ML20237E786
| ML20237E786 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 12/17/1987 |
| From: | Johnson I COMMONWEALTH EDISON CO. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 3979K, NUDOCS 8712290188 | |
| Download: ML20237E786 (15) | |
Text
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.'N Commonwealth Edison l
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) One Fir:: National Plaza. Chicago, Illinois
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7 ~7 Address Reply to: Post Office Box 76'I
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,/ Chicago, Illinois 60690 0767 December 17, 1987 hr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Quad Cities Station Units 1 and 2 Conformance to NRC Regulatory Guide 1.97, Rev. 2 NRC Docket Nos. 50-254 and 50-265
Reference:
Letter from T. Ross to L. D. Butterfield dated September 3, 1987.
Dear Mr. Murley:
The referenced letter provided the Technical Evaluation Report (TER) on Quad Cities Station conformance with Regulatory Guide 1.97, Revision 2, and requested a response to the associated open items. Our responses are provided in the enclosure for each of the items identified in the TER, including any additional information and comments as requested.
No new modifications have been identified as a result of our review.
If there are any further questions regarding this matter, please contact this office.
Very yours I. M. Johnson Nuclear Licensing Adm nistrator 1m Attachment cc:
T. Ross - NRR Regional Administrator - RIII Resident Inspector - Quad Cities p0Y I
8712290188 871217 PDR ADOCK 05000254 gg PDR 3979K P
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1 ATTACHMENT A COMMONWEALTH EDISON COMPANY RESPONSE TO NRC TECHNICAL EVALUATION REPORT NO. EGG-NTA-7761 CONFORMANCE OF QUAD CITIES STATION, UNITS 1 & 2 TO REGULATORY GUIDE 1.97, REVISION 2
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Attachment A 1 of 13 RESPONSE TO NRC REG. GUIDE 1.97, REVISION 2 CONCERNS QUAD CITIES STATION, UNITS 1& 2
. VARIABLE NO. (REF. 1) :
01 NAME:
Neutron Flux TYPE / CATEGORY:
B/l NRC CONCLUSION (REF. 2):
The existing instrumentation is acceptable until Category 1 instrumentation is developed and installed (Section 3. 3.1).
UTILITY RESPONSE:
The design criteria and requirements for the existing instru-mentation are those of the original system design, as provided
- by General Electric Company.
The updated FSAR Figure 7.4.3:3 pg 7.4.3-3 is a chart which shows the relationship of SRM count rate to actual core power level.
Even though the SRMs are fully withdrawn during startup to prevent rod. block, the SRM readings can be correlated to the actual power level based on the attenuation factor of the surrounding materials and the known neutron leakage factors for this BWR design.
The attenuated power levels at the SRM (when fully withdrawn) change directly as the core power level changes.
A period meter is provided which indicates the e-folding time of the exponentially changing power level.
Therefore, this meter is capable of showing power level changes down to minimum sensitivity of the detector even when the detector is fully out of the core.
This gives the operator information on the direction of power changes.
Two SRM channels are powered from 24V/48V Bus lA(2A); the other two channels are powered from 24V/48V Bus 1B(2B).
The sources of power to the SRM drive system are Buses 18-1 and 19-1 (28-1 and 29-1); two SRM drives per bus.
The probability that all four SRM drives would lose power is remote.
It is likely that at least one SRM drive will operate.
The intermediate range (IRM) power level instrumentation may be relied upon as a backup.
These eight detectors (similar to the SRMs) are capable of correlated power level down to about 5X10-4%
j when fully inserted.
The IRMs are also powered from the same buses as the SRMs.
The probability of at least one of the eight operating is very high, as shown in GE Company studies (Reference 4).
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Attachment A 2 of 13 RESPONSE TO'NRC
. REC. GUIDE 1.97, REVISION 2 CONCERNS l
QUAD CITIES STATION, UNITS 1& 2 VARIABLE NO. (REF. 1):
01 (Cont.)
