ML15286A000

NRR E-mail Capture - Cooper - Relief Request (RR) RP5-01

ML15286A000
Person / Time
Site:	Cooper
Issue date:	10/08/2015
From:	Siva Lingam Plant Licensing Branch IV
To:	Shaw J Nebraska Public Power District (NPPD)
References
TAC MF6335
Download: ML15286A000 (5)

Text

NRR-PMDAPEm Resource From: Lingam, Siva Sent: Thursday, October 08, 2015 4:23 PM To: 'jdshaw@nppd.com' Cc: Markley, Michael; Alley, David; Hoffman, Keith; 'dwvande@nppd.com'

Subject:

Cooper - Relief Request (RR) RP5-01 (CAC No. MF6335)

Please note the following official requests for additional information (RAIs) for the subject relief request, and provide your responses within 30 days from the date of this e-mail. We transmitted the draft RAIs to you on October 1, 2015, and we had a clarification call on October 8, 2015. Your timely responses will allow the U.S.

Nuclear Regulatory Commission (NRC) staff to complete its review on schedule.

By letter dated June 9, 2015 (Agencywide Documents Access and Management System Accession No. ML15167A066), Nebraska Public Power District (NPPD, the licensee) submitted Relief Request (RR) RP5-01 for NRC to grant relief from, and authorize alternatives to, certain inservice inspection (ISI) code requirements for the Cooper Nuclear Station (CNS) pursuant to Title 10 of the Code of Federal Regulations (10 CFR),

Part 50, Section 50.55a. This RR requests to use the provisions of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Code Case N-795, Alternative Requirements for BWR [Boiling Water Reactors] Class 1 System Leakage Test Pressure Following Repair/Replacement Activities as an alternative to certain requirements in the ASME Code,Section XI, 2007 Edition through the 2008 Addenda for the Fifth 10-year ISI interval at CNS. To complete the review, the NRC staff requests the following RAIs:

1. Paragraph 55a(z) of 10 CFR 50 states that alternatives to the requirements of 10 CFR 50.55a(g) may be used, when authorized by the NRC, if (1) the proposed alternatives would provide an acceptable level of quality and safety or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

In the licensees June 9, 2015 submittal, the heading for RP5-01 states the Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2) Hardship without a Compensating Increase in Quality and Safety. However, in the body of the request in the Proposed Alternative and Basis for Use section it states, Pursuant to 10 CFR 50.55a(z)(2), relief is requested on the basis that the proposed alternative provides an acceptable level of quality and safety, which is the basis described in 50.55a(z)(1). The precedents cited in the submittal were all granted in accordance with 10 CFR 50.55a(z)(2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Please clarify whether the authorization of RP5-01 is sought in accordance with 50.55a(z)(1) or 50.55a(z)(2).

2. ASME Code,Section XI, Code Case N-795 allows an alternative lower test pressure for certain Class 1 pressure tests following repair/replacement activities at BWR nuclear power plants using a critical reactor core to raise the temperature of the reactor coolant and pressurize the reactor coolant pressure boundary (RCPB). Code Case N-795 was developed because a large majority of BWRs must perform a pressure test that requires the primary system to be isolated (including shutdown cooling) to obtain a test pressure corresponding to 100-percent rated power and allow access for the examination. During this test, the vessel is filled essentially water solid while at a greatly reduced margin to cold overpressure conditions. The licensees have asserted that performance of the primary system pressure test under these conditions places the unit in a position of significantly reduced margin, approaching the fracture toughness limits defined in the Technical Specification Pressure-Temperature [P-T] curves. In addition, reactor pressure corresponding to 100-percent rated power cannot be obtained, at a large majority of BWR units, during normal startup operations at low power levels. This is because the pressure control system does not allow the setpoint to approach the 100-percent pressure value. Also, the core reload 1

analysis does not cover the elevated pressure at low power levels conducive to personnel entry into the drywell.

The NRC has a long-standing prohibition against the production of heat through the use of a critical reactor core to raise the temperature of the reactor coolant and pressurize the RCPB. A letter dated February 2, 1990, from James M. Taylor, Executive Director for Operations, NRC, to Messrs.

Nicholas S. Reynolds and Daniel F. Stenger, Nuclear Utility Backfitting and Reform Group (ADAMS Accession No. ML14273A002), established the NRC position with respect to use of a critical reactor core to raise the temperature of the reactor coolant and pressurize the RCPB. In summary, the NRCs position is that testing under these conditions involves serious impediments to careful and complete inspections, and thus, inherent uncertainty with regard to assuring the integrity of RCPB.

