ML24116A063

Periodic Update of the Updated Final Safety Analysis Report, Appendix B, Table of Contents

ML24116A063
Person / Time
Site:	Point Beach
Issue date:	04/23/2024
From:	Point Beach
To:	Office of Nuclear Reactor Regulation
References
L-2024-063
Download: ML24116A063 (1)

Text

Appendix B Table of Contents FSAR

APPENDIX B TABLE OF CONTENTS

B.1 DOES NOT EXIST B.2 DESIGN PARAMETERS AND PLANT COMPARISONS (Historical) - - - - B.2-1 B.3 INITIAL PLANT DESIGN - - - - - - - - - - - - - - - - - - - - - - - - - B.3-1

UFSAR 2017 Page TOC - B-i of i Plant Comparisons FSAR Appendix B.2 (Historical)

B.2 DESIGN PARAMETERS AND PLANT COMPARISONS The design parameters of the Point Beach Nuclear Plant were initially provided in a comparison with H. B. Robinson, Indian Point 2, and Ginna Station. This comparison provided an informational reference of similar aspects of similar plants constructed during the same time period to demonstrate that the technology employed at Point Beach was proven in multiple applications.

The design parameters listed were considered valid at the time of license issuance, and have been retained for historical context in the following Table B.2-1 and paragraphs. Note: The information provided is not currently reflective of the PBNP design specifications, and therefore should not be used as the basis of any Safety Evaluation without prior verification against current information provided elsewhere in the FSAR.

In 2003 a measurement uncertainty recapture power uprate was performed increasing the rated thermal power level to 1540 MWt. The tables of this section have not been updated since this Appendix is historical.

Design Highlights

The design of each Point Beach unit is based upon proven concepts which have been developed and successfully applied in the construction of pressurized water reactor systems. In subsequent paragraphs, a few of the design features are listed which represent slight variation or extrapolations from units presently operating such as San Onofre and Connecticut-Yankee.

POWER LEVEL - The license application power level of 1518.5 MWt is smaller than the capability of the Prairie Island plant and larger than the capability of the Ginna plant. This level is a reasonable increase over power levels of pressurized water reactors now operating.

REACTOR COOLANT LOOPS - The Reactor Coolant System for each Point Beach unit consists of two loops, the same as the Prairie Island and Ginna Units.

PEAK SPECIFIC POWER - Based on the design hot channel factors, operation at a primary heat output of 1518.5 MWt corresponds to a peak specific power of 16.0 kw/ft. This design rating is slightly lower than that licensed in Ginna (16.5 kw/ft) as well as that of Prairie Island (17.4 kw/ft). The maximum overpower condition is 17.9 kw/ft (112%) compared to 19.6 kw/ft (118%) for Prairie Island and 18.5 kw/ft for Ginna.

FUEL ASSEMBLY DESIGN - The fuel assembly design incorporates the rod cluster control concept in a canless assembly utilizing a spring clip grid to provide support for the 14 x 14 array of fuel rods. This concept incorporates the advantages of the Yankee canless fuel assembly and the Saxton spring clip grid with the rod cluster control scheme. Extensive out-of-pile tests have been performed on this concept and operating experience is available from the San Onofre and Connecticut-Yankee plants.

ENGINEERED SAFETY FEATURES - The engineered safety features provided are similar to those provided for the Connecticut-Yankee plant, augmented by borated water injection accumulators. There is a safety injection system of the Connecticut-Yankee type which can be operated from emergency on-site diesel power. The system design is such that it can be tested

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while the plant is at power. There is air recirculation cooling for post-loss-of-coolant conditions which utilizes the normal ventilation fans. A containment spray system provides cool, chemically-treated, borated water spray into the containment atmosphere for additional cooling capacity, and provides a means of rapidly reducing the concentration of airborne halogen fission products in the containment atmosphere.

EMERGENCY POWER - In addition to the multiple ties to outside sources for emergency power, four diesel generator units are provided as backup power supplies for the case of loss of all outside power. Each generator is capable of operating sufficient safety injection and containment cooling equipment to ensure an acceptable post-loss-of-coolant pressure transient in the affected unit, and safe shutdown of the other unit.

