ML20215C638
| ML20215C638 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 06/08/1987 |
| From: | Gallagher J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20215C618 | List: |
| References | |
| NUDOCS 8706180192 | |
| Download: ML20215C638 (31) | |
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BEFORE THE' UNITED STATES NUCLEAR REGULATORY COMMISSION-
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In the Matter of
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Docket Nos. 50-277 PHILADELPHIA ELECTRIC COMPANY 50-278 1
I APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSES-DPR-44 & DPR-56 2
Edward G. Bauer,~Jr.
Eugene J. Bradley
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2301 Market Street Philadelphia, Pennsylvania 19101.
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Attorneys for Philadelphia Electric Company
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1 8706180192 870612 f PDR ADOCK 05000277:.
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4 BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of Docket Nos. 50-277 PHILADELPHIA ELECTRIC COMPANY 50-278 APPLICATION FOR AMENDMENT OF FACILITY OPERATING LICENSES DPR-44 & DPR-56 I
Philadelphia Electric Company, Licensee under Facility Operating Licenses DPR-44 and DPR-56 for the Peach Bottom Atomic Power Station Unit No. 2 and Unit No.
3, respectively, hereby requests that the Technical Specifications contained in Appendix A of the Operating Licenses be amended by revising Tables 3.2.G and 4.2.G on pages 79 and 88, respectively, as well as the Bases on pages 89 and 93a.
The changes requested herein are, in part, 4-'1ude in the Technical Specifications information and
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'equiren.ents relating to Alternate Rod Injection (ARI) instrumentation to be installed pursuant to 10 CFR Section 50.62(c)(3).
Additional changes requested herein involve revising existing action statements and surveillance requirements 1
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applicable to recirculating pump trip (RPT) instrumentation, which would become applicable to the ARI instrumentation as well.
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l BACKGROUND:
Pursuant to the requirements of 10 CFR Section 50.62(c)(3), concerning anticipated transients without scram (ATWS) events, an Alternate Rod Injection (ARI) system will tua installed at Peach Bottom Atomic Power Station Unit 2 and Unit 3.
(Licensee uses and the proposed Technical Specifications reflect the use of the term Alternate Rod Insertion instead of Alternate Rod Injection; these terms are synonomous.) The proposed ARI modification was described in Licensee's June 30, 1986 and April i
3, 1987 letters to the Director of the Office of Nuclear Reactor Regulation and the Director of BWR Project Directorate #2, respectively.
The ARI system will (1) be independent from the existing Reactor Protection System (roactor trip system) from sensor output to final actuation device, (2) have redundant scram air header exhaust valves, and (3) perform its function in a reliable manner.
The ARI logic will be one-out-of-two twice, energize to trip, with redundant sensors.
10 CFR Section 50.62(c)(5) requires each BNR to have equipment that trips the reactor recirculating pumps under conditions indicative of an ATWS.
Since commercial operation o
began, Peach Bottom Atomic Power Station Units 2 and 3 have been-equipped with a recirculating pump trip (RPT) feature.
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4 actuation occur simultane'ously. -This wi11 be achieved by using the same sensors for'ARI and RPT.
Both RPTrand the proposed ARI' are' actuated on reactor high pressure (1120 psig) or reactor low-low water level (minus 48 inches indicated' level),__ which'are conditions indicative of an'ATWS.
The ARI and RPT systems'..will also-share pressure transmitters.and' level transmitters.
Each
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transmitter will provide a signal to an electronic trip' unit.
l When the signal from a transmitter reaches.the trip setpoint i
.(' corresponding to 1120 psig or minus 48 inches) the electronic f
trip unit will actuate the logic.
Presently, the'RPT logic is actuated directly from pressure switches and level. switches.
The'-
transmitter.and electronic trip unit' combination to be. installed
-will serve the-same function as'the switch, and will simplify:
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i instrument functional tests.
7 The RPT logic will also be modified to minimize the potential of inadhertent actuations and ensure that.a single sensor failure cannot prevent a trip.