Further, the practice is not consistent with basic defense-in-depth safety principles.

The bases for the NRCs position on the first condition are as follows:

a. Nuclear operation of a plant should not commence before completion of system hydrostatic and leakage testing to verify the basic integrity of the RCPB, a principal defense-in-depth barrier to the accidental release of fission products. The assured integrity of the RCPB is fundamental to the safe operation of nuclear power plants and is, therefore, of critical importance in adequately assuring the protection of public health and safety. In addition, adequacy of these inspections is an important factor in assuring adequate protection of public health and safety. In accordance with the defense-in-depth safety precept, nuclear power plant design provides multiple barriers to the accidental release of fission products from the reactor. The RCPB is one of the principal fission product barriers. For this reason, General Design Criteria-14 requires explicitly that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormal leakage, of rapidly propagating failure, and of gross rupture.

Consistent with this conservative approach to the protection of public health and safety, and the critical importance of the RCPB in preventing accidental release of fission products, the NRC has always maintained the view that verification of the integrity of the RCPB is a necessary prerequisite to any nuclear operation of the reactor. Initiation of criticality for the purpose of hydrotesting or leakage testing to verify RCPB integrity is contrary to this basic safety principle.

b. The NRCs historical view has been that hydrotesting must be done essentially water solid so that stored energy in the reactor coolant is minimized during a hydrotest or leak test.

c. The initiation of criticality creates a severe working environment that encumber required inspections to such an extent as to call into serious question the adequacy and ability of those inspections to properly verify reactor coolant boundary integrity. The elevated reactor coolant temperatures result in a severely uncomfortable and difficult working environment in plant spaces where the system leakage inspections must be conducted.

The greatly increased stored energy in the reactor coolant increases the hazard to personnel and equipment in the event of a leak, and the elevated temperatures contribute to increased concerns for personnel safety due to burn hazards, even if there is no leakage. As a result, the ability for plant workers to perform a comprehensive and careful inspection becomes greatly diminished.

The NRCs position established in 1990 was reaffirmed in Information Notice No. 98-13, Post-Refueling Outage Reactor Pressure Vessel Leakage Testing Before Core Criticality, dated April 20, 1998. The Information Notice was issued in response to a licensee that had conducted an ASME Code,Section XI, leakage test of the reactor pressure vessel and subsequently discovered that it had violated 10 CFR part 50, Appendix G, to complete pressure and leak testing before the core is taken critical. A final rule published in the Federal Register on December 19, 1995, clarified the staff position in 10 CFR part 50, Appendix G,Section IV.A.2.d, as follows: "Pressure tests and leak tests of the reactor vessel that are required by Section XI of the ASME Code must be completed before the core is critical." The Information Notice reiterated the NRCs position that 2

under the ASME Code,Section XI, Class 1 and 2 leakage tests provide a level of defense-in-depth for detecting pressure boundary leakage. From a safety perspective, performing this test using nuclear heat defeats the intended purpose of ensuring the integrity of the reactor pressure vessel as a fission product barrier. Further, it noted that a NRC Region III inspector that reviewed the licensee problem identification forms (Inspection Report 50-254/97-27, ADAMS Accession No.

9803170418) found that licensee personnel performing VT-2 examinations of the Unit 2 drywell covered 50 examination areas in 12 minutes calling into question the adequacy of the VT-2 examinations.

Accordingly, the NRC continues to believe that there is a strong basis for the prohibition on the production of heat through the use of a critical reactor core to raise the temperature of the reactor coolant and pressurize the RCPB.

However, as evidenced by the precedents cited in RP5-01, the NRC has authorized pressure tests at a number of plants based on the belief that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Assuming the authorization of RP5-01 is sought in accordance with 50.55a(z)(2) (i.e. compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety) as in the precedents cited, please describe the method(s) that could be used for attaining 100% of normal operating pressure required by IWB-5221(a) in order to perform the Code-compliant system leakage test and describe the hardship or unusual difficulty associated with each.

3. Please describe the method for attaining 90% of normal operating* pressure in order to perform the proposed alternative leakage test.

4. The Basis for Use section of the proposed alternative states that the core decay heat during a maintenance outage is much higher than that after a refueling outage, and that the heat load is difficult to control once shutdown cooling is removed from service.

a. What are the temperature and pressure limits for use of shutdown cooling?

b. Given these limits, please explain why pressurization to 90 percent normal operating pressure, with a hold for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, is possible but pressurizing to 100 percent normal operating pressure is unusually difficult.

