ML20093J554

Large Break LOCA-ECCS Analysis W/ Increased Enthalpy Rise Factor

ML20093J554
Person / Time
Site:	Robinson
Issue date:	07/12/1984
From:	Tahvili T SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML14188B047	List: ML20093J549 ML20093J554
References
XN-NF-84-72, NUDOCS 8407300178
Download: ML20093J554 (75)

Text

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, X N - N F 72

. H. B. ROBINSON UNIT 2 LARGE7 BREAK LOCA-ECCS AN ALYSIS

. .W-lTH INCREA. SED ENTH ALPY RISE FACTOR 1~

l U LY 1984

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RICHLAND WA 99352 m

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r ERON NUCLEAR COMPANY, INC.

l XN-NF-84-72 Issue Date: 7/12/84 H. B. ROBINSON UNIT 2 LARGE BREAK LOCA-ECCS ANALYSIS WITH INCREASED ENTHALPY RISE FACTOR Prepared by: N 22[ty.d T. Tahvili 7/ /d/ft/

PWR Safety Analysis Reviewed by: nyg*

C. E. Leach, Project Manager *

T /r /9[

PWR Safety Analysis

'M bW Concur: 7. / c/ F "

W. V. Kayser, Manager PWR Safety Analysis Concur: -

.rav J. C. Chandler, Lead Engineer Reload Fuel Licensing Approve: S (F i) ht: t, f y

'R. B. Stout, Manager t/

Licensing & Safety Engineering Approve: *NM& // td4'_ IP G. A. Sofer; nager /

Fuel Engi ring & ' Technical Services naa ERON NUCLEAR COMPANY,Inc.

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NUCLEAR REGULATORY COMMISSION DISCLAIMER IMPORTANT NOTICE REGARDING CONTENTS AND USE OF THIS DOCUMENT PLEASE READ CAREFULLY This technical rrport was tierived through research and development progroms sponsored.by Exxon Nuclear Company, Inc. It is being sub-mitted by Exxon Nuclear to the USNRC as part of a technical contri-bution to facilitate safety analyses by licensees of the USNRC which utilize Exxon Nuclear fabricated reload fuel or other technical services provided by Exxon Nuclear for liaht water power reactors and it is true and correct to the best of Exxon Nuclear's knowledge, information, and belief. The information containd herein may be used by the USNRC in its review of this report, and by licensees or applicants before the USNRC which are customers of Exxon Nuclear in trair demor.stration of comoliance wi th the USN RC's regulations.

Without devogeting from the foragoing, neither Exxon Nuclear nor any person acting nn its behalf:

A. Makes any warranty, express or implied, with respect to the accuracy, completeness, or usefultiese of the infor-mation contained in this document, or that the use of any information, apparatus, method, or process disclosed in this document will not infringe privately owned rights:

4 or J .

41 B. Assurras any liabilities with respect to the use of, or for derrages resulting from the use of, any information, ap-paratus, method, or process disclosed in this document.

XN- NF- F00,766

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i-i XN-NF-84-72 TABLE OF CONTENTS Section Page

-1.0 INTR 000CTION................................................ 1 2 . 0 S UMMA R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 L 3.0. L IMITING BREAK LOCA ANALYSIS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6

~

3.l' LOCA ANALYSIS M00EL.................................... 7 3.2 RESULTS................................................ 8 4 . 0 C ON C L U S I ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56 5 .0 R EF ER E N C ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 1

- - , - .- ,, -n-.-. ~,- - --,-,-. ,-, - -,,,,--g, amr- r-- ,-.n, ..g--w, - em, , -

a ii XN-NF-84-72 List of Tables J

. Table Page

' 2.1 H.B. Robinson Uni t 2 LOCA-ECCS Analysis Results . . . . . . . . . . 4 2.2 ENC Licensing -History of H.B. Robinson Uni t 2 . . . . . . . . . . . . 5 3.1. - H.B . : Robi nson Uni t 2 Sys tem Data . . . . . . . . . . . . . . . . . . . . . . . . . 10

. 3. 2 . Fue l Desi gn P arameters . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

11

- 3.3 :H.B. Robinson Unit 2 LOCA/ECCS Event Table for

~L imiting Break (0.8 DECLG and 2. MWD /kgu Exposure) ....... 12 4

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iii XN-NF-84-72 List of Figures Figure Page 3.1 Blowdown System Nod al i zation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13

3. 2 REFL E X Nod al i z at i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 3.3 Axial Peaking Factor vs. Rod Length, 0.8 DECLG Break....... 15 3.4 Normalized Axial De ndenceFactorforFd=2.32 16

-Versus Elevation (F =1.65)................................