The modifie6 RPT logic will bet a c.1e-out-of-two twice, energize to trip. logic with t
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a redundant sensors.
q h,1 DISCUSSION OF PROPOSED CHANGES:
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Introd.uction:
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RPT instrumentation is addressed in Technical Specific'ation 2
s Tables 3.2.G and 4.2..G'(pages 79'and 88). ' Table 3.I
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specifles the required minimum' number of ' operable jnstrtl.nent
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channels, trip leve;l settings and the. action that must be
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taken if:the requirementsfof.the table cannot'be: met.- ' Table 1
4.2.G specifies:the. surveillance,' testing and calibration-i requirements..The11ow-low reactorfwater level;hrips!(page
- 89) and the RPT,'in particular-(page-93a), are discussed.in
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J the Bases of Section 3.2.
Licensee proposes the following-j six categories of Technical. Specification changes (Categories-A,.B, C, D, E and F) and also proposes-to revise.the Bases'of'
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1 Section 3.2 on pages 89 and 93a to address ARI-and 10 CFR j
]q Section 50.62.
Category A:.
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Licensee proposes to add the ARI instrumentation to Technical J
Specification Tables 3.2..G and 4.2.G which' presently apply'to the RPT instrumentati,on.
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Table 3.2.G and its notes specify'the' operability requirements of the trip system and the required actions'to i
be taken if the' operability requirements of the-table:cannot j
be satisfied.
Licensee proposes that the operability requirements and action statements be identical.for"ARI and i
RPT and proposes that the existing operability requirements-l and action statements for RPT be revised as. discussed in the Category B, C and D sections of this Application.
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Table 4.2.G' specifies the test and calibration requirements
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for RPT.
Licensee proposes that the surveillance requirements be identical for ARI and RPT; Also,. Licensee
'propcses changes to the existing RPT surveillance
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t requirements as discussed in-the Category E and F sections of this Application.
Category B:
Licensee proposes to revise Note 1 on Table 3.2.G to reflect the modified RPT logic.
Presently, Note 1 states that "there shall be one operable trip system for each parameter for each operating recirculating pump."
Because the RPT and ARI will be one-out-of-two twice logics, two trip systems, rather'than one, should be required to be operable.
The proposed revision of Note 1 states that "two trip systems, each containing at least one reactor pressure channel and one reactor level channel, shall be operable."
In addition, there is no longer a need to specify that a trip system shall be operable "for each operating recirculating pump" because the modified logic has a common trip for both pumps.
Category C:
Licensee proposes to impose an additional restriction by adding a new Note 2 to Table 3.2.G which is consistent with Peach Bottom protactive instrumentation action statements for other systems addressed in the Technical Specifications.
Typically, if a redundant instrument trip system is taken out of service or becomes inoperable, that trip system is required to be placed in the tripped condition within one hour.
Thus, the proposed Note 2 states: "If the minimum number of operable instrument channels for one trip system L
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.cannot be met,'the affected trip system shall be placed in i
the tripped condition within.one hour."
Category D:
Licensee proposes.to revise the action statement note on l
Table 3.2.G (existing Note 2) because it is deficient.in that
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no time is specified for taking the action.
The revision of i
existing Note 2 is proposed Note 3.
Existing Note 2 states
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" Reduce power and place the mode selector-switch in a mode other than the RUN mode."
The proposed revision of existing Note 2 (proposed Note 3), which would apply to both ARI and RPT instrumentation, states:
"If the minimum number of 1
operable instrument channels for both trip systems cannot'be met, the reactor shall be placed in a mode other than the RUN Mode within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."
Categor,y E:
Presently, the surveillance frequencies for RPT (Table.4.2.G) are "once/ refueling cycle".
However, because the Peach Bottom Technical Specifications lack a definition for
" refueling cycle", and in this case Licensee, interprets
" refueling cycle" as synonymous with " operating cycle", which is defined in the Technical Specifications, Licensee' proposes to specify the surveillance frequencies as "once/ operating cycle".
Therefore, there actually would be no change in the surveillance frequencies.
Similarly, Licensee proposes that the' instrument. functional surveillance requirement for RPT be (
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l specified as an " Instrument Functional Test" instead of
" Instrument Functional Check", which is not defined in the Technical Specifications.
These changes would remove i
uncertainty by using terms which the Technical Foecifications define.
