ML20083K111

LOCA Rept

ML20083K111
Person / Time
Site:	Brunswick
Issue date:	12/31/1978
From:	CAROLINA POWER & LIGHT CO.
To:
References
78NED299, NEDO-24165, NUDOCS 7901080212
Download: ML20083K111 (42)

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i LOSS-OF-COOLANT ACCIDENT ANALYSIS REPORT FOR 5

i BRUNSUICK f STEAM ELECTRIC PLANT L* NIT NO. 1 u py ce. 1 G!r "ff - .

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. NUCLEAR ENERGY PACVECTS DivtSION

• GENERAL ELECTRIC CommeANY SAN JOSS, CAUFO ANI A 95125 GENERAL $ ELECTRIC ,

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NEDO-24165 JMPCSTANT NOTICE REGARDING CCNTENTS OF THIS liEPCR*"

Please Read Carefully The only undertakir.g of Ger.emi Electric Ccr pany respecting infct-ation in this doement are contained in the c:ntmet betxcn Carolin Po;xr and Light Cor pany ard General Electric Company and nothing contained in this doew ent shall be construed as changing the contmet. The use of this l infonation by anycne other tF.1n Carolir.a Fo 'cr and Light Crpany cr for an3 Turpose ather th.~.n that for uhich it is inter.ded, is r.ot cuth:rined; ar.d with respect to any unauthorized use, Gener:Z Elcatric rakes no repre-sentation or :.urmnty, ar.d aesumes no liability as to the cor pleter.ess, accuracy, or usefulness of the infomation contained in this doe:s cr.t.

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NEDO-24165 TABLE OF CONTENTS Pape

1. INTRODUCTION 1-1

2. LEAD PLANT SELECTION 2-1 l

3. INPLT TO ANALYSIS 3-1

4. LOCA ANALYSIS COMPUTER CODES 4-1 4.1 Results of the LAMB Analysis 4-1 4.2 Results of the SCAT Analysis 4-1 4.3 Results of the SAFE Analysis 4-1 4.4 Results of the REFLOOD Analysis 4-2 4.5 Results of the CHASTE Analysis 4-3 4.6 Methois 4-4

5. DESCRIPTION OF MODEL AND INPLT CHANGES 5-1

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6. CONCLUSIONS 6-1 f

7. REFERENCES 7-1  ;.

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NEDO-24165 l

l LIST OF TABLES Table Title g 1

Significant Input Para::eters to the Loss-of-Coolant )

Accident Analysis 3-1 '

2 Su: unary of Results 4-$

3 LOCA Analysis Fig._re Sur::rury 4-6 4A MAPLHCR Versus Average Planar Exposure 4-7 4B MAPLEGR Versus Average Planar Exposure 4-8 4C MAPLHCR Versus Average Planar Exposure 4-9 4D MAPLHGR Versus Average Planar Exposure 4-10 e

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LIST OF ILLUSTRATIONS l Figure Title l

Page l la

! Water Level Inside the Shroud and Reactor Vessel Pressure Following a Recirculation Line Discharge Break. LPCI Injection Valve Failur., Break Area = 80% DBA 6-3 lb Water Level Inside the Shroud a.d Reactor Vessel Pressure Following a Maximum Recirculation Discharge Break, LPCI

' Injection Valve Failure, Break Area = 2.4 ft2 6 -4 Ic Water Level Inside the Shroud and Reactor Vessel Pressure Following a Maximus Recirculation Line Suction Break, LPCI Injection Valve Failure Break Area = 4.2 ft2 6-5.

2a Peak Cladding Temperature Following a Recirculation Discharge Line Break,' LPCI Injection Valve Failure, Break Area = 80% DBA 6-6 2b Peak Cladding Temperature Following a tbximum Recirculation Line Discharge Break, LPCI Injection Valve Failure Break Area = 2.4 f t2 6-7 2e Peak Cladding Temperature Following a Maximum Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area = 4.2 f t2 6-8 3a Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node Following a i Recirculation Discharge Line Break, '.PCI Injection Valve Failure, Break Area = 80% DBA 6-9 3b. Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node for a Maximum Recirculation Line Discharge Break LPCI Injection Valve Failure, Break Arca = 2.4 ft 6-10 3e Fuel Rod Convective Heat Transfer Coefficient During Blowdown at the High Power Axial Node for a Maximum i

