2CAN079202, Application for Amend to License NPF-6,revising TS 2.1.1 Bases,Section 3.2.8 & Bases & Tables 2.2-1 & 3.3-4 to Increase Allowable Pressurizer Pressure Range & Lower Low Pressurizer Setpoint for Reactor Trip & SI
| ML20102B210 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 07/22/1992 |
| From: | Carns N ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20102B212 | List: |
| References | |
| 2CAN079202, 2CAN79202, NUDOCS 9207280262 | |
| Download: ML20102B210 (21) | |
Text
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8_8_8_8. 'Entergy coi. m o s.iion.,io.
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ness.. o nset Tel 501-W+6888 -
.Neil S. " Buzz" Carns.
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'i July 22, 1992 2CAN079202' 4
U. S. Nuclear Regulatory Commission Document Control Desk Hall Station F1-137 Washington. DC 20555 i
SUBJECT:
Arkansas Nuclear One"- Unit 2 l
Docket No.L50-368 License No. NPF-6 g
Technica1' Specification Change Request-Pressurizer Pressure-Gentlemen:
Attached for your review and approval are proposed. Technica1L Specification.
(TS) changes.which revise:2.1.1' Bases. Section ;3.2.8 - and; Bases, and Tab 1's 2.2-1 and.3.3-4 This-change. increases-the; allowable' pressurizer pressate 4
I range.
A lower low pressurizer pressure setpointifor reactor-trip :' safety injection, and containment cooling is:also beingfproposed by this change.
b The proposed change'has been, evaluated in accordanceivith 100FR50.91(a)(1).:
using criteria in 10CFR50. 92(c' - and, it : has ybeen - determined-ithat-i these changes involve no'significant hazards considerations. zThe bases 4forJthese-determinations are included in the enclosed:: submittal.
Although-the, circumstances of, this; proposed.' change ; are :: not! considered review and ; approval -is j requested;: 1The-emergency or exigent, :your prompt c
purpose of. this proposed Technical Specification change isc to : improve !the-r liability ' of the-pressurizerisafety: valves 1by ' reducing Naive simmering :
and subsequent leakage..-
4 We request that Ithe. effective date of-this / change) belupon(1ssuance;of -.the i<
amendment'.
m Very truly yours, f)
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U.S. NRC July 22, 1992 J
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cc:
Mr. ' James L. Hilhoan U. S. Nuclear Regulatory-Commission Region IV.
j-611 Ryan. Plaza-Drive, Suite 1000 l
Arlington, TX 76011 i
NRC Senior Resident Inspector Arkansas Nuclear One - ANO-1 & 2 Number 1, Nuclear Plant Road a
Russellville, AR 72801 Mr. Thomas W. Alexion
~
NRR~ Project Harager, Region IV/ANO-1 U. S. Nuclear Regulatory Commission' j:
= NRR Mail Stop 11-D-23
.One White Flint North l
11555 Rockville Pike i
Rockville, Maryland.20852.
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- Ms. Sheri R. Peterson-NhR Project Manager, Region IV/ANO U. S. Nuclear Regulatory Commission NRR Mail-Stop~11-D * -
One White Flint North 4
11555 Rockville' Pike Rockville, Maryland 20852 t.
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STATE OF ARKANSAS
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SS COUNTY OF LOGAN
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d Affidavit I, N. S. Carns, being duly sworn, subscribe to and say that I am Vice President, Operations _ANO for Entergy Operations, that I have full authority to execute this affidavitt-that I have read the document numbered 2CAN079202 and know the contents thereof; and that to the best i
of my knowledge, information and belief the statements in it-are true.
D Y es.
N. S. Cakns I
SUBSCRIBED AND SWORN To before me, a Notary Public in and for the County and State above named, this MMday of
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- 1992, pf 4t JA Al4Y///4Gl4$fS
- [ No'tary Public
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My Commission. Expires
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ATTACHMENT PROPOSED TECHNICAL'SPECIFICATIONL
.AND'
-RESPECTIVE-SAFETY-ANALYSES IN THE MATTER OF AMENDING j
LLICENSE N0.- NPF-6 ENTERGY OPERATIONS,;INC..-
ARKANSAS NUCLEAR ONE,. UNIT TWO:
DOCKET NO 50-368 i
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PROPOSED CHANGES The proposed amendment would change Arkansas Nuclear One, Unit 2 (ANO-2)'
Technical Specification (TS) - 3. 2.8 and associated bases L to allow plant-operation with pressurizer pressure between 2025 and 2275 psia.
A clarification of TS 2.1.1 Bases -with regards to the application of the-l peak linear heat rate (PLHR) limit to anticipated.operationalf occurrences -
analysis results is also being proposed, i
l Additionally, this proposed amendment would lower the_ANO-2 TS; Table 2.2-1 a
and associated bases reactor protection = low - pressurizer Lpressure - trip setpoint and allowable values to 1717.4 and 1686.3 psla respectively.
