ML20073C318
| ML20073C318 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 09/16/1994 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20073C316 | List: |
| References | |
| NUDOCS 9409230242 | |
| Download: ML20073C318 (19) | |
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ATTACHMENT B MARKED UP PAGES FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF l
FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 l
BYRON STATION UNITS 1 & 2 BRAIDWOOD STATION UNITS 1 & 2 REVISED PAGES:
REVISED PAGES:
3/4 3-1 3/4 3-1 3/4 3-13*
3/4 3-13*
3/4 3-14 3/4 3 l B 3/4 3-1*
B 3/4 3-1*
B 3/4 3-2 B 3/4 3-2 l
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- NOTE:
These pages have no changes but are included for continuity i
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Z 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION I
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LIMITING CONDITION FOR OPERATION i
3.3.1 interlocks of Table 3.3-1 shall be OPERA 8LE.As a minimum, the APPLICABILITY: As shown in Table 3.3-1.
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ACTION:
As shown in Table 3.3-1.
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SURVEILLANCE REQUIREMENTS 3
4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance i
the Reactor Trip System Instrumentation Surveillance Requirements specifie Table 4.3-1.
,I 4.3.1.2 shall be ixt.. aThe REACTOR TRIP SYSTEM RESPONSE TIME of each Reacto be within its limit at least once per 18 months.
4 Eac -4ee6 shall include at least one train such that both trai i
eas once per 36 month i
are tn W t least onc and one channel per function such that all channels i
of redundant channels i every N times 18 months where N is the tota: number Channels" column of Table 3.3-1.-a specific Reactor trip function as
" Total No. o 4
i Verifbe ye,-ified W
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I BYRON - UNITS 1 & 2 3/4 3-1 i
AMENDMENT NO. -ES-j i
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INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTAT LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.
APPLICABILITY: As shown in Table 3.3-3.
ACTION:
With an ESFAS Instrumentation or Interlock Trip Setpoint less a.
conservative than the value shown in the Trip Setpoint column but-t more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip -
Setpoint value.
b.
With an ESFAS Instrumentation or Interlock Trip Satpoint less conservative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Satpoint value.
With an ESFAS instrumentation channel or interlock inoperable, c.
take the ACTION shown in Table 3.3-3.
i BYRON - UNITS 1 1 2 3/4 3-13 AMENDMENT NO. 53
fNSTRUMENTATf0N SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requircments.specified in Table 4.3-2.
i 4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of eac shall be d._.retat:d
- be within the limit at least once per 18 months.
Eac h shall includ at least one train such that both trains are t; t:" at east once per 36 mont s and one channel per function such that all channel t least o :e per N times 18 months where N is the total number o are redundant ch nels in a specific E5FAS function as shown in the " Total No. of Channels" Col of ble 3.3-3.
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BYRON - UNITS 1 & 2 3/4 3-14 gggypyy yq
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i 3 /4.3 INSTRUMENTATION i
5 BASES l
3/4.3.1 and 3.f4.3.2 1EACTOR " RIP SYSTEM and ENGINtiRED SAFFTY FEATUR l
ACTUATION SYs"iM IiniadMENTAT: GN i
Features Actuation System instrumentation and interlocks associated ACTION and/or Reactor trip will be initiated when the parameter (1) the monitored by each channel or combination thereof reaches its Setpoint (2) the specified coincidence logic and sufficient redundancy is maintained to permit l
I a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and i
Engineered Safety Features instrumentation and 3 j
capability is available from diverse parame,ters. ) sufficient system functions 1
The OPERABILITY of these systems is required to provide the overall i
reliability, redundancy, and diversity assumed available in the facility 2
design for the protection and mitigation of accident and transient condittons.
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The integrated operation of each of these systems is consistent with the 3
assumptions used in the safety analyses. The Surveillance Requirements i
specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.
sufficient to demonstrate this capability. periodic surveillance te nimum frequencies are The i
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and surveillance and maintenance outage times have been determined Service Times for the Reactor Protection Instrumenta
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supplements to that report. Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumenta Setpoints specified in Table 3.3-4 are the nominal valu i
j bistables are set for each functional unit.
j adjusted consistent with the nominal value when thA Setpoint.is considered to be j
within the band allowed for calibration accuracy. e "as measured" Setpoint is i
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j tests and the accuracy to which Setpoints can be m i
j Allowable Values for the Setpoints have been specified in Table 3.3-4.
Operation with Setpoints less conservative than the Trip setpoint but within 4
the Allowable Value is acceptable since an allowance has been made in the l
safety analysis to accommodate this error.