3.5 Downcomer Flow Rate During Blowdown Period, 0.8 DECLG Break............................................ 17 3.6 Upper Plenum Pressure During Blowdown Period, 0 . 8 D E CL G B r e a k . . . . . . . . . . . . . . . . . . .' . . . . . . . . . . . . . . . . . . . . . . . . . 18 3.7 Average Core Inlet Flow During Blowdown Period, 0.8 DECLG Break............................................ 19 3.8 Average core Outlet' Flow During Blowdown Period, 0.8 DECLC 8reak............................................ 20 3.9 Total Break Flow During Blowdown Period, 0.8 DECLG Break............................................ 21 3'.10 Break Flow Enthalpy During Blowdown Period, Vessel Side, 0.8 DECLG Break............................... 22 3.11 Break Flow Enthalpy During Blowdown Period, Pump Side , 0.8 DECLG Bre ak . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 3.12 Flow from Intact Loop Accumulator During Blowdown Period , 0.8 DECLG Break . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 3.13 Flow From Broken Loop Accumulator During Blowdown Period , 0.8 DECLG Break . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 3.14 Heat Transfer ' Coefficient During Blowdown Period at PCT t! ode, 0.8 DECLG Break, 2 MWD /Kgu. . . . . . . . . . . . . 26 3.15 Clad Surface . Temperature During Blowdown Period at 27 PCT Node, 0.8 DECLG Break, 2 MWD /Kgu......................

3.16 Depth of Metal-Water Reaction During Blowdown Period at PCT Node, 0.8 DECLG Break, 2 MWD /Kgu. . . . . . . . . . . . 28

iv XN-NF-84-72 List of Figures (Cont.)

Figure Page 3.17 Average Fuel Temperature During Blowdown Period at PCT Node, 0.8 DECLG Break, 2 MWD /KgU.................... 29 3.18 Hot' Channel Average Quality, Center Volume 0.8 DECLG Break, 2 MWD /Kgu.................................. 30 3.19 Heat Transfer Coefficient During Blowdown Period at PCT Node, 0.8 DECLG Break, 9 MWD /Kgu....................... 31

. _3.20 Clad Surface Temperature During Blowdown Period at PCT Node, 0.8 DECLG Break, 9 MWD /KgU....................... 32 3.21 Depth of Metal-Water Reaction During Blowdown Period at PCT Node, 0.8 DECLG Break , 9 MWD /Kgu. . . . . . . . . . . . . . . . . . . . . . . 33 3.22 Average Fuel Temperature During Blowdown Period at PCT Node, 0.8 DECLG Break, 9 MWD /KgU....................... 34 a

3.23 Hot Channel Average Quality, Center Volume 0.8 DECLG Break , 9 MWD /Kgu. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.24 Heat Transfer Coefficient During Blowdown Period at PCT Node,.0.8.DECLG Break, 49 MWD /kgU................... 36 3.25 Clad Surface Temperature During Blowdown Period at PCT Node, 0.8 DECLG Break, 49 MWD /kgu................... 37 3.26 Depth of Metal-Water Reaction During Blowdown Period at 0.8 DECLG Break, 49 MWD /KgU............................. 38 3.27 Average Fuel Temperature During Blowdown Period at PCT Node, 0.8 DECLG Break, 49 MWD /kgu...................... 39 3.28 Hot Channel Average Quality, Center Volume 0.8 DECLG Break, 49 MWD /kgu....................... .... .... 40 3.29 Accumulator (Intact) Flow During Refill and Reflood 41 Periods, 0.8 DECLG Break...................................

3.30 Accumulator (Broken) Flow During Refill and Reflood Periods, 0.8 DECLG Break................................... 42 3.31 HPSI (Intact) Flow During Refill and Reflood Periods, 43 0.8 DECLG Break............................................

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v XN-NF-84-72 List of Figures (Cont.)

Figure Page

-3.32 HPSI (Broken) Flow During Refill and Reflood Periods, 0.8 DECLG Break.............. ............................. 44 3.33 LPSI (Intact) Flow During Refill and Reflood Periods, 0.8 DECLG Break............................................ 45

. 34 3 LPSI (Broken) Flow During Refill and Reflood Periods, 0.8 DECLG Break............................................ 46 3.35 Containment Back Pressure, 0.8 DECLG Break. . . . . . . . . . . . . . . . . 47 -

3.36 Normal i zed Power, 0.8 DECLG Break . . . . . . . . . . . . . . . . . . . . . . . . . . 48

'3.37 Reflood Core Mixture Level, 0.8 DECLG Break . . . . . . . . . . . . . . . . . 49 3.38 Reflood Downcomer Mixture Level ,0.8 DECtG . . . . . . . . . . . . . . . . . . 50 3.39 Reflood Upper Plenum Pressure, 0.8 DECLG Break. . . . . . . . . . . . . . 51 r 3.40 Core Flood ing Rate, 0.8 DECLG Break . . . . . . . . . . . . . . . . . . . . . . . . . 52 t

