ML102030352

Nuclear Generating Plant - Preliminary Review of Progress Energy'S 10 CFR 50.59 Evaluation

ML102030352
Person / Time
Site:	Crystal River
Issue date:	07/28/2010
From:	Jason Paige Plant Licensing Branch II
To:
Paige, Jason C, NRR/DORL,301-415-5888
References
Download: ML102030352 (2)

Text

SUBJECT:

PRELIMINARY REVIEW OF PROGRESS ENERGY'S 10 CFR 50.59 EVALUATION Title 10 of the Code of Federal Regulations (10 CFR), Section 50.59 establishes the conditions under which licensees may make changes to the facility or procedures and conduct tests or experiments without prior NRC approval. Regulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes, Tests and Experiments," and Nuclear Energy Institute (NEI) 96-07, rev. 1, "Guidelines for 10 CFR 50.59 Implementation," provide guidance for evaluating and reporting changes under 10 CFR 50.59. Progress Energy discussed its 10 CFR 50.59 approach for Crystal River's, Unit 3 delaminated containment repair activities during the June 30, 2010 meeting.

The Nuclear Regulatory Commission's (NRC's) inspection program includes a module to assess each licensee's use of 10 CFR 50.59 for evaluating and reporting changes, tests and experiments. The NRC Regional (Region II for Crystal River) inspectors have the responsibility and the lead for monitoring the licensees' activities under 10 CFR 50.59. Due to the complexity and safety significance of the containment delamination and repairs at Crystal River, Unit 3, the office of Nuclear Reactor Regulation (NRR) is providing support to the Region II review of Progress Energy's 10 CFR 50.59 screening and evaluation. Below is NRC feedback generated from the preliminary review of the licensee's 10 CFR 50.59 evaluation and information presented during the June 30, 2010 public meeting.

Summary of NRC feedback General feedback

• Most significant NRC feedback was generated from the three criteria below from 10 CFR 50.59(c)(2).

• The NRC would review the 50.59 evaluation after it is complete by the licensee via inspection. Below are examples of issues we believe may need to be further addressed or explored as you proceed with your 50.59 evaluation.

• The NRC staff supports the licensee's commitment to conduct a structural integrity test to assure the structural integrity of the containment as presented during the June 30 th public meeting.

Result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component (SSC) important to safety previously evaluated in the FSAR (as updated)?

NRC feedback:

• From looking at NEI 96-07, rev. 1, Section 4.4 discusses compensatory actions to address nonconforming or degraded conditions. NEI 96-07 states, "If the licensee corrective action is either to accept the condition "as is" resulting in something different than its as-designed condition, or to change the facility or procedures, 10 CFR 50.59 should be applied to the corrective action, unless another regulation applies, e.g.,

10 CFR 50.55a. In these cases, the final corrective action becomes the proposed change that would be subject to 10 CFR 50.59." Accordingly, the evaluation should address leaving the through-wall vertical cracks "as is" does not impact its current design bases of the containment structure under design loads and load combinations described in the FSAR.

• The licensee should demonstrate that the repaired containment structure meets applicable codes described in the FSAR.

Result in a design basis limit for a fission product barrier as described in the FSAR (as updated) being exceeded or altered?

NRC feedback:

• Containment pressure (which the licensee is not changing) is a design basis limit.

• Even though the minimum design tendon tension force is not included in the FSAR, the parameter is fundamental to the barrier's integrity.

• The design basis calculations and structural integrity test is a means to provide reasonable assurance that the containment after the repair can withstand its design pressure during an accident.

• Also, as stated during the June 30, 2010 public meeting, the licensee is increasing the specified minimum design compressive strength of the existing containment concrete from 5000 psi to 5800 psi, which the minimum design compressive stress is identified in the FSAR.

Result in a departure from a method of evaluation described in the FSAR (as updated) used in establishing the design bases or in the safety analyses?

NRC feedback:

• Per NEI 96-07, rev. 1, Section 4.3.8, if the licensee plans on using a different method of evaluation as described in the FSAR, then the licensee needs to show that the results are essentially the same as, or more conservative than, previous results.

• Similarly, licensees can also use different methods without first obtaining a license amendment if those methods have been approved by the NRC for the intended application.

• In summary, if the licensee changes the methodology from the one used for original calculations and analyses, this item under 10 CFR 50.59 will be answered "yes." However, no licensing action (i.e., license amendment) is needed per NEI 96-07, rev. 1, if the licensee meets the criteria in Section 4.3.8.

• The licensee should evaluate the impact of leaving the through wall vertical cracks "as is" on the design and analysis method of evaluation described in the FSAR.