1 Category F:
1 Licensee proposes to add a note to the RPT Surveillance Requirements, Table 4.2.G, to clarify that the instrument functional test of ARI/RPT may " consist of injecting a simulated electrical signal into the measurement channel."
This clarification is proposed because Licensee plans to test the instrument channels by disconnecting the sensor transmitter from the electronic trip unit and replacing the i
transmitter signal with a test signal.
However, the instrument functional test definition specifies that the simulated signal should be injected ".
. as close to the primary sensor as practicable.
" which could be interpreted as meaning that the transmitter must be tripped during the test.
Thus, as for several safety-related instrument channels in the Technical Specifications (see Note 3 on page 87), Licensee proposes this clarification for ARI/RPT.
This clarification was not previously needed for RPT because the RPT instrument channels consisted of pressure and level switches, which will be replaced with transmitter and electronic trip unit combinations. ;
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- SAFETY ASSESSMENT:
-' Category A The addition of information regarding the ARI' instrumentation to the Technical Specifications reflects the enhancements to minimize-the potential for.an ATWS event, and consequently enhances the protection afforded to the health and safety of the public.
ARI will be administratively controlled in accordance with the Technical Specification requirements to ensure its reliability.
If the minimum trip system requirements of Table 3.2.G cannot be satisfied,~the unit shall be placed in a mode other than RUN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The proposed action statement 5
i was chosen to be consistent with existing' safety feature limiting conditions for operation based on-the likelihood of an actual need for the feature and the importance of the feature.
Experience has shown the Reactor Protection System (RPS) to be Very reliable.
Further reliability is provided by'the' existing backup scram valves (energize to trip). 'The ARI feature will' constitute a second backup to the RPS.
In addition, the Standby Liquid Control System provides an alternative method for shutting down the reactor in the event of a failure to scram.
Licensee.
considers the proposed action statements to be appropriate for ensuring the reliability of the ARI feature and providing additional' assurance that the capability to shutdown the reactor is maintained.
A RPT logic functional test, instrument functional c' heck and instrument calibration are required to belperformed once per
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refueling cycle.by, Table.4.2.G.,
To' assure'an equivalent level of l
reliability for.ARI, the same surveillance requirements will 1
apply to'theEARI/RPT logic and instrumentation.
Category B:
The proposed revision of Note 1-doesLnot alter its intent;' it merely reflects the planned' change of the RPT logic.. Presently, one RPT. trip system can initiate pump trip.
After:the planned' modification of the logic, a' channel from each trip system must be actuated to initiate pump trip, thus minimizing the probability:of an inadvertent pump. trip and unnecessary reactor-transient'.. Consequently, the proposed requirement that."two. trip systems, each containing at least one reactor pressure, channel and one reactor level' channel,;shall be. operable" serves the same
'1 purpose as the existing requirement.
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Category C The addition of proposed Note.2, which requires an inoperable channel to be placed in the tripped condition, ensures that the i
inoperable channel cannot prevent the actuation of a trip.
This i
action statement is typical for protective instrumentation within one-out-of-two twice logics and constitutes an additional
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i restriction on the RPT instrumentation.
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Category D:
To improve the action statement for RPT and be more consistent with the Standard Technical Specifications for General Electric Boiling Water Reactors, NUREG-0123, Revision 3, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time limit is proposed for placing the reactor in a mode other than RUN.
This is an improvement because the existing action statement does not specify a time limit.
The, action statement still establishes the same limiting mode of operation (STARTUP Mode) as Standard Technical Specification 3.3.4.1.
Category E:
Changing the RPT surveillance frequencies on Table 4.2.G from once/ refueling cycle to once/ operating cycle is merely an administrative change.
Licensee considers these terms to be synonymous in this case and, therefore, this change is in the interest of achieving consistency throughout the Technical Specifications by using terms which are defined in the Technical Specifications.
Refueling cycle is not defined in the Technical Specifications.
Operating cycle, in general, is defined on page 5 of the Technical Specifications as the " interval between the end of one refueling outage for a particular unit and the end of the next subsequent refueling outage for the same unit."
Furthermore, under the definition of surveillance frequency on page 8 of the Technical Specifications, a once/ operating cycle surveillance frequency interval is limited to 18 months (plus a 25% grace) for I
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1 instrument and electrical tests.
