ML20087P414
| ML20087P414 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 03/13/1984 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20087P410 | List: |
| References | |
| NUDOCS 8404060410 | |
| Download: ML20087P414 (19) | |
Text
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UNITED STATES j,j ! Sf ^n s
NUCLEAR REGULATORY COMMISSION j
WASHINGTON, D. C. 20555 ysg j j g..v...f SAFETY EVALUATION RY THE OFFICE OF NUCLEAR REACTOR REGtlLATION SUPPORTING AMENDMENT NO. 54 TO FACILITY OPERATING LICENSE NO. NPF-4 VIRGINIA ELECTRIC AND POUER COMPANY OLD DOMINION ELECTRIC COOPERATIVE NORTH ANNA POUER STATION, UNIT NO. 1 DOCKET NO. 50-338
==
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 Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station, Units No. I 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) Averace Temperature of 587.8 degrees Fahrenheit ( F) as opposed to the curtently approved RCS T f 582.8 F.
The licensee's requested chance of av September 29, 1983, would revise the NA-182 TS by changing the fractional thernal power multiplier from 0.2 to 0.3 with a RCS T f 587.8 F.
- Thus, av the proposed change dated September 29, 1983 is~gernane to the requested change dated December 30, 1982, as supplemented. Therefore, these two separate request changes are beina evaluated as one specific licensing action at this time.
6404060410 840313 PDR ADOCK 05000338 P
_PDR,
The requested change dated December 30, 1982 (as supplemented) would implement Phase II of a NA-182 plant upgrade program which would increase secondary steam pressure 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,olant upgrade would increase the RCS T,y by a total of 7.5*F, specifically 580.3 F to 587.8 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 1
electrical output.
The licensee's Phase I plant upgrade increased the RCS T
from 580.3 F to 582.8*F at the licensed reactor thermal power rating of l
av 2775 MWT.
Implementation of the NA-1&2 Phase I Upgrade Program was approved l
at the time the Consnission issued the NA-1 Amendment No. 42 to License NPF-4 (with supporting safety analysis) on October 4,1982 and the NA-2 Amendment 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 change in the RCS T,y for both Phase I and Phase II representing a total change in temperature of 7.5*F even though the requested specific change for Phase II covers a T,y change of 5 F; specificall,v 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 of 587.8 F as opposed to the currently approved Phase I av RCS T,y of 582.8 F.
In addition to increasing the RCS T by 5'F, the net av reactor coolant pump heat input has been measured to be 12 MWT instead of I
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 have been submitted related to the RCS T,y safety limits, the Departure from Nucleate Boiling (DNB) parameters, and the Over Temperature Delta Temperature (OTAT) and Over Pressure Delta Temperature (0 PAT) setpoints. The proposed change would also increase the TS value of core inlet volumetric flow rate based on actual measurements. The currently licensed reactcr thernal 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 hieher secondary cycle thermal efficiency and an approximate 3 MW electrical increase in output.
The licensee'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-l&2 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.
. Finally, the licensee's requested change of September 29, 1983 would revise the fractional thermal power multiplier from 0.2 to 0.3 with a RCS T f
av 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.
Our discussion and evaluation of these changes is provided below.
4 Discussion:
Reanalysis of LOCA and non-LOCA Accidents:
An increase in the RCS T will change the condition of the NSSS in several 3y 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.
Furthernore, the power defect in reactivity is increased.
Finally secondary steam pressure is increased by about 50 psi.
In light of these differences, a reanalysis of LOCA and non-LOCA accidents was submitted by the licensee for NRC staff review and approval.
, Accidents Not Reanalyzed Several transients did not require reanalysis.
Transients at zero power are unchanged because the T,y at hot zero power remains the same.
Similarly, transients which are independent of thermal-hydraulic (Fuel Handling Accidents) and transients which have been shown to be bounded by more serious accidents (Uncontrolled Boron Dilution at Power) were not reanalyzed.