NET LOAD REJECTION - Each of the Point Beach units is designed to accept loss of 50% of external load without a reactor or turbine trip. This is accomplished by an automatic control system which dumps steam to the condenser and atmosphere as a short term supplemental load to provide time for the reactor control system to reduce the reactor output without exceeding acceptable core and coolant conditions. No unique or unproven features are required in the reactor control system to accomplish this.

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Initial Plant Design FSAR Appendix B.3

B.3 INITIAL PLANT DESIGN Research and development (as defined in Section 50.2 of the Commission's regulations) was conducted regarding core design details and parameters, analytical methods for kinetics calculations, thermal shock and its effects on reactor vessel integrity, the safety injection (emergency core cooling) system, xenon stability and related control systems, containment spray additive effectiveness, and capability of reactor internals to resist blowdown forces.

Core Design

The nuclear design, including fuel configuration and enrichments, control rod pattern and worths, reactivity coefficients, and boron requirements are presented in Section 3.2 and the thermal-hydraulics design parameters are also in Section 3.2. Section 3.2 presents the fuel, fuel rod, fuel assembly, and control rod mechanical design. The core design incorporates fixed burnable poison rods (Reference 1) in the initial loading and, when necessary, in subsequent core reloads to ensure a negative moderator reactivity temperature coefficient at operating temperature. This improves reactor stability and lessens the consequences of a rod ejection or loss-of-coolant accident. The mechanical design is presented in Section 3.2.

Development Of Analytical Methods For Reactivity Transients From Rod Ejection Accidents

A control rod ejection accident is not considered credible since it would require the fracture of a control rod mechanism housing. Nevertheless, the reactivity and associated pressure and temperature transients for this accident have been analyzed. Rod ejection analyses for this plant were performed using the CHIC-KIN code(Reference 2), which uses a point reactor kinetics model and a single channel fuel and coolant description. The rod ejection analysis results are given in Section 14.2.6 of this report, together with a brief description of the CHIC-KIN code.

These analyses show that the temperature and pressure transients associated with a rod ejection accident do not cause any consequential damage to the reactor coolant system. The consequences of a rod ejection accident are now lessened because the moderator coefficient of reactivity is always negative at operating conditions. In addition, the effects of rod ejection are inherently limited in this reactor, in which boric acid chemical shim is employed, since full-length control rods need only to be inserted sufficiently to handle load changes.

The initial cores contain fixed burnable poison rods. These, by allowing a reduction in the chemical shim concentration, ensure that the moderator coefficient of reactivity is always negative at operating temperature. The burnable poison rods, contain borosilicate glass. Critical experiments were conducted at the Westinghouse Reactor Evaluation Center using rods containing 12.8 w/o boron and Zircaloy clad UO2 fuel rods, 2.27% enriched. These values are also typical of this plant's initial core. The experiments showed that standard analytical methods can be used to calculate the reactivity worth of the burnable poison rods. The design basis and critical experiments are described in Reference 1. In-core testing completed in the Saxton reactor showed satisfactory performance of these rods.

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Safety Injection System (SI) Design

The design of the safety injection system includes nitrogen-pressurized accumulators to inject borated water into the reactor coolant system to rapidly and reliably reflood the core following a loss-of-coolant accident. Additional analyses have been performed to demonstrate that the accumulators, in conjunction with other components of the emergency core cooling system, can adequately cool the core for any pipe rupture. These analyses are presented in Section 14.3. The computer code, FLASH-R, used for the blowdown phase of the loss-of-coolant accident was modified to take into account the accumulator injection.

Research and development work has also been performed on the integrity of Zircaloy-clad fuel under conditions simulating those during a loss-of-coolant accident. Under the conservatively elevated temperatures predicted for the fuel rods during loss-of-coolant accident, the clad may burst due to a combination of fuel rod internal gas pressure and the reduction of clad strength with temperature. Burst cladding could block flow channels in the core, so that core cooling by the safety injection system would be insufficient to prevent fuel rod melting. Rod burst experiments have, therefore, been conducted on Zircaloy rods. The results of single-rod tests have been presented to the AEC in WCAP-7379-L Volume I (Westinghouse Proprietary) and Volume II.