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4 BYRON - UNITS I & 2 B 3/4 3-1 i
AMENDMENT NO. 55-i s
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INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFET FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) f
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The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor' and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance. Being that there is a small statisitical chance that this will happen, an infrequent excessive drift is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
asurement of response time at the specified frequencies pro
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assurance t eactor trip and the Engineered Safety Feat untion keplace associated with each c completed within the a t assumed in the g
safety analyses. Response time eso y any series of sequential, overlapping or total e urements provided that such WERT' tests demonstrate the tot e response time as Sensor response time verificatio demonstrated by either: 1 in place, or asurements, or (2) utilizing replac(em)ent sensors with ce A,
offsite nse times.
The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded.
If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients.
Once the required logic combination is completed, the system sends actuation signals to those Engineered safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) feedwater isolation, (4) startup of the emergency diesel 9enerators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, 8 (9) auxiliary feedwater pumps start and automatic valv(es) Turbine trip,
- position, (10) containment cooling fans start and automatic valves position, and (11) essential service water pumps start and automatic valves position.
BYRON - UNITS 1 & 2 B 3/4 3-2 AMENDMENT NO. -H-
l 3/4.3 INSTRUMENTATION 3/4. 3.1 REACTOR TRIP SYSTEM INSTRUMENTATION 4
l L]MlllNG CONDITION FOR U"TRAT10N 1
1 3.31 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.
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APPLICABILITY: As shown in Table 3.3-1.
ACTION:
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As shown in Table 3.3-1.
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SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in 4
Table 4,3-1.
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4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be ;.n::t :*rd-o be within its limit at least once per 18 months.
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.Eachf e+4 shall inclu b
atleastonetrainsuchthatbothtrainsarebee4edsqt 4
[' least once per 36 month and one channel per function such that all channels
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least once every N times 18 months where N is the total nun.ber I
of redundant channels i a specific Reactor trip function as shown in the
" Total No. o, Channels" column of Table 3.3-1.
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BRAIDWOOD - UN1151 & 2 3/4 3-1 AMENDMENT NO.
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INSTRUMENTATION 3 /.A. 3. 2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setnoint column of Table 3.3-4.
APPLICABILITY: As shown in Table 3.3-3.
ACTION:
With an ESFAS Instrumentation or Interlock Trip Setpoint less con-a.
servative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-4 adjust the Setpoint consistent with the Trip Setpoint value.
b.
With an ESFAS Instrumentation or Interlock Trip Setpoint less con-servative than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable l
ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
l With an ESFAS instrumentation channel or interlock inoperable, take c.
the ACTION shown in Table 3.3-3.
BRAIDWOOO - UNITS 1 & 2 3/4 3-13 AMENDMENT NO. 42
INSTRUMENTATION SURVEILLANCE REQUIREMENTS l
4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.
4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be d:::n:tr:ted o be within the limit at least once per 18 months.
AEa14e+t shall include at least one train such that both trains are 4e+4ed-at J least once per 36 month and one channel per function such that all channels j
are t::ted at least once per N _ times 18 months where N is the total number of redundant channels in a specific}ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.{
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BRAIDWOOD - UNTTS 1 & 2 3/4 3-14
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3/4.3 INSTRUMENTATION l
i BASES i
3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
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The OPERABILITY of the Reactor Trip System and the Engineered Safety i
Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof' reaches'its Setpoint, (2) the specified coincidence logic and sufficient redundancy is maintained to pemit a channel to be out of service for testing or maintenance consistent with I
maintaining an appropriate level'of reliability of the Reactor Protection and 1
Engineered Safety Features instrumentation, and (3) sufficient syster functions capability is available from diverse parameters.
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The OPERABILITY of these systems is required to provide the overall j
reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencias are sufficient i
to demonstrate this capability. Specified surveillance intervals and i
surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Su..eillance Frequencies and Out of Service l
Times for the Reactor Protection Instrumentation System," and supplements to i
that report. Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and EngineeredgSafety Features instrumentation.
j The Engineered Safety Features Actuation System Instrumentation Trip i
Setpoints specified in Table 3.3-4 are the nominal values at which the i
bistables are set for each functional ut.it. A Setpoint is considered to be i
adjusted consistent with the nominal value when the "as measured" Setpoint is j
within the band allowed for calibration accuracy.
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To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4.
j Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the l
safety analysis to accommodate this error.
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BRAIDWOOD - UNITS 1 & 2 B 3/4 3-1 AMENDMENT NO. 44 i
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INSTRUMENTATION BASES i
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i REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
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INSTRUMENTATION (Continued) g i
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The methodology to derive the Trip Setpoints is based upon. combining all of the uncertainties in the channels.
