ML20035B309
| ML20035B309 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 03/23/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9304010178 | |
| Download: ML20035B309 (14) | |
Text
-
GENuclearEnergy i
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March 23,1993 Docket No. STN 52-001 f
1 Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation I
Subject:
Submittal Supporting Accelerated ABWR Review Schedule - Resolution of TMI-related Outstanding Items
Dear Chet:
Enclosed are SSAR markuups addressing the following DFSER TMI-related outstanding items:
Open items COL Action items 203-6 203-1 203-9 203.1-2 203.1-3 203.1-4 203.1-5 203-2 Sincerely, Y
ack Fox Advanced Reactor Programs cc: Bill Fitzsimmons (GE)
Norman Fletcher (DOE)
Bernie Genetti(GE) 4g.%g now 9304010170=930323 PDR. ADOCK 05200001.
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- #0'3* 6 $ S0 3 4 ffl 3 gev c Standard Plant elevation which would be covered by post LOCA the same time and made from the same sheet flooding for unloading the fuel.
to provide uniformity of relief pressure.
525.23 Pmsure Control (6) The rupture disks are capable of withstanding full vacuum in the wetwell (1) In general, during startup, normal, and vapor space without leakage.
abnormal operation, the wetwell and drywell pressures is maintained greater than 0 psig (7) The piping material is carbon steel. The 2
to prevent leakage of air (oxygen) into the design pressure is 10.5 kg/cm g (150 primary containment from secondary psi), and the design temperature is containment but less than the nominal 2 psig 171*C.
scram set point. Sufficient margin is provided such that normal containment 6.2.5.2.7 Recombiner temperature and pressure fluctuations do not cause either of the two limits to M reached (1) Two permanently installed recombiners are p.
considering variations in initial located in secondary containment. Each {
containment conditions, instrumentation recombiner, as shown in Figure 6.2 40, errors, operator and equipment response takes suction from the drywell, passes the time, and equipment performance, process flow through a heating section, a reactor chamber, and a spray cooler. The (2) Nitrogen makeup automatically maintains a gas is returned to the wetwell.
530 kg/m2 (0.75 psig) positive pressure to avoid leakage of air from the secondary (2) The recombiners are normally initiated on into the primary containment.
high levels as determined by CAMS (if hydrogen is not present, oxygen (3) The drywell biced sizing is capable of concentrations are controlled by nitrogen maintaining the primary containment pressure makeup).
less than 880 kg/m2 (1.25 psig) during the maximum containment atmospheric heating 6.2.5.3 Design Evaluation which could occur during plant startup.
The ACS is designed to maintain the containment in an inert condition except for 6.2.5.2.6 Overpressure Protection nitrogen makeup Leeded to maintain a positive (1) The system is designed to passively relieve containment pressure and prevent air (0;)
the wetwell vapor space pressure at 5.6 leakage from the secondary into the primary kg/cm g. The system valves are capable containment.
2 of being closed from the main control room using AC power and pneumatic air.
The primary containment atmosphere will be inerted with nitrogen during normal operation of (2). The vent system is sized so that residual the plant. Orygen concentration in the primary core thermal pcwer in the form of steam can containment will be maintained below 3.5 volume be passed through the relief piping to the percent measured on a dry basis.
- 20. 2-fo stack.
I M DtT 6 9,$,3 Following an accident, hydrogen concentration (3) The initial driving force for pressure will increase due to the addition of hydrogen relief is assumed to be the expected from the specified design-basis metal water pressure setpoint of the rupture disks.
reaction. Hydrogen concentration will also increase due to radiolysis. Any increase in (4) The rupture disks are constructed of hydrogen concentration is of lesser concern stainless steel or a material of similar because the containment is inerted. Due to dilution, additional hydrogen moves the corrision resistance.
operating point of the containment atmosphere (5) A number of rupture disks are procured a' farther from the envelope of flammability.
6; k Ammendment 16
"r" po.s-s g aus.s.y & M '
^
ABWR 23A6100AB Standard Plant nry c
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q< 'a met. 'The only role of this system related to (5) Prior to initial entry, the drywell shall r
drywell purging for re-entry is to indicate when be purged with air in acccrdance with oxygen levels are high enough to start taking the operating procedure until drywell samples samples that will be used for determining indicate the following conditions are met:
compliance with entry criteria.
(a) Oxygen: Greater than 16.:i percent 6.2.5.6 Personnel Safety content by volume.
Entry into a nitrogen atmosphere is (b) Hydrogen: Less than 14 percent of the particularly hazardous due to ibe fact that the lower limit of flammability, or a body cannot easily detect relative changes in the limit of 0.57 percent hydre gen by nitrogen content of the air. Low oxygen causes volume. (The lower flammability limit blood chemistry changes that can lead to an is 4.1 percent hydrogen content by automatic increase in breathing rate, leading to volume.)
hyperventilation. The individual can lose consciousness in twenty to forty seconds and be (c) Carbon Monoxide: Less than 100 ppm.
totally unable to save himself.