The_
safety injection and containment cooling actuation ' trip setpoint and allowable values given in. ANO-2 Technical Specification Table 3.3-4 would -
I also be lowered to=1717.4.and 1686.3 psia by +his proposed amendment'.
i BACKGROUND i
Small amounts-of : pressurizer - safety. valve _ leakage have ; historicallyf occurred on ANO-2 but not-until recentlyJhas this_ problem'alone resulted in:
plant shutdowns for _ valve replacement' or.i repa it... In the.'past.
small amounts of safety valve : leakage below the: allowable > TS 11mits fori-Reactort Coolant System (RCS). leakage were not considered ~-significant.- _ Also, i
continuous plant operation! runs were sufficiently shorter.. Plant shutdowns
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and forced outages.due to ' other reasonso allowed ' the r safety. valves. to be replaced or repaired if necessary.
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Recently, leakage out of the pressurizer has _ been Lmonitored more closely..
i Modifications have been made to the safety valves ;(Flexi-Disk),? and - to the valves' discharge piping (nozzle load reduction) in ancattempt to minimize-or eliminate valve simmering / leakage.
. However, x these.. changes ihave not.
eliminated the problem.: Initially, valve simmering is: characterized by: low-volume, high velocity, saturated.- steam leakage' 'across ; the valve _ seats.
This:Is condensed _1mmediatelyfwithout causing.a-temperature increase at'the discharge piping temperature sensors-(located;approximately 6 feet from-tho' valve),. nor = a measurable volume - increase in., the Lquenchg tank. - ; Prolonged' F
-simmering tends to allow-an increased ' -steam ;. volume; Lto. the - point of detection.
Continued - exposure _ at-the _h'igher volume and. velocityltypically will cause seat damage.
- Once.seati damage ' occurs,- the valveicannot be rescated into a leak tight ; condition. LAfter' seat; damage,;:the11eakage rate
.tends to increase with time,1soLthe longer-the leakage duration,-the higher:
the leakage L rate.
-As the performance.'of f ANO-2 iimproves p as' indicated by longer continuous plant -operating times, _the totalo safetyL valve ' leakage-increases in an exponential-manner, eventually: causing-plant shutdown. (In October, 1991, ANO-2 was shut ~down to. replace J the safety 7 valves du_e - to
.j leakage concerns. This type =ofEforced outage has;both economic impacts:and associated safety concerns.
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The economic impacts of the high maintenance costs and an' low as. reasonably-achievable (ALARA) considerations to repair / replace the valves, in addition r
to the power loss associated with a forced outage are significant.
There are also costs associated with the treatment of the radwaste generated by
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the leaking valves in add! tion to the requirement for ma'teup water.- Valve leakage within the range currently administratively allowed due to 2
pressurizer heater capacity (1 gpm versus - 10 gpm allowed by Technical Specifications) does not affert the capability of the valve.to perform'its safety function.
Additionally, there-is an exposure - to riske associated with perturbating the RCS for plant shutdown and - cooldown for valve-replacement.
Reducing operating pressurizer pressure, for a short time period when - the valve starts to simmer, is expected to substantially curtail:or.elimincte l
the safety valve simmering problem and subsequent valve leakage.
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operating the RCS at reduced. pressures,
.he - safety ; valves are Lgiven a chance to reach a thermal equilibrium point' at a pressure with sufficient r
margin to the valve lift setpoint (2560 +1,-3% - psia). to avoid simmering.
Presently, plant operation with RCS pressure within-the bounds specified by
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Technical Specification ' 3.2.8, maintains -only an _ approximate fl0% margin to 4
the safety valve lift setpoint.
Small perturbations-in the val.e thermal equilibrium point can initiate valve simmering.
It is postulated i. hat when--
the porturbations occur,'if the RCS pressure-is reduced for a sut Alent time period to allow.the valve to reestablish an equilibrium _ point,.
simmering and valve leakage can be terminated.
To reestablish the.
equilibrium point involves a
significant:
pressure reduction' to approximately 2025 psia for-short durations. To ensure the valve (s)-remain leaktight, it may be appropriate to' maintain continuous ' operation 'at a smaller pressure reduction (approximately 2150 : psia).
By avoiding the simmering, valve seat damage is also precluded which : enhances 2 the valve reliability and lengthens the life of the valve..
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DISCUSSION 1-The proposed change revises -. Technieni; Specification. 3.2.8, s Pressurizer q
Pressure Limiting' Condition for Operation 1(LCO) to allow plant operation in ~
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Mode I with pressurizer pressure between 2025 ~and '2275 ps'ia.
TheseElower pressure limits are consistent with other CE plants.: ' Additionally, this-proposed' change - would reduce-- the. low-pressurizer pressure: Reactor-Protection System (RPS), Safety Injection Actuation _ System (SIAS), J and -
Containment Cooling Actuation J Systemi -(CCAS) trip setpoint and allowable.
values to 1717.4 and 1686.3 psia, respectively.