5. In the Applicable Code Requirement section of the June 9, 2015 submittal of RP5-01 the requirements of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(b)(2)(xxvi), Pressure Testing Class 1, 2, and 3 Mechanical Joints are accurately described, which states, The repair and replacement provisions in IWA-4540(c) of the 1998 Edition of Section XI for pressure testing Class 1, 2, and 3 mechanical joints must be applied when using the 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section. The 1998 Edition of ASME Section XI IWA-4540(c) states "Mechanical joints made in installation of pressure retaining items shall be pressure tested in accordance with IWA-5211 (a)." The last sentence of the second paragraph of this section states, NPPD understands that this means a pressure test is required for a mechanical joint when a new valve or flange greater than NPS-1 is installed as part of the repair/replacement activity, and does not include those items covered by IWA-4132 Items Rotated From Stock.

The NRC staff believes that all mechanical joints made in the installation of Class 1, 2 and 3 pressure retaining items shall be pressure tested in accordance with IWA-5211(a) except those specifically exempted by IWA-4540(c) of the 1998 Edition of Section XI. The ASME Code has taken actions which have eliminated the requirements for pressure testing mechanical joints in editions and addenda of Section XI later than the 1998 Edition, such as the elimination of IWA-4540(c) and the elimination of pressure testing from IWA-4132. The NRC staff has reviewed 3

these actions and determined the need to place a requirement in the regulations as shown in 10 CFR 50.55a(b)(2)(xxvi) which requires the pressure testing of mechanical joints. The NRC staffs objective of the requirement in 10 CFR 50.55a(b)(2)(xxvi) is to ensure the leak tight integrity of joints disturbed by the installation of pressure retaining components during Repair/Replacement activities. The staff notes that activities performed under IWA-4132, Items Rotated From Stock are still Repair/Replacement (R/R) activities because they are still covered by IWA-4000 of Section XI they are merely R/R activities, which Section XI has reduced the requirements on. The NRC notes that IWA-4132 Items Rotated From Stock required pressure testing in accordance with IWA-4500 until the 2001 Edition of Section XI.

The NRC staff understands that ASME Section XI does not prohibit leakage at mechanical connections. However,Section XI would require the evaluation of leakage found at a mechanical joint during a pressure test to determine its acceptability and the NRC believes this should be the final step in the repair/replacement process.

If the licensee does not believe that mechanical joints in pressure retaining components made during rotations from stock need to be pressure tested, please describe the justification and how these joints are evaluated for leakage upon return to service.

6. Will the inspection be performed to the Section XI, IWA-2212 VT-2 Examination requirements by a certified VT-2 examiner?

7. The NRC staff understands that RP5-01 will only be used to perform the pressure test and VT-2 examination for repair/replacement activities performed after the system leakage test required by Table IWB-2500-1, Category B-P has been performed, is that correct?

Siva P. Lingam U.S. Nuclear Regulatory Commission Project Manager (NRR/DORL/LPL4-1)

Cooper Nuclear Station Diablo Canyon Nuclear Power Plant Location: O8-D5; Mail Stop: O8-B3 Telephone: 301-415-1564; Fax: 301-415-1222 E-mail address: siva.lingam@nrc.gov 4

Hearing Identifier: NRR_PMDA Email Number: 2441 Mail Envelope Properties (Siva.Lingam@nrc.gov20151008162300)

Subject:

Cooper - Relief Request (RR) RP5-01 (CAC No. MF6335)

Sent Date: 10/8/2015 4:23:07 PM Received Date: 10/8/2015 4:23:00 PM From: Lingam, Siva Created By: Siva.Lingam@nrc.gov Recipients:

"Markley, Michael" <Michael.Markley@nrc.gov>

Tracking Status: None "Alley, David" <David.Alley@nrc.gov>

Tracking Status: None "Hoffman, Keith" <Keith.Hoffman@nrc.gov>

Tracking Status: None

"'dwvande@nppd.com'" <dwvande@nppd.com>

Tracking Status: None

"'jdshaw@nppd.com'" <jdshaw@nppd.com>

Tracking Status: None Post Office:

Files Size Date & Time MESSAGE 14859 10/8/2015 4:23:00 PM Options Priority: Standard Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:

Recipients Received: