve arrangement for the lincs ,
identified below. For the imlation valve design of each of these lines, provide justifica' ion for i not meeting the explicit requirements of GDC 56, and demonstrate that the guidelines for acceptable alternate containment isolation provisions conta!ned in SRP 6.2.4 are satisfied. The lines in ques-tion are:
o HPCS and RHR test and pump miniflow bypass o RCIC pump miniflow bypass line o RCIC turbine exhaust and pump miniflow bypass lines 1
! o SPCU suction and discharge lines (6.2) l l RESPONSE 430.41 Response to this question will be prosided by November 11,1988. ,
1 QUESTION 430.42 Describe the isolation provisions for the containment purge supply and exhaust lines and discuss ,
design conformance with Branch Technical Position CSB 6 4,
- Containment Purge During Normal Op- j i
erations.
I i RESPONSE 430.42 l l
The containment purge supply and exhaust lines connect to both the drpell and the wetwell. There l is one purge supply penetration for the drywell and one purge supply penetration for the wetwell.
l Similarly, there is one exhaust penetration through each the drpell and wetwell. The purge supply l
line connection to each or both of the drywell and wetwell has two inboard isolation valves in parallel, located outside of but as close as possible to the primary containment. One of these l valves is intended for use for (hlgh volume) inerting and purge. The other, a two. inch val e,is used for any necessary nitrogen makeup during power operation. The outboard isolation valves are l located in each of the lines for purge supply, nitrogen inerting, and nitrogen makeup. The exhaust line has a similar parallel arrangement for the two vabes located nearest to the wetwell penetration l and the two valves located nearest the drywell penetration. Outboard isolation vahes are located in each of the lines to the plant vent and the standby gas treatment system. Altholation valves are air operated and fail in the closed position, the signal sent from the leak detection and isolation system.
As described above, these isolation valves are in conformance to the supplemental guidance of 1 Branch Technical Position CSB6 4 on containment purge during normal operation.
l QUESTION 430.43 Discuss the closure times of isolation valves in system lines that can provide an opeu path from the primary containment to the ensironment (e.g., containment purge system). Also discuss prosisions of radiation monitors in these lines basing the capability of actuating containment isolation. (6.2)
RESPONSE 430.43 Response to this question will be prosided by Decemb.
Amtadmeat 3 20 Mi
M 23A6100AT Standard Plant arx A QUESTION 430.44 Identify the system lines whose containment ise'.tice requirements are covered by GDC 57 and dis-cuss conformance of the design to the GDC reqt nat?.a. (6.2)
RESPONSE 430.44 Response to this question will be presided by Note:nber 11,1988.
QUESTION 430.45 4
For the combustible gas control systems design, identify clearly those areas that may not be part l of the ABWR scope and provide relevant interface requirements. (6.2)
RESPONSE 430.45 There are no combustible gas control system safety.related interfaces for the ABWR Standard Plant. This will be reflected in Section 1.9.
QUESTION 430<M i
According to SRP 6.2.5 specific acceptance criteria related to the concentration of hydcogen or i oxygen in the containment atmosphere among others are the following:
(a) The analysis of hydroi;en and oxygen production should be based on the parameters listed in Table 1 of Regulatory Guide 1.7 for the purpose of establishing the design basis for combus-tible c.ontrol systems.
(b) The fission product decay energy used in the c.?lculation of hydrogen and oxygen production from radiolysis should be equal to or mo<c conscrutise than the decay energy model given in i Branch Techaical Position ASB9 2 in SRP 9.2.5.
2 Proside justification that the assumptions used in the ABWR in catablishing the design basis for .
- the combustible gas control systems are conservative with respect to the criteria a. and b. above. t j (6.2) i i RESPONSE 430.46 l
Response to this question will be prosided by December 31,1083.
j QUESTION 430.47 ,
4 Provide an analysis of the production a::d vcumulation of combustibic gases within the containment
- following a postulated loss.of coolant a ent including all applicable information specified in
- Section 6.2.5.3 of Regulatory Guide 1.70, ..cvision 3. ,
RESPONSE 430.47 Response to this question will be prosided by December 31,19S3.
QUESTION 430.48 l Regarding Containment Type A leakage testing. (6.2.6)
. i i
AmendentM 3 20 N 1 l l
1
4 s O ABWR m6mu Standard Plant REV A QUESTIO'i 430.48a Proside the values for Pa and Pg.
RESPONSE 430.48a P, approximately 40 psig and 0.5 Pa< Pg < Pa-QUESTION 430.48b Include the acceptance criterion for tL during preoperational leakage rate tests, i.e., Lt -
1.a (Ltm/ Lam), for the case when La (Lim / Lam) = 0.7.
RESPONSE 430.48b Response to this question is prosided in reused Subsection 6.2.6.1.1.5.
QUESTION 430.48c Your acceptance criterion for Ltm (SSAR Subsection 6.2.5.1.2.2, item 1) is at variance with the staff's current practice for acceptance of Ltm. Also,it does not comply with the 10 CFR Part 50, Appendix J, Section 111, item A.1.(a) requirement. Therefore, either provide sufficient supporting justification for the exemption from compliance with the above requirement or correct the criterion
. as appropriate to comply with the requirement. Also, correct the stated acceptance criterion (SSAR Subsection 6.2.6.1.2.2, item 3) as appropriate to comply whh Appendix J. Section lit, item A.i(b) l' requirement.
RESPONSE 430.4Rc Response to this question is provided in reused Subsection 6.2.6.1.2.2.
l QUESTION 430.48d Regarding ILRT, identify the systeros that will not be vented or drained and proside reasons for the same.