3.41 T000EE2 Cladding Temperature vs. Time, 0.8 DECLG Break,

-2 MWD /kgU Exposure......................................... 53 3.42 T00DEE2 Cladding Temperature vs. Time, 0.8 DECLG Break, 9 MWD /kgU Exposure......................................... 54 3.43 T000EE2 Cladding Temperature vs. Time, 0.8 DECLG Break, 49_ MWD /kgU Exposure......................................... 55 i

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1 XN-NF-84-72

1.0 INTRODUCTION

In 1975, Exxon Nuclear Company (ENC) performed a Loss-of-Coolarit Accident (LOCA) analysis for H.B. Robinson Unit 2(1), using the ENC WREM-based PWR .ECCS Evaluation Model(2). A follor:ap analysis in December 1976(3) identified the double-ended cold leg guillotine break with a discharge coefficient of 0.8 as the most limiting break. The analysis assumed 6% steam generator tube plugging (SGTP) with a total linear heat generation rate (LHGR) of 13.43 kw/ft, corresponding to a. total peaking I

'(Fg ).of 2.2 at 102% of rated power. In September 1980 and August 1981, additional analyses were performed (4,5) to support the operation of the T

plant with 10% and 15% steam generator tube plugging for an LHGR and Fg gf 13.43 kw/ft and 2.2, respectively. In addition, an analysis was performed

.at.15% steam generator tube plugging which supported an LPGR of 14.16 kw/ft T

and an Fg of 2.32. Additional analyses were performed in 1982 for SGTP of 20%'and in 1983 for SGTP of 30%, at reduced power and temperature (6,7). The 1983 analysis used the ENC WREM-IIA PWR ECCS evaluation model(8,9,10,11) with the NUREG-0630 clad swelling and rupture model and the revised EXEM/PWR steam cooling model(12). Table 2.2 summarizes the ENC licensing history for the H.B. Robinson Unit 2 plant.

This report presents the results of a LOCA ECCS analysis performed.for the previously identified limiting break, of 0.8 DECLG with peak rod exposures up to 49. MWD /kg. This analysis was performed as a result of the decision by Carolina Power and Light Company (CP&L) in 1983 to: (1) replace the H.B. Robinson Unit 2 steam generators, ('?) implement a low i adial leakage fuel management scheme in order to reduce vessel fluences and

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2 XN-NF-84-72 f.

thereby alleviate ~ concerns about thermal shock, and (3) to increase the peak assembly discharge exposure for the H.B. Robinson fuel to 44 MWD /kgu.

To implement the low radial leakage fuel management scheme, the total nuclear enthalpy riis (FT ) was' increased to 1.65. The analysis was

'e performed with an LHGR, including the 1.02 factor for power uncertainty, of T

14.16 kw/ft, corresponding to a total power peaking factor of 2.32 (Fg ),

~

The analysis is applicable for up to 6% steam generator tube plugging with

~

, the~ reactor operating at 100% power, 2300 MWt.

3 XN-NF-84-72 2.0

SUMMARY

The calculational basis and results are summarized in Table 2.1. The maximum calculated peak cladding temperature (PCT) is 20420F, occurring at 60 seconds into the accident at a location 6.0 feet from the bottom of the active core, with a total metal-water reaction less than one percent. The results of the analyses show that within the limits established, the .H.B. Robinson Nuclear Reactor, operating at the rated power level of 2300 MWt, and with steam generator tube plugging up to 6%, satisfy the criteria specified by 10 CFR50.46(13) for peak rod burnups less than 49 MWD /kgU.

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A, Table 2.1 'H. B. Robinson Unit 2 LOCA-ECCS Analysis Results Calculational Basis License Core Power, MWt '2300 Power Used for Analysis, MWt** 2346 Peak Linear power for Analysis, kW/ft** 14.16 Total Peaking Factor, Fg T 2.32 T

Enthalpy Rise, Nuclear, Fg 1.65 Steam Generator Tube Plugging (%) 6.00 2 MWD /kgU 9 MWD /kgU 49 MWD /kgU Analysis Results Peak Rod Exposure Peak Rod Exposure Peak Rod Exposure Peak Clad Temperature (PCT), OF 204E 1815 1785

• Peak Clad Temperature Reached, (sec) 60 139 139 Paak Clad Temperature Location, ft. 6.0 8.5 8.5 Local Zr/H 2O Reaction (max.), %* 4.65 1.93 1.72 Local Zr/H 2 0 Location, ft, from 5.25 5.25 5.25 Bottom Total H2 Generation, % of total <1 <1 <1 Zr Reacted Hot Rod Burst Time, sec. 39.9 45.7 45.2 Hot Rod Burst Location, ft. 6.0 6.0 6.0 5

