ML20108B823
| ML20108B823 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/16/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20108B815 | List: |
| References | |
| NUDOCS 8411160196 | |
| Download: ML20108B823 (19) | |
Text
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UNITED STATES y'
g NUCLEAR REGULATORY COMMISSION 3
4j VfASHINGTON, D. C. 20555 f
SAFETY EVALUATION BY THE OFFICE OF NUCLEC '!EACTOR REGULATION SUPPORTING AMENDMENT NO. 41 TO FACILITY OPERATING LICENSE NO. NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY OLD DC11NION ELECTRIC COOPERATIVE, NORTH ANNA POWER STATION, UNIT NO. 2 DOCKET NO. 50-339
==
Introduction:==
By-letter dated December 30, 1982 as supplemented by letters dated April 25, July 6, and July 11, 1983, the Virginia Electric and Power Company (the licensee) requested a change to the Technical Specifications (TS) to Facility
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Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station, Units No. 1 and No. 2 (NA-1&2). Also, by letter dated September 29, 1983, the licensee requested a change to the NA-1&2 TS.
Specifically, the licensee's requested change of December 30, 1982, as supple-mented, would revise the TS to allow operation with a Reactor Coolant System (RCS) Average Temperature of 587.8 degrees Fahrenheit ( F) as opposed to the currently approved RCS T,y of 582.8 F.
The licensee's requested change of September 29, 1983 would revise the NA-1&2 TS by changing the fractional thermal power multiplier from 0.2 to 0.3 with a RCS T,y of 587.8*F. Thus, the proposed change dated September 29, 1983 is germaie to the requested change dated December 30, 1982, as supplemented. Therefore, these two separate request changes are being evaluated as one specific licensing action at this time.
8411160196 841015 PDR ADOCK 05000339 P
PDR The requested change dated December 30,1982 (as supplemented) would implement Phase II of a NA-l&? plant upgrade program which would increase secondary
T '
steam pressura _in order to maximize'the electrical. output at the currently.
. licensed. reactor thermal power rating of 2775 Megawatts-thermal'(MWT).:
It is noted that the licensee's plant upgrade. program enveloping both a Phase I and Phase II plant upgrade would-increase the RCS.T,y by a total of 7.5'F;.
specifically 580.3 F to 587.F.
This total increase in T,y would increase secondary side steam pressure by 50 psi and result in a 5.6 MVA increase in :
electrical output. The licensee's Phase 'I plant upgrade increased the~ RCS T,y from 580.3*F to 582.8'F at the licensed reactor thermal power rating of- -
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2775 MWT. Implementation of the NA-182 Phase I Upgrade Program was approved at the time the Connission ~ issued the NA-1 Amendment No. 42 to License NPF-4 (with supporting ' safety analysis) on October 4,1982 and the NA-2 Amendment
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No. 32 to License NPF-7 'on October 19.-1983.
It is also noted that the licensee's proposed change relative to the Phase II upgrade is supported in appropriate cases by analyses covering the augmented f
change in the RCS T,y for both Phase I and Phase II representing a total f
change in temperature of 7.5'F even though the requested specific change for i
j Phase II covers a T,y change of 5 F; specifically from the NRC approved Phase I value of 582.8"F to the requested Phase II temperature of 587.8*F.
As stated previously, the proposed change would revise the TS to allow opera-tion with a (RCS) T,y of 587.8"F as opposed to the currently approved Phase I RCS T,y of 582.8'F.
In addition to increasing the RCS T,y by 5'F, the net reactor coolant pump heat input has been measured to be 12 MWT instead of 10 MWT, and this 2 MWT increase would change the currently approved Nuclear
. Steam Supply System (NSSS) rating from 2785 MWT to 2787 MWT. :TS changes l-i have been submitted related to the RCS T,y' safety limits, the Departure from
J 1
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ENucleate Boiling (DNB) parameters, and the Over Temperature Delta Temperature L(OTAT) and Over Pressure Delta Temperature (0 PAT) setpoints. TheLproposed change would also increase th'e TS value of core inlet volumetric flow rate based on actual measurements. The currently licensed reactor thermal. rating of 2775 MWT remains' unchanged.. The proposed 5'F change in the'RCS T,y would provide an increase in the secondary side steam pressure of approximately
- 32. pounds per square inch (psi) and result in a higher secondary cycle thermal efficiency and an approximate 3 MW electrical increase in. output.
