ML103190653

Draft Requests for Additional Information, (Srxb) EPU LAR Review

ML103190653
Person / Time
Site:	Point Beach
Issue date:	11/08/2010
From:	Beltz T Plant Licensing Branch III
To:	Hale S Nextera Energy
kmz1
Shared Package
ML103190659	List: ML103190653
References
TAC ME1044, TAC ME1045
Download: ML103190653 (4)

Text

REQUEST FOR ADDITIONAL INFORMATION POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 EXTENDED POWER UPRATE (LAR 261)

POSTULATED STEAM GENERATOR TUBE RUPTURE

Background

The postulated steam generator tube rupture (SGTR) event, as analyzed for the Point Beach Nuclear Plant (PBNP) extended power uprate (EPU) request, is based on a conservative evaluation of the total reactor coolant mass released, through the steam generator, to the environment. The analyses supporting this evaluation rely on, among others, an assumption that operators can terminate the leakage of reactor coolant into the steam generator shell side, and ultimately, into the environment.

The validity of this assumption relies, in part, on a supplemental analysis that demonstrates that there is adequate margin to steam generator overfill. Although not a part of the PBNP licensing basis, this analysis attempts to validate the mass release evaluation by demonstrating, using a separate set of initial conditions, that following an SGTR, primary to secondary break flow can be terminated and the steam generator can be isolated before it fills with liquid water.

This supplemental analysis, however, is based on assumptions that are non-bounding of permissible plant operation, do not consider uncertainties, and do not include a limiting single active component failure in the mitigating safety system. The analyses also demonstrate a very small available margin to steam generator overfill.

Generic studies documented in WCAP-10698-P-A, SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill, indicate that final liquid volume in the steam generator shell side will be greater, by about 12-13 percent when employing conservative input values.

The NRC staff accepted the technical basis underlying WCAP-10698-P-A, since it described an acceptable methodology for performing a conservative margin-to-overfill analysis for a postulated SGTR.

Adding 12-13 percent to the Point Beach margin-to-overfill analysis results would predict overfill of the ruptured steam generator.

As documented in WCAP-11002-P1, the postulated consequences of a steam generator overfill could include water relief through a safety valve, which then fails due to liquid flow. The valve could either fail to reseat, or fail fully open. In either case, the ruptured steam generator would be unisolable and would continue to discharge effluent from the ruptured steam generator for a period of time that far exceeds the release time assumed in the licensing basis mass release analyses for PBNP EPU. The continued discharge would also impede efforts to depressurize the reactor coolant system to a pressure below the steam generator shell side pressure Consideration of input parameter variability, uncertainties, and limiting single failures, demonstrates that the postulated SGTR event at Point Beach may result in a steam generator 1

Note that the staff discussed WCAP-11002-P in its evaluation of WCAP-10698-P-A, but did not find that it provided an acceptable method for performing a licensing basis safety analysis.

overfill. This information challenges the assumption, employed in the mass release analyses, that the ruptured steam generator is isolable.

Applicable Regulatory Guidance Title 10 of the Code of Federal Regulations, part 50, Section 36 (10 CFR 50.36) promulgates requirements for facility technical specifications.

10 CFR 50.36 contains requirements for limiting safety system settings (LSSS), which are settings for automatic protective devices related to those variables having significant safety functions. Where an LSSS is specified for a variable on which a safety limit has been placed, the setting must be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. Certain LSSS, including those for steam generator water level, provide a basis for and/or establish an allowable range of initial conditions from which postulated accidents are assumed to initiate, and at which automatic equipment actuations are assumed to occur.

10 CFR 50.36 also requires the establishment of limiting conditions for operation (LCOs) of a nuclear reactor, which include, among other things, process variables, design features, and operating restrictions that are initial conditions of design basis accident or transient analyses that either assume the failure of or present a challenge to the integrity of a fission product barrier.

Since the radiological consequences of a postulated SGTR at Point Beach are evaluated using an alternative source term, Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, is applicable. Regulatory Position 5.1.3 states that the numeric values that are chosen as inputs to the analyses should be selected with the objective of determining a conservative postulated dose.

Regulatory Position 5.1.4 states that the NRC staff considers the implementation of an AST to be a significant change to the design basis of a facility that is voluntarily initiated by the licensee.

In order to issue a license amendment authorizing the use of an AST and the TEDE dose criteria, the NRC staff must make a current finding of compliance with regulations applicable to the amendment. The characteristics of the ASTs and the revised dose calculational methodology may be incompatible with many of the analysis assumptions and methods currently reflected in the facilitys design basis analysis. The NRC staff may find that new or unreviewed issues are created by a particular site-specific implementation of the AST, warranting review of staff positions approved subsequent to the initial issuance of the license.

This not considered a backfit as defined by 10 CFR 50.109, Backfitting.