Recirculation Line Suction Break, LPCI Injection Valve Failure, Break Area - 4.2 ft2 6-11 ,

4a Not applicable 6-12 Ab Normalized Core Average Inlet Flow Following a Maximum Recirculation Line Discharge Break, Break Area = 2.4 f t2 6-13 4e Normalized Core Average Inlet Flow Following a Maximum Recirculation Line Suction Break, Break Area = 4.2 f t2 6-14 5a Not Applicable 6-15 Sb Minimum Critical Power Ratio Following a Maximum Recirculation Line Discharge Break, Break Area = 2.4 ft 2 6-16 Sc Minimum Critical Power Ratio Following a Haximum Recirculation Line Suction Break, Break Area - 4.2 ft2 6  !

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NEDO-24165 LIST OF ILLUSTRATIONS (Continued) l Figure Title Page 6a Variation With Break Area of Time for Which I'ot Sode Remains Uncovered (Discharge) 6-18 6b - Variation with Break Area of Time for Which 1:ot Node Remains Uncovered (Suction) 6-19 l

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1. INTRODUCTION i The purpose of this do ucent is to provide the results of the loss-of-ccolant l accident (LOCA) analysis for the Brunswick Steam Electric Plant Unit No. 1 (Brunswick 1). The analysis was performed using approved General Electric (CE) calculational models.

This reanalysis of the plant LOCA is provided in accordance with the NRC requirement (Reference 1) and to demonstrate conformance with the ECCS acceptance criteria of 10CFR50.46. The objective of the LOCA analysis contained herein is to provide assurance that the most limiting break size, break location, and single failure combination has oeen considered for the plant. The required documentation for demonstrating that these objectives have been satisfied is given in Reference 2. The documentation contained in this report is intended to satisfy thase requirements.

The general description of the LOCA evaluation models is contained in Reference 3.

Recently approved model changes (Reference 4) are described in References 5 and 6.

These model changes are employed in the new REFLOOD and CHASTE co=puter codes which have been used in this analysis. In addition, a model which takes into account the effects of drilling alternate flow path holes in the lower tieplate of the fuel bundle and the use of such fuel bundles in a full or partial core loading ,

is described in References 7, 8, and 9. This model was also approved in Refer- /

ence 4. Also included in the reanalysis are current values for input parameters based on the LOCA analysis reverification program being carried out by CE. The specific changes as applied to Brunswick 1 are discussed in more detail in later sections of this document.

Plants are separated into groups for the purpose of LOCA analysis (Referene.e 10).

Within each plant group there will be a single lead plant analysis which provides the basis for the selection of the most limiting break size yielding the highest peak cladding-temperature (PCT). Also, the lead plant analysis provides an expanded documentation base to provide added insigh into evaluation of the details of particular phenomena. The remainder of the plants in that group will 1-1

p;a w n, m ', N M w% a% N N N 6 NEDO-24165 have non-lead plant analyses referenced' to the lead plant analysis. This .bcu-ment contains the non-lead plant analysis for Brunswick 1. which is a B'.lR/4 with low pressure coolant injection (1.PCI) system modification group plant and is consistent.with the requirements outlined in Reference 2.

The same models and computer codes are used to evaluate all plants. Changes to these models will cause changes in phenomenological responses that are similar within any given plant group. The difference in input parameters cre not expected to result in significantly different results for the plants within a given group.

Emergency core cooling system (ECCS) anc geometric dif ferences between plant groups may result in dif ferent responses for different groups but within any group the responses will be similar. Input changes have been made in the new analysis which are essentially an upgrading of the input parameters to the com-puter codes. Thus, the leai plant concept is still valid for this evaluation.

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2. LEAD PIANT SELECTION l

l Lead plants are selected and analyzed in detail to pereit a nure co=prehensive review and eliminate unnecessary calculations. This constittees a generic analysis for each plant of that type which can be ' referenced fn subsequent plant submittals.

The lead plant for Brunswick 1 is James A. FitzPatrick Nuclear Power Plant.

The justifit..' tion for categorizing Brunswick 1 in this group :sf plants and the lead plant analysis for this group is presented in Reference 11.

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3. *NPUT TO ANALYSIS A list of the significant plant input parameters to the LOCA analysis is presented in Table 1.