TheLlicensee's safety evaluation supporting the licensee's proposed changes include the scope of the NSSS Accident Analyses and other. accident analyses specified in Chapter 15 of the NA-182. Final Safety Analysis Report (FSAR).
The safety evaluation also addressed the Balance of Plant (B0P) and NSSS/B0P Interfaces.
Reanalysis of the Emergency Core Cooling System (ECCS) performance.
and the Loss-of-Coolant Accident (LOCA) was performed to verify that the pro-posed changes and the analytical techniques used by~ the licensee were in full compliance with 10 CFR 50, Appendix K.
l Finally, the licensee's requested change of September 29, 1983 would revise l
the fractional thermal power multiplier from 0.2 to 0.3 with a RCS T,y of 587.8 F.
The proposed change would allow optimization of the core loading N
pattern by minimizing restrictions on the fractional power limit, FA, at low power.
On March 13, 1984 Phase II of the Plant Upgrade Program was implemented at NA-1 with the issuance of. Amendment No. 54 to Facility Operating License No. NPF-4 Although our Safety Evaluation supporting Amendment No. 54 stated
.- that we found the Phase II upgrade-to be applicable to'both NA-182, the issuance of an identical amendment for NA-2 was held in abeyance until the
- licensee could implement necessary feedwater valve trim at NA-2 during the Third Refueling Outage (Fall 1984).
By letter dated October.
,1984 the licensee stated that the necessary feed-water valve trim had been implemented at NA-2 to support the Phase II Upgrade Program. Therefore, we are issuing the Phase II Upgrade for NA-2_at this time.
Due to the passage of time since first approved for NA-1&2 and specifically implemented for NA-1 on March 13, 1984, we are restating our safety evalua-tion as originally provided for NA-1&2 to support the Phase II upgrade for NA-2 at this time. Our original discussion and evaluation in addition to our comments on the NA-2 feedwater valve trim are provided below.
I Discussion:
Reanalysis of LOCA and non-LOCA Accidents:
4 An increase in the RCS T,y will change the condition of the NSSS in several ways which can affect plant response to transients and accidents. The RCS subcooling will be reduced by 5 F, and along with it the margin to DNBR.
(This effect is partially offset by the fact that the core inlet flow is higher than previously assumed.) Stored energy in the reactor fuel and in the coolant will also increase proportionally.
Furthermore, the power defect in reactivity is increased.
Finally secondary steam pressure is increased by about 50 psi.
In lie R of these differencesy a reanalgs_is_of__LO_CA an_d_n_on _L_0CA accide_nts l
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w Accidents Not Reanalvzed-
. Several tratisients did not require' reanalysis'.j Transients 'at~ zero pcwer are unchanged? becauseLthe;T,y at. hot :zero. power. remains the same..Similarly,
- transients'which are indepen' dent of thermal-hydraulic 1(Fue1LHandling Accidents)
.and transients which have been shown to'be bounded by more serious' accidents (Ui1 controlled. Baron Dilution-at. Power) wereLnot' reanalyzed. LThe spurious actuation of safety ~ injection was not-reanalyzed:becausefthe original analysis hao shown that DNBR remains-above the initial value.throughout the transient.
Finally', steam generator, tube rupture was not recalculated because the prin-cipal impact'of increasing T w uld be a slight benefit due to increased av initial secondary steam pressure.
~
LOCA Reanalysis The NRC has recently accepted a large Break LOCA (LBLOCA) calculation submitted' for NA-1&2.
The analysis was performed with the approved "1981" Westinghouse.
eveluation.model, assuming F equal to 2.20 and 7% steam generator tube plugging.
0 A peak clad temperature of 2194.7 F was calculated.
The LBLOCA calculation
- submitted with the current amendment recuest used the same evaluation model and boundary conditions: with the following exceptions; (1) T was assuned eaual to 587.8*F instead of 582.8 F, (2) a thermal. design flow of 95,000 Gallons -Per tiinute (GPM) per loop was used rather than 92,800 GPM and (3) 5%
- steam generator tube pluggina was assumed ir place of 75.