The spurious actuation of safety injection was not reanalyzed because the original analysis had 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 would 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-182. The analysis was performed with the approved "1981" Westinghouse evaluation model, assuming F equal to 2.20 and 7% steam generator tube plugging.
g A peak clad temperature of 2194.7 F was calculated.
The LBLOCA calculation submitted with the current amendment request used the same evaluation model and boundary conditions, with the following exceptions; (1) T,y was assumed eaual to 587.8 F instead of 582.8*F, (2) a thermal design flow of 95,000 Gallons Per Minute-(GPM) per loop was used rather than 92,800 GPM and (3) 5%
steam generator tube pluggina was assumed in place of 7%. The calculated peak clad temperature is below 2200*F, and the other acceptance criteria of 10 CFR 50.46 are satisfied.
m.
4 l !
j The assumption of 5% tube plugging is acceptable, but as a consequence, operation at T,y equal to 587.8*F will be permissible only up to 5% tube j
plugging instead of the previously approved limit of 7%.
l The small break LOCA (SBLOCA) has been shown in previous calculations to fall well within the acceptance criteria of 10 CFR 50.46.
For instance, the o
%s worst case break (3 inch diameter) analyzed in the NA-182 FSAR yielded a peak.
clad temperature of 1852*F.
Increased T,y could affect SBLOCA in two ways; l
j (1) more stored energy in the primary system and (2) higher initial pressure i
j on the secondary side.
Both of these effects have minimal impact on SBLOCA, i
I and consequently the licensee is justified in not reanalyzing the accident.
t j
Non-LOCA Transients and Accidents The reanalysis of non-LOCA trans'ients and accidents was performed in confor-l mance with the Standard Review Plan, using analytical methods which have been approved by the staff.
Because increased T,y would lead to higher' stored energy in the primary system, the change had little effect on transients-involving increased heat renoval.
i Accidental steam generator depressurization and minor steam line breaks are i
bounded by the major steam line' break.at hot zero power, for which the cal-culated DNBR does not drop below 1.30.
Accidents due to excessive load-increase, 'and excessive heat removal due to 'feedwater malfunctions continue:
to meet Standard Review Plan criterion of DNBR greater'than 1.30.
l L
For events involving decreased heat removal, the increase in T,y results in a 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 more serious feedline rupture event, the primary pressure and temperature tran-sient is considerably less severe than in the original FSAR.
This is pri-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 removal by the auxiliary feedwater system is sufficient to prevent overpressurization of the Reactor Coolant System and prevent core uncovery.
The complete loss of forced coolant flow accident continues to meet the DNBR criterion, even though violation of the limit is acceptable for this class of accident. The locked RCP rotor event yields slightly higher peak pressures and clad temperatures with increased T,y, but the calculated results are still within acceptable limits. These results are reasonable for a 5'F increase in T,y.
Accidental depressurization of the primary system with the higher T leads av to a slightly lower calculated DNBR, but the DNBR criterion is still exceeded by a sizable margin.
L
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1 N
a t
, Thermal Hydraulic Design Evaluation of Coolant System Parameters 4
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.
From i
the table it can be-seen that the 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 performance of the reactor coolant pumps.
The total ccre inlet thermal flow rate is the TS minimum flow limit utilized for thermal and hydraulic analyses (e.g., DNB evaluations).
Based on NA-182 calorimetric data, the measured core inlet volumetric flow 3
rate is 302,100 gpm with 2.8 percent of the steam generator tubes plugged.
I 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 4
calorimetric -aT method to determine the core inlet flow rate.
For this i
flow measurement technique the maximum uncertainty in the total flow measure-ment is 2.0 percent. Accounting for a 5 percent steam generator tube plugging.
level and the maximum flow measurement error of 2.0 percent, a total core inlet thermal 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,y increase and in. fact was used by the licensee in their_ design analyses to set thermal limits.