RFSPONSE 430.48d Response to this question will be prosideo by December 31,198S.
QUESTION 430.48e Proside P&lDs and process now drawings for systems that will be wnted or drained.
i RESPONSE 430.48e i
Response to this question will be prosided by December 31,19sS.
QUESTION 430.49
, Regarding Type B teus,(6.2.6)
I i
Amendment 3 .%W ,
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_ _ _ . . _ , .--,. -. __ . _ - - . ~ _ . . _ . _ _ . _ , _ . _ , _
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- l MM 33A6100AT Standard Plant nry A QUESTION 430.49a Clarify how air locks opened during periods when containment integrity is required by plant's Tech-nical Specifications will be tested to comply with Appendix J, Section 111, item D 2.(b).(iii).
RESPONSE 430.49a Response to this question is prosided in revised Subsection 6.2.6.2.3.
QUESTION 430.49b Provide the frequency for periodic tests of air locks and associated inflatable seals.
RESPONSE 430.49b Response to this question is prosided by resised Subsection 6.2.6.2.3.
(UESTION 430.4?c Proside the acceptance criteria for air lock testing and the associated inflatable seal testing.
RESPONSE 430.49c Response to this question will be prosided by December 31,1988.
! QUESTION 430.49d 1
) List all containment penetrations subject to Type B tests.
t i RESPONSE 430.49d l
Response to this question will be provided by December 31,1988.
l ,
i QUESTION 430.49e List all those penetrations to be excluded from Type B testing and the rationale for excluding them.
RESPONSE 430.49e Response to this question will be prosided by December 31,1988.
l QUESTION 430.50
) Regarding Type C tests (6.2.6) a I (a) Correct the statement (Subs ch
- 6.2.6.3.1, Paragraph 1) as appropriate to ensure that the
- hydraulic Type C tests are . sormed only on those isolation valves that are qualified for such tests per Appendix J. The current statement implies that these tests are not necessar.
ily restricted to the vahes that qualify for such tests.
(b) List all the r:rimary containment isolation valves subject to Type C tests ud proside the necessary P&lDs.
Amendment 3 20.344 3
b er
'P S ABM 234sioo41 Standard Plant _ arv A (c) Provide the list of valves that you propose to test in the reverse direction and justifica-tion for such testing for each of these valves.
(d) Identify the valves that you propose to test hydrostatically based on their ability to main-tain a 30 day water leg seat. Also, identify other valves which you propose to test hydro-statically and provide the basis for such tests. Provide the test pressure for all the valves mentioned above.
(c) Indicate test pressures for MSIVs (with justification if it is less than P ) and isolation valves sealed from a scaling system.
(f) Indicate how you will perform Type C leak tests for ECCS systcrus and RCIC system isolation uh es.
(g) Confirm that the interval between two consecutive periodic Type C tests will not exceed 2 years as required by Appendix J.
(h) State what testing procedures you will follow regarding the valves that are not cosered by Appendix J requirements.
RESPONSE 43030 kesponse to this question will be prosided by December 31,19SS.
QUESTION 43031 Identify the reporting requirements for the tests. Note that your response should address compli-ance with requirements in this regard as stated in Appendix J, Sections III.A(a),IV,A and V. (For example, regarding follow up tests after containment modification. you have not included Type C test-ing for affected areas). (6.2.6)
RESPONSE 430Jl Response to this question is presided in resised Subsection 6.2.6.4.
QUESTION 43032 Regarding Secondary Containment, (6.2.6)
(a) Identify the speciat testing procedures you will follow to assure a maximum allowable in leakage of $0 percent of the secondary containment free volume per day at a differential pressure of 0.25" water gauge with respect to the outdoor aimosphere (see Section 6.5.1.3.2).
(b) Identify all potential leak paths which bypass the secondary containment. (For such identi-fication, see (BTP) CSB 6 3, Determination of Bypass Leakage Paths in Dual Containment Plants).
l (c) Identify the total rate of secondary containment bypass leakage to the ensironment.
RESPONSE 430J2 l
Response to this question wi'l be prosided by December 31,19SS.
i j AmeMmm 3 W5 1
e .
ABM 2346i004 1 igii3 dard Plant RD' A QUES 110N 43033 Identify all the laterface requirements relating to containment leak testing. (6.2.6) .
RESPONSE 43033 There ere no containment leak testing safety.related interfaces for the ABWR Standard Plant. This will be reflected in Section .1.9.
l 1
QUESTION 430.54 ;
Regarding Control Room Habitability systems, (6.4)
(a) Provide the minimum positive pressuie at which the control building envelope (which includes ,
the mechanical equi,iment room) will be maintained with respect to the surrounding air spaces !
when makeup air is supplied to the envelope at the design basis rate (295 CFM).
(b) Provide the periodicity for verification of control room pressurization with design flow rate of makeup air.
(c) Clarify whether all the potential leak paths (to be provided in Section 9.4.1) include damp.
ers or vahes upstream of recirculation fans.
(d) Identify the action to be taken when there is no flow of the equipment room return fan and consequently the equipment room is ove pressurized (Tabic 6.41 contains no information on 4
the above).
l (e) Provide the actual minimum distances (lateral and vertical) of the control room ventilation ,
inlets from major potential plant release points that have been used in your control room dose analysis. Also, provide a schematic of the location of control room intake vents.
(f) Proside Figure 6.4 5 (plan siew) which you state shows the release paints (SGTS vent).