If needed to monitor increasing power levels, the eight APRM channels may ~ be usable since they come on scale at less than 1% of full power'and overlap the.IRMs through approximately
.30% of full power when the IRMs reach their upper range limit.
However, the need'for post-accident monitoring with APRMLis remote due to.the strong negative' reactivity available from control rods, the Doppler coefficient, and the moderator void coefficient..Under the conditions of a design basis LOCA event, the scram system is assumed to operate properly.
A scram can be verified by diverse means, such as:
a)-
indication of scram relay operating b) scram valve position indication c)
CRD scram accumulator low pressure indication d) scram discharge volume high level alarms
~
e) indication of expected responses; i.e.,
makeup to the vessel, pressure decay,' torus temperature. rise, etc.
Redundant indications powered from reliable power sources and diverse indications assure that the operator can monitor the reactivity status of the reactor.
When proven SRM (or wider range) instrumentation is available for continuous incore use, CECO will consider that option for imple-mentation at Quad Cities.
However, existing-instrumentation utilized as Category 3 is considered sufficient to monitor this variable.
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3~of'13 RESPONSF TO NRC
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REG. GUIDE 1.97, REVISION 2 CONCERNS E
QUAD CITIES STATION, UNITS 1 & 2 VARIABLE NO.
(REF. 1):
A3 NAME: 'Drywell Pressure TYPE / CATEGORY:
A/l NRC CONCLUSION (REF. 2):
The licensee should record this variable (Section ',3.2).
UTILITY RESPONSE:
Drywell pressure is recorded in the Main Control Room, both as B1 2
variable No. 07 and as Cl variable No. 18, on GE Model 531 Pressure Recorder 1(2)-8740-12 (-5 to 70 psig).
This recorder was qualified to the original General Electric design specifications.
Drywell pressure is also recorded as a B1 variable on two safety-related (ESF Division 1 and II),
Westronics Model D4E wide range recorders 1(2)-lt10-13A and 13B (-5 to 250 psig).
These recorders have been seismically qualified by type testing in accordance with IEEE Standard 344 after environmental aging in accordance with IEEE s
Standard 323.
Both the wide range and narrow range recorders meet the range requirements of NUREG 0737, Item II.F.1 for maximum internal drywell design pressure of 62 psig with sufficient reso-lution over the 0 to +5 psig range provided by the narrow range l
recorder 1(2)-8740-12.
Since the variable is already recorded on qualified, independent, I
separate and diverse recorders, another degree of redundancy l
afforded by recording it separately as an Al variable is unnecessary.
Therefore, the existing instrumentation is considered adequate to meet the intent of RG 1.97 and NUREG 0737 regarding this variable.
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Attachment A 4 of 13 RESPONSE TO NRC REG. GUIDE 1.97, REVISION 2 CONCERNS, QUAD CITIES STATION, UNITS 1 & 2 VARIABLE NO.
(REF. 1):
51 NAME:
Primary Containment Area Radiation TYPE / CATEGORY:
E/l NRC CONCLUSION (REF. 2):
The licensee should show that this instrument meets the original station seismic c'riteria (Section 3.3.8).
UTILITY RESPONSE:
The components of the existing Iligh Range' Radiation Monitoring "ystem are qualified as identified in the attached table.
The.
documents listed verify that the existing instrumentation meets and exceeds the original station seismic design criteria speci-fled for the drywell and reactor building-in the FSAR Sections 5.2.3.5 and 12.2.2, respectively.
It has also been verified in station surveillance Procedure OIS 45 that the instrument is calibrated over the range'from 1 to 103 R/hr using the manufacturer's calibration procedure.