Therefore, specifying the surveillance frequency as once/ operating cycle ensures compliance
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with this restriction of 18 months (plus a 25% grace) for ARI/RPT.
l Changing the RPT instrument functional surveillance from a j
" Check" to a " Test" ensures that the functional intent of this surveillance is met.
The Technical Specification definition of an Instrument Check is ".
. a qualitative determination of acceptable operability by observation of instrument or channel I
i behavior during operation."
However, the Instrument Functional Test definition specifies the ".
injection of a simulated signal into the channel or instrument to verify the proper instrument channel response."
Correcting this inconsistency ensures compliance with the functional intent of the surveillance requirement.
Category F:
The proposed note clarifying the method of instrument functional testing is consistent with testing methods found acceptable for safety-related protective instrumentation channels of a similar design to that of RPT after the ATWS Rule modification.
Examples I
of this are Primary Containment Isolation System low-low-low i
reactor water level, main steam tunnel high temperature and main
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steam high flow; Core Standby Cooling System reactor water level, I
drywell pressure, reactor pressure; and Neutron Monitoring System 3
control rod block actuation instrument channels (see Technical 1 a
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- 4. 2.B and ' 4. 2.C and ' Note :L on. page Specification. Tables 4.2.A, 87).
The addition of this note for RPT does'not degrade'the effectiveness of the instrument functional ~ test; it merely allows the instrument channels to be tested in the more convenient manner"which the new transmitter - electronic. trip unit combination permits.
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SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION:
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i The NRC has provided guidance concerning the application of the standards for determining whether license amendments involve no significant. hazards considerations by providing examples (see 51 Fed. Reg.'7751).
Each category of revisions requested here is individually discussed below with regard to these examples (where applicable) followed by a'significant hazards determination.
Category A:
An example (vii) of a change that' involves no significant 4
hazards considerations is "a change to conform;a1 license to changes in the regulations, where the license change results in very minor changes to facility operations clearly in keeping with the regulations."
The installation of the ARI trip system fits this example because it is pursuant to the requirements of 10 CFR Section 50.62 and the proposed ARI-Technical Specifications do not change normal facility operations.
This system enhances safety by providing
- additional-redundancy.forothe capability'to scram the reactor.
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Another example (ii) of a change'that involves no signif'icant.
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hazards considerations is "a change'that constitutes an additional-limitation, restriction or control not presently
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included incthe technical specifications,Le.g'., a more-stringent. surveillance requirement."-- The; addition of-the ARI.
instrumentation to the Technical Specifications fits this
.q example.
The reliability of ARI is assured by imposing the proposed operability and' surveillance requirements of Tables' 3.2.G and 4.2.G.
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It has been determined that the Category.A changes; requested i
herein do not involve a significant hazards. consideration, based on the foregoing discussions, for the following reasons:
1)
The proposed revisions do not involve a significant-increase in the probability or-consequences'of an accident previously evaluated because the installation-
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of the ARI system and establishment of Technical Specification controls decrease the probability of a reactor transient without a scram.
The ARI system does not adversely affect any safety-related equipment.
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11)
The proposed revisions do not create the possibility of a'new orLdifferent kind of accident from any. accident previodsly evaluated because the ARI system and
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3-4 associated Technical Specification requirements do^not alter facility operations-or. adversely affect any.
accident analyses. 'The-ARI trip. systems ldofnot affect
- the reactivity characteristicsLof a-reactor' scram and'
. utilize the'same_ parameter' values asLthe normaliseram; initiation system.
A trip of both. recirculation' pumps-
. coincident'with'a scramLduring? reactor operation'isLless severe'than a t' rip of~both pumps without a scram,fwhich.
-has-been analyzed in the FSAR_(14.5.5.3).
-lii) The' proposed revisions do not involvefa significant reduction in a margin of' safety because the JuuL system decreases'the probability of an ATWS. event without adversely affecting any other safety margin.
The ARI' Instrumentation will'be properly; isolated,from safety -
related circuits.
Category B:
The revision of Note 1 is necessitated.by the planned modification of the RPT logic, a's part of the ARI installation modification, and does not change the intent or effect of the note.