(d) Carbon Dioxide: Less than 5000 ppm.
A general procedure which outlines the critical items to be included in any procedure (c) Airborne Activity: Less than controlling purged drywell entry is provided applicable limits in 10CFR20, or below. This procedure is intended to be a equivalent.
framework of minimum requirements for drywell entry and for general guidance. S p e cific, (6) During the purge, drywell atmosphere detailed site procedures and administrative samples shall be drawn from a number of controls must be developed by each utility to locations when the drywell oxygen analyzer i
meet the specific needs of each particular indicates an oxypn concentration of 16.5 physical plant and administrative setup.
percent or greater.
General Prncedure Drvwell Entrance Control Samples shall be analyzed for oxygen, Followine De-inertine hydrogen, ca'rbon monoxide, carbon dioxide and airborne activity.
(1) Inerting and de-inerting of the drywell OpfN shall be in conformance with applicable When the results of two successive samples y9 technical specifications. IN57# T 4,M4 taken at least one-half hour apart are found to be within the conditions in (2) Personnel access to :he drywell is normally Subsection 6.2.5.6(5), initial entry may be prohibited at all times when the drywell has authorized.
an oxygen. deficient atmosphere, unless an emergency condition arises in which case the (7) Criteria for entry are:
procedure outlined in subsection 6.2.5.6(8) should be followed.
(a) The initial entry will require a minimum of two (2) persons.
(3) The status of the drywell atmosphere @=11 be posted at the drywell entrance at all (b) Initial entry will require, in times, and the entrance locked, except when addition to normal protective clothing cleared for entry.
protective equipment consisting of 6 elf-contained arcathing apparatus (4) Suitable authorization, control and (such as Scott Air Pack), portable air recording procedures shall be established sampling and monitoring equipment, and and remain in effec throughout the entry portable radiation survey meters.
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oh'i ' Jo.2-6 { J 0,3 - 9 fy 3of3 f;V S / R T G, S, f, 3 Q During normal operation, nitrogern makeup and containment pressure c in accomplished using only the 50 na supply lines. The large valves (550 mm) closed and flow to the plant stack ventilation lines are is prevented by the rupture the containment through the overpressure protection line (350 sa) disk.
The following conditions assure that the large (550 mm) contsinnent purge and vent lines will be isolated following a LOCA:
all times during normal operation and will (1) The valves remain closed at the beginning and end of +>
only be opened for inerting or de-inerting at 3 $hvtdour rs, (2) The valves and piping provide redundancy such that no single failure can estage.
prevent isolation of the 06-4be purge and vent lines.In the even (3)
(4) The valves fail in the closed position. If solenoids is lost or the pneumatic pressure fails, the valves will close.
lySER T for 3 C &The sizing of the piping and rupture disk for overpressure protection is discussed in ier,gr gh 19E 2.8.1.
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ABWR nA6too^c i
. Standard Plant REV C j
response which describes additional tests to be ing or terminating vent usage.
2 conducted during the preoperational and/or startup phase.
Response
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The specific training requirements for reactor The capability to vent the ABWR reactor cool-g3 operators are discussed in Section 13 2 of the SRP ant system is provided by the safety relief valves and p s,3_f which is outside the scope of the ABWR Standard reactor coolant vent line as well as other-systems.
j Plant. See Table 1.91 for COL license information The capability of these systems and their satisfaction,7m 7 of item II.B.1is discussed below, fn5 requirements The additional tests specified in Appendix E of The ABWR design is provided with eighteen
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the BWROG response are contained within the power-operated relief valves which can be manually initial test program described in Chapter 14. See operated from the control room to vent the reactor g, specifically Subseetions 14.2.12.1.1 ( 3 ) ( a ),
pressure vessel. The point of connection to the main
/ 14.2.12.1.9(3)(j), and 14.2.12.1.44(3)(a) for the steam lines which exits near the top of the vessel to relevant testing.
these valves is such that accumulation of gases above that point in the vessel will not Lifect removal of IA.2.5 Reactor Coolant System Vents gases from the reador core region.
[II.B.1)
These power-operated relief valves satisfy the in-NRC Position tent of the NRC position. Information regarding the design, qualification, power source, etc., c,f these Each applicant and licensee shall install reactor valves is provided in Subsection 5.2.2.
coolant system (RCS) and reactor vessel head high point vents remotely operated from the control The BWR Owners' Group position is that the re-room. Although the purpose of the system is to vent quirement of single-failure criteria for prevention of noncondensible gases from the RCS which may in-inadvertent actuation of these valves, and the re-hibit core cooling during natural circulation, the quirement that power be removed during normal op-vents must not lead to an unacceptable increase in cration, are not applicable to BWR's. These dual-the probability of a loss-of-coolant accident (LOCA) purpose safety / relief valves serve an important pres-or a challenge to containment integrity. Since these sure relief function in mitigating the effects of trans-vents form a part of the reactor coolant pressun ients *.nd concurrently provide ASME code over-boundary, the design of the vents shall conform to pressure protection via their independent safety tbc requirements of Appendix A to 10 CFR Part 50, mode of operation. Therefore, the addition of a General Design Criteria. The vent system shall be second " block" valve to the vent lines would result in designed with sufficient redundancy that assures a a less safe design and a violation of the code.
