ML20246L803
| ML20246L803 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 07/12/1989 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20246L806 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737 GL-83-36, NUDOCS 8907190003 | |
| Download: ML20246L803 (9) | |
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ATTACHMENT 1 i
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.i PEACH BOTTOM ATOMIC POWER STATION.
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UNITS 2 AND 3 j
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Docket Nos. 50-277.
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'50-278
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DPR-56' j
.q TECHNICAL SPECIFICATIONS CHANGE REQUEST j
No. 89-04
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" Reactor Water Level and Reactor-Pressure Surveillance Instrumentation Calibration Frequency Change" 1
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Docket Nos. 50-277
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b 50-278 License Nos. DPR-44 m
DPR-56 Description of Change l
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i This Technical Specification Change Request proposes
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that the calibration frequency of the Reactor Water Level (narrce range) and Reactor' Pressure Surveillance Instrumentation be changed from once/6 months to once/ operating cycle:(Table 4.2.F, page 86).
Operating Cycle is defined in the Technical Specifications as the " interval between the end of one refueling outage for a particular unit and the end of the next subsequent
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i refueling outage for the sama unit".
Typically, a Peach Bottom
'i operating cycle is less than 22 months.
l The instruments involved, with the exception of two instruments, are part of the Feedwater Control Sybtem and are described in Section 7.10.3.1 of the Updated FSAR.
The feedwater control room indication involved (LR-2(3)-6-96;LI-2(3) 94A,B,C;PR-2(3)-6-96;PI-2(3)-90A,B,C). receives its signals from the Feedwater ReactorLVessel Water Level Measurement system and aids the operators when performing normal operations.
The two remaining instruments are accident monitoring (NUREG-0737) reactor pressure recorders (PR-2(3)-2-3-404 A,B) that receive their signals from instruments that also provide signals to control / initiate Core Standby Cooling Systems (CSCS).
The primary instruments which feed the accident monitoring reactgr pressure recorders are covered by Technical - _ _ _ _ _ _ _ _ _
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Docket Nos. 50-277 h-50-278 j]y/
' License Nos. DPR-44 I
- 9 Sr ".fications Table 4.2.B_as a. result of a'recent modification which expanded'their function to the CSCS. 'Previously, these-1 u
instrumentsLonly served the accident monitoring. indication l
function.
Table 4.2.B specifies a calibration frequency of once/ operating cycle foi these instruments; therefore, this proposed Technical Specifications Change would achieve i
consistency between these specifications.
1 This change is requested to avoid calibrating the
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1 feedwater level and pressure transmitters associated with the control room indication during reactor operation..The valving.
operations necessary to return these transmitters to service following calibration could perturb other instruments and consequently cause a scram and/or primary containment isolation.
Because a less frequent calibration will not-significantly compromise the function of the instrumentation and will aliminate I
a potential for a plant transient or forced outage,;this'
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i Technical Specification change is considered appropriate.
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4 Safety Assessment:
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Calibration of this instrumentation is not discussed i
in the Technical Specification Bases or Updated FSAR.
Section j
I 7.10.4 of the Updated FSAR, "Feedwater-Control System -
i Inspection and Testing" states:
"All feedwater ficw control j
system components can be tested and inspected prior to plant
4 Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 operation and during scheduled shutdowns".
Section 7.20.6 of the Updated FSAR, " Accident Monitoring-Inspection and Testing",
merely states that periodic testing is in accordance with the Technical Specifications.
Narrow range reactor water level and pressure surveillance indication is compared at least once per shift with indication from other sensing devices to detect instrument malfunctions as required by Table 4.2.P (instrument check).
A review of historical calibration records has revealed that this instrumentation experiences minor drift.
The instruments are typically found within tolerance when tested.
The Feedwater instrumentation involv;d is not safety-related nor does it initiate any safety features or control any safety-related systems.
The Feedwater instrumentation is not NUREG-0737 accident monitoring instrumentation (as specified in the Technical Specification Bases on page 93).
This instrumentation is only associated with Feedwater and the Main Turbine.
The accident monitoring instruments that, as a result of a recent modification, now provide CSCS signals are comparable to the instruments that previously served the same function.
These instruments were made by the same manufacturer and have similar setpoint drift characteristics.
Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5) typically specify a calibration frequency of once/18 months for instrumentation that does 'ot Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 perform a safety function.
NRC Generic Letter 83-36, "NUREG-0737 Technical Specifications", recommends a once/18 months calibration frequency for accident monitoring reactor pressure indication.
Because a 25% extension to 22 1/2 months is also permissible, in effect, there is little difference between a once/18 months and once/ operating cycle requirement.
Based on the above, the proposed change is considered to be safe, appropriate and in the interest of avoiding unnecessary plant transients and forced outages.
Significant Hazards Consideration Determination:
The Company proposes that the change requested herein does not involve a significant hazards consideration based on the foregoing discussion for the following reasons:
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The proposed revisions do not involve a significant increase in the probability or consequences of an accident previously evaluated.
The proposed Technical Specification change reduces the calibration frequency of the narrow range reactor water level indicators and the reactor pressure indicators from once/6 months to once/ operating cycle.
The Feedwater instruments provide indication as part 1
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Docket Nos. 50-277 50-278 License Nos. DPR-44 acR-56 of the feedwater level control loops.- This instrumentation is not safety-related.
l The proposed frequency of once/ operating cycle is consistent with industry standards and NRC guidelines, and ensures an acceptable level of reliability for the instrumentation.
Based on a review of historical calibration data, feedwater level control and accident monitoring will not be adversely affected.
The Feedwater instruments share manifolds with other instruments which generate scram and/or primary containment isolation signals.
If the calibration is done at power, valving the narrow range level and pressure instruments back into service following calibration may cause a pressure transient which could result in a reactor scram or isolation.
The proposed frequency would eliminate the need to perform the calibration at power or to shut the plant down for the Purpose of calibration.
Because the proposed change does not alter the function of the instrumentation, the change does not increase the probability of occurrence or the consequences of an accident or malfunction previously evaluated.
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Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 ii)
The proposed revisions do not create the possibility of a new or different kir-of accident from any accident previously evaluated.
l The proposed change does not involve any hardware changes to the instruments or changes to their ranges.
The proposed change effects only the frequency of calibration, and does not involve any new testing or calibration methods or configurations.
Additionally, the proposed change does not effect the redundancy, electrical separation or equipment qualification of the instruments.
Therefore, the proposed change does not create the possibility for an accident or malfunction of a different type than any previously evaluated.
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The proposed revisions do not involve a significant redt.ction in, 'ergin 74f safety.
The Fes?dt: ter reactor water level and Feedwater reactor pressure indicatcrs which are the subject of the proposed Technical Specification change do not initiate or control safety-related systems, and are not part of accident monitoring.
Their function is to provide indication as part of the feedwater level control loops.
Feedwater level control is discussed in Section 7.10 of the UFSAR.
The accident monitoring Docket Nos. 50-277 50-278 License Nos. DPR-44 DPR-56 instruments involved are similar to numerous other instruments which serve more significant safety functions and are calibrated once/ operating cycle.
Thus, accident monitoring capability will not be degraded such that any margin of safety could be decreased.
Accident Monitoring is discussed in Section 7.20 of the UFSAR.
Surveillance intervals for the instrumentation involved are not discussed in the UFSAR or Technical Specification Bases.
The proposed change does not affect the function or operability of the indicators or their associated transmitters and, therefore, does not reduce any safety margins.
Environmental Impact:
An environmental impact assessment is not required for l
the amendment requested by this change request because the requested change conforms to the criteria for " actions eligible for categorical exclusion" as specified in 10 CPR 51.22(c)(9).
The requested change will not change the amou ;s or types of effluents that are postulated for design basis events because the calibration frequency change will not impact any design basis events previously evaluated.
The change involves no significant hazards consideration as demonstrated in the preceding section.
l The change involves no significant change in the types or significant increase in the amounts of any effluents that may be
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1 Docket Nos. 50-277 50-278 l
License Nos. DPR-44 DPR-56
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I released offsite, and there will be no significant increase in i
i individual or cumulative occupational radiation exposure.
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Conclusion:==
l The Plant Operations Review Committee and the Nuclear Review Board'have reviewed the proposed change to the Technical
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Specifications and have concluded that they_do.not involve _an i
i unreviewed safety question, and will not endanger the health and safety of the public.
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