This note as it is presently worded would become ineffective aftersthe~RPT logic is-modified..
It has been determined that the. Category B change requested herein does not involve a significant hazards consideration, based on'the foregoing discussions, forEthe following reasons.
1)
The proposed revision does not involve a significant i
increase in the probability or consequences of an i
accident previously evaluated because the change ensures 1
that the action statement retains its purpose following the modification of the RPT logic.
This change does not affect any safety-related equipment.
ii)
The proposed revision does not create the possibility of i
a new or different kind of accident from any accident i
previously evaluated because changing the operability requirement of the RPT trip system merely to reflect the j
modified logic cannot create the possibility of any 1
accident.
The modified logic is more reliable and less
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likely to spuriously actuate.
iii) The proposed revision does not involve a significant J
l reduction in a margin of safety because the purpose and effect of the note is not being changed, h
j Category C:
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An example (ii) of a change that involves no significant hazards considerations is "a change that constitutes an additional limitation, restriction or control not presently l
included in the Technical Specification, e.g., a more stringent surveillance requirement."
The addition of proposed Note 2 fits this example because it constitutes an -
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additional ~ restriction or control on the ARI/RPT trip systems.
It has been determined that the Category C change requested-herein does not involve a significant hazards consideration, i
based on the foregoing discussions, for the following reasons:
1)
The proposed revision does not involve a significant l
i increase in the probability or consequences of an accident previously evaluated because placing an inoperable channel in the tripped condition will reduce the probability of a failure of the ARI/RPT trip systems, thereby decreasing the probability or consequences of an accident.
11)
The proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated because compliance with this proposed action statement does not affect any accident analyses.
Placing one trip system in the tripped I
condition cannot initiate an accident because it merely increases slightly the probability of an inadvertent scram coincident with two recirculation pump trips.
This transient is less severe than tripping of both pumps without a scram which has been analyzed in the FSAR (14.5.5.3) and would not cause an accident.
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c 111) The proposed revision does not involve a significant reduction in'a margin of safety because compliance with this proposed action statement will increase the probability that the.ARI/RPT trip system will perform as designed.
Category D:
An example (ii) of a change that involves no significant hazards considerations is "a change that constitutes an additional limitation, restriction or control not presently included in the technical specifications, e.g., a more stringent surveillance requirement".
The revision of existing Note 2 (proposed Note 3) on Table 3.2.G fits this example because the proposed action statement imposes a time limit for taking the action, where previously no time limit I
existed.
r It has been determined that the Category D change requested herein does not involve a significant hazards consideration, based on the foregoing discussions, for the following reasons:
1)
The proposed revision does not involve a significant increase in the probability or consequences of an accident previously evaluated because the revision of the action statement is more restrictive by imposing a time limit for the required action while still requiring the reactor to be placed in a mode other than RUN.
This ;
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action? statement remains consistent with Standard Technical Specification 3.3.4.'l'.b for General Electric BWRs (NUREG-0123, Rev. 3) in that at least~the.STARTUP Mode ("a mode other than'RUN")-is~ required to.be.
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established, whichLis_ sufficient.to mitigate-the consequences of an'ATWS event.
Changing this action statement does not affect the' probability of an ATWS d
event.
11)
The proposed revision does not create theLposs'ibility of i
a new or different kind of accident from any accident previously evaluated'because the change:does not. affect any accident analyses.
The operational' mode in which the reactor shall be placed if RPT-instrumentation becomes inoperable has not been changed.
iii) The proposed revision does not involve a significant reduction in a margin of safety because this' change is conservative by imposing a time limit for the required action.
Category E:
An example (1) of a change that involves no significant hazards considerations is "a purely administrative change to technical specifications: for example, a change-to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature".
The proposed changes to specify the surveillance frequencies as once/ operating cycle instead of once/ refueling cycle, and to specify the instrument functional surveillance requirement as an " Instrument Functional Test" instead of " Instrument 1
Functional Check" is purely administrative because Licensee considers the existing terms and the proposed new terms to be synonymous in these cases and the changes would achieve consistency by using the term which is defined in the i
Technical Specifications.
