ML20079G213

Forwards Pages 1.5-1 Through 1.5-5 of Facility FSAR Re Status of R&D Programs.Pages Will Be Included in Next FSAR Amend

ML20079G213
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	05/26/1982
From:	Tramm T COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
4211N, NUDOCS 8206080356
Download: ML20079G213 (6)

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) one First N;tionit Ptzs, Chic?go. Illinois O' Address Reply to: Post Office Box 767 Chicago. Illinois 60690 Ma y 26, 1982 4

Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission l Washington, DC 20555 t

Subject:

Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Status of Research Programs NRC Docket Nos. 50-454, 50-455, 50-456 and 50-457

Dear Mr. Denton:

This is to provide advance copies of Byron /Braidwood FSAR information regarding the status of research and development 4

programs discussed in the PSAR.

I Enclosed are fifteen copies of pages 1.5-1 through 1.5-5 of the Byron /Braidwood FSAR. These will be included in the next amendment. One signed original and fif teen copies of this letter are also provided.

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Please address questions regarding this matter to this i office.

Very truly yours, t /A-& W T. R. Tramm l

Nuclear Licensing Administrator

. Im I

4211N gh 8206080356 820526 PDR ADOCK 05000454 A PDR

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B/B-FSAR 1.5 REQUIREMENTS FOR FURTHER TECHNICAL INFORMATION The design of the Byron /Braidwood units is based upon proven concepts which have been developed and successfully applied to the design of pressurized water reactor systems. There are currently no areas of research and development which are required for operation of this plant.

At the time of issuance of construction permits for the Byron /Braidwood units , the Preliminary Safety Analysis Report (PSAR) and the standard design report which it referenced, RESAR-3, identified certain research and development programs which were incomplete. These programs, which have been suc-cessfully completed, have provided technical information which has been used either to demonstrate the safety of design, more sharply define margins of conservatism, or lead to design improvement s . Reference 1 presents descriptions of those safety-related research and development programs which have been carried out for, or by, or in conjunction with, Westing-house Nuclear Energy Systems, and which are applicable to Wese-inghouse Pressurized Water Reactors. The discussion which follows in section 1.5 documents the completion of the Con- -

s truction Permit stage research programs. -

1.5.1 Programs Required for Plant Operation Two programs were identified as required for plant design and operation in the PSAR:

a. Core Stability Evaluation and,

b. Fuel Rod Burst Program.

Both progams are complete. The Fuel Rod Burs t Program was completed at the time of the PSAR. The core Stability Evalu-ation Program was not. A discussion of the Core Stability Evaluation Program follows.

1.5.1.1 Core Stability Evaluation The program to establish means for the detection and control of potential xenon oscillation and for the shaping of the axial power distribution for improved core performance has been satisfactorily completed. See item 1, reference 2, for a fur-ther discussion of the tests and results.

1.5.2 Other Programs Not Required for Plant Operation The following programs were not complete at the time of the PSAR but are now satisfactorily complete.

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. , 1 B/B-FSAR 1.5.2.1 Fuel Development Program for Operation at High Power Dens itie s The program to demonstrate the satisfactory operation of fuel at high burnup and power densities has been satisfactorily completed. See item 8, reference 2, for a further discussion of the program and its results.

1.5.2.2 Blowdown Forces Program Westinghouse has completed BLODWN-2 an improved digital compu-ter program for the calculation of local fluid pressures, flows and density transient in the primary coolant systems during a LOC A.

BLODWN-2 is used to evaluate the effects of blowdown forces in this application. Refer to item 15 in reference 4 fo,r a fur-ther discussion of the tests and results.

1. 5 . 2.3 Blowdown Heat Transfer Testing (Formerly Titled Delayed Departure Ocom Nucleate Boiling)

The NRC Acceptance Criteria for Emeroency Core Cooling Systems for Light-Water ' Power Reactors was i .aed in section 50.46 of 10CFR50 on December 28, 1973. It uafines the basis and conser-vative assumptions to 'Ina used in the evaluation of the erfor-mance of Emergency Core Cooling Systems (ECCS) . Westin house believes that some of the conservatism of the criteria s asso-ciated with the manner in which transient DNB phenomena are treated in the evaluation models. Transient critical heat flux data presented at the 197 2 specialists meeting of the Committee on Reactor Safety Technology (CREST) indicated that the time to DNB can be delayed under transient conditions. To demonstrate the conservatism of the ECCS evaluation models, Westinghouse initiated a program to experimentally simulate the blowdown phase of a LOCA. This testing ia part of the Electric Power Research Institute (EPRI) sponsor ld Blowdown Heat Transfer Program, which was started early in 1976. Testing was com-pleted in 1979 A DNb correlation developed by Westinghouse

Ecom these test results is used in the ECCS analyses for l

Byron / Braidwood .