The RCS T,y.has been increased from 582.8*F to 587.8*F. The variations in inlet temperature and temperature rises are attributable to the thermodynamic properties of compressed liquid
.o.
water and the increased core inlet volumetric flow rate. The overall impact of these changes 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 conditions expected after increasing T,y 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.
Based on this data, the staff finds that the key parameters in W-3, which have been previously reviewed and approved by the staff, acceptably bound the thermal conditions anticipated after the increase in T,y.
I
. TABLE 1 COMPARIS0N OF REACTOR COOLANT SYSTEM PARAMETERS Thermal and Hydraulic Design Parameters Design Conditions Current Proposed NSSS Power, MWt 2785 2787 Net Reactor Coolant Pump Heat Input, MWt 10 12 Reactor Core Heat Output, MWt 2775 2775 l
System Pressure, Nominal psia 2250 2250 l
System Pressure, Min., Steady State, psia 2220 2220 Total Core Inlet Thermal Flow Rate, gpm 278,400 285,000 6
6 Total Core Inlet Thermal Flow Rate, 1bm/hr 105.1 x 10 106.3 x 10 Core Effective Flow Rate for Heat Transfer, 6
6' r
lbm/hr 10C.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 i
l I
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. TABLE 2 W-3 CORRELATION LIMITS l
REF.
PRESSURE MASS EQUIV.
LOCAL AXIAL INLET
. CORRELATION NO.
RANGE VELOCITY DIAMETER QUALITY HEIGHT TEMP.
(psia)
(M1b/h-f2)
(in.)
(in.)
(*F)-
W-3 1,2 1000-1.0-0.2-1 18 10-
>400 0
2400 5.0 0.7 144 F-factor 1,2
.1000-1.0-0.2-
-<0.15 10-2400 3.0 0.7 144 Coldwall 1,2 1000-1.0-10.15
>10 Factor 3,4 2400 5.0 Spacer 3,4 1490-1.5-10.15 96-404-Factor 2440 3.7 168 624 TABLE 3 CORE CONDITION WITH TAVG INCREASE core inlet temp. (*F) 555.5 mass velocity (mlb/h-f2) 2.442 pressure (psia) 2250-
. Containment Safety Marcin 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 T conditions of the NSSS:
av (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 duration of 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, th;t the proposed uprated NSSS conditions will have a negligible -impact on the containment functional design.
~
Subcompartment analyses for the reactor cavity and steam generator and pres-surizer compartments were not redone.. The licensee's calculations confirm
that, for a subcooled reactor coolant system, mass and energy releases wculd decrease with increased reactor coolant temperature.
Therefore, the analyses documented in the NA-182-FSAR are bounding for the uprated conditions.
We concur with this finding.
The licensee did not reanalyze the 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 release for a MSLB at power because of the uprated NSSS conditions, the no-load condition would remain the limiting case. We concur J
with this finding since the steam generator inventory at no-load conditions would continue to dominate any additional energy release that would occur for a MSLB at power.
Main Steam System Consideration of t'he change in the RCS T f r the main steam system involved av main steam safety valve capacity and main steam isolation capability.
The main steam safety valves have a total relieving capacity of 12,826,269 pounds per hour (lb/hr) which is more than the total uprated main steam flow of 12,251,367 lb/hr. The main steam trip and non-return valves were evaluated for rapid closure impact loads applied subsequent to main steam system pipe rupture at uprated conditions (increased steam pressure) by the licensee.
The results of the computer runs that modeled the transients effect on the
_ 14 valves showed that these valves would close as required without jeopardizing the integrity of the pressure boundary.
Auxiliary Feedwater System Consideration of the change in the RCS T for the auxiliary feedwater (AFil) av system involved AFW ability to provide adequate flow for decay heat removal.
The AFW pumps are designed to deliver rated flow to the 5, team generators at a static head equivalent to the set pressure af the lowe";t main steam sufety valve. Because this setpoint pressure will not change, the resistance param-eters associated with the AFW system will remain the same, and this AFW flow i
l l
requirement (based on 2910 MWT core power plus 2%) for NA-182 remains unchanged.