Inherent to the determination of the
- Trip Setpoints are the magnitudes of these channel uncertainties.
Sensor and rack instrumentation utilized in these channels ~are expected to be capable of operating withis, the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not j
met its allowance.
Being that there is a small statisitical chance that this will happen, an infrequent excessive drift is expected.
Rack or sensor drift,
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i in excess of the allowance that is more than occasional, may be indicative of I
more serious problems and should warrant further investigation.
5, The m ent of response time at the specified frequencie es 9 g/u c assurance that the trip and the Engineered Safe res actuation 3 g/p associated with each channe leted with 1me limit assumed in the safety analyses.
Response time may trated by any series of sequential, 41Dr+ A,, overlapping or total channe maasurements that.such tests demon-j strate the total response time as defined.
Sens nse time veri-ficatio e demonstrated by either: (1) in place, onsite, or o test j
rements, or (2) utilizing replacement sensors with certified response TheEngineeredSafetyFeaturesActuationSystemsensesselectedhlant
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i parameters and determines whether or not predetermined limits are being exceeded.
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If they are, the signals are combined into logic matrices sensitive to 1
combinations indicative of various accidents, events, and transients.
Once i
the required logic combination is-completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition.
As an example, the following actions 4
may be initiated by the Engineered Safety Features' Actuation System to mitigate j
the consequences of a steam line break or loss of coolant accident: (1) Safety Injection purnps start and automatic valves position, (2) Reactor trip, (3) feed-water isolation, (4).startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, and (11) essential service water pumps start and 2
automatic valves position.
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l INSERT A The verification of response time at the specified frequencies provides assurance that the reactor trip and the engineered safety features actuation associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be verified by actual tests in any series of sequential, overlapping or total channel measurements, or by summation of allocated sensor response times with actual tests on the remainder of the channel in any series of sequential or overlapping measurements. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632,
" Elimination of Pressure Sensor Response Time Testing Requirements, Revision 1, "provides the basis and methodology for using allocated sensor response times in the overall verification of the Technical Specifications channel response time. The allocations for sensor response times must be verified prior to placing the sensor in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. One example where time response could be affected is replacing the sensing assembly of a transmitter.
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l ATTACHMENT C EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATIONS FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 l
Commonwealth Edison (Comed) has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to Title 10 Code of Federal Regulations, Part 50, Paragraph 92, Section c (10 CFR 50.92 (c)), a proposed amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed amendment would not:
j 1.
Involve a significant increase in the probability or consequences of an accident previously evaluated; or l
2.
Create the possibility of a new or different kind of accident from any i
accident previously evaluated; or 3.
Involve a significant reduction in a margin of safety.
A. INTRODUCTION Response Time Testing (RTT) of Reactor Trip System (RTS) instrumentation and Engineered Safety Features Actuation System (ESFAS) instrumentation has been required by Technical Specifications since the mid 1970's. The purpose of RTT was to demonstrate that the instrumentation met the response time performance requirements assumed in the plant safety analyses.
j A data review conducted by the Electric Power Research Institute (EPRI) has shown that RTT has not detected response time failures. This can be attributed in a large part to the fact that a calibration surveillance is typically performed first and has discovered failures that would affect response time. Additionally, since its inception, i
RTT has proven to be resource intensive. RTT is generally performed in discrete steps, with the sensor response time being one of the steps. RTT of sensors is especially expensive, since many of the tests require special equipment and technical skills in addition to extensive test times.
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i As a result, EPRI initiated a program to determine if RTT could be eliminated for specific pressure and differential pressure transmitters and switches. The results of l
the EPRI program are delineated in EPRI Report NP-7243, " Investigation of Response l
Time Testing Requirements, Revision 1."
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WCAP-13632, " Elimination of Pressure Sensor Response Time Testing Requirements, Revision 1," provides the technical justification for deletion of periodic response time testing of selected pressure sensing instruments. The program described in WCAP-13632 uses the recommendations contained in EPRI Report NP-7243 for justifying elimination of response time testing surveillance requirements on certain pressure and differential pressure sensors. To address other sensors installed in Westinghouse designed plants, WCAP-13632 contains a similarity analysis to sensors in EPRI Report NP-7243 or a Failure Modes and Effects Analysis (FMEA) to provide justification for elimination of response time testing requirements. The specific sensors installed at Byron and Braidwood are listed below.