Iow probability of inadvertent or irreversible ac-Moreover, the inadvertent opening of a relief valve in a BWR is a design-basis event and results in a
- tuation, controllable transient.
Each license shall provide the following infor-mation concerning the design and operation of the In addition to these automatic (or manual) relief high point vent system.
valves, the ABWR design includes various other means of high-point venting. Among these are:
(1) Submit a desription of the design, location, size, and power supply for the vent system along (1) Normally closed reactor vessel head vent valves, with results of analyses for loss-of-coolant operable from the control room, which discharge accidents initiated by a break in the vent pipe.
to the drywell. The reactor coolant vent line is The results of the ant. lyses should demonstrate located at the very top of the reactor vessel as compliance with acceptance criteria of 10 CFR shown in the nuclear boiler P&lD (Figure
/
50.46.
5.1-
. This 2-inch line contains two safety.
related Class IE motor-operated valves that are (2) Submit procedures and supporting analysis for operated from the control room. The location of operator use of the vents that also include the this line permits it to vent the entire reactor core information available to the operator for initiat-tA.2 3
^****f Imrnr is,g, g The COL Applicant will develop plant-
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i ABM nasiooxc Standard Plant nev c 1A.2.17 Instrument for Monitoring 1A 2.19 Review and Modify Procedures Accident Conditions [II.F.3]
for Removing Safety-Related Systems From Service [II.K.I(10)]
NRC Position j
NRC Position Provide instrumentation adequate for monitor-j ing plant conditions following an accident that in-Review and modify (as required) procedures for cludes core dama6e.
removing safety-related systems from service (and 1
restoring to service) to assure operability status is COL known. M& EAT /A S./f
,w, 3, f-2
Response
3 The ABWR Standard Plant is designed in accor-
Response
dance with Regulatory Guide 1.97, Revision 3. A de-tailed assessment of the Regulatory Guide, including See Subsection 1A.3.2 for COL license the list of instruments, is found in Section 7.5.
information requirements.
1A.2.18 Safety-Related Valve Position 1A.2.20 Describe Automatic and Manual Indication [II.K.1(5)]
Actions for Proper Functioning of Auxillary Heat Removal Systems When FW NRC Position System Not Operable [II.K.1(22)]
(1) Review all valve positions and positioning re-NRC Position quirements and positive controls and all related test and maintenance procedures to assure For boiling water reactors, describe the auto-proper ESF functioning, if required.
matic and manual actions necessary for proper func-tioning of the auxiliary heat removal systems that are (2) Verify that AFW valves are in open position.
used when the main feedwater system is not operable (see Bulletin 79 08, item 3).
Response
Response (1) The ABWR Standard Plant is equipped with status monitoring that satisfies the requirements lithe main feedwater system is not operable, a of Regulatory Guide 1.47. See Subsection 7.1.2 reactor scram will be automatically initiated when for detailed information on the status monitor-reactor water level falls to J evel 3. The operator can ing equipment and capabi3ty provided in the then manually initiate the RCIC system from the ABWR Standard Plant design.
main control room, or the system will be automati-d (41.affliedef cally initiated as hereinafter described. Reactor In addition to the status monitoringplant spe-water level will conttaue to decrease due to boli-off cific procedures (see Subsection IA.3.2) will until the low low icvel setpoint, Level 2, is reached.
assure that independent verification of system At this point, the reactor core isolation cooling
?
line ups is applied to. valve and electrical (RCIC) system will be automatically initiated to sup-line ups for a5 safety-related equipment, to ply taakeup water to the RPV. This system will D
surveilla edures,Mo restoration continue automatic injection until the reactor water 70<$I'l followi Through these proce-level reaches Level 8, at which time the RCIC steam ures,a wi i be required for the supply valve is closed.
performance of surveillance tests and mainte-ance, including equipment removal from service In the nonaccident case, the RCIC systern is nor-and return to service.
mally the only makeap system utilized to furnish sub-sequent makeup water to the RPV. When level (2) This requirement is not applicable to the reaches Level 2 again due to loss of inventory ABWR. It applies only to Babcock & Wilcox through the main steam relief valves or to the main designed reactors.
condenser, the RCIC system automatically restarts as described in Subsection 1A.2.22. This system then BMo comply wth IE B,Ilefik gpg, IA.2-11 Amendment Il MN /#M The COL applicant must verify the operability of safety-related systems after performing maintenance or tests as part of the test to restore a system to service.
