ML19256D285
| ML19256D285 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/29/1971 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910170794 | |
| Download: ML19256D285 (11) | |
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Peter A. Morris, Director, Division of Reactor Licensing THREE MILE ISMND NUCLEAR STATION, UNIT 1, DOCKET No. 50-289 The information saluaitted by the applicant, including Amendment 18 dated March 22, 1971, has been reviewed and evaluated I
j by the Material Engineering Branch, DRS. Our evaluation of the l
immaas reviewed is enclemed. Tentative cenelusions are enclosed in perantheses; the summary of actions to be taken to resolve open issues is emelesed is brackets.
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s Three Mile Island-1
_ REACTOR COOLANT SYSTEM Fracture Touchness Criteria The reactor vessel was designed in accordance with the ASME Boiler and Pressure Vessel Code,Section III. Current ASME Section III Code rules permit that a vessel be pressurized only above a temperature equal to the sum of the Nil Ductility Transition (NDT) temperature and 60*F.
The NDT temperature, according to paragraph N-331 of the Code, can be obtained by either the dropweight test (DWT) or the Charpy V-notch (C ) impact test.
y Recent fracture toughness test data indicate that these rules do not always guarantee adequate fracture toughness of ferritic materials. The Charpy V-notch tests are adequate to measure the upper shelf fracture energy value; however, they generally, do not correctly pradict the NDI temperature. The latter, therefore, must be obtained from other tests, such as the DWT test.
Quite often, also, it is difficult to define the transition temperature region in which fracture toughness of ferritic materials increases rapidly with temperature from the C test curves.
In addition, this transition y
temperature region shif ts to higher temperature when the thickness of the specieen tested is incressed (si:e effect).
We have reviewed the available fracture toughness. data for the reactor vessel and applied our proposed fracture toughness criteria to arrive at a lowest pressurization temperature of 265*F.
}k[)0 222
Three Mile Island-l We intend to specify the following limits in the Technical Specifications:
1.
For the first two years, the reactor coolant system should be operated and hydrotested in such a manner that at temperatures below 265'F, (a) the pressure does not exceed 550 psig (i.e.,
i'5 percent of the norral operating pressure), and (b) the rate of temperature change does not exceed 50'F/hr.
2.
Full pressurization of the reactor coolant system is acceptable during the first two years at temperatures above 265'F.
These limits are based on the effect of component thickness on the fracture ;caghness transition temperature and will apply only to the initial two yetrs of plant operation. The limits will be reviewed when data are made available from the material surveillance program test results following withdrawal of the first capsule.
(We are discussing these operating limitations with the applicant and expect to resolve this issue before the ACRS meeting.)
Three Mile Island-1
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REACTOR COOLANT SYSTEM Sensitized Stainless Steel The applicant has stated that the sensiti 3ation of non-stabilized stainless steel (s.s.) will be avoided. Ferritic nozzle and pipe ends will be buttered with stainless steel applied by the weld deposition 4
technique. This procedure will eliminate use of separate wrought austenitic s s. safe-ends which could become furnace sensitized. The applicant has also stated that no nitrogen bearing s.s. was used in the reactor coolant pressure boundary. We conclude that the steps taken to avoid sensitization of austenitic s.s. during fabrication are acceptable.
Three Mile Island-l REACTOR COOLANT SYSTEH Electroslag Welding The electroslag (E-S) welding process was used for longitudinal shell seams in th". steam generators and in the pressurizer shell. The E-S velding procedures were identical to those for the Oconee Unit 1 j
(DOCKET NO. 50-269) steam generators and the Dresden 2 and 3 reactor vessels (Appendix F to DOCKET NOS. 50-237 and 50-249). Appendix A to the ACRS REPORT for the Dresden 2 and 3 Reactor Vessels E-S Welding summarizes the process variables and the controls applied. We conclude that the E-S welding process as used for the steam generators and pressurizer is acceptable.