Table 1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-C00LAST ACCIDENT AN/. LYSIS Plant Parameterc:

Core Thermal Power 2531 HWt, which corresponds to 105: of rated steam flow 6

Vessel Steas Output 10.96 x 10 lbe/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure IdSS psia Recirculation Line Rreak Area for large Breaks - Discharge 2.4 ft2 (DSA) 1.9 ft (80% DBA)

- Suction 4.2 ft2 Nu=ber of Drilled Bundles $60 Fuel Parameters:

Peak Technical Initial Specification Design Minimum Linear Heat Axial Critical Fuel Bundle Generation Rate Peaking Power f

Fuel Type Geometry (kW/ft) Factor Ratio

• A. IC Type 1 8x8 13.4 1.4 1.2 B. IC Type 2 8x8 13.4 1.4 1.2 i

C. 8DRB265L 8x8 13.4 1.4 1.2 D. 8DRB283H 8x8 13.4 1.4 1.2 To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is per formed with an HCPR of 1.18 (i.e.,1.2 divided by 1.02) for a bundle with an initial HCPR of 1.20.

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4. LOCA ANALYSIS' COMPUTER CODES 4.1 RESULTS OF THE LAMB ANALYSIS This code ,is used to analyze the short-term blowdosTi phenocer1 for large postu-lated pipe breaks (breaks in which nucleate boiling is lost before the water level drops and uncovers the active fuel) in jet pump reactors. The LAMB output (core flow as' a function of time) is input to the SCAT code for calculation of blowdown heat transfer.

The LAMB results presented are:

o Core Average Inlet Flow Rate (normalized to unity at the beginning of the accident) following a Large Break.

4.2 RESULTS OF THE SCAT ANALYSIS This code completes the transient short-tern thermal-hydraulic calculatiun for large breaks in jet pump reactors. The CEXL correlation is used to track the boiling transition in time and lo stion. The post-critical heat flux heat transfer correlations are built 'into SCAT which calculates heat transfer coefficients for input to the core heatup code CHASTE.

The SCAT results presented are:

i e Minimum Critical Power Ratio following a Large Break, e Convective Heat Transfer Coefficient following a Large Break.

4.3 RESULTS OF THE SAFE ANALYSIS f.

This code is used primarily to' trac *n the vessel inventory and to model ECCS performance during the LOCA. The application of SAFE is identical for all break '

sizes. The code is used during the entire course of the postulated accident, but af ter ECCS initiation SAFE is used only to calculate reactor system pressure and ECCS flows, which are pressure dependent.

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. NEDO-24165 The SAFt results presented are:

e Water Level inside the Shroud (up to the time REFLOOD initiates) and a Reactor Vessel Pressute 4.4 RESULTS OF REFLOOD ANALYSIS This code is used across the break spectrum to calculate the system inventories after ECCS actuation. The models used for the design basis accident (DBA) application ("DBA-REFLOOD") was described in a supplement to the SAFE code description transmitted to the L'SNRC December 20, 1974. The "non-DBA REFLOOD" analysis is nearly identica. to the DBA version and employs the same major assumptions. The only dif ferences stem from the fact that the core may be partially covered with coolant at the time of ECCS initiation and coolant levels change slowly for smaller breaks by comparison with the DBA. More precise modeling of coolant level behavior is thus requested principally to determine the contribution of vaporization in the fuel asserblies to the countercurrent flow limiting (CCFL) phenomenon at the upper tieplate. The differences from the DBA-REFLOOD analysis are:

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(1) The non-DBA version calculates core water level more precisely than f

the DBA version in which greater precision is not necessary. ,

5 (2) The non-DBA version includes a heatup model similar to but less i

detailed than that in CilASTE, designed to calculate cladding temper-ature during the small break. This heatup model is used in calcu-lating vaporization for the CCFL correlation, in calculating swollen level in the core, and in calculating the peak cladding temperature.

l l The REFLOOD results presented are: t e Water Level inside the Shroud e Peak Cladding Temperature and Heat Transfer Coefficient for breaks calculated with small break methods I

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NEDO-24165 4.5 REST ~LTS OF THE CHASTE ANALYSIS 6

I This code is used, with suitable inputs from the other codes, to calculate the l

fuel (ladding heatup rate, peak cladding temperature, peak local cladding oxidation , and cere-wide cetal-water reaction for large breaks. The i ailed fuel model in CHASTE considers transient gap conductance, clad swelling and rupture, and metal-vater reaction. The e=pirical core spray heat transfer and channel vetting correlations are built into CHASTE, which solves the transient heat transfer equations for the entire LOCA transient at a single axial plane in a single fuel asse=bly. Iterative applications of CHASTE determine the caximum permissible planar power where required to satisfy the requirements of 10CFR50.46 acceptance criteria.