The results of multi-rod tests have been reported to the AEC in WCAP-7495-L.

Systems For Reactor Control During Xenon Instabilities

Extensive analytical work has been performed on reactor core stability(Reference 3, Reference 4, and Reference 5). These indicated that a core of this size may be unstable against axial power redistribution, but is nominally stable against transverse (denoted X-Y) power oscillations. The plant was, therefore, provided with instrumentation and control equipment which would allow the operator to detect and suppress the axial power oscillations.

The original plant design provided for part-length control rods to control axial power oscillations which could result from the potential of power spatial redistribution caused by instabilities in local xenon concentration. Initial plant operations established that part-length control rods were not necessary for control of axial power oscillation. The part-length control rods at Point Beach Nuclear Plant Units 1 and 2 were subsequently removed.

Control information for axial power oscillation suppression is obtained from four long ion chambers, each divided into an upper and lower section mounted vertically outside the core. Both calculation and experimental measurements at SENA, San Onofre, and Haddam Neck have shown that this out-of-core instrumentation represents in-core power distribution adequately for power distribution control(Reference 5).

The control strategy is based on the difference in output between the top and bottom sections of the long ion chambers. If the operator allows axial power imbalance to exceed operating limits, various levels of protection are invoked automatically. These include generation of alarms, turbine power cutback, blocking of control rod withdrawal, and reactor trip. This capability is described in Section 7.0.

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Containment Spray Additive For Iodine Removal

Initially, sodium thiosulphate, Na2S2O3, was proposed as the iodine removal additive to the boric acid containment spray, but an evaluation program led to the selection of sodium hydroxide, NaOH. The results of the evaluation program are detailed in Reference 6 and are summarized briefly below:

1. Chemical Characteristics The Na2S2O3 solution was found to be oxidized by air at the post-accident temperatures in containment. NaOH was not unstable in this way.

2. Iodine Removal Characteristics The removal efficiency of the NaOH solution (at pH not less than 9.5) was comparable to that of the Na2S2O3 solution.

3. Materials Compatibility Corrosion rates of copper and copper-alloy heat exchanger tubing were reduced by more than an order of magnitude compared with high pH Na2S2O3 solution and were acceptably low (<0.01 mils/month at 1000F) for the application. These tests showed that pitting or local corrosion did not occur.

4. Radiolysis The NaOH solution was radiolytically stable, and liberates significantly less net hydrogen than the unstable Na2S2O3 solution.

Therefore, further testing has centered on the use of NaOH as the spray additive leading to the development of a technical basis for its inclusion in the plant engineered safety features as a means of fixing absorbed iodine, enhancing the natural rate of deposition of I2, and thus lowering the calculated off-site thyroid dose resulting from a postulated release of fission products to the containment atmosphere.

Section 6 gives a further discussion of iodine removal by the containment spray system.

Blowdown Capability Of Reactor Internals

The forces exerted on reactor internals and the core following a loss-of-coolant accident are computed by employing the BLODWN-2 digital computer program developed for the space-time-dependent analysis of multi-loop PWR plants. This program and the models used are discussed in Section 14.3.3.

Reactor Vessel Thermal Shock

Research was performed prior to and following the issuance of the Point Beach Operating Licenses to determine the effect of the addition of cold water from the accumulators to the reactor

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pressure vessel. This research considered three failure modes: the ductile failure mode, the fatigue yielding mode and the brittle failure mode. Analysis of the ductile and fatigue modes determined that reactor vessel integrity is maintained following addition of the accumulator water.

Extensive analysis of the brittle failure mode demonstrated adequate reactor vessel fracture toughness to prevent brittle failure for a period of several years of plant operation.

Subsequently, but before the end of the analyzed period, the NRC issued 10 CFR 50.61, Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events.