The actual cali-brated range exceeds the required range by one decade and is, 7 R/hr range.
therefore, adequate to monitor the 1 to 10
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Attachment A 6 of 13 RESPONSE TO NRC REG. GUltL 1.97, REVISION 2 CONCERNS QUAD CITIES STATION, UNITS 1& 2 VARIABLE NO. (REF. 1) :
24 NAME:
Radiation Exposure Rate TYPE / CATEGORY:
C/2 NRC CONCLUSION (REF. 2):
The licensee should'show that the ranges supplied for this variable encompass the radiation levels expected at the instrument locations (Section 3.3.10).
UTILITY RESPONSE:
The area radiation monitoring system has a range of 10-1 to 103 mR/hr (10-4 to 1 R/hr).
The original intent of the area radiation monitoring system was to indicate normal' operation dose rates and to alarm under abnormal cir-cumstances in a non-LOCA situation.
The calculated post-LOCA dose rates in the reactor building (secondary containment) exceed 1R/hr-(NUREG 0578 shielding report, Ref. 3, and calculations for similarly designed stations).
This post-LOCA dose
. rate is due to postulated airborne leakage from the primary containment and to post-LOCA coolant water being circulated through exposed piping
-in the' reactor building.
Basad on post-LOCA zone maps developed in response to NUREG 0578, the dose rate in the reactor building exceeds 500 R/hr for the first seven days of the design basic accident.
Accident mitigation, however, does not include entry into this radiation environment.
The calculated post-LOCA doses in the reactor building are of a magni-tude which precludes entry in the building for equipment repair during the initial accident phase.
During the recovery phase the dose rates 3
in the reactor building are expected to be less than 10 R/hr. and entry would be governed by health-physics access control using portable meters which have an upper end limit of 103 R/hr.
Therefore, the station will rely on access control and portable survey meters with an upper end limit-of 103 R/hr. for entry into the reactor building.
The present area raciation monitoring system is expected to continue to meet its originally intended function.
At post-LOCA times when entry into the reactor building is expected to be feasible, the dose rates should be less than 103 R/hr.
This allows the use of portable survey meters, with 3
an upper range limit of 10 R/hr., to encompass the expected radiation levels.
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Attachment A 7 of 13 e
RESPONSE TO NRC REG. GUIDE 1.97, REVISION 2 CONCERNS QUAD CITIES' STATION, UNITS 1 & 2 VARIABLE NO.
(REF. 1):
47 NAME:
Cooling Water Flow to ESF System Components TYPE / CATEGORY:
D/2 The licensee should provide Category 2 flow instrumentation for the diesel generator cooling water system (Section 3.3.13).
UTILITY RESPONSE:
A review of the Quad Cities Units 1 and 2 diesel generator cooling water system conducted subsequent to the issuance of Reference 1 has indicated that flow instrumentation already exists.
The flow elements, which are original equipment, and indicators are as follows:
Diesel generator cooling water pump No. 1(2)-3903 has one annubar flow element 1(2)-3941-26 (0 to 2000gpm) in the pump discharge piping and a second annubar flow element 1(2)-3941-28 (0 to 1800gpm) in the emergency room cooler piping.
These elements permit the monitoring of total pump flow and the~ flow to the diesel generator heat exchanger and the emergency room coolers.
Pump No. 1/2-3903 has one annubar flow element 1/2-3941-27 (0 to 2,000 gpm) in the discharge piping.
This element permits the monitoring of total pump flow to the diesel generator room and room coolers from the swing pump.
Each flow element has an associated local Barton Model 227A flow indicator.
The flow indicator (FI) equipment piece number (EPN) is identical to that of its associated flow element (FE) number given above.
The annubar Type 733 flow elements were installed as original equipment.
No seismic qualifi-cation of these elements was required since the installation antedated RG 1.97, Rev.
2.
The flow elements are located in the diesel generator pump rooms in mild environmental Zone No. 35 (Specification 13524-069-N202, Rev. 0) and the associated local flow indicators are readily accessible to the operations personnel after LOCA during the recovery phase.
Hence, the existing flow instrumentation meets the intent of RG 1.97, Revision 2.