The CHASTE results presanted are: -

e Peak Cladding Teeperature versus time e Peak Cladding Temperature vers Break Area e Peak Cladding Temperature and Peak Local Oxidation versus Planar Average Exposure for the most limiting break size e Maximus Averags Planar Heat Generation Rate (MAPLHCR) versus Planar f Average Exposure for the most limiting break size A su==ary of the analytical results is given in Table 2. Table 3 lists the figures provided for this analysis. The MAPI.HCR values for each fuel type in the Brunswick I core are presented in Tables 4A through 4D.

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4.6 MET 110CS In the following sections, it will be useful to refer to the methods used to analyze DBA, Jarge breaks, and small breaks. For jet-pump reactors, these are defined a,s follows:

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a. DBA Methods. LAMB / SCAT / SAFE /DBA-REFLOOD/ CHASTE. Areak size: DBA. '

b. Large Break Methods (LBM). LAMB / SCAT / SAFE /non-DBA REFLOOD/ CHASTE.

Break sizes: 1.0 ft2 < A < DBA.

c. Small Break Methode (SBM). SAFE /non-DBA RE.'100D. Heat transfer coefficients: nucleate boiling prior to core uncorery, 25 Btu /hr-ft - *F af ter recovery, core spray when appropriate.

Peak cladding tcmperature and peak local oxidation are calculated in non-DBA-REFLOOD. Break sizes A < l.0 ft .

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NED0-24165 l Table 2 l

SUMMARY

OF RESULTS e Break Size Core-Wide e Location Peak Local Metal-Water e Single Failure PCT ('F) Oxidation (%) Reaction (!)

e 1.9 ft2 (80% DBA) 2183 III 2.9 0.19 e Recirc Discharge e LPCI Injection Valve e 2.4 ft2 (Ed.' 2066(I) Note 2 Note 3 e 3*cire Discharge e trJ1 Inj?.-tion Valve l

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1. PCT from CHASTE.

2. Less than most limiting break (2.9%).

3. Less than most limiting break (0.19%).

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Table 3 LOCA ANALYSIS FICURE

SUMMARY

Limiting Discharge Break Maximum Discharge Break Maxisum Suction Break (LPCI Injection (LPCI Injection (LPCI Injection Valve Failure) Valve Failure) Valve Failure)

(80% DRA) (2.4 ft2) (4.2 ft2)

Water Level Incide Shroud and Reactor Vessel Pressure la lb ic Peak Cladding Temperature 2a 2b 2c Heat T *nsfer Coefficient 3a 3b 3c Core Average Inlet Flow 4b 4b 4c .,

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Minimum Critical Power Ratio Sb 5b Sc fy 5

Peak Cladding Temperature u of HIRhest Powcred Plane Experiencing Boiling Transition 2a Uncovered Time Versus Break Area for Discharge Breaks 6a Uncovered Time Versus Break Area for Suction Breaks 6b

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i NEDO-24165 Table 4A HAPLHGR VERSUS AVERAGE PLANAR EXPOSL*RE Plant: Brunswick 1 Fuel Type: Initial Core - Type 1 Average Y1anar Exposure MAPLilGR PCT Oxidation (Mwd /t) (ktJ/f t) (*F) Fraction 200 11.7 2113 0.022 1,000 11.8 2115 0.022 5,000 12.0 2128 0.023 10,000 12.1 2126 0.023 15,000 12.3 2166 0.026 20,000 12.0 2141 0.024 25,000 11.1 2014 0.016 20,000 10.1 1876 0.009 I

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NEDo-24165 Table 4B MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE Plant: Brunswick 1 Fuel Type: Initial Core - Type 2 Average Planar MAPLHGR PCT 0xidation Exposure (kW/ft) (*F) Fraction

( %'/t) 11.0 2017 0.018 200 1,000 11.1 2016 0.018 5,000 11.7 2078 0.020 10,000 12.2 2146 0.024 15,000 12.2 2163 0.026 20,000 12.0 2149 0.025 25,000 11.1 2026 0.017 30,000 10.1 1886 0.010 4-A