Objective -

The objective of the Blowdown Heat Transfer Test was to deter-mine the time that DNB occurs under LOCA conditions. This information was used to confirm a new Westinghouse transient DNB correlation. The steady-state DNB data obtained from 15 x 15 and 17 x 17 test programs was used to assure that the geometrical differences between the two fuel arrays is cor-rectly treated in the transient correlations.

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B/ B-FSAR Program The progr am was divided into two phases. The Phase I tests started from steady state conditions, with sufficient power to maintain nucleate boiling throughout the bundle, and progressed through controlled ramps of decreasing test section pressure or flow initiated DNB. By applying a series of controlled condi-tions, investigation of the DNB was studied over a range of qualities and flows, and at pressures relevant to a PWR blow-down.

Phase I provided separate-effects data for heat transfer cor-relation development.

Typical parameters used for Phase I testing are shown below.

Parameters Nominal Value Initial Steady State Conditions Pressure -

1250 to 2250 psia Test section mass velocity 1.12 to 2.5 x 106 lb/hr-ft2 Core inlet temparature 550 to 6000F -

Maximum heat flux 306,000 to 531,000 Beu/hr-ft2 Transient Ramp Conditions Pressure decrease 0 to 350 psi /sec and subcooled depressurization from 2250 psia Flow decrease 0 to 100 percent /sec Inlet enthalpy Constant Phase II simulated PWR behavior during a LOCA to permit defini-tion of the time delay associated with onset of DNB. Tests in this phase covered the large double-ended guillotine cold leg break. All tests in Phase II were also started after estab-lishment of typical steady state operating conditions. Th e fluid transient was then initiated, and the rod power decay was programmed in such a manner as to simulate the actual heat input of fuel rods. The test was terminated when the heater rod temperatures r'eached a predetermined limit.

Typical parameters used for Maase II testing are shown below.

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B/B-FSAR l

l Parameter Nominal Value Initial Steady State Conditions Pressure 2250 psia Test section mass velocity 2.5 x 106 lb/hr-f t2 Inlet coolant temperature 5450F '

Maximum heat flux 531,000 Btu /hr-f t2 Transient Conditions Simulated break Double-ended cold leg guillotine breaks Test Description [1 I

The experimental program was conducted in the J-Loop at the

• Westinghouse Forest Hills Facility with a full length 5 x 5 rod bundle simulating a section of a 15 x 15 fuel assembly to determine DNB occurrence under LOCA conditions.

The heater rod bundles used in thisfprogram were internally-heated rods, capable of a maximum power of 18.8 kW/ft, with a total power of 135 kW (for extended periods) over the 12-foot heated length of the rod. Heat was generated internally by means of a varying cross-sectional resistor which approximates a chopped. cosine power distribution. Each rod was adequately instrumented with a total of 12 clad thermocouples.

Results The experiments in the DNB Facility resulted in cladding tem-perature and fluid properties measured as a function of time throughout the blowdown range from 0 to 20 seconds.

1 Facility modifications and installat' ion of the initial test bundle were completed. A series of ' shakedown tests in the J-Loop were performed. These tests provided data for ins tru-mentation calibration and check-out, and provided information regarding facility control and performance. Initial program tests were performed during the first half of 1975. Under the sponsorship of EPRI, testing was reinitiated during 1976 on the same test bundle. The testing was terminated in November and plans were made for a new test bundle and further testing dur-ing 1978-19 79. These tests were completed in December of 1979.

1. 5 .3 References

1. F. T. Eggleston, " Safety-Related Research and Development for Westinghouse Pressurized Water Reactors , Program Sum-maries ," WCAP-8768, Oc tober 1978.

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2. F. T. Eggleston, " Safety-Related Research and Development for  ;

Westinghouse PWR's Program Summaries ," WCAP-8 768. Spring 1976 '

Edition.

3. " Safety-Related Research and Development for Westinghouse PWR's Program Summaries, WCAP-8458. Fall 1977 Edition.

4. " Safety-Related Research and Development for Westinghouse PWR's Program Summaries, WCAP-8004. Fall 197 2 Edition. -

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