Thus, these changes fit this J
- example, j
It has been determined that the Category E change requested herein does not involve a significant hazards consideration, t
based on the foregoing discussions, tor the following reasons:
1 i)
The proposed revisions do not involve a significant increase in the probability or consequences of an j
accident previously evaluated because the changes are l
purely administrative and do not affect the probability or consequences of an accident.
The changes reduce the i
probability of improper interpretation of the requirements.
ii)
The proposed revisions do not create the possibility of a new or different kind of accident from any accident previously evaluated because the changes are purely administrative and do not affect any plant operation or safety-related equipment and, therefore, the changes cannot create the possibility of an accident.
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iii) The proposed revisions do not involve a significant reduction in a margin of safety-because the changes'are purelyfadministrative and do.not affect any; margin of, safety'.
Category F:
Adding the note to Table 4.2.G clarifying the.new testing method will allow the modified RPT instrumentation to be I
functionally tested in the same manner.as previously found acceptable by the NRC for similar protective instrumentation.
It has'been' determined that the Category F change requested-herein does'not involve a significant hazards consideration, based on the foregoing discussions, for the following reasons:
1)
The proposed revision does not involve a'significant increase in the probability or consequences,of an accident previously evaluated because only the. method of performing an instrument functional test is~ involved and.
the effectiveness of the test is not degraded.
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Furthermore, the change in the test method does not increase the probability of a reactor transient and does not adversely affect any. safety ~related equipment.
Thus, the probability or consequences of an accident'are not affected. ;
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- 11). The7 proposed revision doesTn'ot create 1theLpossibilitynof-t a'new or different kind ofraccident'from'any: accidents q
previously, evaluated because. changing the~ method of l
testing the logic is facilitated'byfthe new. logic: design.
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- and does: not -involve any unreviewedfoperation: or
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procedure.f This change.doesinot introduceLanyinew'k'ind of equipment failure, mode or accident l precursor and,:
therefore,;does'not create the'possibil'ityfof a;new-or:
' different kind of accident.
iii) The proposed revision does not involve'a significant i
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reduction-in a margin of safety because'the i
effectiveness of the instrument' functional t'est is not:
degraded and no safety related equipment;or activity isi adversely affected.
i
'l lThe Plant Operating Review Committee and the' Nuclear
-Review Board have reviewed these proposed changes to the 1
i Technical Specifications and have concluded that they do not i
involve an unreviewed safety question or a significant; hazards-consideration and will not endanger the health and safety of the public.
Respectfully submitted, PHILADELPHIA ELECTRIC COMPANY?
By b d
CVice. Pres Fdent 1 !
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COMMONWEALTH OF PENNSYLVANIA ~
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COUNTY OF PHILADELPHIA J. W. Gallagher, being first duly sworn, deposes'and says:
That he is Vice' President _of Philadelphia, Electric Company; that'he has read'the foregoing Application for Amendment of_ Facility-Operating Licenses DPR-44 and DPR-56 and knows the contents thereof; and that the statements and matters set-forth-therein are true and correct to the best of hic knowledge, information and-belief.
h1J b-U Vice President Subscribed and sworn to before me this T
day of June, 1987.
f f-M(us Notary Public
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MELANlE R. CAMPANELLA Notary Pubile, Philadelphia, Philadelphia Co.
My Commission hpires February 12,1990
4 CERTIFICATE OF SERVICE' I hereby certify that1 copies of the foregoing' Application were served on the.following by-deposit in the United States mail, first-class postage prepaid, on this 12th day of June, 1987.
Regional Administrator U.S. Nuclear Regulatory Commission Region 1 1
631 Park. Avenue King'of Prussia, PA 19406
- T. P. Johnson, Resident Inspector.
U. S. Nuclear Regulatory Commission Peach Bottom Atomic Power Station P. O. Box 399 l
Delta, PA 17314 Mr. Thomas Gerusky, Director Bureau of Radiation Protection Department of Environmental Resources Fulton Bank Building, 5th Floor Third G Locusts Streets Harrisburg, PA 17120 I4 gl Eug ne J Bradley Attorney for Philadelphia. Electric Company l
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Ct m
r PBAPS Unit 2-3.2' BASE 3 l
In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.