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200 ,11.6 2128, 0.02's 1,000 11.6 2129 0.025 5,000 12.1 2178 0.029 ,

10,000 12.1 2169 0.028 15,000 12.1 2183 0.029 20,000 11.9 2170 0.029 25,000 11.3 2101 0.023 30,000 10.7 2020 0.017 4

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Table 4D

!!APLHGR VERSUS AVERAGE PLANAR EXPOS 1*RE Plant: Brunswick 1 Fuel Type: 8DRB283H Average Planar '

Exposure MAPLHGR PCT Cxidation (Mwd / t) (kW/ft) ('F) Fraction 200 11.2 2090 0.023 1,000 11.2 2083 0.022 5,000 11.8 2149 0.027 10,000 12.0 2161 0.028 15,000 12.1 2180 0.029 20,000 11.8 2164 0.028 25,000 11.3 2096 0.023

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j 5. DESCRIPTION OF MODEL AND IKPUT CHANCES This ser . ion provides a general description of the input and model changes as they relate to the break spectrus calculations. It provides a general background so that the more specific calculated results shown in subsequent sectione can be more easily understood, particularly as they relate to how well trends observed in specific lead plant break spectrum analyses can be l

applied to the general nonlead plant case. The most limiting break ize resalts are not discussed in this context (except to the extent that they affect the shape of the break rpectrum) because detailed limiting break size calculational results will be presented 'or each plant.

The majority of the input and model changes primarily affect the amount of ECCS flow entering the lower plenum as a restle of the countercurrent flow lir.iting (CCFL) effect. These changes as app.ied to Brunswick I are listed

, below.

1. Input Chances

a. 'La: rect'ed Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOO3 code were' corrected.\

s b. Incorporated more accurate bypass areas - The bypass areas -,

in the top guide were recalculated using a more accurate ttyhnique.

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c. Corrected Core Power in REFLOOD - The core power in REFLOOD was corrected to 102% of rated power.

d. Corrected guide tube thermal resistance.

e. Correct heat capacity of reactor internals heat nodes.

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2. Model Chanto

a. Cere CCFL pressure dif ferential = 1 psi - Itcorporate the assumption that flow from the bypass to lower plenum cust avercome a 1 pei pressure drop in core.

b. Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed. -

A few of the changes affect the accident calculation irrespective of CCFL.

These changes are listed below.

1. Input Char.ce

a. Break Areas - The DBA break area was calculated more accurately.'

b. Core Power - The core power in REFLOOD has been corrected to .

1021 of rated.

2. Model Change

a. Improved Radiation and Conduction Calculation - Incorporatior.

of CHASTE 05 for heatup calculation.

b. Suction Line Friction in Discharge Valve Closure Assumption -

Took credit for friction due to irreversible losses in the suctish line. _

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6. CONCLUSTONS Analyses have demonstrated that failure of the LPCI is the cost severe failure among the low pressure ECCS because, unlike the core spray which must pass through the CCFL regions at the top of th'e core, LPC2 is injected into the lower plenum through the jet pumps. Thus, the LPC1 injection valve is the worst single failure in the large break region. This is the case for a break occurring in either the suction or discharge piping. For a break in the discharge piping, this failure results in no LPCI flow, and for a suction line failure, LPCI flow is minimized.

Comparison of the calculated PCT's for the maximum size break in the suction aPd discharge piping determines which is the DBA and which is the second most limiting location. For Brunswick 1 the discharge break is the most limiting location (Table 2). The characteristics that determine which is the most limiting break area at the DBA location are:

a. the calculated hot node reflooding time

b. the calculated hot node uncovery time, and i

c. the time of calculated boiling transition. '

l The time of calculated boiling transition increases with decreasing break i size, since jet pump suction uncovery (which leads to boiling transition) is determined primarily by the break size for a particular plant. The calcu-lated hot node uncovery time also generally increases with decreasing break size, as it is primarily determined by the inventory loss during the blow-down. The hot node reflooding time is determined by a num.ber of interacting

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.henomena such as depressurization rate, countercurrent flow limiting and a combination of available ECCS.

