ML20028E040

Forwards Addl Info Re Analysis of Min Containment Pressure. Revised Response to Question 022.24 & Revs to FSAR Sections 6.2,15.6 & 15.B. Encl.Rev Necessary Due to Changes in Initiation Times of RCFC & Containment Spray

ML20028E040
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	01/14/1983
From:	Tramm T COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
5800N, NUDOCS 8301200285
Download: ML20028E040 (10)

Text

I

/^\ Commonwealth Edison

' ) one First N:tional Plaza. Chicago, lihnois

. O Address R ply to: Post Office Box 767 Chicago. Illinois 60690 January 14, 1983 Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Minimum Containment Pressure Analysis NRC Docket Nos. 50-454, 50-455, 50-456, and 50-457

Dear Mr. Denton:

This is to provide additional information regarding the analysis of minimum containment pressures at Byron /Braidwood. NRC review of this information should close Confirmatory Issue 17 of the Byron SER.

Enclosed is a revised response to Question 022.24 and revisions to FSAR Sections 6.2, 15.6, and 15.B. These revisions were necessary due to changes in the initiation times of the RCFCs and containment spray used in the minimum containment pressure analysis for ECCS performance evaluation. Computer drawn curves corresponding to FSAR Figures 15.6-7, 9, 10, 14, 15, 18, and 20 were also generated in the reanalysis. The differences between the revised curves and the curves presently in the FSAR are negligible.

Therefore, revised curves will not be submitted. However, the figures presently in the FSAR will be resubmitted, along with the enclosed pages, in Amendment 41 in order to completely document that the reanalysis was performed.

One signed original and fifteen copies of this letter and the enclosures are provided for your review. Please direct further questions to this office.

Very truly yours, O ,..

l 7 "h[} &

T. R. Tramm Nuclear Licensing Administrator bgl 1m Attachment 8301200285 830114 PDR ADOCK 05000 5800N

)

B/B-FSAR .

QUESTION 022.24

" Provide the assumed essential cooling wate'r temperature l used in FSAR Table 6.2-25 to verify that the minimum ]

essential service water temperature has been used to  !

maximize the heat removal capacity of the reactor con-

. tainment fan coolers used in the minimum containment pressure analysis for ECCS performance evaluation (Ref-erence Branch Technical Position CSB 6-1) ."

RESPONSE ,

A reanalysis has been performed for the Byron /Braidwood

. Stations, reflecting revised fan cooler performance based on a reduction in the minimum essential service cooling water temperature to 45' F. The large break LOCA analysis -

performed is identical to the previously docketed (Amendment ,

32) analysis for the worst break except for the use of revised fan cooler data and for the use of more accurate fuel temper-aturg data. The containment f an cooler start time used in this revised analysis is 15 seconds vs.45 seconds in the original analysis. Westinghouse has obtained NRC approval (Reference 1) for use of the more accurate fuel temperature data on a generic basis. Figure 6.2-25 has been revised to reflect performance of the containment fan cooler with an inlet cooling water temperature of 45' F..

F'igures 6.2-24 and 15.6-12 have been revised to show the minimum ECCS containment backpressure results for the revised analysis. The revised containment pressure results are essentially identical to the previous results for the first 100 seconds, then begin a gradual decline and are approximately 0.5 psig lower than the previous results at 400 seconds.

Table 15.6-3 has been revised to incorporate the revised large break LOCA results. Since the peak clad temperature occurs at 133 seconds, and there is very little change in the pressure transient up to this time, the effect on the

.large break LOCA results is minimal. The 7' F benefit in peak clad temperature is due to the more accurate fuel temper-ature data.

Reference:

1. Letter f rom John F. Stol: to Thomas M. Anderson, March 27, 1980.

Q22.24-1

B/B-FSAR TABLE 6.2-52

.DECLG BLOWDOWN MASS AND ENERGY RELEASES (CD= 0 . 6 )

TIME MASS FLOW ENERGY FLOW (sec) (lbm/sec) (million/ Btu /sec) 42.081 0.00 0.000 42.706 .02 .030 42.906 .02 .030 43.006 .02 .030 43.206 .02 .030 43.406 .81 1.058 47.850 33.58 43.856 56.562 3368.91 422.893 70.462 293.42 199.662 87.562 313.92 198.247 106.'662 327.57 193.452 127.562 338.90 187.620 174e.862 360.23 174.348 231.162 383.04 159.612 304.762 415.20 144.088 467.862 459.33 121.602 I

~

6.2-162 r ------- - - - , - - - - - - - -

pm__

?

B/B-FSAR TABLE 6.2-54 .