The -following.. discussion 'is. divided into three parts.. First,s ithe, pressurizer -pressure reduction.1 justification? is ; provided ' based on the-h Safety Analysis Report (SAR). Chapter 115 evaluations.and plant safety; system -
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Core Protection Calculatorj (CPC) and fCore - Operating. Limit- _ Supervisory -
verification. _ L ext,. the > clarification-to the ! PLilR System (COLSS). range N
Technical Specification. Basis ' is. ' discussed a Finally',- the-- proposed r low
_ pressurizer pressure RpS, SIAS, and.CCAS trip setpoint:and allowable valueL changes are presented.
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Pressurizer Pressure Reduction Safety analyses supporting Chapter 15 of the ANO SAR - presently bound plant operation with actual pressurizer pressure ;(i.e., including -pressure -
instrumentation measurement uncertainty) between 2200 and 2300; psia. These analyses were reviewed to identify which would - be adversely affected by plant operation at a lower pressurizer. pressure. Tabic 1 presents those events affected by lower pressurizer pressure.
The affected analyses identified in Table 1 have been _ reperformed assuming an initial pressurizer pressure of 2000 psia.' The plant response to)those events was simulated with the NRC approved CESEC-III-computer code.
Departure from nucleate boiling ratio (DNBR) analyses were performed based on the TORC computer code, the CE-1 critical heat flux correlation, and the CETOP code.
Calculation factors - were -' combined - statistically with. other ' uncertainty factors at the 95/95 confidence / probability level-for Cycle. 2 to define _ a specified acceptable fuel-design limit (SAFDL) nn1 the CE-1 minimum DNBR.
The DNBR limit was - again revised ' in Cycle 5 to 'directly incorporate NRC penalties and credit a reduced-rod bow penalty.
The Cycle.5 DNBR. limit of'
'1.25 remains applicable for current. analyses..Results were verified to be within acceptance criteria, including'the SAFDLs.
_No. fuel-cladding damage is predicted for any event, therefore,.no changes to the radiological doses-were calculated.
Over the past 9 cycles, changes have been made in some of:the conservative assumptions utilized in the accident analysis that'.is presented in T9bles
.i 2,
5, 7,
10, and 12.
These changes are a_"esult of changing' plant conditions, cycle specific parameters, conservative enhancements, crediting' 1
more limiting TS, and plant modifica t. ions.
This' proposed TS change only -
j relates'to the reduced RCS pressure; therefore, unrelated assumptions _will-not be - discussed in this submittal..
ASEA > Brwn ' Boveri - Combestion Engineering Nuclear Power maintains ' all the recorded calculations for the reanalyzed events._
These: calculations docu'ent T the conservative. input assumption bases and are available for review.
The Loss of External Load / Turbine Trip boundingf analysis was performed using an initial RCS pressure of 2000 psia.
A lower Linitial ' pressurizer pressure delays the_ reactor trip on-high pressurizer pressure,' allowing the pressurization rate to increase prior ' to-trip, -- and thereby forcing _ a greater _ pressure increase following the trip.
This event was' reanalyzed.
with the conservative _ assumptions - listed ' in Tablet 20 The : results n are presented in Table 3 and compared to Cycle :2 results in Table 4.
A peak' RCS pressure of 2744 psia -and - steam generator. pressure _ of 11160 psia were.
calculated.
These values are within 110. percent of thei design ' limit
. pressures, 2750 psia and 1210 psia - for the RCS and steam generator, respectively.
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An uncontrolled CEA bank' withdrawal event ~ from both suberitical _and 1%
power was evaluated.
The 100% power case is not-adversely af fected by the lower RCS pressure due to credit-taken for the CPC low DNBR trip, which as indicated below is valid over the proposed pressurizer pressure range. The subcritical CEA bank withdrawal evaluation, which is terminated by.the liigh Logarithmic Power Level-Trip, produced acceptable results. _ Table 5 lists the conssrvative assumptions used for the Cycle 10 analysis with resulta-presented in Tabic 6.
The analysis determined that at the time of-minimum DNBR, a three dimensional power peaking factor (Fq) of 5.78 corresponds to a DNBR of 1.25.
The analysis also determined that' for Fq < 10,.the_ fuel centerline temperature is-_less than l3450'F.
Since thi maximum Fq_
calculated for this. event is 4.5, the specified acceptable fuel design limits (DNBR 2 1.25 and fuel centerline temperature below 4900*F) are met.
CEA withdrawal from 1% power event' was also performed at an initial RCS-pressure of 2000 psia.
The liigh Prossurizer Pressure Trip was-originally credited in the Final Safety Analysis Report (FSAR analysis) to' terminate this event.
Since the original analysis, a Variable:Over: Power Trip (V0PT)~
has been added to the CPCs as part. of = the CPC Improvement Program implemented'during Cycle _5.
A list of the conservative' assumptions used in this analysis are given in Table 7.
The sequence of events is presented in Table 8 and the results la Table 9.
As--indicated,- the V0PT is the'first-trip encountered - and - terminates the reactor power excursion at-a lower.
level than previously calculated.
As a result, : the - minimum DNBR remains-above'1.8, and the maximum linear heat rate remains below 17 kw/f t, each 5
within the acceptance criteria.of 1.25 and 21: kw/ft.respectively.
Crediting the V0P1 more than offsets the sma111 adverse effects of. the lower RCS pressure; hence, acceptable "osults - were "obtained when the CEA bank withdrawal from 1% power was evaluated.
The V0PT trip condition for a CEA withdrawal event for 1% power will' occur prior to 40% power based on _ the current: CPC addressable constants,.
These constants are determined for each: cycle consistent-' with the sof tware ' and methodology-established in the~ CPC Improvement: Program (CEN-304-P, CEN-308-P, CEN-305-P, and CEN-310-P) and the-Lcurrent-cycle : design, performance, and safety analyses.
The CEA ejection events from llot Full Power; (liFP)' and-HotE Zero Power (IlZP) were both reevaluated utilizing a new lower __' limit' of -: 2000 psia.
The-STRIKIN-II computer prograin was, used in this 1 evaluation " to ' simulate the heat conduction within a reactor.' fuel rod and its associated surfac.e. heat, transfer.
Conservative ' assumptions ' used in. the CEA : ejection analysis are given in Table _10 and results are presented in - Table a 11;
- The. maximum.
centerline enthalpy decreaaed for both cases; however, 'a -slight-increase in '-
the number of fuel pins having; incipient centerline; melting for the IIFP was noted-(a. total of-0.32%);7but no fuel pins were calculated as having clad damage-or fully molten centerline.--llence, the results from this evaluation were considated acceptable.
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A single part length CEA (PLCEA) drop incident was reevaluated to determine-the effects of reduced RCS pressure.
Only positive reactivity insertions resulting from a ' PLCE A _ d r op _ a re of concern.
With the PLCEA insertion limits imposed by Section : 3.1.3.7 o_f-the ANO-2 Technical Specifications,-
positive reactivity insertions can only be postulated for PLCEA drops below 50% power.
It should be noted that the FSAR analysis was performed prior to the addition of TS section - 3.1.3.7 in which - 100% power was originally assumed to be conservative.
The positive reactivity increases core power, and consequently, pressurizer - pressure.
The_ pressure and _ power ' increases of the High Pressurizer Pressure Trip.
A are constrained by the action c reduction in the initial pressure can delay the high pressurizer: pressure trip,. thereby allowing e greater power increas e ', and a : correspondingly larger decrease in the fuel thermal margin.
However, sufficient initial-thermal margin will.. be preserved by 1the _ COLSS, _ which is verified _ every -
cycle in the reload analvees, to assure ' that the DNBR SAFDL is met throughout the PLCEA drop event.
The conservative assumptions for ' the PLCEA drop, which produce the maximum power increase that avoids the high pressurizer pressure trip are IIsted In Table 12.
The corresponding plant response is presented in Table 13.
The data and algorithms of the CPCs-were verified = to be _ valid for a range =
of pressurizer pressures which cover;thm proposed operating pressure range.
Additionally,-- the CPCs generate a Lreactor trip signal -if-the RCS pressure -
exceeds 2375 psia or drops below ;1860 psia.
COLSS monitors and initiates alarms if the LCOs on DNBli, peak linear heat rate, core power,- axial shape index, or core azimuthal tilt-are exceeded.
Only the DNBR LCO is dependent on pressurizer pressure.
An _ uncertainty factor is. applied in the COLSS cniculation - for DNBR to. account-for instrument uncertainty on the measured parameters (RCS temperature,.
- pressure, flow, etc.)
used 'as inputs to -the COLSS calculations.
Additionally, the'COLSS calculations-are benchmarked against more detailed-cstculations over the allowed operating i ranges.
-Differences. between the calculation results ;were also included in _ thWuncertainty factors. -The uncertainty factors have been reviewed and verified to be conservative over the proposed expanded pressurizer pressure range down : o '2000. psia.
2)
PLHR Clarification-The uncontrolled' CEA bank withdrawal from subcritical conditions analysis results indicated a transient. peak linear: heat rate l in excess of - the 21 kw/ft-limit given'in TS 2.1.1.2 (A PLHR less than-28 kw/ft and exceedlag.21'
. kw/ft for less than one secondywas calculated).
The ' intent of this TS is t.o prevent the. fuel from melting. during Lnormal _ operation : andi following1 anticipated -operational : occurrences.
iA limit of D 21 - kw/f t ; is _ specified -
based on fsteady state '; operation fuel centerline melting temperatures;-
hence, higher linear heat? rates can occur-under transient-' conditions without resulting in fuel molting'. As. indicated above, 'the fuel. centerline melt temperature was acceptable for the subcritical CEA' bank withdrawal. -A clarification to-the Bases of TS 2.1.1.2 is' proposed in this amendmect to 1
ensure the appropriate application of the peak-linear' heat rate limit.
