Document: NRC Regulatory Guide
Document ID: c55ba6c5-aa2d-4ad6-aba2-2001e16524ab
Document Type: regulatory_guide
Title: Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations (Rev. 1)
Source: NRC Regulatory Guide Division 1
Source URL: https://www.nrc.gov/docs/ML1221/ML12216A015.pdf
Revision Date: 2023-06
Chapter: 
Section ID: RG-1.150
CFR Part: 
CFR Title: 

Content:
is guide is derived from the safety requirements of the Atomic Energy Act of 1954, as amended, and the Energy Reorganization Act of 1974, as implemented by the Commission's regulations. In particular, § 50.55a, "Codes and Standards," of 10 CFR Part 50 requires, in part, that structures, systems, and components be designed, fabricated, erected, constructed, tested, and inspected to quality standards commensurate with the importance of the safety function to be performed. 3.2 Need for NEPA Assessment The proposed action is not a major action, as defined by paragraph 51.5(a)(10) of 10 CFR and does not require an environmental impact statement. 1.150-19 4. RELATIONSHIP TO OTHER EXISTING OR PRO- POSED REGULATIONS OR POLICIES Recommendations of this guide would be supplemental to the requirements of Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," of the ASME Code, which is adopted by § 50.55a, "Codes and Standards," of 10 CFR Part 50. 5. SUMMARY This guide was initiated as a result of a request from NRR. Preliminary results of the round robin UT examination procedures following ASME Code procedures indicate a need for additional guidelines to the existing ASME Code procedures to control equipment performance, calibration block specifications, and scanning procedures to improve the reproducibility of results and detectability of through-thick- ness flaws. Minimum ASME Code requirements do not specify the details of recording requirements that are essential to evaluate flaws. This deficiency in the Code rules makes it difficult for the NRC staff or their consultants to review, analyze, and assess the UT data to determine the flaw size and evaluate the system safety when the data are made available to NRC at a later date. The present data obtained from UT equipment of uncertain and unspecified performance lead to discussions and delays in the review process resulting in loss of NRC staff time and loss of plant availability and power