(g) Section 6.4.2 4 and Figure 6.41 indicate ontv one air inlet for supplying makeup air to the emergency zone. However, Tables 6.4 2 and 15.6 8 and Section 15.6.5.5.2 indicate that there are two automatie air inlets for the emergency zone. Correct the above discrepancy as appro. :
l priate. Also describe the characteristics of these inlets with respect to their relative 1o. i
' estions and automatic selection control features. State h>w both flow and isolation in each I inlet assuming single active component failure will be ensured. l l
(b) Describe the design features for protecting against confined area releases (e.g., multiple l barriers, air flow patterns in ventilation zones adjacent to the emergency zone).
! (i) Describe the specific features for protecting the control room operator from airborne radio.
activity outside the control room and direct shine from all radiation sources (e.g., shield.
ing thickness for control room structure boundary, two-door vestibules).
]
(j) Clarify what you mean by "sustained occupancy * (see SSAR Section 6.4.1.1, item 3) for 12 per, sons, s
(k) Provide justification for not specifying any unfiltered infiltration of contaminated air ,
I into the cor. trol room in SSAR Table 15.6 8.
' l
$ Skb '
, . 1 ABM a346ioo41 l
Standard Plant nry A 1
QUESTION 43033 identify all the interface requirements relating to containment leak testing. (6.2.6)
RESPONSE 43033 There are no containment leak testing safety related interfaces for the ABWR Standard Plant.
QUESTION 43034 Regarding Control Room Habitability systerns, (6.4) !
(a) Provi?e the minimum positive pressure at which the control building envelope (which includes the mechanical equipment room) will be maintained with respect to the surrounding air spaces
, when makeup air is supplied to the envelope at the design basis rate (295 CFM).
(b) Provide the periodicity for verification of control room pressurization with design flow l rate of makeup air. .
1 (c) Clarify whether all the potential leak paths (to be provided in Section 9.4.1) include damp- -
ers or valves upstream of recirculation fans. ,
(d) Identify the action to be taken when there is no flow of the equipment room return fan and consequently the equipment room is over pressurized (Table 6.41 contains no iaformation on the above). ,
(c) Provide the actual minimum distances (lateral and vertical) of the control room ventilation inlets from major potential plant release points that have been used in your control room a dose analysis. Also, provide a schematic of the location of control room intake vents.
(f) Proside Figure 6.4 5 (plan siew) which you state shows the release points (SGTS vent).
] (g) Section 6.4.2.4 and Figure 6.41 indicate onIV one air inlet for supplying rnakeup air to the i
emergency zone. Howeves, Tables 6.4 2 and 15.6-8 and Section 15.6.5.5.2 indicate that there are two automatic air inlets for the emergency zone. Correct the above discrepancy as appro-priate. Also describe the characteristics of these inlets with respect to their relative lo-cations and automatic selection control features. State how both flow and isolation in each inlet assuming single actisc component failure will be ensured.
(h) Describe the design features for protecting against confined area releases (e.g., multiple barriers, air flow patterns in ventilation zones adjacent to the emergency zone).
(i) Describe the specific features for protecting the control room operator from airborne radio-activity cutside the control room and direct shine from all radiation sources (e.g., shield-ing thickness for control room structure boundary, two door vestibules).
(j) Clarify what you man by ' sustained occupancy' (see SSAR Section 6.4.1.1, item 3) for 12 per-sons, r
(k) Provide justification for not specifying any unfiltered infiltration of contaminated air .
into the control room in SSAR Table 15.6 8. l t
l l
l Amendment 3 h6 l i
ABWR mm Standard Plant arv.A (1) Provide Subsection 63.1.1.6 which you state (SSAR Section 6.4.6) contains a complete de-scription of the required instrumentation for ensuring control room habitability at all times.
(m)Give schematics for control room emergency mode of operation during a postulated LOCA (this is required for calculating control room LOCA doses).
(n) The source terms and control room atmospheric dispersion factors (X/O values) used in the control room dose analysis (see SSAR Tables 15.6-8 and 15.612) to demonstrate ABWR control room compliance with GDC 19 are non conservative. Therefore, reevaluate control room doses during a postulated LOCA using RG 13 source terms and assumptions and the methodology given in Referer.cc 4 of SSAR Section 15.6.7. Include possible dose contributions from containment shine, ESF filters and airborne radioactivity outside the control room. Also check and cor.
rect as appropriate the recirculation rate in the control room (22.4 M*/sec) given in ,
Table 15.6 8.
- (o) Section 6.4.7.1,
- External Temperature,' provides design maximum external temperatures of 100 F and .10 F. How are these values used in the design ar 3 assessments related to the ABWR? What factors, such as insulation, heat generation from control room personnel and equipment and heat losses, are taken into account? Do these values represent *instanta-neous' values or are they temperal and/or spatial averages?
(p) Clarify your position on potential hazardous or toxic gas sources ossite of an ABWR.11 ap-plicable, indicate the special features provided in the ABWR design in this regard, to en-sure control room habitability.
4 (q) Identify all the interface 'quirements for control room habitability systems (e.g., instru-
- mentation for protection sgainst toxic gases in general and chlorine in particular; poten-tial toxic gas release points in the environs).
1 RESPONSE 430.54 Response to this question will be prosided by December 31,1988.
QUESTION 430J5 Regarding ESF Atmosphere Cleanup Systems,(6.5.1) [
(a) Provide a table listing the compliance status of the Standby Gas Treatment System (SGTS) with suh of the regulatory posi' ions specified under C of RG 1.52. Provide justifications for each of those items that do not fully comply with the corresponding requirements. In
, this co3 text, you may note that the lack of redundancy of the SGTS filter train (the staff considers that filter trains are also active components See SRP 6.4, Acceptance Criterion II.2.b) is not acceptable. Further, the described sizing of ths charcoal adsorbers based on assumed decontamination factors for various chemical forms of iodine in the suppression pool is not acceptable (RG 1.3 assumes a decontamination factor of I for all forms of iodine and RG 1.52 requires compliance with the above guide for the design of the adsorber section).