The calculated peak' clad temperature is below 2200 F, and the other acceptance criteria of-10_CFR 50,46'are.satisfi@ds..
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The -assumption-of 5% tube plugging-is acceptable, but :as a consecuence, i
operation at T,yl equal to 587.8"F will-be _ permissible only up _to-5%- tube
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- pluoging instead of the previously. approved limit of 7%.
The small-break LOCA (SBLOCA) h'as been shown in previous : calculations to
' fall well - within the acceptance criteria of'10 'CFR -50'.46.
For. instance,-the-
- worst case $reak (3_ inch diameter) analyzed in the NA-1&2 FSAR yielded a; peak:
clad temperature of_1852*F. ~ Increased T could affec,t SBLOCA-in'two ways; 3y (1) more stored eneray in the primary system and (2) higher initial pressure-
- on-the secondary side
'Both of;these effects have minimal impact on SBLOCA, and consequently the licensee is justified in not reanalyzing:the accident.-
9 Non-LOCA Transients and Accidents-The reanalfsis of non-LOCA transients and accidents'was performed in confor-
. mance with the Standard Review Plan, using analytical methods'which have been approved by the staff.
Because increased T w uld lead' to higher stored energy in the primary system, av the change had little effect on transients involving increased heat renoval.
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I Accidental steam generator depressurization and minor steam line breaks are bounded by the major steam line break at hot zero power, for which the cal-p l
culated.DNBR does not drop below 1.30.
Accidents due to excessive load increase, and ' excessive heat removal due to feedwater malfunctions continue jo meet ' Standard Review Plan criterion of DNBR oreater than 130
For events involving decreased heat removal, the increase in T results in a av slightly lower calculated DNBR.
Nonetheless, the criterion for DNBR greater than 1.30 is still satisfied.
This category includes the loss-of load, loss-of-main feedwater and loss-of-offsite power transients.
For the 5 ore serious feedline rupture event, the primary pressure and temperature tran-sient is considerably iess severe than in the original FSAR.
This is pri-i marily due to taking credit for an auxiliary feedwater system design improve-ment which established a one-to-one relationship between auxiliary'feedwater pumps and steam generators.
As in the original FSAR, heat remcval by the auxiliary feedwater system is sufficient to prevent overpressurization of the Reactor Coolant System and prevent core uncovery.
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The complete loss of forced coolant flow accident continues to meet the DNBR criterion, even though violation of the limit is acceptablt for this class of accident.
The locked RCP rotor event yields sliphtly higher peak pressures and clad temperatures with increased T but the calculated results are still 3y, within acceptable limits.
These results are reasonable for a 5 F increase av*
y Accidental depressurization of the primary systen with the higher T leads a/
to a slightly lower calculated DNBR, but the DNBR criterion is still exceeded by a sizable margin.
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-8 Thermal Hydraulic Desian Evaluation of Coolant System Paraneters At rated thermal load, increasing the RCS T to 587.8 F on the primary side.
av of the steam generator tubes will increase the temperature of the steam on the secondary side by approximately 6.8 F, which corresponds to a 50 psi increase in steam pressure.
Table 1 provides a comparison of the current and proposed RCS temperatures and flow rates at rated thermal power.
Fron the table it can be seen that th.e reactor core thermal rating, pressure and "no load" temperature remain at current values.
The core inlet volumetric flow rate has been increased based on the actual per#ormance of the reactor coolant pumps.
The total core inlet thermal flow rate is the TS minimum flew limit utfrized for thermal and hydraulic analyses (e.g., DNB evaluations).
Based on NA-1&2 calorimetric data, the measured core inlet volumetric flow rate is 302,100 gpm with 2.8 percent of the steam generator tubes plugged.
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If the steam generator tube plugging level was increased to 5 percent, the measured flow would decrease by less than 1 percent.
The NA Units employ a calorimetric -aT method to determine the core inlet flow. rate.
For this flow measurement technique the maximum uncertainty in the total flow measure-ment is :2.0 percent.
Accounting fnr a 5 percent steam generator tube plugging level and the maximum flow measurement error of 2.0 percent, a total core inlet thernal flow rate of 785,000 gpm is conservatively low.
