**RSICC CODE PACKAGE PSR-445**

**1. NAME AND TITLE**

VISA2: Code System to Calculate Probability of Reactor Vessel Failure.

**2. CONTRIBUTORS**

Pacific Northwest Laboratory, Richland, Washington through the Energy Science and Technology Software Center, Oak Ridge, Tennessee.

**3. CODING LANGUAGE AND COMPUTER**

FORTRAN 77; DEC VAX or IBM PC (P00445MNYCP00).

**4. NATURE OF PROBLEM SOLVED**

VISA2 (Vessel Integrity Simulation Analysis) was developed to estimate the failure probability of nuclear reactor pressure vessels under pressurized thermal shock conditions. The deterministic portion of the code performs heat transfer, stress, and fracture mechanics calculations for a vessel subjected to a user-specified temperature and pressure transient. The probabilistic analysis performs a Monte Carlo simulation to estimate the probability of vessel failure. Parameters such as initial crack size and position, copper and nickel content, fluence, and the fracture toughness values for crack initiation and arrest are treated as random variables. Linear elastic fracture mechanics methods are used to model crack initiation and growth. This includes cladding effects in the heat transfer, stress, and fracture mechanics calculations. The simulation procedure treats an entire vessel and recognizes that more than one flaw can exist in a given vessel. The flaw model allows random positioning of the flaw within the vessel wall thickness, and the user can specify either flaw length or length-to-depth aspect ratio for crack initiation and arrest predictions. The flaw size distribution can be adjusted on the basis of different inservice inspection techniques and inspection conditions. The toughness simulation model includes a menu of alternative equations for predicting the shift in the reference temperature of the nil-ductility transition.

VISA2 is an upgraded release from the original VISA program developed by U.S. Nuclear Regulatory Commission staff. Improvements include a treatment of cladding effects; a more general simulation of flaw size, shape and location; a simulation of inservice inspection; a revised simulation of the reference temperature of the nil-ductility transition; and treatment of vessels with multiple welds and initial flaws.

**5. METHOD OF SOLUTION**

The solution method uses closed form equations for temperatures, stresses, and stress intensity factors. A polynomial fitting procedure approximates the specified pressure and temperature transient. Failure probabilities are calculated by a Monte Carlo simulation.

**6. RESTRICTIONS OR LIMITATIONS**

Maximum of 30 welds. VISA2 models only the beltline (cylindrical) region of a reactor vessel. The stresses are a function of the radial (through-wall) coordinate only and are independent of the axial and circumferential coordinates. The polynomial approximation of pressure-temperature transients requires the user to avoid (or smooth) transients with pressure spikes and other discontinuities.

**7. TYPICAL RUNNING TIME**

Typical problems require five minutes on a DEC VAX11/780. NESC executed the sample problem in 50 seconds on a DEC VAX6220.

**8. COMPUTER HARDWARE REQUIREMENTS**

Both DEC Vax and PC versions were submitted to NESC in February 1989. In 1999 they were transmitted to RSICC through the ESTSC and merged into one package. The sample problem required 61 Kbytes of memory on a DEC VAX6220.

**9. COMPUTER SOFTWARE REQUIREMENTS**

The DEC Vax version runs under VMS and requires a Fortran 77 compiler. The PC version was tested at NESC with Microsoft Fortran 3.3 and 4.1 compilers. RSICC tested the included PC executable in a DOS window of Windows95.

**10. REFERENCES**

**a) included in documentation:**

F.A. Simonen, K.I. Johnson, A.M. Liebetrau, D.W. Engel, and E.P. Simonen, "VISA-II A Computer Code for Predicting the Probability of Reactor Pressure Vessel Failure," NUREG/CR-4486, PNL-5775 (March 1986) with corrections.

**b) background information:**

D.L. Stevens, F.A. Simonen, J. Strosnider, Jr., R.W. Klecker, D.W. Engel, and K.I. Johnson, "VISA, A Computer Code for Predicting the Probability of Reactor Pressure-Vessel Failure," NUREG/CR-3384, PNL-4774 (September 1983).

E.P. Simonen, K.I. Johnson, and F.A. Simonen, Vessel Integrity Simulation Analysis (VISA) Code Sensitivity Study, NUREG/CR-4267, PNL-5469 (December 1985).

**11. CONTENTS OF CODE PACKAGE**

Included are the referenced documents in (10.a) and one DS/HD diskette which includes source, a PC executable, and sample problem input and output.

**12. DATE OF ABSTRACT**

April 2000.

** KEYWORDS:** FRACTURE MECHANICS; HEAT TRANSFER; REACTOR SAFETY