Based on Regulatory Position 5.1.4, the staff questions the licensing basis assumption that the SGTR results in only a steam release to the atmosphere, despite that margin to steam generator overfill is not included in the Point Beach licensing basis. By referencing the extended power uprate margin to overfill analysis in response to an NRC RAI associated with the AST application, the licensee introduced the previously unreviewed issue concerning steam generator margin to overfill. The licensee attempted to demonstrate that Point Beach has margin to overfill, but the staff did not accept this analysis due to the fact that it relied on unacceptably non-conservative inputs and assumptions. The approved staff position regarding an acceptable margin to overfill analytic methodology is communicated in the SER approving WCAP-10698-P-A and in SERs approving plant-specific implementation of the WCAP-10698-P-A in toto.

Issue Based on the NRC staffs independent assessment, consideration of the full range of permissible plant operation, consistent with the above regulatory and licensing basis requirements, and consideration of parametric uncertainties, for a postulated SGTR would result in a predicted overfill of the steam generator.

In order for the NRC staff to accept the proposed mass release analysis, the licensee must provide an acceptably conservative evaluation of margin to steam generator overfill to validate the assumption that the steam generator will not fill with water prior to the termination of an atmospheric release.

During a November 2, 2010, public meeting between NextEra Energy and the NRC staff, a draft version of this request for additional information was discussed. The licensee provided the following information for the staffs consideration:

• The licensing basis for Point Beach includes a steam generator tube rupture concurrent with a loss of offsite power (LOOP), an event expected to occur with a frequency of approximately 3E-6 per year.

• Assuming a concurrent, limiting single equipment failure results in an event of very low frequency, 3E-9 per year.

Also during the November 2, 2010, meeting, the licensee provided results of an analysis to demonstrate that even with the occurrence of a postulated SGTR event concurrent with a LOOP and a limiting single failure, the resulting radiological consequences would remain within NRC regulatory limits.

The licensee stated that the liquid release radiological consequences analysis was an added assurance analysis, and it was not intended for incorporation into the licensing basis. The licensees reasoning included the fact that the steam generator tube rupture concurrent with a limiting single failure and a LOOP is beyond the design basis of Point Beach, and that such an event has a very low probability.

While the NRC staff agreed that the licensees technical approach to answering the staffs RAI was reasonable, the NRC staff communicated that the licensee would need to provide additional justification for not incorporating the limiting results into the Point Beach licensing basis and that the staff would continue to consider the acceptability of this approach.

The NRC staffs position remains that current regulatory guidance directs the staff to ensure that analytic assumptions contained in the radiological consequences analysis are appropriately conservative, and the information provided by the licensee to date does not provide adequate validation of the assumption that environmental releases during the postulated SGTR event would include only steam.

Request Provide a thermal-hydraulic analysis for Point Beach, at both current licensed thermal power conditions and at the proposed, uprated conditions, for a limiting margin-to-overfill/overfill scenario. Once acceptable methodology would be for the analysis to align as closely as possible to what is approved in WCAP-10698-P-A; however, since the licensee has asserted that a limiting single failure is not in the Point Beach licensing basis, this exception to the WCAP-10698-P-A methodology would be acceptable.

In addition to providing the analytic results, please address the following:

1. Ensure that the limiting liquid release pathway and scenario are identified. Include consideration of the steam line equipment water-release failures discussed in WCAP-11002. If a liquid release is predicted, provide analyses of the static and dynamic structural effects in the main steam system and of the consequences of passing water through the steam pressure relief valves.

2. Under the assumed LOOP conditions, address the functionality of each atmospheric discharge valve (ADV). Discuss what, if any, mitigating function the ADV provides and its capability to perform that function under the assumed LOOP conditions. If valve actuation is manually performed, provide information to demonstrate that the operator is capable of causing the valve actuation within the analytically assumed time.

3. For the CLTP case, provide recent trending data for full-power steam generator water level to demonstrate that a conservative initial steam generator water level has been selected.

4. Identify operator actions credited in the analysis. Where an operator action is credited, confirm that such action is consistent with station procedures.

5. Update the information contained in response to RAI SRXB-5 in the September 28, 2010, supplemental letter to reflect assumptions used in the analysis performed in response to the above request.

6. Should the radiological consequences from the analyses requested in this RAI be more severe than the currently proposed radiological analyses, then update the licensing basis radiological consequence analyses for both the AST and EPU license amendment requests to reflect these results. Since the staff is allowing the single failure exception to the WCAP-10698-P-A methodology, the above requested analysis represents an event that has a significantly higher likelihood of occurrence.

Please also provide the following additional information, which was discussed during the November 2, 2010, meeting:

1. Identify procedures addressing steam generator overfill conditions. What parameters do operators monitor to help ensure that overfill does not occur?

2. For any revised radiological consequence analyses, provide the basis for the assumed flashing fraction if it is less than 100%.