Therefore, a thermal flow rate of 285,000 gpm may be utilized as a design thermal flow rate for the proposed RCS T increase and in fact was used by th'e licensee av in their design analyses to set thermal limits.
The RCS T has been increased av from 582.8 F to 587.8 F.
The variations in inlet temoerature and temperature rises are attributable tr the thernodynamic properties of compressed liquid
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F waterandthe-increasedcoreinletvolumetriclflowrate.:The'overall.. impact
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of these charoes 'in the thermal: hydraulic. performance of the core has been
. evaluated and found.to be acceptable.
Confirmation of'W-3 ~ DNB Correlation Bounds -
The staff requested that the licensee confirm that the applicable range 'for the key parameters in the W-3 DNBR correlation bounds the Leonditions expected '
after increasing Tav to 587.8 F.
The licensee supplied Tables 2 and 3 and-associated references which demonstrate the applicability of W-3 for the proposed temperature conditions of the core. -Base'd on this data,1the staff.
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' finds that the key parameters in W 3, which have-been oreviously reviewed and approved by the staff, acceptably bound the thermal conditions anticipated after the increase in T,y.
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- 10 TABLE 1 COMPARISON OF REACTOR COOLANT SYSTEM PARAMETERS
' Thermal and Hydraulic Design Parameters Design Conditions
-Current Proposed.
2785 2787 NSSS Power, MWt 10 12 Net-Reactor' Coolant Pump Heat Input, MWt Reactor Core Heat Output, MWt 27,'5 2775 System Pressure, Nominal psia 2250 2250 System Pressure,, Min., Steady State, psia 2220 2220
' Total Core Inlet Thermal ~ Flow Rate, gpm 278,400 285,00b'
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6 6
Total Core In_let Thermal Flow Rate, lbm/hr 105.1 x 10 106.3 2 10
-Core Effective Flow Rate for Heat Transfer, 6
6 lbm/E'r 100.4 x 10 101.'5 x 10
. Reactor Coolant System Temperatures, F
Nominal Reactor Vessel / Core Inlet 546.9 '
555.5 Average Rise in Vessel 66.9 64.5 Average Rise in Core 69.7 67.2 Average in Core 583.6 591.1 Average in Vessel 580.3 587.8 No Load 547.0 547.0
' TABLE 2 W-3 CORRELATION LIMITS REF.
PRESSURE MASS EQUIV.
LOCAL AXIAL INLET CORRELATION NO.
RANGE VELOCITY DIAMETER OUALITY HEIGHT TEMP.
(psia)
(Mlb/h-f2)
(in.)
(in.)
( F)
W-3 1,2 1000-
- 1. 0.-
0.2-
-<0.18 10-
>400
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2400 5.0 0.7 144 F-factor 1,2 1000-1.0-0.2-2400 3.0 0.7
-<0.15 10-144
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Col dwall 1,2 1000-1.0-10.15
>10 Factor 3,4 2400 5.0 2er 3,4 1490-1.5-10.15 96-404-Tactor 2440 3.7 168 524 TABLE 3 CORE CONDITION WITH TAVG INCREASE core inlet temp. (*F) 555.5 nass velocity (mlb/h-fr) 2.442 pressure (psia) 2250 l
,. s. -. yy ; ! -
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~ ' 1.2-Containment ~ Safe'.y Maroin 8
The'following acceptance criteria.for subatmospheric containment. functional
~
design form'the basis for the licensee's evaluation of : containment safety.
margin for the ~uprated RCS Tavl conditions of the NSSS:
(1)
The calculated peak containment pressure shall not exceed the design pressure of 45 psig; (2) 'The containment shall be depressurized to below one atmosphere absolute pressure in less than 60 minutes; (3)
Once depressurized, the containment shall_ be maintained at a pressure less than one atmosphere absolute for the durationaof the accident.
The licensee has re-analyzed the postulated loss of coolant accident (LOCA) for the uprated NSSS conditions assuming a pump suction double ended rupture (PSDER), and evaluated the effect on the Net Positive Suction Head Available (NPSHA) for the Recirculation Spray (RS) and Low Head Safety Injection (LHSI) pumps.