10 CFR 50.61 contains screening criteria for material fracture toughness, such that, if the materials of construction of the reactor vessel for a nuclear power plant maintain fracture toughness in compliance with the screening criteria, the functional integrity of the reactor vessel is ensured. It has been demonstrated that Point Beach Units 1 and 2 have adequate fracture toughness to be in compliance with the screening criteria of 10 CFR 50.61 through the end of their Operating Licenses. Therefore, brittle failure of the Point Beach reactor vessels is not a credible failure mode.

Identification Of Contractors

The Licensee engaged or approved the engagement of the contractors identified below in connection with the design and construction of the Point Beach Nuclear Plant. However, irrespective of the contractual arrangements discussed below, Wisconsin Electric Power Company is the sole holder of the operating licenses and, as the Licensee, is responsible for the design, construction, and operation of the Point Beach Nuclear Plant.

Point Beach Nuclear Plant was designed and built by Westinghouse Electric Corporation as prime contractor for the Licensee. Westinghouse contracted to provide a complete, safe, and operable nuclear power unit ready for commercial service. The project was directed by Westinghouse from the offices of its Atomic Power Divisions in Pittsburgh, Pennsylvania, and by Westinghouse representatives at the plant site during construction and plant startup. Westinghouse engaged the engineering firm of Bechtel Corporation, San Francisco, California, to provide the design of the structures and non-nuclear portions of the plant and to prepare specifications for the purchase and construction thereof. The Licensee reviewed the designs and specifications prepared by Westinghouse and Bechtel to assure that the general plant arrangements, equipment, and operating provisions were satisfactory.

The plant was constructed under the general direction of Westinghouse through Bechtel as the general contractor who was responsible for the management of all site construction activities and who either performed or subcontracted the work of construction and equipment erection.

NUS Corporation, Washington, D.C., was engaged as consultants on general site studies and meteorology. The firm of Murray and Trettel, Inc. assisted on meteorology. The firm of Dames and Moore, Chicago, Illinois, was engaged as consultants on earth science and geology. The engineering firm of Sargent and Lundy, Chicago, Illinois, was engaged to design cooling water facilities.

In addition, specialists in environmental sciences have participated in developing information concerning the Point Beach site. Harza Engineering Company of Chicago, Illinois, provided assistance in hydrology and the firm of John A. Blume and Associates of San Francisco,

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California, provided assistance is assessing the seismic history of the sites and establishing the ground accelerations associated with the design earthquake.

Stone and Webster Engineering Corporation of Boston, Massachusetts, provided assistance in system planning and site studies.

The Licensee had qualified representatives at the site throughout construction and, with their own personnel and consultants, inspected major components and construction installations. The Licensee's initial operating force performed acceptance testing of all structures and equipment.

REFERENCES

1. Wood, P. M., Baller, E. A., et al, Use of Burnable Poison Rods in Westinghouse Pressurized Water Reactors, WCAP-7113 (October 1967).

2. Redfield, V. A., CHIC-KIN... A Fortran Program for Intermediate and Fast Transients in a Water Moderated Reactor, WAPD-TM-479 (January 1, 1965).

3. Poncelet, C. G. and Christie, A. M., Xenon Induced Spatial Instabilities in Large Pressurized Water Reactors, WCAP-3680-20 (March 1968).

4. McGaugh, J. D., The Effect of Xenon Spatial Variations and the Moderator Coefficient on Core Stability, WCAP-2983 (August 1968).

5. Westinghouse Proprietary Report, Power Distribution Control in Westinghouse PWRs, WCAP-7208 (October 1968).

6. Westinghouse Confidential Report, Investigation of Chemical Additives for Reactor Containment Sprays, WCAP-7153 (March 1968).

7. Westinghouse Customer Report, Fracture Mechanics Evaluation of the Wisconsin Electric Power Company and Wisconsin Michigan Power Company Point Beach Nuclear Plant Unit 2 Reactor Vessel, WCAP-8737 (February 1977).

8. Westinghouse Customer Report, Fracture Mechanics Evaluation of the Wisconsin Electric Power Company and Wisconsin Michigan Power Company Point Beach Nuclear Plant Unit 1 Reactor Vessel, WCAP-8742 (February 1977).

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