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4 CPCs monitor linear heat. rate and analysis results are typically given in kw/ft, thereby making the peak linear heat rate limit of 21 kw/ft appropriate in most. situations.
- But, f t. t anticipated operational occurrences with transient peak ifnear heat rates, the more appropriate limit is the specified acceptable fuel design limit centerline molting temperature, which is the basis for the peak linear heat rate.
Thin clarification is consistent with other CE plant frterpretations of thin Technical Specification.
Additionally, this is consistent with the interpretation utilized for the CpCs as documented in the methodology and software manuals.
3)
Lower Low Pressurizer Pressure Setpoint The ANO-2 SAR briefly addresses an inadvertent operation of Emergency Core Cooling System (ECCS) during powet operatien.
In this scenario, an SIAS is-inadvertently actuated during power = operation.
_ Operating _ with the pressurimer pressure below approximately 1150 psia may result _ in an undesira'le SIAS following a reactor trip uom significant power ' levels.
The prouability of an-inadvertent-actuation of an SIAS during power
@ tation is not increased, but the likelihood of-receiving an' SIAS "dlowing a reactor trip is increased during reduced pressure operation, n should be noted that if an undesirable SIM is received following a rea-tor trip, the consequences of - that event would be _ bounded _ by the inadvertent actuation during power operathn.
1 Post-t rip pressure response from reduced initial pressurizer pressure is.
expected to be comparable to higher pressure trips, with the.- minimum post-trip. pressure being correspondingly lower.
These: lower post-trip.
responsen will come closer to th e_
low pressuriner pressure 1.o..SIAh>
setpoint. To minimize the impacts of-reduced pressure operation post-trip.-
previous - plant data was reviewed.to" determir.o Lthe L minimum - acceptable ptessurizer pressure fer. avoiding an SI AS post-trip.; 'Asca' result, a lower low pressurizer pressure trip setpoint is being. proposed.
All of the safety analyses identified -above :as. being ' adversely _. impacted by the reduced pressurizer. pressure were reevaluated down to'2000. psia. 'CPCs and i OLSS were also verified to be" applicable-down to : 20CO, pala.. Ilowever,.
due to the _ potential. for an undesirable SIAS actuatJonifollowing a r3 actor -
trip 'when operating with thof pressurizer. pressure = below 12150 : psia, = a-minimum continuous _ operating pressure of_2150; psia will be administrative'i controlled._ ;0perat, ton. down to: the : proposed.L ministim-pressurizer pressue limit-2025 psia will';also be-administrative 1yf controlled, to-short f.
.' durations',;thereby: allowing operator flexibility when_attemptingTto rescatn simmering - prenurizer L code:. safeties; yet1 minimize the - exposure 4 to-(ait undesirable SI AS actuation.
Operation above 2150 ; psia will: not ' increase' the probability of having an undesirable SIAS-(postatrip). The duration of1 steady state operation below 2150 psia L will _ be administrat.voly ' controlled to a.value-(in this. case 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)-that-will ensure that the probabilitytofE 1
inadvertent SIAS actuation sill ( not - be significantly) increased. 'The;2925
- psia -11mito is based ion; thel analysisi assumption.'of; 2000 psia: plus 25 pst which bounds pressure measurement uncertainties..
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Roductions in the. low pressurizor pressure RPS and Engineered Safety Featuro Actuation System (ESPAS) trip setpoint and allowabic values are also being proposed.
Those reductions will help prevent an undesirabic SIAS following a reactor trip when operating at reduced pressuren.
The new i
low pressurizer pressure notpoints and - allowable values are based on now instrument error calculations; the safety analysis setpoint assumptions nave not changed.
A reduct ion in the calculated ' instrument errot is duo primarily from the use of more ron11stic environar ntal assumptions at the time of actuation.
A new instrument error calculation has -been generated using ANO instrument error procedures.
- The statistical method of the square root of t120 sum of squares was used to determino the random error on 4
a component levn1 and for tho loop.-
Non-random errors were combined i
algebraically with the random error.torm to establish total nrror.
Although ANO has not committed to strict compliance with the. requirements..
of ISA-S67.04-1988 "Setpoints for Nuclear Sa fety-Rolatod ' instrumentation Used in Nuclear Power Plants," theno guidelines wero considered in calculating the loop orrors, periodic. test arrors (PTE), and - albwabin values associated with the low pressurizer pressure sotpoints... This calculation supports tho. proposui low pressurizer pressure setpoint = of e
1717.4 psia and allowable value of 1686.3spaia-for'the RPS, SIAS, and CCAS tt g functions.
r In the new instrument error calculation, eeveral noh-roa11 stile assumptions are removed with regard to the containment i conditions ; - specifically, the conditions et which the low' pressurizer pres _suro instruments. reach the trip setpoint. The original batrument error calculation conservatively assumed-worst case long-term harsh environment-insido containment based.on'a large break loss of coolant accident (IOCA).