~
7 ABM u s1, c Standard Plant
,ey c and engmeered safeguards systems.
high water-level signal. The HPCI system will restart on low water level but the RCIC system will (4) Fuel-zone, water level range: This range used not. The RCIC system is a low flow system when for its RPV taps the elevation above the main compared to the HPCI system. The initiation levels steam outlet nozzle and the taps just above the of the HPCI and RCIC system should be separated internal recirculation pump (RIP) deck. The so that the RCIC system initiates at a higher water zero of the instrument is the bottom of the ac-level than the HPCI system. Further, the initiation 3
tive fuel and the instruments are calibrated to be logic of the RCIC system should be modified so that accurate at 0 psig and saturated condition. The the RCIC system will restert on low water level.
water level measurement design is the conden.
These changes have the potential to reduce the sate reference type,is not density compensated, number of challenges to the HPCI system and could and uses differential pressure devices as its pri-result in less stress on the vessel from cold water mary elements. These instruments provide injection. Analyses should be performed to evaluate input to water-levelindication only, these chann The analysis should be submitted to i
the NRC staff and changes should be implemented if There are common condensate reference cham-justified by the snalysis.
bets for the narrow range; wide-range; and fuel-zone, water-level ranges.
Response
The elevation drop from RPV penetration to the The ABWR Standard Plant design is consistent t
drywell penetration is unifwm for the narrow range with this position. The high pressure core flooder and wide ruge water-level instrument lines in order (HPCF) system is initiated at Level 1 1/2, and the I
to minimize the change in water level with changes RCIC system is initiared at Level 2. At level 8, the in drywell temperaturei injection valves for the HPCF and the RCIC steam supply and injection valves wi!! automatically close in Reactor water.levelinstrumentation that ini-order to prevent water from entering the main steam tiates safety systems and engineered safeguards is lines.
shown in Figure 113.
In the unlikely event that a subsequent low level i
1A.2.21.1 Fallure of PORY or Safety to recurs, the RCIC steam supply and injection valves Close [II.K.3.(3)]
will automatraDy reopen to allow continued flooding i
of the vessel. The HPCF injection valves will also l
NRC Position automatically reopen unless the operator previously closed them manually. Refer to Subsections COL a*3'h3 Assure that any failure of a PORV or safety 7.3.1.1.1.1 (HPCF) and 7.3.1.1.1.3 (RCIC) for valve to close will be reported to the NRC promptly, additional details regarding system initiation and All challenges to the PORVs or safety valves should isolation logic.
l be documented in the annual report. This j
requirement is to be met before fuelload.
1A.2.23 Modify Break. Detection Logie to Prevent Spurious Isolation of HPCI
~
And RCIC Systems [II.K.3(15)]
Response
See Subseidios 1A.3.4 for COL license NRC Positlen 6
information E 4 e i
The high-pressure coolant injecuon (HPCI) and C
1A.2.22 Separation of HPCI AND RCIC reactor core isolation cooling (RCIC) systems use l
System Initiation Levels [II.K.3(13)]
differential pressure sensors on elbow :aps in the l'
steam lines to their turbine drives to detect and f
isolate pipe breaks in the systems. The pipe-break-NRC Position detection circuitry has resulted in spurious isolation Currently, the reactor core isolation cooling of the HPCI and RCIC systems due to the pressure l
(RCIC) system and the high-pressure coolant injec-spike which accompanies startup of the systems. The tion (HPCI) systems both initiate on the same low-pipe-break-detection circuitry should x modified to water-level signal and both isolate on the same 1A.2-13 f
Amendment D 6
23A6100AC Standard ?'
arv c differential pressure signals which isolate the RCIC (5) Earlier initiation of ECC systems,
' turbine are processed through the leak detection and isolation system (f DS). Spurious trips are avoided (6) Heat removal through cmergency condensers,
~
because the PCIC has a bypass start system con-trolled by valves F037 and F045 (see Figure 3.4-8, (7) Offset valve setpoints to open fewer valves per RCIC P&lD).
challenge, On receipt of RCIC start signals, bypass valve (8) Installation of additional relief valves with a F045 opens to pressurize the line downstream and block or isolation-valve feature to eliminate accelerate the turbine. The bypass line via F045 is opening of the safety / relief valves (SRV's),
small (1-inch) and naturally limits the initial flow consistent with the ASME Code, surge such that a differential pressure spike in the upstream pipe will not occur.
(9) Increasing the high steam line flow serpoint for main steam line isolation valve (MSIV) closure, After a predetermined delay (approximately 510 seconds), steam supply valve F037 opens to ad-(10) Lowering the pressure setpoint for MSIV Clo-mit full steam flow to the turbine. At this stage, the
- sure, M
line downstream is already pressurized. Thus, it is 90J,/"/ highly unlikely that a differential pressure spike (11) Reducing the testing frequency of the MSIV's, could occur during any phase of the normal start up process. Thd /racris /s As/r/ M f*4tec/fon (12) More stringent nive leakage criteria, and
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IA.2.24 Reduction of Challenges and (13) Early remove W leaking valves.
Falleres of Relief Valves - Feasibility Study and Systern Modification An investigation of the feasibility and constraints
[II.h'.3(16)]
of reducing challenges to the relief valves by use of the aforementioned methods should be conducted.