Therefore, revise charcoal weight and charcoal lodine loading given in SSAR Table 6.51 as appropriate.
(b) Specify the laboratory test criteria for methyl iodine penetration that will be identified as an interface requirement to be qualified for the adsorber efficiencies for iodine gisen
- in SSAR Table 15.6 8. Also, provide the depth of the charcoal beds for the control room emergency system, t
Amendment 3 20 1-s?
1
MM 23A6100AT Standard Plant arv 4 (c) Proside a table listing the compliance status of the instrumentation provided for the SGTS for read out, recording and alarm prosisjons in the control room with sach of the instrumen-tation items identified in Table 6.5.11 of SRP 6.5.1. For partial or non compliance items, provide justifications.
(d) Clarify whether primary containment purging during normal plant operation when equired to limit the discharge of contaminants to the ensironment will always be thrc.'gh the SGTS (See SSAR Section 6.5.1.2.3.3). Clarify whether such a release prior te 1*ue purge system isola-tion has been considered in the LOCA dose analysis.
(c) Proside the corupliance status tables referred to in items (a) and (c) above for the control room ESF filter trains. (The staff notes that you have committed to discuss control room habitability system cannot be complete until the information identified above is provided.
the above information is requested now.)
(f) Identify the applicable interface requirements for the SGTS and the control room ESF atmo-sphere cleanup system.
RESPONSE 430.55 Response to this question will be prosided by November 11,1988.
QUESTION 430.56 Regarding Fission Product Control Systems and Structures,(6.5.3)
(a) Provide the drawdown time for achieving a negative pressure of Oc25 inch water gauge for the secondary containment with respect to the environs during SGTS operation. Clarify whether the unfiltered release of radioactivity to the environs during this ti:ac for postulated LOCA has been considered in the LOCA dose analysis. (Note that the unfiltered release need not be considered provided the required negative pressure differential is achieved within 60 sec-onds from the time of the accident).
(b) Provide justification (See SRP Section 6.5.3, 11.4) for the decontamination factor assumed in SSAR Tab!cs 6.5 2 and 15.6 8 for iodine in the suppression pool, correct the elemental, particulate and organic iodine fractions giu;n in the tables to be consistent with RG 1.3, and incorporate the correction it. the LOCA analysis tables. Alternatively, taking no credit for iodine retention lo the suppression pool, revise the LOCA analysis tables. Note that the revision of the LOCA analysis tables (this also includes the control room doses) men.
tioned above is strictly in relation to the lodir.e retention factor in the suppression pool (also, there may be need for revision of other parameter (s) given in the tables and these will be identified under the relevant SRP Sections questions). ;
(c) Identify the applicable interface requirements.
RESPONSE 430.56 ;
Response to this question will be prosided by November 11,1988.
Amendmer.: 3 203-as
ABWR miooar Standard Plant arv A QUESTION 0037 Regarding SSAR Section 6.7, the staff notes that the Nitrogen Supply System has been discussed un.
der this section,instead of the hiain Steam isolation Valve Leakage Control System (htSIV LCS) as re-quired by the Standard Format for SARS. The staff will review the material presented in SSAR Section 6.7 along with the material that will be presented in SSAR Section 9.3.1.
Regarding htSIV LCS, the staff notes that you are committed to provide a non safety related htSIV leakage processing pathway consistent with those evaluated in NUREG 1169,"Resolution of Generic Is-sue C 8,* August 1986. Since the staff has not finalized its position so far on the acceptability of the NUREG findings with regard to the design of the htSIV LCS,, provide pertinent information on the system design including interface requirements to evaluate the to be proposed design against the ac-ceptance criteria of SRP 6.7. (6.7)
RESPONSE C0.57 In accordance with Section 8.9 of the GE ABWR Licensing Review Bases (hlurley to Artigas dated Au-gust 7,1987) GE committed to a design that provides a non safety related main steam isolation valve J (htSIV) leakage process pathway consistent with those evaluated in NUREG 1169. Accordingly, the drains and vents are routed to the main condenser for leakage control to take advantage of fission product platcout and holdup in the main steam line, drain line, and the main condenser. Fission products are removed by plateout on the relatively cool condenser tubes. The earlier BWR designs, where the fission products are routed through the reactor building to the standby gas treatment sys-tem, had the disadvantage of increasing the dose rate to plant persorinel. In addition there was no holdup or removal of noble gases, so that dose rate to the public may be higher.
The earlier BWR designs also had the disadvantage of being ineffective if the htSIVs greatly ex- !
, cceded the design leak rate (typically 11.5 standard cubic feet an hour). Because of no uncovery in .
the ABWR design, the ABWR would have less fission product generation during a postulated loss of coolant accident than earlier BWR designs, As a consequence the ABWR design is better able to handle leakages beyond the technical specification limits. '
The ABWR design is also passisc requiring no operator actiers. The valves on the drain lines open t automatically when the reactor is at less than 40 percent steam flow to vent to the main condenser. ;
in addition, the valves fall open on loss of air or elsctrical power to ensure that this pathway ex- ;
ists during an accident. These valves and drain lines are illustrated in the Nuclear Boiler system l P&lD (Figures 5.13).