The period between hot node uncovery and reflooding is the period when the hot node has the lowest heat transfer. The break that results in the longest period during which the hot node remains uncovered usually results 6-1

A,l$ ,W ' h&S 9 Y&&W' hW$'~'s' NEDo-24165 in the highest calculated PCT. If two breaks have similar times during which the hot node remains uncovered, then the larger of the two breaks will be limiting as it would have an earlier boiling transition time (i.e., the larger break would have a more severe LAMB / SCAT blowdown heat transfer analysis) .

Figures 6a and 6b show the variation with discharge and suction break size of the calculated time the hot node remains uncovered for Brunswick 1.

Based on these results the 0.8 discharge DBA was de crained to be the break that results in the highest calculated PCT in the 1.0 ft to DBA region. The determination of the 0.8 discharge DBA being the most limiting break was based on the reasoning discussed above. The detailed calculations for the Icad plant (Reference 11) confirmed that this procedure for determining the most limiting break is justified.

The conservative approach that was used for the lead plant of using the dis-charge DBA LAMB / SCAT results with the 0.8 discharge DBA SAFE /REFLOOD results for calculations for the 0.8 discharge DBA was used in all calculations for the analysis to determine the HAPLHCR's in Tables 4A through 4D.

The DBA (the complete severance of the recirculation discharge piping) results are shown on Figures Ib through $b. ,

The second most limiting location for the LOCA is the recirculation suction line. From Figure 6b, the suction DBA was determined to be the most limiting suction break. The results of the DBA suction break are shown on Figures Ic through Sc.

The single failure evaluation showing the remaining 2CCS following an assumed failure and the ef fects of a single failure or operator error that causes any manually controlled, electrically cperated valve in the ECCS to move to a position that could adversely affect the ECCS are presented in Reference 12.

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Figure 2a. Peak Cladding Temperature Following a Recirculation Discharge Line Break LPCI Injection Valve Failure, Break Area = 80% DBA

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Break Area = 2.4 ft2 i

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l NEDO-24165

7. REFERENCES

1. Letter, A. Schwencer (NRC) to J. A. Jones (CP&L), "Re: Brunswick Steam Electric Plant, Unit Nos.1 & 2," dated March 11, 1977.

2. Lett'er. Darrel C. Eisenhut (NRC) to E. D. Fuller (CE). " Documentation of the Reanalysis Results for the Loss-of-Coolant Accide=t (LOCA) of Lead and Non-Lead Plants." June 30, 1977.

3. General Electric Company Analytical Model for Loss-of-Ceolant Analysis in Accordance with 10CFR$0 Appendix K, NEDO-20566. (Draf t), submitted August 1974, and General Electric Refill Reflood Calculation (Supplement to SAFE Code Description) transmitted to the USAEC by letter, G.L. Gyorey (CE) to Victor Stello, Jr. (NRC), dated December 20, 1974.

4. " Safety Evaluation for General Electric ECCS Evaluatien Model Modifications,"

letter from K. R. Coller (NRC) to G.G. Sherwood (CE), dated April 12, 1977.

5. Letter, A. J. Levine (CE) to D. F. Ross (NRC) dated January 27, 1977,

" General Electric (CE) Loss of Coolant Accident (LOCA) Analysis Model revisions - Core Heatup Code CHASTE 05.

6. Le ter, A. J. Levine (CE) to D. B. Vassallo (NRC), dated March 14, 1977,

" Request for Approval for Use of Loss of Coolant Accident (LOCA) Evaluatior.s Model Code REFLOOD05."

7. "Supplerental Information for Plant Modification to Eli=inate Significant In-Cere Vibrations," Supplement 1, NEDE-21156-1 Septe=ber 1976.

8. "Supplerental Information for Plant Modification to EIL=inate Significant In-Core Vibrations," Supplement 2. NEDE-21156-2, January 1977. '

l'

9. Letter, R. Engel (CE) to V. Stello (NRC), " Answers to 53tC Questions on '

I NEDE-21156-2 " January 24, 1977.

I

10. Letter, G. L. Cyorey (CE) to V. Stello, Jr., dated May 12, 1975, d

" Compliance with Acceptance Criteria for 10CTR50.46."

f

11. Letter, George T. Berry (PASNY) to Robert k'. Reid (NRC), " James A. '!

' FitrPatrick Nuclear Power Plant ECCS Analysis Docket No. 50-333," dated l July 29,1977.

j

12. Brunswick Steam Electric Plant Final Safety Analysis Report.

.f 7-1/7-2 1

-