In. reality these instruments would perform their function prior to the worst. caso conditions occurring.'= Based-on this, sma11' break LOCAs, large break LOCAs, and; steam lino' breaks (SLBs) were evaluated separately.
Soismic error terms wero-removed in 'all cases as a concurrent initiating event sindo this is beyond the, ANO design basis..
Rcdiatfori terms were also removed-due Lto tho quick L response of the pressurizer pressure following LOCAs.
No ifuni -damsgo 'isi prod [cted i
following a ShB and the primary system -reunins intact; hence, no radiation terms are assumed for a SLB.
The maximum containment temperatup-assumed for the LOCA instrument. error calculations was l'ased on. specific analyses and the availability -of the - redundant high. containment pressure trip instruments.. Crediting thosn more realistic containment conditions'results in-a better defined and more realistic. low pressurizer pressure setpoint.-
DETERMINATION OF SIGNIFICANT llAZARDS An analysis of the propos'ed changes han ' boen performod; int accordance with 10CFRSO.91(a)(1), regarding no significant ; hazards consideration using the standards-in 10CFR50.92(c).
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A discussion of those standards es they relate to this amendment request follows:
Criterion 1 - Does not involyn a Significant increano in the Probability or Consequences of an Accident Previously Evaluated.
Previously analyzed accidents and anticipated operational occurrences that are affected by this change have been reviewed.
This change has no impact on probability of occurrence of these accidents.
Pressurizer pressure and low pressurizer pressure setpoints are only used as input paramotors to the accident analyses which do not affect probability.
No physical changes to the plant are being proposed; therefore thorn is no impact on the probability of accidents.
One item, "Ina Nortent Operation of ECCS During Power Operation" is considered to o potentia'ly impacted sinco operation at RCS prosaures below 2150 pala may result in undesirable ECCS (SIAS) actuation post-trip (not actually at power).
This _ susceptibility ~ is not -considered a significant increase in probability cince the reduction-in pressure does not impact the probability of ECr', initiation signal generation, rather, if a plant trip occurred for any res,on and thn plant was operating at reduced pressures, the statistical combination of uncertainties shows that the SIAS initiation setpoint may be reached.
Since 'administrativo controls-will ensure that operation below 2150 psin is limited to very short durations (i.e. less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> steady state operati%n) any postulated increaso in prchability of an inadvertent operation of ECCS is not considered significant.
Previously evaluated accidents and anticipated operationD occurrences which were determined to be adversely impacted by thn red %ed 3.ro c rizer pressure have been evaluated with no significant increase 'in the consequences.
As indicated in the discussion, the ' SAFDI-and ~ acceptance criteria were veriflod to be maintained.-
Additionally, no' fuel cladding.
damage is predicted for any event and_ no ' changes to the radiological doses wera - calculated.
Thorofore, no ' increases _in. tho. consequences of. any accident are predicted.
J valunq: is Changing the low pressurizer pressurn _setroint -- and Lallowable based on the refinement of the instrument' error calculations.- No chango to
-tho analyzed events is. proposed duo ~ to. tha t new setpoint' and allowable values.
Those new limits still ensure 'the~ analysis assumptions are valid;
- hence, thern is no increase in the consequences:of the~ accidents.proviously evaluated.
gjiterion ? - Does Not Create the Possibility;of -a New or Different Kind of' Accident from-any Previously Evaluatod'.
' The _ proposad changes do not _ involve? any physical! modifications 1(i.e. : new
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systems, new-components, etc.) to the plant.1 The normal oporating.valun at(
which RCS pressure is hold: remains within the rangesntypical 'of. pressurized-water - reactors ' and _ moro Lspecifically, _ CE designed' nuclear steam supply 7 sy's tems.
The results of tho > accidenti re-analysoso-suggest no 1 dif forent phenomena or plant behavior'than previously_ considered.1 The change to tho-
- 5, Page.8
1 1
bases fc.c Technical Specification 2.1.1 are for c.larificetion and result in consistency with other CF, plants' approved Technical Specifications and
-j ses.
The low pressure setpoint change does not < create any new or different system actuations or interactions than evaluated previously. The i
slightly increased potential for SIAS initiatfori post-trip does not suggest l
new or different type accidents since inadvertent S1AS (at power) is i
already assessed in SAR Chapter 15 and is bounding for any post-trip SIAS.
Therefore, the proposed ci anges do not create the possibility of a new or differt.nt kind of accident from any previously evaluated.
l l
Criterion 3 - Does Not involve a Significant Reduction-in a Margin of Safety
-1 Any accident or anticipated operational' occurrence which was determined to -
^
be adversely impacted by the reduced pressurizer pressure was evaluated -to.
ensure acceptable results' are maintained.
The refined-instrument - error a
i calculations supporting the lownr low - pressurizer pressure setpoint and l
allowable values were verified to ensure the ; preAent accident _ analysis _
assumption are still niaintained.
Based on these evaluatfons, the proposed 4
changes do not involve any significant reduction in a margin of _ safety.