This set of specifications provides the limiting conditions of operation for the primary j
system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.
The objectives of the Specifications are (i) to assure the effectiveness cf the protective instrumentation when required even during periods when portions of such systems are out-of-service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.
When-necessary, one channel may be made inoperable for brief intervals to conduct required functional tests'and calibrations.
Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances 1
explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to 3
prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.
Such instrumentation must be available whenever primary containment integrity is required.
)
i The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
l The low water level instrumentation set to trip at zero inches indicated level (538 inches above vessel zero) closes all isolation valves except those in Groups 1, 4 and 5.
a Details of valve grouping and required closing times are given in Specification 3.7.
For valves which isolate at this level, this trip setting is adequate to prevent the core from being uncovered in the case of a break in the largest line assuming a 60 second valve closing time.
Required closing times are less than this.
The low-low reactor water level instrumentation is set to trip when reactor water level is minus 48 inches indicated
)
level (490 inches above vessel zero).
This' trip initiates
]
HPCI, RCIC, Alternate Rod Insertion and trips the l
l recirculation pumps.
The low-low-low reactor water level instrumentation is set to trip when the reactor water level is minus 160 inches indicated level (378 inches above vessel 1
zero).
This trip closes Main Steam Line Isolation Valves, j
Main Steam Drain Valves and Recirc Sample Valves (Group 1),
j activates the remainder of the CSCS subsystem, and starts a "This page is effective upon completion of the ATWS Rule i
AR1/RPT Modification (Modification 865)."
PBAPS Unit 3 og o
'.2 BASES In addition to. reactor protection instrumentation which initiates a' reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences..This set'of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core-cooling systems, control rod block and standby gas treatment systems.
F The objectives of the Specifications are (1) to assure the effectiveness of the protective instrumentation when required even during periods when portions of such systems are out-of-service for maintenance, and (ii) to prescribe the trip i
settings required to assure adequate performance.
When necessary,.one channel may be made inoperable for brief intervals to conduct required functional tests and
)
calibrations.
E Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or i
low-end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
I Actuation of primary containment valves is initiated by protective instrumentation shown in' Table 3.2.A which senses the conditions for which isolation is required.
Such instrumentation must be available whenever primary containment integrity is required.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
The low water level instrumentation set to trip at zero inches indicated level (538 inches above vessel zero) closes all isolation valves except those in Groups 1, 4 and 5.
Details of valve grouping and required closing times are given in Specification 3.7.
For valves which isolate at this level, this trip setting is adequate to prevent the core from being uncovered in the case of a break in the largest line assuming a 60 second valve closing time.
Required closing times are less than this.
The low-low reactor water level instrumentation is set to
[
trip when reactor water level is minus 48 inches indicated level (490 inches above vessel'zero).
This trip initiates HPCI, RCIC, Alternate Rod Insertion and trips the' I
recirculation pumps.
The low-low-low reactor water level instrumentation is set to trip when the reactor water level is minus 160 inches indicated level (378 inches above vessel zero).
This trip' closes Main Steam Line Isolation-Valves, Main Steam Drain Valves and Recirc Sample Valves (Group 1),
activates the remainder of the CSCS subsystem, and starts "This page is effective upon completion of the ATWS Rule ARI/RPT Modification (Modification 865)."
'l Unit 2
- c. '.
PBAPS 3.2 BASES'(Cont'd.)
4
'l The recirculation pump trip limits the consequences of an enticipated transient without scram (ATNS) event.
The response of the plant to.this postulated event is within the bounds of
)
study events given in General Electric Company Topical Report, a
NEDO-10439, dated March,1971.
An alternate rod insertion scram limits.the consequences of a l
Reactor Protection System failure to scram during an anticipated j
transient.. The recirculation pump trip and alternate rod insertion systems are required by 10 CFR 50.62.
In the event of a loss of twa reactor building ventilation. system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200 degrees F.
Restoration of the main steam line tunnel ventilation flow momentarily exposes the temperature sensors to high gas temperatures.
The momentary temperature increase can cause an unnecessary main steam line isolation and reactor scram.