ACTIVE HEAT SINK DATA FOR MINIMUM POST-LOCA CONTAINMENT PRESSURE I Containment Spray System Parameters A. Maximum spray system flow, total 8118 gpm D. Fastest post-LOCA initiation of spray system Assuming offsite power loss at start of LOCA 35 see II Containment Atmosphere Recirculation Fan Coolers ,

A. Maximum number of fan coolers operating 4 B. Fastest post-LOCA initiation

{

Assuming offsite power loss at start of LOCA 15 sec C. Performance data j See Figure 6.2-25 for fan cooler temperature versus heat load curve.

/

i 6.2-164  !

B/B-FSAR

d. Mass released to Containment during blowdown. (Figure 15.6-16)

e. Energy released to Containment during blowdown.

(Figure 15.6-17)

f. Fluid quality in the hot assembly during blowdown.

(Figure 15.6-18)

g. Mass velocity during blowdown. (Figure 15.6-20)

h. Accumulator water flow rate during blowdown. (Figure f 15.6-19)

1. Pumped safety injection water flow rate during l reflood. (Figure 15.6-21) 1 The maximum clad temperature calculated for a large break is l

2096 *F which is less than the Acceptance Criteria limit of 2200*F of 10CFR50.46. The maximum local metal-water reaction is 5.46%, which is well below the embrittlement limit of 17% as l required by 10CFR50.46. The total core metal-water reaction is

, less than 0.3% for all breaks, as compared with the 1%

criterion of 10CFR50.46, and the clad temperature transient is i terminated at a time when the core geometry is still amenable to cooling. As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.

Small Break Results As noted previously, the calculated peak clad temperature resulting f rom a small break LOCA is less than that calculated for a large break. Based on the results of the LOCA sensi-tivity studies (Ref erence 21) the limiting small break was found to be less than a 10 inch diameter rupture of the RCS cold leg. Therefore, a range of small break analyses are presented which establishes the limiting break size. The results of these analyses are summarized in Tables 15.6-1 and 15.6-4.

Figures 15.6-34 through 15.6-47 present the principal parameters of interest for the small break ECCS analyses. For all cases analyzed the following transient parameters are presented:

a. RCS pressure. (Figures 15.6-34, 15.6-41, 15.6-42)

b. Core mixture height. (Figures 15.6-35, 15.6-43, 15.6-44) 15.6-21 y . - _ _ _ m  % - ,. ,. --- - , . , . -

B/B-FSAR TABLE 15.6-3 LARGE BREAK LOCA RESULTS FUEL CLADDING DATA l'

CD = 0.8 C = 0.6 C = 0.4 DECLG bECLG bECLG Results for N Loop i Peak clad temperature (*F) 2016 2096 1657 Peak clad temperature location (ft) 7.5 7.5 7.25 Local Zr/H 2 O reaction, maximum (%) 4.29 5.46 0.40 Local Zr/H2 O location (ft) 7.5 7.5 7.25 j i

Total Zr/H 2O reaction (%) <0.3 <0.3 <0.3 Hot rod burst time (sec) 56.4 40.6 152.86 i

Hot rod burst location (ft) 6.25 6.25 7.25 j l

1 1

i t

I t

i i

I l

15.6-35 i

, B/B-FSAR CORE PEAKING FACTOR 2.3.

HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAD - 1758.4*F = PCTB  !

ELEVATION - 6.25 feet.

HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR A NON-RUPTURED REGION OF THE CLAD - 2096.4 *F = PCTN ELEVATION - 7.5 feet CLAD STRAIN DURING BLOWDOWN AT THIS ELEVATION 0.224%

MAXIMUM CLAD STRAIN AT THIS ELEVATION - 5.073%

Maximum temperature for this non-burst node occurs when the core reflood rate is less than 1.0 inch per second and reflood heat transfer is based on the (STEAM COOLING) calculation.

AVERAGE HOT ASSEMBLY ROD BURST ELEVATION - 7.5 feet HOT ASSEMBLY BLOCKAGE CALCULATED - 0.0%

1. BURST NODE The maximum potential impact on the ruptured clad node is expressed in letter NS-TMA-2174 in terms of the change in

. the peaking factor limit (FQ) required to maintain a peak clad temperature (PCT) of 2200*F and in terms of a change in PCT at a constant FQ. Since the clad-water reaction rate increases significantly at temperatures above 2200*F, individual effects (such as APCT due to changes in several fuel rod models) indicated here may not accu-rately apply over large ranges, but a simultaneous change in FO which causes the PCT to remain in the neighborhood of 2200*F justifies use of this evaluation procedure.

From NS-TMA-2174:

For the Burst Node of the clad:

0. 01 A FQ + # 150*F BURST NODE APCT

- Use of the NRC burst model and the revised Westinghouse burst model could require an FQ reduction of 0.027

- The maximum estimated impact of using the NRC strain model is a required FQ reduction of 0.03.

15.B-3

e B/B-FSAR Therefore, the maximum penalty for the Hot Rod bu.mt node is:

APCT1= (0.027 + .03) (150*F/.01) = 855'F Margin to the 2200*F limit is:

APCT2 = 2200*F - PCTB = 441.6*F l The FQ reduction required to maintain the 2200*F clad temper-ature limit is:

AFO B (APCT y - APCT ) (.01 AFQ))

2 150*F

=

(855 - 441.6) ( . 01)9

" 0.02756 (but not less than zero).

2. NON-BURST NODE The maximum temperature calculated for a non-burst section of clad typically occurs at an elevation above the ccre mid-plane during the core reflood phase of the LOCA tran-sient. The potential impact on the maximum clad tempera-ture of using the NRC fuel rod models can be estimated by examining two aspects of the analyses. The first aspect is the change in pellet-clad gap conductance resulting from a difference in clad strain at the non-burst maximum clad temperature node elevation. Note that clad strain all along the fuel rod stops after clad burst occurs and use of a different clad burst model can change the time at which burst is calculated. Three sets of LOCA analysis results were studied to establish an acceptable sensitivity to apply generically in this evaluation. The possible PCT increase resulting from a change in strain (in the Hot Rod) is +20*F% decrease in strain at the maximum clad tempera-ture locations. Since the clad strain calculated during the reactor coolant system blowdown phase of the accident is not changed by the use of NRC fuel rod models, the maximum decrease in clad strain that must be considered here is the difference between the " maximum clad strain" and the " clad strain at the end of RCS blowdown" indicated above.

Therefore:

~~ ' '

15.B-4

- . _ _ = _ _ _

B/B-FSAR APCT 3

=

(0 strain) ( AN- LOMTN STRAIN) 2

=

(0)01(0.05073 - 0.00224)

= 96.98'y The second aspect of the analysis that can increase PCT is the flow blockage calculated. Since the greatest value of blockage indicated by the NRC blockage model is 75%, the maximum PCT increase can be estimated by assuming that the current level of blockage in the analysis (indicated above) is raised to 75 percent and then applying an appropriate sensitivity formula shown in NS-TMA-2174.

Therefore,

APCT 4

= 1.25'F (50 - % CURRENT BLOCKAGE)

+ 2.36*F (75-50)

= 1. 25 - (50-0) *F + 2.36 (75-50) *F

. = 121.5'F The total potential PCT increase for the non-burst node is then:

APCTS = APCT3 + APCT4 = 9 6 . 9 8 + 121. 5 = 218 . 4 8 'F l

Margin to the 2200*F limits is APCT6 = 2200*F - PCTN = 2200 - 2096.4 = 103.6*F l

'The FQ reduction required to maintain this 2200*F clad temperature limit. is (from NS-TMA-2174)

AFQ N =

(APCT 5 - APCT6 ) ( ) - (218.48 -103.6) (0.01) 9 = -0.115 l - i 10'F APCT l

l

= 0.115 AFQ N

i l

l 15.B-5

. B/B-FSAR The peaking factor reduction required to maintc$n the 2200'F clad temperture limit is therefore the greater of AFQB and AFQN '

or; AFQPENALTY = 0.115 l B. The effect on LOCA analysis results of using improved analytical and modeling techniques (which are currently approved for use in the Upper Head Injection plant LOCA analyses but which are also applicable to Byron /Braidwood) in the reactor coolant system blowdown calculation (SATAN computer code) has been quantified via an analysis which has recently been submitted to the NRC for review. Recog-nizing that review of that analysis is not yet complete and that the benefits associated with those model improvements can change for other plant designs, the NRC has estab-lished a credit that is acceptable for this interim period to help offset penalties resulting from application of the NRC fuel rod models. That credit for two, three and four loop plants is an increase in the LOCA peaking factor limit of 0.12, 0.15 and 0.20 respectively 0.20 - 0.0384 = 0.1616 > 0 C. The peaking factor limit adjustment required to justify plant operation for this interim period is determined as the appropriate AFQ credit identified in section (B) above, minus the AFQPENALTY calculated in section (A) above (but not greater thaa zero)

FQ ADJUSTMENT = 0.2 - 0 .115 = 0 . 0 8 5 > 0 , l Therefore FQ ADJUSTMENT = 0.0.

15.B-6