The analysis results were compared'with the appropriate design criteria.
We conclude, based on these results. that the proposed uprated NSSS conditions will have a negligible impact on the contairment functional design.
i Subcomoartment analyses for the reactor bavity and steam generator and pres-
"suriker compartments were not redone.
The licensee's calculations confirm E
J
s
. 13 -
that, for.a-subcooled reactor coolant system, mass and energy releases' would 1
decrease with increased reactor coolantitpmperature.
Therefore, the' analyses
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documented in the NA-152 FSAR are bounding'for the uprated conditions.
We-concur with this, finding.-
The licensee did not reanalyze t'he main steam line break (MSLB) accident for the uprated conditions.
The current design basis'MSLB is a full.guil-
'lotine break at the no-load (hot shutdown) condition and this analysis remains unchanged for the uprated NSSS conditions.
Although there would be some additional energy relee e for a MSLB at power because of the uprated NSSS
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conditions, the no-load condition would remain the limiting case.
We concur.
with this finding since the steam generator inventory at no-load conditions would continue to dominate.any additional energy release that would occur
'or a MSLB at power.
Main Steam Svstem Consideration of the change in the RCS T for the main steam system involved 3y
' main steam safety valve capacity and main steam isniation capability.
The main steam safety valves have a total relievino capacity of 12.826.269 pe u,nd s per hour (Ib/hr) which is more thar the total uprated main steam t!ow ef' 12,251,367 lb/hr.
The main steam trip and non-return valves were ovaluated for ranid closure impact loads applied subsecuent to main steam svstem pipe ructure at uprated conditinns (increased steam presy.urel by tho licensee.
The results of the computer runs that modeled the transients effec
- nn the E
a i=
T i r valves showed that these valves would close as required without jeopardizing W
the integrity of the pressure boundary.
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V Auxiliary Feedwater System a
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E Consideratio, of the change in the RCS T for the auxiliary feedwater (AFU) av system invofved AFW ability to provide adequate flow for decay heat removal.
r The AFW pumps are designed to deliver rated flow to the steam generators at 9{
a static head equivalent to the set pressure of the lowest main steam safety Eg valve.
Because this setpoint pressure will not. change, the resistance param-EF eters associated with the 'AFW system will remain tile same, and this ARiflow
[
requirement (based on 2910 MWT core power plus 2%) for NA-182 remains unchanged.
b Therefore, the existing AFW system will be adequate _ at the uprated conditions.
EL
=P Condensate and Feedwater System
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m-e5 Consideration of the change in the RCS T f r the condensate and feedwater av system involved its isolation capability following transients and accidents.
The small decrease in feedwater pressure (by approximately 2 psi) does not 57 af#ect the closure. capability of the feedwater isolation valves, e
s Component Cooling and Service Water Systens Zn h
Consideration of the chance in the RCS T f r the component coolina system av
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and service water system involved their ability to remove beat # rom safety V
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-related ecufpment... The increased RCS col' leg temperature increas.es the d
heat -loadings.on the-component c'ool.ing sater _(CCW) system during normal
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' operating.conditionsdue;totheglightlyincreasedheatloadfromthe chemical and volume control system heat exchangers.
The affected heat ex-
- changers are the non-regeneration,iexcess : letdown'and seal water return heat exchangers.. The cumulative-heat loadings to the'CCW system at the.
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prated-operating conditions. remain 'less than the ' design.value used for the original plant design. -Heat removal capability.for safaty'relatedLequipment ccoled by the CCW system is~not affected by this change.
Consequently, the service water system is also not impacted by the uprating.
Soent Fuel Pool Cooling System There is no impact on the spent fuel pit heat loads as a result of the up--
rating sin'e core thermal power and the associated decay heat levels for c
spent fuel remain unchanged.
Fractional Thermal Power Multiplier The licensee has proposed to revise the TS by changing the fractional thermal power multiplier from 0.2 to 0.3 with a RCS T equal to 587.8 F.
The proposed av l>
change would allow optimization of the core loading pattern by minimizing restrictions on the' fractional power limit, Fa, at low power.
At. full power, the Fa limit will remain unchanged.
In the expression for Fs, as specified in the ffA-18 2 TS, Fa = 1.55 [1+0.3(1-P)1.