Permission is provided to increase the temperature trip setpoint to 250 degrees F for 30 minutes during restoration of ventilation system to avoid an unnecessary plant transient.
l The Emergency Aux. Power Source Degraded Voltage trip i
function prevents damage to safety-related equipment in the event l
of a sustained period of low voltage.
The voltage supply to each j
of the 4kV buses will be monitored by undervoltage relaying.
j With a degraded voltage condition on the off-site source, the undervoltage sensing relays operate to initiate a timing i
sequence.
The timing sequence provides constant and inverse time voltage characteristics.
Degraded voltage protection includes:
(1) An instantaneous relay (ITE). initiated at 90% voltage which initiates a 60-second time delay relry and a 6 second time delay relay.
The 6-second time delay relay requires the presence of a safety injection signal to initiate transfer; (2) An inverse time voltage relay (CV-6) initiated at 87% voltage with a maximum 60 cecond delay and operates at 70% voltage in 30 seconds; and (3) 1 An inverse time voltage relay (IAV) initiated at approximately 60% voltage and operates at 1.8 seconds at zero volts.
Whan the timing sequence is completed, the corresponding 4kV emergency circuit breakers are tripped and the emergency i
buses are transferred to the alternate source.
Tha 60-second timing sequences were selected to prevent unnecessary transfers during motor starts and to allow the automatic tapchanger on the j
startup transformer to respond to the voltage condition.
The 6-
)
necond timing sequence is necessary to prevent separation of the Emergency buses from the off-site source during motor starting transients, yet still be contained within the time envelope in FSAR Table 8.5.1.
-93a-t "This page is effective upon completion of the ATWS Rule
.ARI/RPT Modification (Modification 865)."
l Unit 3 I
.o PBAPS
(
4 3.2 BASES (Cont'd.)
{
i l
The recirculation pump trip limits the consequences of an I
anticipated transiendswithout scram (ATWS) event.
The response of the plant to this postulated event is within the bounds of 3
study events given in General Electric Company Topical Report, NEDO-10439, dated March,1971.
An alternate rod insertion scram limits the consequences of a l
Reactor Protection System failure to scram during an anticipated transient.
The recirculation pump trip and alternate rod insertion systems are required by 10 CFR 50.62.
l In the event of a loss of the reactor building i
ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200 degrees F.
Restoration of the main steam line tunnel ventilation flow momentarily exposes the temperature sensors to high gas temperatures.
The momentary temperature increase can cause an unnecessary main steam line isolation and reactor scram.
Permission is provided to increase the temperature trip setpoint l
to 250 degrees F for 30 minutes during restoration of ventilatioit j
system to avoid an unnecessary plant transient.
I t
The Emergency Aux. Power Source Degraded Voltage trip i
function prevents damage to safety-related equipment in the event I
of a sustained period of low voltage.
The voltage supply to each of the 4kV buses will be monitored by undervoltage relaying.
With a degraded voltage condition on the off-site source, the undervoltage sensing relays operate to initiate a timing I
sequence.
The timing sequence provides constant and inverse time voltage characteristics.
Degraded voltage protection includes:
(1) An instantaneous relay (ITE) initiated at 90% voltage which 1
initiates a 60-second time delay relay and a 6 second time delay I
relay.
The 6-second time delay relay requires the presence of a j
safety injection signal to initiate transfer; (2) An inverse time
{
voltage relay (CV-6) initiated at 87% voltage with a maximum 60 second delay and operates at 70% voltage in 30 seconds; and (3) i An inverse time voltage relay (IAV) initiated at approximately j
60% voltage and operates at 1.8 seconds at zero volts.
j i
When the timing sequence is completed, the corresponding 4kV emergency circuit breakers are tripped and the emergency i
buses are transferred to the alternate source.
The 60-second timing sequences were selected to prevent unnecessary transfers during motor starts and to allow the automatic tapchanger on the l
startup transformer to respond to the voltage condition.
The 6-second timing sequence is necessary to prevent separation of the emergency buses from the off-site source during motor starting i
transients, yet still be contained within the time envelope in FSAR Table 8.5.1.
i
-93a-
"This page is effective upon completion of the ATWS Rule ARI/RPT Modification (Modification 865)."
!