ML20057E866
| ML20057E866 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 10/05/1993 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9310130279 | |
| Download: ML20057E866 (4) | |
Text
.
GENucle rEnergy i
GeneralElettnc Company l
115 Curtner Avenue. San hse. CA 95125 r
l October 5,1993 Docket No.52-001 t
k y
f
?
t t
Chet Posiusny, Senior Project Manager Standardization Project Directorate.
}
Associate Directorate for Advanced Reactors i
i and License Renewal Office of the Nuclear Reactor Regulation j
Subject:
Submittal Supporting Accelerated ABWR Schedule - Plant Systems i
Branch Items 1
r
Dear Chet:
i Enclosed are SSAR markups covering: (1) classification of the spent fuel pool line, (2) appropriate figure for crack leak rate, and (3) inlet temperature protection for the Fuel Pool l
Cooling System filter demineralizers. These items were discussed with Plant Systems l
Branch.
F Please provide a copy of this transmittal to Butch Burton.
f Sincerely, f
b i
J ek Fox Advanced Reactor Programs I
cc:
Alan Beard (GE) i Norman Fletcher (DOE) t I
i JIE2rd I
120059 j
9310130279.931005 M
PDR ADOCK 05200001-h q
I A.
PDR j3
-i i
23A6100 Rev.1 ABWR standardsarety Analysis Report 4
l Table 3.2-1 Classification Summary (Continued)
Quality Quality Assur-Group ance Safety Classi-Require-Seismic f
d ment
- Category Notes b Location
- fication Principal Component
- Class FG Fuel Storage Equipment se s us.p~sE 1.
Fuelgstorade t'acks-N SC E
I (bb) new and spent 2.
Defective fuel N
SC E
(bb)
-w gy ca.hswe,-
E I
I s I4 Y b d aal P
N Sc 3.
bw.-
(bb)
F7 Under-Vessel Servicing N
SC E
Equipment E
F8 CRD Maintenance Facility N
SC F9 Internal Pump Maintenance N
SC E
1 Facility F10 Fuel Cask Cleaning Facility N
SC-E
)
F11 Plant Start-up Test N
M E
Equipment 1
1 F12 Inservice inspection N
M E
Equipment l
G1 Reactor Water Cleanup System 1.
Vessels including N
SC C
E supports (filter /
demineralizer) 2.
Regenerative heat N
SC C
E exchangers including supports carrying reactor water 3.
Cleanup recirculation N
SC C
E pump, motors Notes and footnotes are listed on pages 3.2-53 through 3.2-60 Classification of Structures. Components, and Systems - Amendment 31 3.2-30 i
A
~
23A6100 Rev. 2 ABWR standardSafety Analysis Report steam SRVs, discharging to the suppression pool,is monitored by temperature sensors mounted in thermowells in the individual SRV exhaust lines. The thermowells are located several feet from the valve bodies so as to prevent false indication. These temperature sensors transmit signals to the control room for monitoring. Any temperature increase detected by these sensors, that is above the ambient temperatures, indicates SRV leakage.
5.2.5.5 Unidentified Leakage inside the Drywell 5.2.5.5.1 Unidentified Leakage Rate The unidentified leakage rate is the portion of the total leakage rate received in the drwell sumps that is not identified as previously described. A threat of significant compromise to the nuclear system process barrier exists if the barrier contains a crack that is large enough to propagate rapidly (critical crack length). The unidentified leakage rate limit must be low because of the possibility that most of the unidentified leakage rate might be emitted from a single crack in the nuclear system process bamer.
An allowance for leakage that does not compromise barrier integrity and is not identifiable is established for normal plant operation.
The unidentified leakage rate limit is established at 19 liters / min to allow time for corrective action before the process barrier could be significantly compromised. This unidentified leakage rate is a small fraction of the calculated flow from a critical crack in a primarv sptem pipe (Appendix SE).
5.2.5.5.2 Margins of Safety W - 2. ?
The margins of safety for a detec le flaw to reach critical size are presented in Subsection 5.2.5.5.3. Figure 5.2-8 shows general relationships between crack length, leak rate, stress, and line size using mathematical models.
5.2.5.5.3 Criteria to Evaluate the Adequacy and Margin of Leak Detection System For process lines that are normally open, there are at least two different methods of I
detecting abnormalleakage from each system comprising the nuclear system process barrier, located both inside the primary containment (drywell) and external to the drywell,in the reactor building the steam tunnel and the turbine building (Tables 5.2-6 and 5.2-7). The instrumentation is designed so it can be set to provide alarms at j
established leakage rate limits and isolate the affected system if necessary. The alarm j
points are determined analvtically or based on measurements of appropnate j
parameters made during startup and preoperational tests.
The unidentified leakage rate limit is haced,vrith ar. adequate margin for contingencies, on the crack size large enough to propagate rapidly.
1 5.2-46 Integnty of Reactor Coolant Pressure Boundary - Amendment 32
y 23A6100 Rav.1 ABWR standardsoreryAnso sisneporr
)
r u
Penetrations through shielding walls are located so as not to compromise radiation shielding requirements.
The filter-demineralizers are controlled from a local panel. A differential pressure and l
conductivity instruments provided for each filter-demineralizer unit indicate when backwash is required. Suitable alarms, differential pressure indicators and flow j
. indicators monitor the condition of the filter-demineralizers.
l l
Svstem instrumentation is provided for both automatic and remote-manual operations.
l A low-low level switch stops the circulating pumps when the fuel pool drain tank reserve.
j capacity is reduced to the volume that can be pumped in approximately one minute with one pump at rated capacity (250 m /hr). A level switch is pr vided in the fuel pool f
3 to alarm locally and in control room on high and low level emperature elementy oR a
sw
@ prmided toiiispla.gpoo temperatur 'n the main control room,in addition, leakage.
l flow detectors in the pool drains and p ol liners are provided and alarmed in the nvid M.d-4aymbv4 km O l
control room.
A M oe lovv-m e h p 3 l
The circulating pumps are controlled from the control room and a local panel. Pump low suction pressure automatically shuts off the pumps. A pump low discharge pressure l
alarm is indicated in the control room and on the local panel. The circulating pump i
motors are powered from the normal offsite sources backed by the combustion turbine l
generators.
The water level in the spent-fuel storage pool is maintained at a height sufficient to i
provide shielding for normal building occupancy. Radioactive particulates removed l
from the fuel pool are collected in filter-demineralizer units which are located in l
shielded cells. For these reasons, the exposure of plant personnel to radiation from the FPC System is minimal. Further details of radiological considerations for this system are provided.in Chapter 12.
The circulation patterns within the reactor well and spent-fuel storage pool are established by placing the diffusers and skimmers so that particles dislodged during refueling operations are swept away from the work area and out of the pools.
Check valves prevent the pool from siphoning in the event of a pipe rupture.
Heat from pool evaporation is handled by the building ventilation system. Makeup
]
water is provided through a remote-operated valve.
1 9.1.3.3 Safety Evalustion The maximum possible heat load for the FPC System upon closure of the fuel gates (21 days) is the decay heat of the full core load of fuel at the end of the fuel cycle plus j
the remaining decay heat of the spent fuel discharged at previous refuelings upon closure of the fuel gates; the maximum capacity of the spent-fuel storage pool is 270%
Fuel Storage and Handhng - Amendment 31 9.1-13
,,,