NRC Position Other methods should also be included in the feasi-bility study. Those changes which are shown to The record of relief-valve failures to close for all reduce relief-valve challenges without compromising boiling water reactors (BWRs) in the past 3 years of the performance of the relief valves or other systems plant operation is appproximately 30 in 73 reactor-should be implemented. Challenges to the relief years (0.41 failures per reactor-year). This has dem-valves should be reduced substantially (by an orJer onstrated that the failure of a relief valve to close of magnitude).
would be the most likely cause of a small-break '. ass-of-coolant accident (LOCA). The high failure rate is
Response
the result of a high relief-valve challenge rate and a relatively high failure rate per challenge (0.16 fail-General Electric and the BWR Owners' Group l ures per challenge). Typically, five valves are chal-reponded to this requirement in Reference 6. This lenged in each event. This results in an equivalent response, which was based on a revi w of existing failure rate per challenge of 0.03. The challenge and operating information on the chall ' e rate of relief failure rates can be reduced in the followmg ways:
valves, concluded that the BWR/ product line had already achieved the ' order of magnatode' level of (1) Additional andeipatory scram on loss of feedwa-reduction in SRV chat![ge rate. The ABWR relief p
valve system also has similar design features which
- tre, also r.doce the SRV challenge rate. With regard to (2) Revised relief-valve actuation setpoints, inadvertently opened relief valves (IORV), the BWR/6 p t design evaluated for the Owners' (3) Increased emergency core cooling (ECC) flow, Group r ort reflected a reduced leveljf'lORC compare with previous design because of of (4) Lower operating pressures, W-N Amendment 25
y l,0l L ce L 90* % l ~ 5' gg Standard Plant nev c es 4 h eeraea, u m, g,,,,
1A3.25 Report on Outages of Emergency See Subseetion 1A.3.5 forqnterf ase-Core Cooling Systems Licensee Repott kequirements and Proposed Technical Specification Changes [II.K.3(17)]
1A.2.26 Modification of Automatic De.
Pressurization System Imgic - Feasi-l NRC Position bility for Increased Diversity for Some Event Sequences [II.K.3(18)]
Several components of the emergency core
[
cooling (ECC) systems are permitted by technical NRC Position specifications to have substantial outage times (e.g.,
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for one diesel-generator; 14 days for the The automatic depressurization system (ADS)
HPCI system). In addition, there are no cumulative actuation logic should be modified to eliminate the outage time limitations for ECC systems. Licensees need for manual actuation to assure adequate core should submit a report detailing outage dates and cooling. A feasibi'ity and risk assessment study is lengths of outages for all ECC systems for the last 5 required to deterraine the optimum approach. One i
years of operativa. The report should also include possible scheme that should be considered is ADS the causes of the outages (i.e., ontroller failure, actuation on low reactor-vessel water level provided spurious isolation).
no high. pressure coolant injection (HPCI) or high-pressure coolant ;ystem (HPCS) flow exists and a l
i Clarification low-pressure emergency core cooling (ECC) system is running. This logic would complement. act The present technical specifications contain replace, the existing ADS actuation logic.
limits on allowable outage times for ECC systems and components. However, there are no cumulative
Response
outage time limitations on these same systems. It is possible that ECC equipment could meet present An 8 minute high drywell pressure bypass timer technical specification requirements but have a high has been added to the ADS initiation logic to address unavailability because of frequent outages within the TMI action item II.K.3.18. This timer will initiate on allowable technical specifications, a Low Water Level 1 signal. When it times out, it bypasses the need for a high drywell signal to initiate The licensees should submit a report detailing the standard ADS initiation logic.
outage dates and length of outages for all ECC systems for the last 5 years of operation, including For all LOCAs inside the containment, a high causes of the outages. This report will provide the -
drywell signal will be present and ADS will actuate staff with a quantification of historical unreliability 29 seconds after a Low Water Level 1 signalis 3
due to test and maintenance outages, which will be reached. All LOCAs outside the containment requirements in the technical specifications.
become rapidly isolated and any one of the three high pressure ECCS can control the water level The Based on the above guidance and clarification, a high drywell pressure bypass ti:ner in the ADS detailed report should be submitted. The report initiation logic will only affect the LOCA response if should contain (1) outage dates and duration of all high pressure ECCS fail followmg a break outside outages;(2) causes of the outage; (3) ECC systems the containmen' For this case the ADS will or components involved in the outage; and (4) automatically initiate within 509 seconds (8 minute corrective act',
s.
Tests and maintenance timer plus 29 second standard ADS logic delay) outages should, included in the above listings following alow Water Level 1 sigant which are to cover the last 5 years of operation. The licensee sl ould propose changes to improve the availability of ECC equipment,if needed.
Applicants for an operating license shall establish a plan to meet these requirements.
Response
1A.315 Amendment 25 i
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/f id/1 ABWR 2mioarc Standard Plant nevc 1A5 COLLICENSEINFORMATION changes, proposed or implemented, deemed appropriate, to improve the availability of the emer.