The proposed change would increase
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w the partial power multiplier from 0.2 to 0.3 in the expression above~ however, at full power, P becomes 1.0 and the multiplicative Leffect of the 0.3 partial.
N multiplier is zero (0).
The increase in the. fraction power Fa will be com-pensated for by more restrictive fractional power core thermal limits.
These more restrictive core thermal. limit lines will maintain the current design bases DNB criteria.
Analyses supporting the proposed change used analytical techniques consistent with North Anna design bases and previously NRC-approved Westinghouse fractional power multiplier analyses which are appropriately applied to NA-1&2.
Therefore, we find the proposed change.to be
' acceptable.
Evaluation:
Based on the above, we have determined that the licensee has satisfactorily reexamined the impact of increasing the RCS T, to 587.8*F for a full range of 3
transients and accidents.
We have further determined that the licensee's
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proposed change encompasses the analysis of all transients and accidents specified in the Standard Review Plan.
Although there is some loss of I
'ma rg f.: in many of the events, the relative ac'ceptance criteria are met.
In addition, all ac.cep,tance criteria of 10 CFR 50.46 are satisfied and the analvtical techniques as used by the licensee are in full compliance with 10 CFR 50, Appendix K.
We have also reviewed and evaluated the thermal-hydraulic aspects of the licensee's proposed change and conclude the proposed increase in RCS T and 3y
associated increase in core design flow rate are acceptable.
The licensee has provided acceptable documentation regarding containment functional design.
We have determined that the increase in the RCS T does not result in any av containment safety concern.
We have further reviewed the potential effects of the proposed change re-garding 80P/NSSS interfaces and find that predicted changes are small and are within the envelope of the approved NA-182 system design.
Finally, we have determined that increasing the partial power multiplier from 0.2 to 0.3 for a RCS T f 587.8 F will be compensated for by more av restrictive core thermal limits.
These limits will maintain the current DNB cri te ria.
In addition, the proposed change osed analytical techniques pre-viously approved by the NRC which are appropriately applied to NA-1&2 and therefore we find the proposed change to be acceptable.
Based on all of the above, we find the proposed change to be acceptable.
We further find that the proposed changes to the NA-2 TS regarding these natters are acceptable.
^ s noted above, the licensee's submittal of the large break LOCA calcui. ien submitted in support nf the proposed RCS T of 587.8 F assumed only P' <.toan 3
gene.*ator tube plugging.
Therefore, operation at a RCS T of 587.8*F is 3
ipornved for only up to 5% stean cenerator tube plucainn.
1 w
- We stated in our safety evaluation supporting the Phase 'II Upgrade for NA-1 (issued March 13,1984) that the Phas'e II Upgrade for NA-2 would be held in abeyance until such time that necessary feedwater valve + rim could be implemented at NA-2 to compensate for a decrease in feedwater valve opera-tional flexibility at the Upgraded Phase II conditions.
By letter dated
.5 October 3, 1984, the licensee stated that necessary feedwater valve trim
' modifications had been completed to support the NA-2 Phase II Upgrade. We requested that Region II inspection verify the completion of these modifica-tions. On October 4, 1984, we were so notified by Region II that the appropriate feedwater valve trim modifications were complete in support of the NA-2 Phase II Upgrade.
Therefore, based on all of the above, we find implementation of the Phase II Upgrade to be acceptable for NA-2.
Environmental Consideration:
This amendment involves a change in the installation or use of a facility component located within the restricted &rea as defined in 10 CFR Part 20.
The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in in-dividual or cumulative occupational radiation exposure.
The Commission has previously issued a proposed finding that this amendment involves no signifi-cant hazards consideration and there has been no public coment on such finding.
Accordingly, this amendment meets the eligibility criteria for categorical
c
)
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exclusion set-forth 'in '10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance'of this amendment.
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Conclusion:==
We have concluded, based on the considerations discussed above, that:
(1) there-is. reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regula-tions and th'e issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date: October 16, 1984 Principal Contributors:
L. Engle, DL/0RB#3 R. Barret, DSI/RSB l
G. Schwenk, DSI/CPB J. Guo, DSI/CSB R. Gael, DSI/ ASB l
l
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