• Computer value at 380 seconds. y

• • Including 1.02 factor for power uncertainties. E

. Tabla 2.2' ENC Licensing Histtry of.H. B. Robinsun Unit 2' XN-NF-75-41,. XN-NF-80-43, XN-NF-82-18(P) Current-XN-NF-76-54 XN-NF-81-54 .XN-NF-82-18 Supplement 2 XN-NF-84-72 Plant Parameters 1765-& 1976 1980 & 1981 1982 1983- 1984 Fd- 2.2 2.2 2.2 2.32 2.32 2.32 2.32 T

Fgj 1.55 1.55 1.55' .1.55 1.60 1.60 1.65 LHGR (kw/ft) 13.43 13.43 13.43 14.16 12.04 12.04 14.16

% of rated power 102 102 102 102 87 87 102 Primary coolant flow 89965 89965 89965 89965 82700 80000 88330 (GPM/ Loop) m Vessel Tave ( F) 579.5 579.5 579.5 579.5 537.1 537.1 579.5 i SG Tube Plugging (%) 6 10 15 15 20 30 6 Break Type Break Limiting Limiting Liiniting Limiting Limiting Limiting Spectrum Break Break Break Break Break Break 0.8 0.8 0.8 0.8 0.8 0.8 DECLG DECLG DECLG DECLG DECLG DECLG 5,

n

i 6 XN-NF-84-72 3.0 LIMITING BREAK l.0CA ANALYSIS This report documents the results of the LOCA-ECCS analysis performed for H.B. Robinson Unit 2 with a steam generator tube plugging up to 6%. The analytical techniques used are in compliance with Appendix K of 10 CFR 50, and are described in the ENC WREM models(2), and the Emergency Core Cooling System Evaluation Model Updates: WREM-II(14), WREM-IIA (10), and EXEM/PWR(12),

A - LOCA break spectrum analysis was performed and reported in XN 54(3), . The limiting LOCA break was. determined to be a large double-ended guillotine break of the cold leg, with a discharge coefficient of 0.8 (0.8 DECLG). The analyses performed and reported herein used the following LOCA/ECCS models:

(1) . Fuel Rod .Model - The RODEX2(15) stored energy and fission gas release model in place of the previously approved GAPEX(16) model.

(2) Blowdown Model - The RELAP4-EM code with NUREG-0630 clad swelling and rupture model(12), and fuel rod model consistant with RODEX2 gap conductance model(17),

(3) 'Reflood Model - The REFLEX code with the EXEM/PWR core outlet enthalpy model(12) and the ENC WREM carry rate fraction (CRF) correlation (2),

(4) Heatup Model - The TODDEE2 code with the EXEM/PWR steam cooling model(12), the NUREG-0630 clad swelling and rupture model(12), and the WREM heat transfer correlations (2),

(5) All other model revisions documented in XN-NF-82-20(P), Revision 1(12),

7 XN-NF-84-72 3.1 LOCA ANALYSIS MODEL The Exxon Nuclear Company EXEM/PWR ECCS evaluation model(12) was used_to perform the analyses. This model consists of the following computer codes: R0DEX2(15) for initial rod stored energy and internal fuel rod gas inventory; RELAP4-EM(17,18) for the system blowdown and hot channel blowdown calculations; CONTEMPT-LT/22 as modified in CSB 6-1(19) for computation of containment backpressure; REFLEX (10,12) for computation of system reflood; and T000EE2(12,20,21) for the calculation of final fuel rod heatup.

The H.B. Robinson Unit 2 nuclear reactor is a three-loop Westing-house pressurized water reactor with dry containment. The reactor coolant system is nadalized into control volumes representing reasonably homogereous regions, interconnected by flow-paths or " junctions" as described in XN 57(1). -The system nodalization is depicted in Figure 3.1. The single failure

'is assumed to be the loss.of or,e of three HPSI pumps in addition to a loss of f

one LPSI pump. The reactor coolant pump performance characteristic curves are the Westinghouse pump curves built into the RELAP4 code. Six percent of the tubes in ~ each steam generator are assumed to be plugged. The transient behavior was determined from the governing conservation equations for mass, 5 energy, and momentum. Energy transport, flow rates, and heat transfer are o

O determined from appropriate correlations. System input parameters are given shows the REFLEX nodalization in the reflood in Table 3.1. Figure 3.2

. calculation of the H.B. Robinson Unit 2.

L

. 8 XN-NF-84-72 The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The chopped cosine axial power profile _used for the analyses is shown in Figure 3.3, with a maximum axial peaking factor of 1.365, corresponding to a total peaking factor of 2.32, and T T

.Fei of 1.65.- The Fg determined with this axial profile in combination with

.the current K(Z) function developed originally by the NSSS vendor is used to T

define the envelope for Fg , where the K(Z) curve is limited by large break LOCAs. Where small break LOCAs are limited, the K(Z) curve was modified such that Linear Heat Generation Rates (LHGRs) were determined by the NSSS vendor analyses. The K(Z) curve is represented in Figure 3.4. The analysis of the loss-of-coolant accident is performed at 102 percent of rated power. The fuel design parameters are shown in Table 3.2.