1A3.1 Emergency Procedures and gency core cooling equipment. (See Subsection Emergency Procedures Training Program 1A.2.2.5).
Emergency procedures, developed from the
/A.1 b 8ee d vres for feged y Veg h N emergency procedures guidelines, shall be provided and implemented prior to fuelloading. (See Subsec-frvced'arr$ flJ/[M dfWolo
[#r bd 0** 1^ 21)*
0jwahv'.s UJe 0{ llf fe Y W Y41 E.
1A3.2 Review and Modify Procedures for
[feefulfcc//or/42,5)
Removing Safety-Related Systems From Service Procedures shall be reviewed and modified (as required) for removing safety-related systems from service (and restoring to service) to assure operabil.
l ity status is known. (See Subsections 1A.2.18 and 19).
1A.3.3 In Plant Radiation Monitorf.ng Equipment and training procedures shallbe provided for accurately determining the airborne io-dine concentration in areas within the facility where l plant personnel may be present during the accident.
(See Subsecnon 1A.2.35).
1A.3.4 Reporting Failures of Reactor System Relief Valves Failures of reactor system relief valves sht!! be reported in the ual report to the NRC. (See Sul>-
t/
seaion L1).
1A.3.5 Report on ECCS Ontages Starting from the date of commercial opera-tions, an annua, report shou'd be submitted wi+.h in-cludes inomnce of emergency core coohng system un-availability because of component failure, mainte-nance outage ced or planned), or testing, the following
- shallbe colleced:
(1) Outage (2) Duration outage 90.5.H (3) Cauw of outage (4) Emergency core cooling system or component involved (5) Corrective aedon taken The above informanon shall be assensbled into a report, which will also include a discussion of any lA 31 Amendment 23
W AO.3 *A hf 1 k A}
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33A6100AS 1
Standard Plant ne x
Response
valves are opened during power operation. These are air-operated valves with rapid closure times, present-j This requirement is not applicable to the ABWR.
ing little opportunity for substantial releases from the i
It applies only to PWR-type reactors.
PCV in the event of a transient requiring containment isolation. Note that under the technical specifications, 19A.2.26 Isolation Dependability [ Item (2) containment inerting and purging with the larger (xiv)]
ventilation lines is permitted during power operation above 15% for limited periods at either end of the NRC Position operating cycle. The process of purging the contain-Provide containment isolation systems that:
ment with air also serves to remove any potential
[II.E.4.2]
activity for ALARA considerations prior to actual personnel entry into the PCV.
(A) Ensure all non-essential systems are isolated automatically by the containment isolation The large ventilation valves will be tested regularly
- system, and after any valve maintenance to assure that closing egt times are within the limits assured in the radiological pg,3 2 (B) For each non-essential penetration (except in-design basis. gee Subsection 19A33 for COL license strument lines) have two isolation barriers in information T 7g fpg,4, ye pf or fje
- series, toservice trSt frotraM de failed in Sdsrefin 29, 19A.2.28 Design Evarustor [ Item (2) (xvi)]
(C) Do not result in reopening of the containment isolation valves on resetting of the isolation NRC Position
- signal, Establish a design criterion for the allowable (D) Utilize a containment set point pressure for ini-number of actuation cycles of the emergency core tiating containment isolation as low as is com-cooling system and reactor protection system consis-patible with normal operation, tent with the expected occurrence rates of severe over cooling events (considering both anticipated tran-(E) include automatic closing on a high radiation sients and accidents). (Applicable to B&W designs signal for all systems that provide a path to the only.) [II.E.5.1]
ensrons.
Response
Response This requirement is not applicable to the ABWR. It This item is addressed in Subsection IA.2.14.
applies only to PWR-type (B&W designed) reactors.
19A.2.27 Purging [ Item (2) (xv)]
19A.2.29 Additional Accident Monitoring Instrumentation [ Item (2) (xvil))
NRC Position NRC Position Provide a capability for containment purg-ing/ venting designed to minimize the purging time Provide instrumentation to measure, record and consistent with ALARA principles for occupational readout in the control room: (A) contamment pres-exposure. Provide and demonstrate high assurance sure, (B) containment water level, (C) containment that the purge system w!!! reliably irotate under acci-hydrogen concentration, (D) containment radiation dent conditions. [II.E.4.4) intensity (high level), and (E) noble gas effluents at all potential, accident release points. Provide for con-
Response
tinuous sampling of radioactive iodines and particu-lates in gaseous effluents from all potential accident The ABWR primary containment vessel (PCV) op-release points, and for onsite capabihty to analyze and crates with an inert atn ame During normal op-measure these samples. [II.F.1]
cration, all large valves in containment ventilation lines are closed. Only small,2", nitrogen-makeup 19A.2-6 Amendment 23
Co t-94 3 -1 fy $ o/ y ABM n462oare -
Stanslarti Plant nEv e f
3.9.7' COL License Information Subsection 3.9.3.1.)