As required by NUREG 0737, Supplement 1, Section 6.2,
" Documentation and NRC Review," the instrumentation used to implement the measure-ment of diesel generator cooling water flow will meet the following design requirements, as shown in the attached revision to the variable summary table (Reference 1, Pages 37 & 38).
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RESPONSE TO NRC REG. GUIDE 1.97, REVISION 2 CONCERNS QUAD CITIES STATION, UNITS 1 & 2 VARIABLE NO. (REF. 1):
49 l
l NAME:
Emergency Ventilation Damper Position TYPE / CATEGORY:
B/2 NRC CONCLUSION (REF. 2):
The licensee should verify that the alternate instrumentation (used to monitor the diesel generator room temperature and main control room dampec and fan status) is Category 2 (Section 3.3.14).
UTILITY RESPONSE:
l In lieu of position switches on the diesel room ventilation dampers, an alternative room-high-temperature monitor and alarm are provided.
Each loop comprises a pneumatic temperature sensor, a local indi-cator, a temperature switch and a temperature alarm.
These instru-ments are not safety-related and are located in a mild environment (Zone 18).
f The main control room damper valves 1/2-5741-326A and 1/2-5741-326E are provided with non-safety-related position indicating switches and lights.
These switches and lights are also located in. mild environmental Zone 18.
The switches and lights are' seismically qualified _and are suitable for Category 2 application.
The differential pressures of fans 1/2-5795-10 and 1/2-5795-30 are monitored by a pneumatic differential. pressure transmitter, local indicator, pressure switch and status light with an alarm.
Panel No. 2212-35X in which these components are mounted is also located in a mild environmental zone.
Since the alternate instrumentation involved is in a mild environ-ment and is not safety-related, environmental and seismic qualifi-cation is not required in accordance with RG 1.97, Rev.
2, Paragraph 1.3.2(a).
Therefore, the existing instrumentation is considered by the station to be adequate to meet the intent of RG 1.97 regarding this variable.
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l Attachment A 10 of 13 RESPONSE TO NRC REG. GUIDE 1.97,. REVISION 2 CONCERNS QUAD CITIES STATION, UNITS 1 & 2 VARIABLE NO. (REF. 1):
52 NAME:
Secondary Containment Radiation
~
TYPE / CATEGORY:
E/2 The licensee should show that the ranges supplied for this variable encompass the radiation levels expected at the instrument locations (Section 3.3.15).
UTILITY RESPONSE:
The area radiation monitoring system has a range of '10-1 to 103 mR/hr (10-4 to 1 R/hr).
The original intent of the area-radiation monitoring system was to indicate normal' operation dose rates and to alarm under abnornal circumstances in a non-LOCA situation.
The calculated post-LOCA dose rates in the reactor building (secondary containment) exceed 1 R/hr (NUREG 0578 shielding report, net.
3, and calculations for similarly designed stations).
This post-IfcA dose rate is due to postulated airborne leakage from the primary containment and to post-LOCA coolant water being circulated through exposed piping in the reactor building.
Based on post-LOCA zone maps developed in response to NUREG 0578 the dose rate in the reactor building exceeds 500 R/hr for the first seven days of the design basis accident.
Accident mitigation does not include entry into this radiation environment.
The calculated post-LOCA' doses in the reactor building are of such a value as to preclude entry in the building for equipment repair during the initial accident phase.
During the recovery phase the expecteg dose rates in the reactor building would likely be less R/hr and entry would be governed by health-physics accesg
/hr.
than 10 control using portable meters with an upper end range limit of 10 R
Therefore, the station will rely on access control and portable survey 3
meters with an upper end range limit of 10 R/hr for entry into the reactor building.
The present area radiation monitoring system is expected to continue to meet its originally intended function.
At post-LOCA times when entry into the reactor building is expected to be feasible, the dose rates should be less than 103 R/hr.
This gliows the use of portable survey meters, with an upper range limit of 10 R/Nr., to encompass the expected radiation levels.