Three cases of LOCA-ECCS calculations were performed with input which bounds the fuel history up to 49,000 MWD /kgU peak rod exposure. The most

limiting fuel conditions were determined and used in each calculation. Decay power, internal rod pressure and the fission gas releases were highest at EOL (third case) for the hot rod, while stored energy was calculated f.o be highest at lower exposure (first case). The combination of highest stored energy, rod pressure,' and . decay power were used in the LOCA-ECCS analyses over the exposure ranges shown.

3.2 RESULTS Table 3.3 presents the timing and sequence of events as determined for the large guillotine break with a discharge coefficient of 0.8. Figures 3.! through 3.34 present plotted results for system blowdown analysis. Unless

9 XN-NF-84-72 otherwise'noted on the figures, time zero correenonds to the time of break initiation.. Figures 3.14 through 3.28 present results for the hot channel blowdown calculations. Figure 3.35 presents calculated containment back-pressure time history. Figure 3.36 shows the normalized power calculation

-results. The reflood calculation results are shown in Figures 3.37 through 3.40.

The maximum peak cladding temperature (PCT) calculated for the 0.8 DECLG break occurs at 2 MWD /KgU and is 20420F (Figure 3.41). The maximum local metal-water reaction is 4.65 after 380 seconds, and the total core metal-water reaction is less than 1%. The PCT location is at an elevation of 6.0 feet from the bottom of active core. For the 9 MWD /kgU exposure, the calculated PCT is 18150F (Figure 3.42) occurring at 139 seconds at an elevation of 8.5 feet relative to 'the bottom of the active core. For the EOL case, the calculated PCT is 17850F occurring at 139 seconds at 8.5 feet elevation relative to the bottom of the active core (Figure 3.43).

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10- -XN-NF-84-72

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N Table 3.1 H. B. Robinson Unit 2 System Data Primary-Heat Output,TMWt 2346*

~ Primary Coolant Flow, lbm/hr 100.3-x 106

~

Primary Cool ar.t Volume, ~f t3 9768**

~

10perating Pressure, psia 2,250.

-Inlet Coolant = Temperature, OF 546.2

~

- Reactor : Vessel . Volume, f t3 3660 Pressurizer Volume; Total, f t -3 1300 Pressupizer Volume, Liquid,-ft3 780 L Accumulator Volume,- Total, f t3 (each of'three) 1200 iAccumulator' Volume, Liquid,ft(3) 825

. AccumulatorfTrip Point Pressure, psia 615 Steam Generator Heat Transfer Area, ft(2) 40,859**

tSteam Generator Secondary F1ow, 1bm/hr I3.37 x 106

Steam Generator Secondary pressure,-psia 800 o = Reactor Coolant: Pump Head, f t _

264 Reactor ' Coolant' Pump Speed, rpm 1180

. Moment of-Inertia, lbm-ft /2rad 70,000 Cold Leg Pipe,.I.D., in 27.5.

" Hot-Leg Pipe,.I.D., in '29.0

-  : Pump Suction Pipe, I.D., in

,31.09

• Primary Heat Output used in RELAP4-EM Model = 1.02 x 2300 = 2346 MWt.

'** Includes 6% SG tube plugging, a

11 XN-NF-84-72 Table 3.2 Fuel Design Parameters

' Parameter ENC Fuel Cladding, 0.D, in. 0.424 C1 adding, I.D., in. 0.364 o

. Cladding Thickness, in. 0.030

Pellet.0.0.,.in..

0.3565

..Diametral Gap, in. 0.0075

. Pellet Density, % TD 94.0 Active Fuel Length, in. 144

- Enriched.U0 2

, in. 132 Upper Blanket,-in. 6.0 Lower Blanket, in. 6.0

~-

Cell' Water / Fuel Ratio 1.76

. .. Rod Pitch 0.563 r

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12 XN-NF-84-72 Table 3.3 H.B. Robinson Unit 2 LOCA/ECCS Event Table for Limiting Break (0.8 DECLG and 2 MWD /kgU Exposure)

Time Event (seconds)

Start. 0.0 Initiate Break 0.1 Safety Injection Signal- 0.6 Accumulator Injection, Broken Loop 3.1 Pressurizer Empties 9.0

' Accumulator Injection, Intact Loop 12.0 End-of-Bypass 22.57 Safety Pump' Injection, HPSI . 25.6

. Safety Pump Injection, LPSI (Broken) 32.17

' Start of Reflood 45.79 Accumulators Empty 48.76 ,

Safety Pump Injection, LPSI (Intact) 48.87 Peak Clad Temperature Reached (sec) 60.0 fh

13 XN-NF-84-72

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4.0 CONCLUSION

For breaks up to and including the double-ended severance of a reactor - . .

coolant pipe, the Emergency Core Cooling System for H.B. Robinson Unit 2 will meet the Acceptance Criteria as presented in 10 CFR 50.46, with the ?.32 F[

and1.65Fhlimits. The criteria are as follows:

(1) The calculated peak fuel element clad temperature doos not evceed the 22000F limit.