i 3.9.7.1 Reacter Internals Vibration Analysis, 3.9.7.3 Pump and Valve Inservice Testing Measurement and laspection Program Program Col-l The first COL applicant will provide, at COL a'pplicants will provide a plan for the ;g,3 y r
the time of application, the results of the detailed pump and valve inservice testing and j
vibration assessment program for the ABWR inspection program. This plan will l
prototype internals. These results will include the following information specified in Regulatory (1) Include baseline pre-service testing to l
Guide 1.20.
support the periodic in-service testing of.
the components required by technical
[
R. G.1.20 Subiect specifications. Provisions are included to disassemble and inspect the pump, check C.2.1 Vibration Analysis valves, and MOVs within the Code and Program safety-related classification as necessary, C.2.2 Vibration Measurement depending oa test resutts. (See Program Subseetions 3.9.6, 3.9.6.1, 3.9.6.2.1 and C.2.3 Inspection Program 3.9.6.2.2)
C.2.4 Documentation of Results (2) Provide a study to determine the optimal.
l frequency for valve stroking during NRC review and approval of the above inservice testing. (See Subsection i
information on the first COL applicant's docket 3.9.6.2.2) will complete the vibration assessment program requirements for prototype reactor internals.
(3) Address the concerns and issues identified in Generic Letter 89-10; specifically the In addition to the information tabulated method of assessment of the loads, the above, the first COL applicant will provide the method of sizing the actuators, and the information on the schedules in accordance with setting of the torque and limit switches.
I the applicable portions of position C.3 of (See Subsection 3.9.6.2.2) j Regulatory Guide 1.20 for non-prototype internals.
3.9.7.4 Asdit of Design Specificatlos ans Design Repoets Subsequent COL applicants need only provide the information on the schedules in accordance COL applicants will make available to the with the applicable portions of position C.3 of NRC staff design specification and design Regulatory Guide 1.20 for non prototype reports required by ASME Code for vessels, f
I internals. (See Subsection 3.9.2.4).
pumps, valves and piping systems for the purpose of audit. (See Subsection 3.9.3.1) j l
3.9.7.2 ASME Class 2 w 3 or Quality Groep D Components with 48 Year DesignIme 3.9.8 References i
COL appnamate will identify ASME Class 2 1.
BWR Fuel Channel Mechanical Design and or 3 or Quality Group D camponents that are Deflection, NEDE-21354-P, September 1976.
l
)
subjected to cyclic loadings, including operating vibration loads and thermal transientr, effects,.
2.
BWR/6 Fuel Assembly Evaluation of Combined of a magnitude and/or duration so severe the 6G Safe Shutdown Earthquake (SSE) an'd l
year design life can not be assured by required Loss-of-Coolant Accident (LOCA) Loadingr, l
Code calculations and, if similar designs have NEDE-21175-P, November 1976.
j i
not already been evaluated, either provide an appropriate analysis to demonstrate the required 3.
NEDE-24057-P (Class III) and NEDE-24057 design life or provide designs to mitigate the (Class I) Assessment of Reactor Internals.
j magnitude or duration of the cyclic loads. (See Vibration in BWR/4 and BWR/S Plants, 3945 Amendment 3 I
i
- " 3 -1 ty r of at ABWR 2mm^e Standard Plant Rev B Table 3.9-8 (Continued)
IN-SERVICE TESTING SAFETY-RELATED PUMPS AND VALVES T22 Standby Gas Treatment System Valves Safety Code Valve Test Test SSAR Class Cat. Fune. Para Freq. Fig.
No. Qty Description (h)(I)
(a)
(c)
(d)
(c)
(f)
(g)
F012 2 Fdter train DOP samplingline valve 3
B P
El 651(2,3) downstream of after HEPA F014 2 STGS sample line valve 3
B P
El 65-1(2,3)
F015 2 PRM discharge to stack valve 3
B P
El 63-1(2,3)
F500 2 Fdter unit vent line valve 3
B P
El 651(2.3)
F501 2 Fdter unit drain line valve 3
B P
El 651(2,3)
F504. 2 Futer unit vent line valve 3
B P
El 651(2,3) i F505 2 Exhaust fan vent line valve 3
B P
El 651(2,3)
F506 2 Futer train vent line valve 3
B P