Rather, an overall increase in the margin of safety is anticipated, as discussed above, by increasing t he safety valve' re11 ability and decreasing tl.e exposure to risks of plant i utdown and cooldown for valve replacement-j j
or repair.
l
' CONCI'JSION Based on the above safety evaluation, it is concluded - that. the - proposed change does not c.onstitute a significant har.ards consideration as defined i
by 10CFRSO.92.
i c
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t TABLE 1 EVENTS AFFECTED BY LOWER PRESSURIZER PRESSURE 1.
Loss of External Lond/Turbino Trip i
1 2.
Uncontrolled Control Element Assembly (CEA)
H Withdrawal i
3.
Single Part Length CEA Drop 4.
CEA Ejection i
~
l TADt.E 2 CONSERVATIVE ASSUMPTIONS FOR.
LOSS OF EXTERNAL LOAD / TURBINE TRIP
-j Cycle 2 Cycle 10 Parameter Value -
Value Core Power, Hwt.
2900 2910-Core Inlet Temperature, 'F 540
- 540 L
RCS Pressure, psia 2200'
?2000 i
- Steam Generator Pressure, psia 610 796=
f i-CEA Worth at Trip, %AK/K
- 5. 4'
/-5.4:
Moderator Temperature 0.5 0.0 Coefficient, 10-* AK/K/*F-Number U-Tubes Assumed 0
~(
841 Plugged por Steam Generator.
r CLppler Coefficient Multiplier 0.85
.0.85; n
Pressurizer Safety, Valves
~2500 2525 e
Opening Pressure,' paia E
c-a I..
h
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_.. _. _._ _,. _. _ _.. _ _. _.. _. _ _ _ _ ~
d TABLE 3 SEQUENCE OF EVENT TOR Tile LOSS OF EXTERNAL LOAD / TURBINE TRIP Setpoint j
Time,_sec
, Event or Value 0.
Turbine Stop Valvos Close, and Main Feedwater Valyns Close 7.5 liigh Pressurizer Pressure Trip 2422 psfa-Condition Occurs i
8.4 Trip Breakers Open, Pressurizer Safety Valves Open 12525 psfa' 8.7 Main Steam Safety Valves--Open 1093 psla :
l t
l 9.0 CEAs Begin to' Drop _
10.6 Haximum RCS Pressure' Occurs
<2750 psia i
14.8 Pressurizer Safety Valves Close.
2424 psia 15.0-Peak Secondary Pressure Occurs
<1210 pain, I
, TABLE 4 LOSS OF EXTERNAL LOAD /TURDINE TRIP.
COMPARISON OF RESULTS r
Cycle 2 Cycle'10-Parameter Value-Value Maximum RCS Pressure, psia 2671
'2744 Haximum Steam Generator ~
'1144 11160 Pressure, psia P
Page lif
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=:
,n,
e TABLE 5 CONSERVATIVE ASSUMPTIONS FOR SUBCRITICAL CEA WITilDRAWAL Cycle 2 Cycle 10 Parameter Vclue
_Value.
Core Power, Hwt 2.9x10'1 1.52x10-'
Core Inlet Temperature, 'F 54A.6 545 RCP Pressure, psia 2200 2000 RCS Flow, gpm 322,000 321,200 Steam Generator (SG)_
990 1003 Pressure, psia CEA Worth'at Trip, %AK/K
-5,0
-1.44t Doppler Coefficient'Fultiplier 0.85 0.85 Moderetor Temperature 0.5 0.5 Coefficient, x10** AK/K/*F CEA Withdrawal Worth, 0.00025 0.00025 AK/K/second I
t Only the reactivity that was withdrawn was credited on reactorLtrip.
TABLE 6 SUBCRITICAL.CEA WITIIDRAWAL COMPARISON OF RESULTS-Cycle 2:
l Cycle'10 Paramotor
_Value-Valce Minimum DNBR 1.281
> 1,' 2 5*
Peak Fuel Centerline 3 BOO
.<3459**
Temperature, 'F
- Based on a cycle:10 Fq < 5.78.
'** Based on a-Fq"< 10.
t Page 12:
/
4 TABLE 7 CONSERVATIVE ASSUMPTIONS FOR 1% POWER CEA '/ITIIDRAWAL Cyclo 2 Cycle 10 Parameter Valun Value Core Power, Hwt 29 29 Coro Inlet Temperature, 'F
$44.6 545 RCS Pressure, psia 2200 2000 RCS Flow, gpm 322,000 321,200 Steam Generator Pressure, psia 978 994.