In conclusion, the ABWR design provides a passive non safety related means for controlling and mitigating the release of fission product leakage through the htSIVs and meets the GE ABWR Licensing Resiew Bases.
QUESTION 00.58 The accident analyzed under this section considers only the airborne radioac:ivity that may be re-
- leased due to potential failure of a concentrated waste tank in the radwaste enclosure. The SRP ac.
ceptance criteria, however, requires demonstration that the liquid radwaste concentration at the !
nearest potable water supply in an unrestricted area resulting from transport of the liquid radwaste to the unrestricted area does not exceed the radionuclide concentration limits specified in 10 CFR Part 20, Appendix B Table 11, Column 2, Such a demonstration will require information on possible j dilution and/or decay during transit which, in turn, will depend upon site specific data such as l
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Amendment 3 20349 l ,
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ABM 334sioaar Standard Plant RFV A surface and ground water hydrology and the parameters governing liquid waste movement through the soll. Additionally, special design features (e.g., steel liners or walls in the radwaste enclosure) may be provided as part of the liquid radwaste treatment systems at certain sites. The staff will, therefore, review the site specific characteristics mentioned atove lodividually for each plant i referencing the ABWR and confine its review of ABWR, only to the choice of the liquid radwaste tank.
Therefore, provide information on the following: (15.73)
(a) Basis for determining the concentrated waste tank as the worst tank (8 bis may very well be .
the case, but in the absence of information on the capacities of major tanks, particularly the waste holdup tanks, it is hard to conclude that the above tank both in terms of radionu:lide concentrations and inventories will turn out to be the worst tank).
(b) Radionuclide source terms, particularly for the long lived radionuclides such as Cs 137 and Sr 90 (these may be the critical isotopes for sites that can claim only decay during tran sit) in the major liquid radwaste tanks.
RESPONSE 434.58 Response to this question will be prosided by December 31,1988.
QUESTION 440.1 SRP 4.6 identifies the following GDCs 23,25,26,27,28, and 29 in the acceptance criteria. Con-firm that the reactivity system, described in Section 4.6 of the SSAR, meet the requirements of the i above GDCs.
RESPONSE 440.1 Section 4.6 has been revised to reference the evaluation of the reactivity system against the re-quirements of the above GDCs contained in Subsection 3.1 i
QUESTION 440.2 l
In Section 4.6.2.3.2.2 analysis of malfunction relating to rod withdrawal, it is stated, 'There ,
are known single malfunctions that cause the unplanned withdrawal of even a single control rod.' Con. l firm that this is a editorial mistake and correct it if so. Otherwise, explain in detail the basis for this statement and why this is acceptable, kESPONSE 440.2 i
This editorial error has been corrected in Subsection 4.6.23.2.2. l l
QUESTION 4403 .
In Section 4.6.1.2 it is stated that CRD system in conjunction with CRC &lS and RPS systems pro-vides selected control rod run in (SCRRI) for reactor stability control. Describe in detail how SCRRI works.
RESPONSE 4403 Response to this question is prosided in resised Subsections 4.6.1.2(10) and 7.7.1.2.2(2).
l Arnendment 3 20350
ABM 33A6t00AT Standard Plant REY A l
QUESTION 430.
In Figure 4.6 8a, CRD system P&lD, sheet 1, piping quality classes AA D. FC.D, FD.D, FD.B, etc. i are shown. Submit the document which explains these classes and relates them to ASME code classes. !
RESPONSE 440.4 l
l This information is scheduled to be included in Section 1.7. Essentially, the first two letters l of the codes specify the pipe primary pressure rating (150 lb.,900 lb., etc.) the type of service l (condensate or reactor water, steam, etc.), and material (carbon or stainless steel). The symbols l *A*, *B' and 'C' represent ASME Section 111, code Classes 1, 2, and 3, respectively. The symbol *D' l
represents ASME Section 8, or ANSI B31.1 or other equivalent codes.
QUESTION 440J in Figure 4.6 8b, the ieck receiver tank is shown. What is the function of this tank? }{ow big is this tank? Will a high 14 vel in the tank impact the operation of the control rod drive? r RESPONSE 440J This leakage collection tank is no longer part of the design. The intent of the leakage collec.
tion system was to assist the operator in identifying which drives were potential candidates for seal replacement during plant outages, which would facilitate plant maintenance planning. liowever, the de.
sign could not provide the level of differentiation of leakage between individual drives needed for this purpose and was therefore deleted. An updated P&lD (Figure 4.6-8b) will be provided by December 31,19S8 to document this change. L l
QUESTION 440.6 Identify the essential portions of tbc CRD system which are safety related. Confirm that the safety related portions are iselable from non.cssential portions. (4.6)
RESPONSE 440.6 The essential portions of the CRD system which are safety related are:
(a) The hydraulic control units (l{CUs),
I (b) The scram insert piping from the l{CUs to the fine motion coritrol rod drives (FMCRDs), and l
(c) The FMCRDs (except the motors)
The non. essential portions of the CRD system interface with the essential portions at the follow.
ing connections to the llCUs:
(a) The accumulator charging water line ;
1 (b) The FMCRD purge water line, and I (c) The scram valve air supply from the scram air header.