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(2) The amount of fuel elemen cladding that reacts chemically with ;j ) y' .

water or steam does not exceed 1 percent of the total amount of zircalov in the .4,-*,'-ec[4 - -

. - - . p- .,

.. i S r. . n e reactor.

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,.x .,. ,o (3) The cladding temperature transient is terminated at a time why the .}. h core geometry is st'll amen abl e to coolino. The hot fuel rod claddina [J'/.[. . :Q f,t

? .- ;p oxidatior, limits of 17% are not exceeded durina c after quenching. , i.j g V =

wff P1, .>. ' .

(4) The core temperature is reduced and decay heat is removed for an iy...A

)L .R= ,7 '

= M. .,5 extended per iod of time, as reouired by the 1one-1ived radicactivity

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~~

XN-NF-84-72 -

~

5.0 REFERENCES

1. "H.B. Robinson Unit to. 2 LOCA Analyses Using the ENC WREM-Based PWR ECCS - -

Evaluation Model (September 26, 1975 Version )," XN-75-57, October 1975, -

E and Revision 1, November 1975, Exxon Nuclear Company, Richland, WA.

2. " Exxon Nuclear Company WREM-Basea Generic PWR ECCS Evaluation Model,"

XN-75-41, July 1975, and Supplements and Revisions thereto, Exxon Nuclear Company.

3. "LOCA Analyses for H.B. Robinson Unit No. 2 Using WREM-Based PWR ECCS Evaluation Model with Reduced LPSI Flow, Steam Generator Plugging and Increased Upper Head Temperature," XN-76-54, Exxon Nuclear Company, Richland, WA, December 1976.

4. "ECCS and PTS Analyses for H.B. Robinson Unit No. 2 Reactor with 5%,10%

and 15% Steam Generator Tube Plugging," XN-NF-80-43, Exxon Nuclear Company, Richland, WA, August 1981.

6. "ECCS and Plant Tr ar, s i en t Analyses for H.B. Robinson Unit 2 Reactor Operating at Reduced Primary Temperature," X.1-NF-82-18, Exxon Nuclear Company, Richland, WA, March 1982.

7. "ECCS and Plant Transient Analyses for H.B. Robinson Unit 2 Reactor Operating at Reduced Primary Temperature; Supolement 1: Ane'ysis of the Seized Primary Coolant Pump Accident for Asymmetric Steam Generator Tube Plugging," Exxan Nuclear Company, Richland, WA, March 1983.

8. Letter, G.F. Owsley (ENC) to D.F. Ross (NRC), " Description of RELAP4-EM ENC 28B," dated October 30, 1978.

9. Letter, Thomas A. lppolito (NRC) to Warren S. Nechodom (ENC), " ENC-EM Update Evaluation," March 1979.

10. " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Updated WREM-IIA," XN-NF-78-30( A), Exxon Nuclear Company, Richland, WA, May 1979.

11. Letter, Thomas A. Ippolito (NRC) to W.S. Nechodom (ENC), " Topical Report Evaluation," dated March 30, 1979.

12. Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," XN-NF-82-20(P), Feb. 1982; Rev. 1, August 1982; Supplement 1, March 1957i and Supplement 2, March 1982, Exxon Nuciear Company, Richland, WA.

13. " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50; Federal Register, Volume 39, Number 3, January 4, 1974.

?

E 53 XN-NF-84-72 3

14. " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model r Update ENC WREM-II," XN-76-27, July 1976, Supplement 1, Eptember 1976, and Supplement 2, November 1976, Exxon Nuclear Company.

f

! 15. "RODEX2: Fuel Rod Thermal-Mechanical Response Evaluation Model," XN-NF-

- 81-58(P), Revision 2, Exxon Nuclear Company, February 1983.

=

16. "GAPEXX: A Computer Program for Predicting Pellet-to-Cladding Heat c Transfer Coefficients," XN-73-25, Exxon Nuc lear Company, Richland, WA,

[ August 1971

17. " Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates,"

i XN-NF-82-20(P), Supplement 2 to Rev.1, Exxon Nuclear Company, Richl;nd,

-- WA, June 1984

18. Letter, T.A. Ippolito (NRC) to W.S. Necnodom (ENC), "SER for ENC RELAP4-

EM Updcte," March 1979.