5.0
-5,5 CEA Worth at Trip, %AK/K Moderator Temperature 0.5 0.5 Coefficient, x10-' AK/K/'F Reactivity Addition, x10 '
1., 5
.1. 5 AK/K/second Doppler Coefficient Multipljer 0.85 0.85 i
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9m y
. -. - - -. ~. _.. -. -
)
TABLL 8 i
l SEQUENCE OF EVENT FOR Tile
)
CEA WITilDRAWAL (BANK) FROM 1% POWER EVENT 4
f Sotpoint Time, see Event or Valun 0.0 CSA Bank Begins to Withdraw from the Core 20.8 CPC V0PT Condition Occurs 40.0%-
i 21.4 Trip Breakers Open, l
i CEA Bank Withdrawal-is Torminated' 1
3 21.7 Maximum Peak Core Power, 50.4% of 2815 Mwt Hiximum PLllR Occurs
$21 kw/ft 22.0 CEAs Begin to Drop-l 22.3 Maximum Coro llent Flux, 34.5% of 2815 Hwt j
Minimum'DNBR Occurs 21.2S 25.1 Maximum RCS Pressurn Occurs
<<27$0 psia TABLE 9 CEA WITilDRAWAL FROM 1% POWER COMPARISON OF RESULTS Cyclo 2
/ Cycle-10 Parameter Va ly 2.,g.
Value
' 91 50.4 Maximum Coro Power
% of-2815 Mwt Maximum Core Hoat Flux 76 34.5-
% of 2815 Mwt
}
Maximum RCS Pressure, psia 2662
-2303 Maximum L11R, kw/ft 16.5 Minimum DNBR' l'.89
- These results from the Cycle 2 reload analyses woro not originallyL transmitted and are not readily available.
?
Page'14-9
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TABLE 10 1
CONSERVATIVE ASSUHPTIONS FOR Tl!E CEA EJECTION EVENTS Cycle 2 Cycle 10 Parameter Valun Value l
a i
Corn Power, Hwt.
l lIFP 2900 2910 IIZP 1
1 A
l-Core Inlet Temperature, 'F ilFP
$56,7 556.7 ilZP 544.6 545 RCS Pressure, psia _
'2200 2000 f-RCS Flow, gpm 322,000.
322,000 l
CEA Worth at Trip, %AK/K liFP
-5.4
-5.4
+
IlZP
-2.4
-2.4 1
Doppler Coefficient Multiplier 0.85 0.85 j.
Moderator Temperature Coefficient, x10
- AK/K/'F i
liFP 0.5 0.0 llZP 0.5 0, r-Ejected-CEA Worth, AK/K IIFP 0.0030~
.0.0028 IlZP 0.0082
.0.0082 l'
Ejected Peak, Fq i
IIFP 5.145-5.208 llZP 20.58 IL 71 s
Axial Power Peak Fz HFP 1.75 1.75.
IlZP -
2.5
' 2. 5 -~
Delayed: Neutron = Fraction D~,.
p Total i'
HFP.
O.00'482' 0.00536
!!ZP ~
0.00482-0.00536:
1+.
High Linear Trip Setpo' int, 124.8?
^128.1=
% of=2815 Hwt CEA Ejection Time, see HFP 0.05 0.05 IlZP.
0.05 0.05 PageLISL i:
4..
TADLE 11 CEA EJECTION RESULTS Cycle 2 Cycle 10 parametsr Value_
_Value Total Average Enthalpy of the Hottest Fuel Pin, cal /gm IIPP 156 157.2 IlZP 164 94.5 Total Conter11ne Enthsipy of the llottest: Fuel Pin,' cal /gm liFP 267 264.1 IlZP 296_-
141.9-Number of Fuel Pins llaving Clad Damage (E,,g(avg) 2 200 cal /gpm), %
liFP 0
0 llZP 0
0 Number of Fuel Pins llaving Incipient Centerline Halting (Etot(b) % 250 cal /gm), %
liFP (0.5 0.32 IlZP
$0.5.
0-Number of Fuel Pine liavi.1g Fully Holten Centerline Holting (E'ot(b) 2 310 cal /r,m), %
t
~ 11FP 0
0-IlZP 0'
0 4
i Tage'16L
TABLE 12 CONSERVATIVE ASSUMPTIONS FOR A SINGLE PART LENGTil CEA DROP Cycle 6 Cycle 10 Parameter Value Value Core Powtr, Hwt 1450*
1460 Core Inlet Temperature
'F-551 551 RCS Pressure, psia 2200 2000 RCS Flow, gpm 322,000 321,200 Steam Generator Pressure, psia 958 941 CEA Worth at Trip, %AK/K
-5,0:
Not Applicable Moderator Temperature 0.5-0.20 Coefficient, x10** AK/K/'F.
Reactivity Addition, %AK/K 0.04
.0.04-Nuuber U tubes Assum6d Plugged 0
841 per Steam Generator 4
- Yhe FSAR analysis was based on 100% power; hence is net included for-comparison.
'lABLE 13 NOMINAL SEQUENCE.0F EVENT FOR Tile PLCEA DROP FROM.50% POWER EVcNT.
Setpoint Time 'nec Event or Value 0.0 Single.PLCEA Drops into the Core' 90.4 Main Steam Safety Valves.Open
.1093 psia-107.8 Haximum RCS Presaure Occurs 52422 paia
~
ll 300.0
- Asymptotic Core Power and lleat 66'5% of' Flux Reached.
2815:Mwt "r
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