Amendment 3 20 3-$1 I -. _ _ - _ _ _ - _ _ _ - _ _ - - - - _ - _ _ _ _ - _ - _ _ _ _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ - _ - _ - - - _ _ _ _ - - - -
\ 23A6100AT Sgndard Plant mA The safety related portions of the HCU and the scram function are protected against failure in the non essential portions of the charging water and purge water lines by check valves. Also,instrumen-
, tation in the charging water line provides signals to the reactor protection system to cause reactor scram in the event of loss of charging water pressure. Loss of pressure in the scram air header causes the scram valves to actuate, resulting in reactor scram. This fail safe feature is the same as provided on current BWR designs using locking piston type control rod drive.
QUESTION 440.7 la the old CRD avstem, the major function of the cooling water was to cool the drive mechanism and its seals to pretlude damaEe resulting f om long term exposure to reactor temperatures. What is the function of purge water flow to the drives? (4.6)
RESPONSE 440.7 The functi on of the purge water flow to the firti m.otion control rod drives is to prevent reactor L t
xater from entering the drive housing during operation. Thh will minimize crud buildup in the drise housing and reduce operator exposure during diive maintenance.
QUESTION 440.8 We understand that the LaSalle Unit 2 fine motion control rod drive demonstration test is stillin
- progress. Submit the test results as soon as it is available.
RESPONSE 440.8 [
At the current time, the LaSalle Unit 2 fine motion control rod drive demonstration test is ex.
pected to be terminated in October 19SS. The final report for the FhlCRD In Plant Test Program, which will include the LaSalle Test results, will be formally issued in September 19S9. .
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QUESTION 440.9 In the present CRD systeta design, the ball check valve ensures rod insertion in the event the accu-
- mulator is not charged or the inlet scram value fails to open if the reactor pressure is above 600
- psig. Confirm that this capability still exists in the ABWR design. (4.6) I 1
RESPONSE 440.9 l The ABWR control rod design does not have the capability of the locking piston control rod design ,
to insert hydraulically using reactor pressure in the event of a failure in the hydraulic control i
units (Lc., scram valve fails or accumulator is not charged). Ilowever, the fine motion control rod drive (Fh!CRD) has a diverse means of insertir.g the control rod using electric inotor run.in if hydrau-lic scram fails. This feature provides the Fh!CRD with the capability to insert the control rod over the entire range of reactor operating preuures.
QUESTION 440.10 l
l In section 4.6.2.3.1, it is stated the scram time is adequate as shown by the transient analyses of Chapter 15. Specify the scram time. (4.6.2.3.2.1) 1 1
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AmenJment 3 20 M 2 i
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MM 23AstoaAT Standard Plant arV A RESPONSE 440.10 The average maximum scram time of all control rods in the core under the reactor conditions with accumulator available and reactor steady state pressure as measured at the vessel bottom below 76.3 Kg/cm's (1085) psig) shall meet the following requirements: (all times are after deenergizing of scram solenoids) lasertion % Time (seconds) 10 1 0.42 40 1 1.00 t 60 1 1.44 1m 11M QUESTION 440.11 l For both the low (*zero*) and operating power region describe the patterns of the control rod ,
groups that are ctpected to be withdrawn simultaneously with the new tod system, and estimate the maximum for the total and differential reactivity worth of these groups. What sort of margin to pc.
riod scram will exist in the low power range. (4.6) 4 RESPONSE 440.11 l (1) Summary of rod withdrawal stratecy The ABWR rod groups are assigned as shown in Figures 20.31 and 20.3-2. The FMCRD step size is 18.3 mm (0.57 of full CRD stroke), with a nominal speed of 30mm/sec. The number of rods per gang for rod groups #1,2,3,4 is 26, i.e., the whole group of 26 rods will be moved simulta-neously as one gang. Group 1 and 2 will be mosed continuously from full in to full out. Group l 3 and 4 which coser the rod pattern condition from cold critical to hot critical, will be moved I in jog cycle in one step at a time. The peripheral rods of group 5 and 6 will be mosed as onc !
gang. For the remaining 7,8,9,10 groups, rods are divided into 4 rod gangs and 8 rod gangs.
j A BWR/6 type banked position witWrawal sequence (BPWS) constraint, called grouped wir*
i quence (GWS),is applied in ABWR as the rod withdrawal sequence guideline, it is in et, ,
j to the low power setpoint (LPSP), or 25cc power. Abose LPSP, the rod withdrawal sequence is t based on core. management pie descloped rod withdrawal sequence in 4 and 8 rod gangs, f i (2) Typical rod catter s at various onwer level l
l (a) Hot criticality after hot recovery (EOIC, rated condition Xc) l Rod pattern: Figure 20.3 3 (quarter core only, same for all Figure 20.3 4 thru 20.3 9)
Rod position of each group: Table 20.3-4 (b) 5% power', Cold Startup, equilibrium Xe, BOEC i
i Rod pattern: Figure 20.3-4 l
2 dfR4RdfM46% ) eY Ns
MM '
33AHMAT Standard Plant arv A (c) 10% power *, Cold Startup, equilibrium Xe, BOEC Rod pattern: Figure 203 5 (d) 25% power *, Cold Startup, equilibrium Xe, BOEC Rod pattern: Figure 203-6 (c) 40% power *, Cold Startup, equilibrium Xe, BOEC Rod pattern: Figure 203 7 (f) 537, power *, Cold Startup, equilibrium Xe, BOEC Rod pattern: Figure 203-8 i
(g) 100% power 100% flow, Cold Startup, equilibrium Xc, BOEC ,
l
) Rod pattern: Figure 203 9 ,
- ctinimum core flow (3) Estimates of maximum reactisity stirth Reacthity Worth Estimates l I
i Group Nie Gruup Mas. Worth 1st Rod Max. Woeth IJt Gang j
- 1 .