El 651(2,3)
F507 2 Fdter train vent line valve 3
B P
El 651(2,3)
F508 2 Futer train vent line valve 3
B P
El 651(2,3)
F509 2 Fdter train vent line valve 3
B P
El 651(2,3)
F510 2 Fdter train vent line valve 3
B P
El 651(2,3)
F511 2 Exhaust stack drain line valve 3
B P
El 651(2,3)
F 00 2 Futer unit demister dp instrument line valve 3
B P
El 651(2,3)
F701 2 Futer unit demister dp instrument line valve 3
B P
El 651(2,3)
F705 2 Futer train prefilter dp instrument line valve 3
B P
El 651(2,3)
F706 2 Fdter train prefilterdp instrument line valve 3
B P
El 651(2,3)
F707 2 Fdter train preHEPA dp instrwnent line vabr 3 B
P El 651(2,3)
F703 2 Fdter train preHEPA dp instrument line valve 3 B
P El 651(2,3)
F709 2 Fliter train charcoal adsorber dp inst. line viv 3 B
P El 651(2,3)
F710 2 Fdter train charcoal adsorber dp inst line viv 3
B P
El 651(2,3) r F711 2 Fdter train after HEPA dp inst line ulve 3
B P
El 651(2,3)
F712 2 Fdter train after HEPA dp icst line valve 3
B P
El 651(2,3)
F713 2 Fdter train exhaust flow instrument line valve 3 B
P El 6.5-1(2,3)
F714 2 Fdter train exhaust flow instrument line valve 3 B
P El 651(2,3)
T31 Atmospheric Control System Valves F001 1 N2 supply line from Reactor Building HVAC 2 A
I,A LP 2 yrs 6.2-39(1)
S 3mo F002 1 N2 supply line to drywellinboard cont-2 A
I,A 1,P 2 yrs 6.2 39(1) aimment isoaltion valve S
3 mo F003 1 N2 supply line to werwc!! inboard coat-2 A
1,A 1,P 2 yrs 6.2-39(1) ainment iscaltion valve S
3 mo V F004 1 Containment atmosphere exhaust line from 2
A I,A 1,P 2 yrs 6.2-39(1)
S 3 mo drywellisoaltion valve F005 1 Drywell atmosphere exhaust line valve 2
A I,A I,P 2 yrs 6.2-39(1)
S 3 mo i
T31-F004 bypass line y F006 1 Containment atmosphere exhaust line form 2
A 1,A 1,P 2 yrs 6.2 39(1) i wetwellisolation valve S
3 mo
/ F007 1 Wetwell overpressureline vahr 2
A P
1,P 2 yrs 6.2-39(1)
M $83 Amendment 24
co1 po,3 - 2 Py4cQ gQ e
zware Standard Plant Rev B Table 3.9-8 (Continued)
IN-SERVICE TESTING SAFETY-RELATED PUMPS AND VALVES T31 Atmospheric Control System Valves Safety Code Valve Test Test SSAR Class Cat. Fune. Para Freq. Fig.
No. Qty Description (h)(I)
(a)
(c)
(J)
(c)
(f)
(g)
V F008 1 Containment atmosphere exhaust line 2
A I,A 1,P 2 yrs 6.2-39(1) to SGTS S
3 mo
/
F009 1 Containment atmosphere exhaust line to 2
A I.A I,P 2 yrs 6.2-39(1)
R/B HVAC S
3 mo V
F010 1 Drywc!! overpressure line valve 2
A P
1,P 2 yrs 6.2-39(1)
F025 1 N2 supply line frorn K 5 outboard cont-2 A
I,A 1,P 2 yrs 6.2-39(1) aimnent isolation valve S
3 mo F039 1 N2 supply line from K-5 outboard cont-2 A
I,A L,P 2 yrs 6.2-39(1) ainment isolation valve S
3 mo F040 1 N2 supply line from K-5 to drywell inboard 2
A 1,A I,P 2 yrs 6.2-39(1) isolation valve S
3 mo F041 1 N2 supply line from K-5 to werwell inboard 2
A 1,A 1,P 2 yrs 6.2-39(1) isolation valve S
3 mo F044 8 Drywell/wetwell vacuum breaker valve 2
C A
P RO 6.2-39(2)
R E3 F050 1 N2 supply line to drywell test line valve 2
B P
El 6.2-39(1)
F051 1 Containment atmosphere exhaust line test 2
B P
El 6.2-39(1) line valve F054 1 Drywell personnel air lock hatch test 2
B P
El 6.2-39(2) line valve F055 1 N2 supplyline from test line valve 2
B P
El 6.2-39(1)
F056 1 Wetwell personnel air lock hatch test 2
B P
El 6.2-39(2) line valve F703 1 N2 supplyline to drywellFE upstream 2
B P
El 6.2-39(1) instrument line F701 1 N2 supply line to drywell FE downstream 2
B P
El 6.2-39(1) instrument line F702 1 N2 supplyline to werwellFE upstream 2
B P
El 6.2-39(1) mstrument line F703 1 N2 supply line to wetwell FE dommtream 2
B P
El 6.2-39(1) instrument line F720 8 DW/WW vacuum breaker valve N2 supply 2
A I,P L
RO 6.2-39(2)
Nas
- alatina valve
=
F730 1 Drpuell pressure instrument line isolaton 2
B P
El 6.2-39(2) valve F731 1 Drywell pressure instrument line solenoid 2
A I,P 1,P RO 6.2-39(2) isolation valve F732 2 Drywell pressure instrument line valve 2
B P
El 6.2-39(2)
F733 2 Drywell pressure instrument line solenoid 2
A 1,P 1,P RO 6.2-39(2) isolation valve F734 4 Drywell pressure instrument line for NBS 2
B P
El 6.2-39(2) valve 3.95823 Amendment 23
.