F b 19. " Exxon Nuclear Company ECCS Evaluation of a 2-Loop Westinghouse PWR with i Dry Containment Using the ENC WREM-II ECCS Model - Large Break Example

[ Problem," XN-NF-77-25( A), Exxon Nuclear Company, Richland, WA, September

= 1978.

E

_ 20. G.N. Lauben, NUREG-75/057, "T00DEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program," May 1975.

{

F 21. " Exxon Nuclear Company ECCS Cladding Swelling and Ruptut e Model," XN-NF-

) 82-07(P), Exxon Nuclear Company, Richland, WA, March 1982.

~

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XN-NF-84-72 Issue Date: //12/84 H. B. ROBINSON UNIT 2 LIMITING BREAK LOCA-ECCS ANALYSIS WITH INCREASED ENTHALPY RISE FACTOR Distribution

= MJ Ades l GJ Busselman

JC Chandler

=

._f RE Collingham GC Cooke 2 LJ Federico

SE Jensen l WV Kayser M CE Leach j MR Killgore j GA Sofer d IZ Stone a

29 RB Stout i -

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] NRC/CP&L/TJ Helbling (82) 3 Document Control (5) .. . , ..

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33

SINGLE FAILURE ANALYSIS Attachment il TABLE 1: SLMMARY OF THE RESULTS Transients Worst Single Failure Transients Not Analyzed Comment -

Feedwater Malfunctions that 15.1.1 This increase in heat removal by result in a decrease in feed- the secondary is not severe enough water temperature to drop Reactor Coolant System pressure to the SI setpoint. Reactor trip and turbine trip prevent drastic cooldown of reactor coolant system.

15.l.2 Feedwater System Malfunctions This increase Is heat removal by that result in an increase in the secondary is not severe enough feedwater flow to drop Reactor Coolant System pressure to the Si setpoint. Reactor trip and turbine trip prevent drastic j

cooldown of the reactor coolant system.

) 15.1.3 Excessive increase in Steam Driven Auxillary l Secondary Steam Flow Feedwater Pump f alls to deliver flow 15.1.4 Inadvertent Opening of SG Bounded by Excessive increase In Rollet or PORV Secondary Stear Flow (15.1.3) and hand calculati,>ns.

15.1.5 Main Steamilne Break For fue! thermal limits

with Loss of Offsite failure of one of two diesel Power generators to start For containment Integrity for MSLB inside containment:

failure of one of two diesel generators to start Main Steamline Break For fuel thermsl limits: failure with Offiste Power of one of three Si pumps available For of f site dose for MSLB outside containment: continued normal feedwater injection at reduced flow (3610NH/pgp)

TABLE 1:

SUMMARY

OF THE RESULTS (Continued)

Transients worst Single Failure Transients Not Analyzed Comment 15.2.1 Loss of External Steam driven Auxillery Feedwater Electric Load Pump falls to deliver flow

'15.2.2 Turbine Trip Bounded by Loss of Load (15.2.1) 15.2.3 Loss of Condenser Vacuum and Bounded by i.oss of Load (15.2.1) I other events resulting in Turbine Trip 15.2.4 Inadvertent Closure of MSIV's Bounded by Loss of Load (15.2.1) 15.2.6 Loss of Non-Emergency AC Power Bounded by Complete Loss of Flow l

to Station Auxillaries (15.3.1) and Loss of Normal I Feedwater Flow (15.2.7) l I

i 15.2.7 Loss of Normal hteam driven Auxillary Feedwater l

Feedwater Flow Pump falls to deliver flow 15.2.8 Feedwater System Pipe Break Bounded by Steamline Break (15.1.5)

O 15.3.1 Loss of Forced Primary Steam drive'n Auxiliary Feedwater Coolant Flow Pump falls to deliver flow 15.3.3 Reactor Coolant Pump Steam driven Auxiliary Feedwater Shaf t Seizure Pump falls to deliver flow 15.3.4 Reactor Coolant Pump Broken Bounded by Shaf t Selzure (l$.3.3)

Shaft 15.4.1 Uncontrolled RCCA Steam Driven Auxillary Feedwater Withdrawal from Pump falls to deliver flow SubcritIcai or Low Power Startup Condition 15.4.2 Uncontrolled RCCA Steam Driven Auxillary Feedwater Withdrawal from Power Pump falls to deliver flow (3610NH/pgp)

TABLE le

SUMMARY

OF THE RESULTS (Continued) .i Transients worst Single Failure Transients Not Analyzed Comment l 15.4.3 RCCA Misoperation Steam Driven Auxillary Feedwater Pump falls to deliver flow 15.4.4 Startup of an inactive Power operation with less than Coolant Loop at incorrect three loops is not allowed Temperature l

CVCS Malfunction that The operational modes of refueling 15.4.6 Results in a Decrease and startup are analyzed to show that in Boron Concentration adequate time exists to secure inadvert-In the Reactor Coolant ont boron dilution before criticality occurs. No ESF systems are involved.