Byron 1 & 2 Braidwood 1 & 2 Pressurizer Water Level Barton 764 Barton 764-Steam Generator Water Level Barton 764 Barton 764 Pressurizer Pressure Barton 763 Barton 763A Steamline Pressure Barton 763 Barton 763 Containment Pressure Barton 752 Barton 752 Reactor Coolant Flow Barton 752 Barton 752 Reo. tor Coolant System Wide Range Pressure Barton 763 Barton 763 l
Tobar 32PA2 Tobar 32PA2 l
Refueling Water Storage j
Tank Level Barton 752 Barton 752 The basis for eliminating periodic response time testing for each of the above sensors is discussed in WCAP-13632 and/or EPRI Report NP-7243. These reports provide justification that any sensor failure that significantly degrades response time will be detectable during surveillance testing such as calibration and channel checks.
Based on these results, Comed proposes to amend Surveillance Requirements 4.3.1.2, 4.3.2.2 for Specifications 3.3.1 Reactor Trip System Instrumentation, and 3.3.2 Engineered Safety Feature Actuation ~ System Instrumentation respectively and the bases sections for these specifications for Byron Units 1 and 2, and Braidwood Units 1 and 2. This revision will indicate that the system response time shall be verified using a sensor response time justified by the methodology described in WCAP-13632 Revision 1.
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NO SIGNIFICANT IIAZARDS ANALYSIS i
1.
The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
This change to the Technical Specifications does not result in a condition where the design, material, and construction standards that were applicable prior to the change are altered. The same RTS and ESFAS instrumentation is being used; the time response allocations /modeling assumptions in the Updated Final Safety Analysis Report (UFSAR), Chapter 15, Accident Analyses, are still the same; only the method of verifying time response is changed. The proposed change will not modify any system interface and could not increase the likelihood of an accident since these events are independent of this change. The proposed activity will not change, degrade or prevent actions or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the UFSAR. Therefore, the proposed amendment does not result in any increase in the probability or consequences of an accident previously evaluated.
2.
The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
This change does not alter the performance of the identified pressure and differential pressure transmitters and switches used in the plant protection systems. All sensors will still have response time verified by test before placing the sensor in operational service, and after any maintenance that could affect response time. Changing the method of periodically verifying instrument response for these sensors (assuring equipment operability) from time response testing to calibration and channel checks does not result in any design, installation, or operational changes and thus will not create any new accident initiators or scenarios. Periodic survcillance of these instruments will detect significant degradation in the sensor msponse characteristics. Implementation of the proposed amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
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3.
The proposed change does not involve a significant reduction in a margin of safety.
This change does not affect the total system response time assumed in the safety analyses. The periodic system response time verification method for the identified pressure and differential pressure sensors and ' switches is modified to allow use of(1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite or offsite (e.g.
vendor) test measurements, or (3) using vendor engineering specifications.
The method of verification still providss assurance that the total system response is within that defined in the safety analyses, since calibration tests will-detect any degradation which might significantly affect sensor response. time.
Based on the above, it is concluded that the proposed license amendment request does not result in a reduction in margin with respect to plant safety.
Therefore, based upon the above evaluation, Comed has concluded that these changes involve no significant hazards considerations.
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ATTACHMENT D i
i ENVIRONMENTAL ASSESSMENT FOR PROPOSED CHANGES TO APPENDIX A TECHNICAL SPECIFICATIONS OF l
FACILITY OPERATING LICENSES NPF-37, NPF-66, NPF-72, AND NPF-77 l
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Commonwealth Edison Company (Comed) has evaluated this proposed license amendment request against the criteria for identification oflicensing and regulatory j
actions requiring environmental assessment in accordance with Title 10, Code of Federal Regulations, Part 51, Section 21 (10 CFR 51.21). Comed has determined that this proposed license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9). This determination is based upon the following:
1.
The proposed licensing action involves the issuance of an amendment to a license for a reactor pursuant to 10 CFR 50 which changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or which changes an inspection or a surveillance requirement. This proposed license amendment request eliminates the requirement to perform response time l
testing for pressure and differential pressure transmitters used in Reactor Trip System Instrumentation and Engineered Safety Features Actuation System Instrumentation applications; 2.
this proposed license amendment request involves no significant hazards considerations; 3.
there is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite; and 4.
there is no significant increase in individual or cumulative occupational radiation exposure.
Therefore, pursuant to 10 CFR 51.22(b), neither an environmental impact statement
)
nor an environmental assessment is necessary for this proposed license amendment request.
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ATTACIIMENT E WCAP-13632 (Proprietary)
ELIMINATION OF PRESSURE SENSOR RESPONSE TIME TESTING REQUIREMENTS WOG Program MUHP-3040 R.evision 1 i
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ATTACIDENT F WCAP-13787 (Non-Proprietary)
ELIMINATION OF PRESSURE SENSOR RESPONSE TIME TESTING REQUIREMENTS WOG Program MUHP-3040 Revision 1 l
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