1A50l!26
Three Mile Island-1 REACIOR COOLANT SYSTEM Pump Flywheel Inteerity The primary system pump flywheels for the Three Mile Island Unit I have been manufactured by Allis-Chalmers and are similar to those used in other plants. The flywheel material is ASTMA-516-66 Grade 65; the plate was normalized to refine the grain structure and to improve tough-ness. The finished flywheels were subjected to 100% volumetric UT inspection. Finished machine bores were subjected to magnetic particle inspection.
The applicant has stated that the pump flywheels will be accessible for inservice inspection. An opening at the bottom of the motor provides access to the bottom surface of the lower flywheel, allowing ultrasonic
. inspection in the axial direction. Handholes in the upper part of the motor give access to the top and rim of the large diameter flywheel segment and allow ultrasonic inspectior. of this segment in both the.
axial and radial direction.
We conclude, on the basis of the initial inspection performed and the proposed inservice inspection,that the flywheels are acceptable.
(To determine that the fracture toughness of the flywheels is adequate, we have asked the applicant to furnish the mechanical properties of the flywheel plates. We expect to obtain and review this information prior to the ACRS meeting.)
1450 226
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Three Mile Island-1 REACTOR COOLANT SYSTEM Inservice Inst,ection Program The components of the primary coolant system subject to inservice inspection include the reactor vessel, pressurizer, steam generators, pumps, valves, and pi, ping. All dissimilar velds will be accessible for inspection. Representative longitudinal welds and circumferential velds on piping, steam generators, pressurizer, and pump casings are inspectable. Representative welds on the vessel head are inspectable, as are longitudinal and circumferential welds of the vessel shell. A 42-inch annulus exterior to the vessel is provided for external access for~ f nspection. The vessel insulation stands 15 to 18 inches clear of the shell.
We conclude that the inservice inspection program satiafica tha provisions of the July 31, 1969 AEC document, " Inservice Inspection Requirements for Nuclear Power Plants Constructed with Limited Accessi-bility for Inservice Inspection" and is acceptable.
1450 227
Three Mile Island-l
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REACIOR COOLANT SYSTEM Reactor Vessel Material Surveillance Program The material surveillance program is consistent with programs that have been accepted on previous PWR plants. The reactor vessel material surveillance program submitted by the applicant includes provisions with respect to total number of specimen capsules placed in the reactor vessel, number of capsules scheduled to be withdrawn and tested, archive material available for additional specimens if required later in the service life of the vessel, and material chemistry documentation.
We conclude that this program will adequately monitor radiation-induced changes in material fracture toughness properties of the ferritic mate-rials of the reactor vessel, during its service life.
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Three Mile Island-l REACTOR COOLANT SYSTEM Leak D tection Svstem The leak detection system provided for the reactor coolant pressure boundary is cuitably sensitive, redundant, diverse in operation, and is provided with suitable control room alarms and readouts. The major i
components of the system are the containment atmosphere radioactivity monitors and the level and flow indicators on the containment sump and cooling coil condensate. Indirect indication of leakage can be obtained from the containment pressure and temperature indicators.
The system is similar to the one recently approved at Oconee. We conclude that the leak detection system is adequate to detect small cracks in the reactor coolant pressure boundary in a timely manner.
1450.229
Three Mile Island-l CONTAINHINT Incegrated Leak Test Program (The integrated leak test program presented in the FSAR is incomplete and does not meet the AEC's suggested test practices in regard to test frequency. The applicant is aware of our concern and an acceptable t
program will be developed in the Technical Specifications before licensing of the plant is completed.)
1450 230 I
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Three Mile Island-l (ISSILE PROTECTION Containment and Engineered Safety Features The applicant has provided missile protection for the reactor building, and liner, and for the components of the Engineered Safety Features.
Missiles considered i,ncluded valve stems and bonnets, instrument thimbles, nuts and bolts, and control rod drive mechanisms. The design of missile protection shielding was based on the following criteria:
a) The thickness of missile shields provided is at least 2 times the potential penetration depth.
b) The local collapse effects were analyzed and sufficient
' reinforcing has been provided.
c) The missile was considered as a point load on the structure.
d) Proper application was made of penetration formulas and analytical methods.
We conclude that the design criteria used by the applicant will provide a basis for adequate missile protection for this plant.
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