2 m 3 2.1%
4 1.5%
i 5 -
6 -
l 7E i 8 J max 3.29 11.2G 1 1.5 %
4 9 i !
10 j l I (4) Marain te aeriod Scram estimates ,
i {
, For 3% total rod worth (full in to full out), the shortest period per sicp is -60 seconds. For l 2% total rod worth, the shortest period per step is -100 seconds. ;
So, for step wise withdrawal therr is plenty of margin to period scram (10 second scram 1 setpoint)
QUESTION 440.12 J
Describe the relative core location of control rods sharing a scram accumulator. Can a failure of the scram accumulator fail to insert adjacent rods? If so, discuss the consequences of that failure.
(4.6)
Amendment 3 20154
7 MM 33AnooAT Standard Plant arv A RESPONSE 440.12 The gros. ped HCU to contiol rod drive assignment and their relative core locations are shown in Figure 20.310. As can be seen, the two control rods sharing a scram accumulator are separated by several core cell locations. A failure of an HCU cannot result in the failure to insert adjacent rods, i
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ABM . 2mioasr Standard Plant REV A TABLE 20.3 2 CORE DECAY HEAT (l) FOLLOWING LOCA SHORTTERM ANALYSES !
(Response to Question 430.21)
Time (see) Normallard Core Heat (2) 0 1.084 2 0.5566 6 0.5501 10 0.3'859 20 0.1239 30 0.0772 I- 31 0.0771 3
60 0.0472 100 0.0427 120 0.G4 121 0.039 200 0.0358
. 600 0.0279 1000 0.0245 ,
J NOTES ,
4 (1) Based on 1973 ANS Standard with 20% margirs.
(2) Nomsali:cd to 102% of rated themtalponer.
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] Amendment 3 :o m J
1
e 0 ABWR mamar Standard Plant ny A TABLE 20.3 3 INTEGRATED CORE DECAY HEAT VALUES (1)
SHORT TERM ANALYSES (Response to Question 430.21)
Integrated Decay Heat h is h ts) la Full Power Seconds (2) 0.0 0.0 0.1 0.1093 1 1.0172 2 1.7083 4 2.8408 1 6 4.0372 l
8 5.3096 101 6.4256 2 10.092 4 12.554 6 13.736 8 14.700 102 15.569 2 19.445 4 26.045 6 31.V4 8 37.234 103 42.293 NOTES (1) Ba:cd cut 1973 ANS Staindard shh 2(Es rnaqbt.
(2) FullPomer = 3.797x 106 Bat /sec.
Amendment 3 20137
e O ABWR amm1 Standard Plant ni v A TABLE 20.3 4 HOT STARTUP CRITICALITY ROD SEQUENCE (Response to Question 440.11)
Groun # Ganqt # Red _Mithdyn To Notch Position 1 (1) 0 48 (12 ft) 2 (2) 0 48 >
3 (3) 0 48 4 (4) 0 48 5 (5) 0 18 (4.5 ft) 6 (6) 0 18 7 A (7) 0 12 (3 ft)
B (8) 0 12 C (9) 0 10 (2.5 ft) 7 (10) 0 12 8 A (11) 0 12 B (12) 0 10 C (13) 0 10 9 A (14) 0 B (15) O C (16) 0 10 A (17) 0 B (18) O C (19) 0 D (20) 0 E (21) 0 Amendment 3 20338
ABWR numu Standard Plant REV A 67 0 p- 1 2 q A p- 3 4 3 q p 2 1 2 1 2 1 q 4 3 4 3 4 3 4 -
51 1 2 1 2 1 2 47 -
3 4 3 4 3 4 3 -
1 2 1 2 1 2 1 2 39 3 4 3 4 3 4 3 4 3 35 2 1 2 1 2 1 2 1 31 4 3 4 3 4 3 4 3 4 27 1 2 1 2 1 2 1 2 23 4
4 1 4 3 4 3 4 -
19 1 2 1 2 1 2 1 15 1 -
3 4 3 4 3 4 3 -
11 L 2 1 2 1 2 1 d
~
I 1 4 3 4 I I 1 2 B C l
2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 62 66 8847744 4
Figure 20.31 ROD GROUPS 1-4, SEQUENCE A (Response to Question 440.11)
Amendment 3 .g339
ABMR zwioast Standard Plant nrw A 67 D 5 q A l
63 59 p- 5 B C
O C 5 ,
8 p 6 10 E C 10 E 6 q 55
- 7 8 7 7 8 7 ~
0 8 C C 8 0 51 10 9 IOC 8 10 0 6 6 0 8 8 47 5 8 8 8A B A 7 8 5 -
8 8 B 10 9 10 g 9A 10 9 10 E 8 8 8 E 8 7 8A 7 OA I O C C A A C C 5 9C IO 8 10 OA 10 E 5 C A A C C 31 B I 8A I 7 8A 7 8 C C A A C C 27 10 8 10 9A 10 98 10 E B 8 8 E 23 5 8 I O O 7 O 5 8 B A A 8 B 19 6 10 8 IO 8 10 6 0 B C 8 0 15
- 7 O 7 I Sg 7 ~
D B C C 0 11 b 6 10 E
8 C 10 E 6 d 7
I 5 B C 8 C 5 d 3 -
L_ 5 i
B C 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 62 66 88477-03
- Figure 20.3 2 ROD GROUPS 5-10, SEQUENCE A (Response to Question 440.11)
Amendment 3 y, mi
ABWR 2mmu Standard Plant arv 4 CONTROL ROD CONF 10UR ATION IN NOTCHE8 WITHDRAWN 1 3 5 7 9 11 13 15 17 10 21 23 25 27 29 31 33 67 18 1 63 18 10 3 59 18 0 0 5 55 12 10 10 7 51 18 0 0 0 9 47 18 10 12 12 11 43 0 0 0 0 13 39 10 10 12 12 15 35 18 0 0 0 0 11 4 31 19 Figure 20.3 3 HOT RECOVERY CRITICALITY CONTROL
, (Response to Question 440.11)
CONTROL ROD CONFIGURATION IN NOTCHES WITHDRAWN 3
1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33
, 67 0 1 j 63 0 8 3 59 0 0 0 5 55 6 8 6 7 51 0 0 0 0 9 1 47 0 8 6 8 11 !