15.4.7 Inadvertent Loading of a Administrative procedures preclude Fuel Assembly Into en occurrence of this event.

Improper Location 15.4.8 Spectrum of RCCA Steam driven Auxillary Feedwater

! Ejection Accidents Pump falls to deliver flow l

15.5.1 Inadvertent Operation of FOCS J Shutof f head of high pressure Si pumps is 1500 psia < 1750 psia trip setpoint pressure.

l 15.5.2 CVCS Malfunction that increases Effect on Reactor Coolant System Pressure Reactor Coolant Inventory is completely mitigated by the Reactor Protection System and relief valves.

15.6.I Inadvertent Opening of Failure of one of three Sl Pressurizer Safety or pumps to deliver flow PORV l

15.6.2 Loss of Reactor Coolant from Bounded by large break LOCA (15.6.5)

Rupture of Small Plpes or from Cracks in Large Pipes which actuate the ECCS (3610NH/pgp)

_._ ..m .

y;;- __

+

TADtE 1: StsetARY OF THE RESULTS (Continued)

Transients' ilforst Single Fellure Translents Not Analyzed Comment 15.6.3 - Steen Generator Tube Steam driven Auxillery Feedwater Rupture . Pump faIis to'dellver fIow 15.6.5 LOCA' Failure of one diesel generator-to start (3610NH/pgp)

ATTACINENT 12

.~

XN-NF-84-68 (P)

H. B. ROBINSON UNIT 2 RADIOLOGICAL ASSESSMENT OF POSTULATED ACCIDENTS l-

1 AFFIDAVIT

-STATE OF Washington )

ss.

COUNTY OF Benton )

I, Richard B. Stout, being duly sworn, hereby say and de-pose:

1. I am Manager, Licensing and Safety Engineering, for Exxon Nuclear Company, Inc. (" ENC"), and as such I am authorized to execute this Affidavit.

2. I am familiar with ENC's detailed document control system and policias which govern the protection and control of information.

3. I am familiar with the document XN-NF-84-68(P) entitled "H.B.. Robinson Unit 2 Radiological Assessment of Postulated Accidents" referred to- as " Document". Information contained in this Document has been classified by ENC as proprietary in accordance with the control

- system and policies established by ENC for the control and protection of information.

4. The Document contains information of a proprietary and confidential nature'and is of the type customarily held in confidence by ENC and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in the Document as proprietary and confidential.

5. The Document has been made available to Carolina Power and Light Company and the U.S. Nuclear Regulatory Commission in confidence, with the request that the information contained in the Document not be 6 disclosed or divulged.

L p .

( .

f

6. The thcument contains information which is vital to a com-petitive advantage of ENC and would be helpful to competitors of ENC when competing with ENC..

7. The information contained in the Document is considered to be proprietary by ENC because it reveals certain distinguishing aspects of radiological assessment procedures which secure competitive advantage to ENC for fuel design optimization and improved marketability, and includes information utilized by ENC in its business which affords ENC an opportunity to obtain a competitive advantage over its competitors who do not or may not know or use the information contained in the Document.

~

8. The disclosure of the proprietary information contained in the Document to a competitor would permit the competitor to reduce its e.;penditure of money and manpower and to improve its competitive position by giving it extremely valuable insights into radiological assessent procedures ~ and would result in substantial harm to the competitive position.of ENC.

9. The Document contains proprietary information which is held in. confidence by ENC and is not available in public sources.

10. In accordance with ENC's policies governing the protection and control of information, proprietary information contained in the Document has been made'available, on a limited basis, to others outside ENC only as required and under suitabl(. agreement providing for non-disclosure and limited use of the information.

11. ENC policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

12. This Document provides information which reveals the radiological assessment procedures developed by ENC over the past several years. ENC has invested thousands of dollars and many man-years of effort in developing the BWR thermal hydraulic analysis methods revealed in the Document. Assuming a competitor had available the same background data and incentives as ENC, the competitor might, at a minimum, develop the information for the same expenditure of manpower and money as ENC.

13. Based on my experience in the industry, I do not believe that the background data and incentives of ENC's competitors are suf-ficiently similar to the corresponding background data and incentives of ENC to reasonably expect such competitors would be in a position to duplicate ENC's proprietary information contined in the Documents.

THAT the statements made hereinabove are, to the best of my knowledge, information, and belief, truthful and complete.

FURTHER AFFIANT SAYETH NOT.

I l

'}

ll],L_.JkJ SWORN TO AND SUBSCRIBED before me this & day of

'A . lu , 19M.

q 3 b cm 1. cb NOTARY PUBLIC i

l l