ML19256D341
| ML19256D341 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/29/1970 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910170845 | |
| Download: ML19256D341 (6) | |
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Jui. : 0 4D Feter A. Morris, Director I O Division of Reactor Licensing THREE MILE ISLAND NUCLEAR STATION UNIT NO.1 - DOC:tET NO. 50- 289 Adequate responses to the enclosed request for additional information aro required before we can completa our review of the subject application.
These requests, prepared by the DRS Structural Engineering Branch, concern the reactor internal structures, reactor coolant pressure boundary, and Class I mechanical equipment material presented in Sections 1, 3, 4, 5, 6, 9, 11.
-3 14 of the application.
The review of topical reports IAW-10006, RAW-10008, and GAI No.1729, which are partiment to this application, is underway by the DRS staff and DRS consultante. Additional requests for information will follow the complatics of these reviews.
Eeview of the Isak detection system limits and the inservice inspection program will be completed when proposed Technical Specifications are submitted by the applicant.
Edson G. Case, Director Division of Reactor Standards Enclosare:
Request for Additional Information for Three Mile Island Dmit No. 1 cc w/ancl:
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REQUEST FOR ADDITIONAL INFORMATION THP2E MILE ISLRiD STATION - UNIT NO. 1 DOCKET NO. 50-289
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A.
Reactor Coolant Pressure Boundary 1.
Identify, for all components within the reactor coolant pressure boundary, those stainless steel parts of the pressure-containing membrane and load-bearing stainless steel members which are vital to the structural integrity of the reactor vessel and core that may have become furnace-sensitized by the fabrication sequence or partially-sensitized due to a slow cooldown rate from the solution heat treatment temperature. Such sensitized steels are recognized as susceptible to stress corrosion cracking.
Also identify if any stainless steel type 304 or 316 has been used for which the addition of nitrogen to enhance the strength of the material has been specified. Under certain conditions, such steels may also be susceptible to stress corrosion cracking.
Describe the plans which will be followed in the light of the susceptibility of these stainless steel cocponents or parts to stress corrosion cracking.
2.
If the process of electroslag velding was used in the fabrication of components within the reactor coolant pressure boundary, identify such components and describe the respective process specifications, 1454 317
control of variables, and quality control procedures which were applied in production to achieve the physical properties in the velds and heat affected zones comparable to those obtained in the veld procedure qualification tests.
B.
Reactor Internals Identify the reactor from which vibration test data vill be applicable to evaluate the adequacy of the Three Mile Island Unit 1 core support structure to withstand vibrations, and specify the vibration surveil-lance program which will be applied to the Three Mile Island Unit 1 design to demonstrate comparable performance.
C.
Other Safety-Related Systems and Components 1.
With respect to seismic ground motion and differential relative settlement, describe the design criteria which will be employed for critical piping buried or otherwise located outside of the contain-ment structure and for the situations where this piping enters the various structures.
2.
The seismic design of a nuclear power plant represents one of the more complex problems involving design interfaces between design organizations. To provide us with information to assess whether the seismic designs bases are correctly translated into the required specifications, drsvings, procedures, and instructions so that the necessary structures, systems, and components can withstand seismic loads combined with the other appropriate concurrent loads, turnish the following information:.
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3-a.
Describe the design organizations that are involved in the seismic design of all structures, systems, and components of the plant that are related to safety, b.
Describe the responsibilities of the involved design organi-zations in connection with the seismic design and the extent to which these responsibr.lities have been promulgated to thm organizations in writing. Identify the design organization that has been assigned overall responsibility for the adequacy of the seismic design.
c.
Describe the documented procedures that have been or vill be promulgated to provide for the interchange of needed design information and changes thereto and the coordination of the various facets of the seismic design among the involved design organizations, d.
Describe the manner by which you assure that the design pro-cedures described in c. above have been or are being followed, e.
Describe the design control measuras that have been or vill be instituted to verify or check the adequacy of the seismic design and by whom they will be performed. Describe the design procedures that have been or will be promulgated to provide for these measures.
f.
Describe the requirements that are or will be included in the purchase specifications for safety-related equipment to assure 1454 319
that this equipment is adequately designed to withstand and can function under the seismic design conditions. Describe the provisions that have been included in the purchase specifi-cation to permit the purchaser to verify that these require-ments are satisfied.
3.
Identify the Category 1 (01 ass I in the FSAR) piping systems and equipmeat, other than the primary coolant loop:. which have been designed to withstand the loads that would result from the combined I
Design Basis Earthquake (Maximum Hypothetical in the PSAR) and pipe
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rupture loads of the Design Basis Accident. Identify the loading combinations and the stress and deformation limits applicable to these loading combinations which were included in the derign criteria for these systems.
D.
Leak Detection Provide the sensitivities and response times of the leak detection systems described in Section 4.2.3.8 of the FSAR, including an estimate of time to.easure the smallest detectable unidentifiable coolant leak from the reactor coolant system.
E.
Reactor Vessel Material Surveillance Progra=
1.
The FSAR states that the material surveillance program will be in accordance with Topical Report BAW 10006, Reactor Vessel Material Surveillance Program. For Three Mile Island Unit 1, Table 6 of the report, indicates only two scheduled capsule withdrawals are 1454
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planned. A planned withdrawal schedule of a minimum of four capsules, is co u idered essential for the plant, and the nate-rial surveillance should be independent of all other Babcock and Wilcox nuclear systems. Unless such a withdrawal schedule, and material surveillance progran are adopted for the Three Mile Island Unit 1, additional justification is required for consid-eration of the proposed program.
2.
If the astimated inservice transition temperature shif t of the reactor vessel beltline arterial is based on data which are related to control of residual elements (specified ir ASTM-E-185-70 Section 3.1.3) including control of copper and vanadium, specify the results of the chemical tests of these vessel materials includ-ing the residual element content in weight percent to the nearest 0.01%..
F.
Missile Protection With respect to the primary pump motor flywheels discussed in Section 4.2.2.6 of the FSAR, furnish the proposed inservice inspection program which will be conducted to detect flaws which may develop in service.
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