43 0 0 0 0 13 39 8 6 8 6 15 l
]
35 0 0 0 0 0 17 ;
31 19 ,
8847746 Figure 20.3 4 5% POWER CONTROL ROD PATTERN !
(Response to Question 440.11) !
i Amendment 3 20 M1 I
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MM 23A61 COAT Standard Plant RTV A CONTROL ROD CONFIGURATION
. IN NOTCHES WITHDRAWN i 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 l ,.
67 12 1 63 12 8 3 l 59 12 0 0 5 1
1 55 8 8 8 7 pm l $1 12 0 0 0 9 1 47 12 8 8 8 11 43 0 0 0 0 13 39 8 8 8 8 15 35 12 0 0 0 0 17 31 19 I Fi9ure 20.3 5 10% POWER CONTROL ROD PATTERN (Response ta Question 440,11)
CONTROL ROD CONFIGURATION IN NOTCHES WITHDRAWN l 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 f
67 1 l
63 20 3 !
59 0 0 5 !
55 20 20 20 7 l 51 0 0 0 9 47 20 20 20 11 43 0 0 0 0 13 39 20 20 20 20 15 35 0 0 0 0 17 31 19 e847741 Figure 20.3 6 25% POWER CONTROL ROD PATTERN (Response to Question 440.11)
Amendment 3 N .tc
MN 23A6100AT Standard Plant nry A CONTROL ROD CONFl0URATION IN NOTCHES WITHDRAWN 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 .
67 1 .
l 63 3 59 8 10 5 55 7 51 8 10 6 9 r 47 11 43 8 10 6 to 13 39 15 35 10 6 10 6 17 31 19 f
Figure 20.3 7 40% POWER CONTROL ROD PATTERN (Response to Question 440,11)
CONTROL ROD CONFIGURATION IN NOTCHES WITHDRAWN 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 67 1 J 63 3 I 59 10 22 5 55 7
! 51 to 22 8 9 47 11 43 10 22 10 22 13 39 15 35 22 8 22 8 17 31 19 8847706 Figure 20.3-8 53% POWER CONTROL ROD PATTERN (Response to Question 440,11) i Amendment 3 20 % )
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ABM 23461004r Samadard Plant wa 4
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CONTROL MOO CONFIGUR ATION IN NOTCH 88 WITHDRAWN 1 3 5 7 9 11 13 15 17 19 21 23 25 27 29 31 33 67 1 63 3 59 5 f i
55 7 ,
51 8 10 9 I
47 11 I
43 8 13
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39 15 35 10 8 17 31 19 8447747 Figure 20.3 9 100% POWER CONTROL ROD PATTERN (Response to Question 440,11) !
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l' Amendment 3 2am I 1
ABWR 2numxt Standard Plant RFV A DIVISION HCU NO, C_3 A 1 25 50 8 26 51 51 C 52 77 52 D 78 103 52 P
0' 67 D 102 8 22 A
- y 63 100 83 86 14 13 3 15 l_
$9 p 78 93 98 95 81 7 1 12 11 to 4 q 6 ;
55 -
81 84 79 101 94 87 2 16 20 19 7 1 51 87 82 85 80 92 38 103 17 23 6 8 2 16 47 -
92 88 96 83 86 102 89 82 18 22 9 3 24 21 23 --
43 80 89 97 100 99 90 98 96 14 13 25 15 17 18 6 39 99 90 78 93 94 95 91 101 97 24 21 12 11 10 4 9 25 2700 900 35 85 91 84 79 103 30 44 45 51 20 19 6 77 53 58 65 59 31 50 34 29 35 36 37 46 49 71 75 65 69 68 67 52 64 73 27 31 43 42 40 50 38 39 70 72 64 73 74 71 63 54 23 -
48 46 49 28 34 47 43 56 63 76 60 57 70 62 66 -
19 41 27 33 31 48 42 77 62 66 54 59 56 61 15 -
26 32 30 44 45 41 27 61 68 75 53 58 55 -
33 b 29 35 36 37 26 32 55 69 72 67 52 d i I I y 40 28 38 39 60 57 74 I I 3 8 47 33 76 C 180' l
J/I 2 6 to 14 18 22 26 30 31 38 42 46 50 54 58 62 66
, saa n 02 i
Figure 20.310 GROUPED HCU TO CONTROL ROD DRIVE ASSIGNMENTS '
(Response to Question 440.12)
, l Amendment 3 to M.5
,4 ABWR mamu Standard Plant ni v A
20.4 REFERENCES
- 1. Dino C. Scaletti to Ricardo Artigas, Request for Additionalinformation Regarding the General Electric Company Application for Certification of the ABit'R Design, February 22,1988.
- 2. Dino C. Scaletti to J.S. Gay, Request for Additional information Regarding the General Electric Company Application for Certification of the ABit'R Design, July 7,19S8.
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AmenJment 3 20 4,1