I Therefore, the existing AFW system will be adequate at the uprated conditions.
f Condensate and Feedwater System Consideration of the change in the RCS T,y for the condensate and feedwater system involved its isolation capability following transients and accidents.
The small decrease in feedwater pressure'(by approximately 2 psi) does not' affect the closure capability of the feedwater. isolation valves.
Component Coolino and Service Water Systems:
Consideration of the change in _the RCS T f r the component cooling system, av and service water system. involved their ability to remove heat from safety --
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.h*------
^ --'-
B I
i related equipment.
The increased RCS cold leg temperature increases the heat loadings on the component cooling water (CCW) system during normal operating conditions due to the slightly incre.. sed heat load from the chemical and volume control system heat exchangers.
The affected heat ex-changers are the non-regeneration, excess letdown and seal water return heat exchangers. The cumulative heat loadings to the CCW system at the uprated operating conditions remain less than the design value used for the original plant design. Heat removal capability for safety related equipment cooled by the CCW system is not affected by this change.
Consecuently, the service water system is also not impacted by the uprating.
Spent Fuel Pool Cooling System There is no impact on the spent fuel pit heat loads as a result of the up-rating since core thermal power and the associated decay heat levels for 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 change would allow optimization of the core loading pattern by minimizing N
restrictions on the fractional power limit, Fa, at low power. At full power, the Fa" limit will remain unchanged.
In the expression for Fa, as specified N
N in the NA-1& 2 TS, FA = 1.55 [1+0.3(1-P)).
The proposed change would increase l
t I
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 effect 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 av transients and accidents. We have further determined that the licensee's proposed change encompasses the analysis of all transients and accidents specified in the Standard Review Plan. Although there is some loss of margin in many of the events, the relative acceptance criteria are net.
In addition, all acceptance criteria of 10 CFR 50.46 are satisfied and the analytical techniques as used by the licensee are in full compliance with' 10 CFR 50, Appendix X.
We have also reviewed and evaluated the thermal-hydraulic aspects of the licensee's proposed change and conclude the proposed increase in RCS Tav.and
. associated increase in core design flow rate are acceptable. The licensee has provided acceptable documentation regarding containment functional design.
We have deternined that the increase in the RCS T,y does not result in any containment safety concern.
We have further reviewed the potential effects of the proposed change re-garding BOP /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,y of 587.8 F will be compensated for by more restrictive core thermal limits. These limits will maintain the current DNB criteria.
In addition, the proposed change used analytical techniques pre-viously approved by the NRC which are appropriately applied to NA-182 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-182 TS regarding these matters are acceptable.
As noted above, the~ licensee's submittal of the large break LOCA calculation submitted in support of the proposed RCS T,y of 587.8 F assumed only 5% steam generator tube plugging. Therefore, operation at a RCS T,y of 587.8"F is approved for only up to 5% steam generator tube plugging.
. Finally, it is noted that the above safety evalution is for both NA-182.
However, at this time, the proposed change is applicable to NA-1 only. The licensee has noted that the 80P review for the Phase II upgrade conditions I
at NA-2 identified a decrease in feedwater valve operational flexibility at the uprated conditions.
Necessary modification in feedwater valve trim will be completed during the NA-2 Third Refueling Outage (Fall 1984).
Therefore, issuance of the Phase II upgrade program for NA-2 will be held in abeyance until such modifications are completed and verified by the NRC.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 651.5(d)(4), that an environmental impact statement or negative declara-l tion and environmental impact appraisal need not be prepared in connection l
with the issuance of the amendment.
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 regulations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Date:
March 13, 1984 Principal Contributors:
L. Engle, DL/0RB#3 R. Barret, DSI/RSB G. Schwenk, DSI/CPB J. Guo, DSI/CSB R. Goel, DSI/ASB
,