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Attachment A 11 of 13 RESPONSE TO NRC REG. GUIDE 1.97, REVISION 2 CONCERNS QUAD CITIES STATION, UNITS 1 & 2 VARIABLE NO.
(REF. 1):
60 NAME:
Particulate and Halogens TYPE / CATEGORY:
E/3 NRC CONCLUSION (REF. 2):
Tr.e licensee should demonstrate that the provided range encompasses the recommended range (Section 3.3.16).
UTILITY RESPONSE:
-The main stack effluent monitoring consists of a particulate filter /
iodine cartridge assembly filter followed by an Eberline SPING-4, an NAI detector and a Victoreen PAARM (post-accident airborne radiation monitor).
This combination of instruments is for normal, abnormal (spike) and post-accident situations.
During normal operations, station technical specifications require that a particulate filter / iodine cartridge assembly be counted once every seven days.
This requirement is fulfilled by removing the particulate filter / iodine cartridge assembly which precedes the SPING-4 and counting it on a Ge Li detector.
A typical measured operational concentration is on the order of 10-12 pCi/cc.
In a design basis accident situation, a high radiation signal on the SPING skid transfers the process stream to the Victoreen PAARM from which a filter is collected every 30 minutes or as required.
This iuodification was installed to meet the NUREG 0737 requirement 2
for measurement of iodine and particulate effluent streams of 10 pCi/cc concentration for 30-minute sampling periods.
Quad Cities Station uses grab sampling particulate filter and iodine cartridge counting to quantify iodino and particulate releases for normal and post-accident applications.
This methodology quantifies releases on the order of 10-14 to 10+2 pCi/cc.
Therefore, the required range of 10-3 to 10+2 p Ci/cc is encompassed by Quad Cities methodology.
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l Attachment A 1
12 of'13 1
RESPONSE TO NRC REG. GUIDE.1.97, REVISION 2 CONCERNS QUAD CITIES STATION, UNITS 1 & 2 i
VARIABLE NO.
(REF. 1):
63 NAME:
Plant & Environs Radiation I
TYPE / CATEGORY:
E/None NRC CONCLUSIONS (REF. 2):
The' licensee should provide instrumentation that covers the recommended range (Section 3.3.17).
UTILITY RESPONSE:
The reactor building is expected to be inaccessible for an extensive post-accident period.
As shown in Reference 3, the preponderance of sources containing the pcstaccident radioactivity; i.e.,
- drywell, recirculation piping and standby gas treatment system, are located in the reactor building.
Therefore, access to other areas of the plant are governed by the shine from the airborne activity on the refueling floor and penetrations in the reactor building walls.
Most other areas of the plant are expected to be less than 500 R/hr (NUREG 0578 shielding report, Ref. 3, andcalculationsforsimilarlydesignedstationsg..
Therefore, the present survey equipment has adequate range (0-10 R/hr.).
'The present area monitoring system is expected to continue to meet its originally intended function.
At post-LOCA times when entry into the reactor building is expected to be feasible, the dose rates shouldLbe less than~103 R/hr.
This gliows the use of' portable survey meters, with an upper range limit-of 10 R/hr., to encompass the expected radiation levels.
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Attachment A 13 of 13 J
REFERENCES
'l.
CECO Summary Report, Quad Cities Station, Units 1
&.2, Compliance to Regulatory Guide 1.97, Revision 2, dated July 31, 1985.
2.
A.
C. Udy, NRC Technical Evaluation, Report No. EGG-NTA-7761, Conformance to Regulatory Guide.l.97, Quad Cities Station, Units 1 & 2, August,.1987.
3.
Quad _ Cities Station, Updated Final Safety Analysis' Report, Appendix G, Postaccident Radiation Levels, A Review of the Quad Cities Station in Response to NUREG 0578 2.1.6.b.
4.
General Electric, GEK 9597, Chapter.29.
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