ML071630250

Response to Request for Additional Information - Request for Authorization to Extend Third 10-Year ISI Interval for Reactor Vessel Weld Examination

ML071630250
Person / Time
Site:	Palisades
Issue date:	05/31/2007
From:	Schwarz C Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MD3059
Download: ML071630250 (4)

Text

Entergy Nucjlear Operations, Inc.

EntergyPalisades Nipclear Plant 27780 Blue ,tar Memorial Highway Covert, MI 4?043 May 31, 2007 10 CFR 50, Appendix A U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Palisades Nuclear Plant Docket 50-255 License No. DPR-20 Response to Request for Additional Information - Request for Authorization to Extend Third 10-Year ISI Interval for Reactor Vessel Weld Examination (TAC No. MD3059)

Dear Sir or Madam:

By letter dated September 15, 2006, NMC (the former licensee for Palisades Nuclear Plant (PNP)) requested Nuclear Regulatory Commission (NRC) approval for the use of an alternative to the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, paragraph IWB-2412, Inspection Program B, for the Palisades Nuclear Plant. PNP submitted this relief request because the Westinghouse Owners Group Topical Report, WCAP-16168-NP, "Risk-Informed Extension of Reactor Vessel Inservice Inspection Interval," dated October 2003, is currently being reviewed by the NRC and not yet approved.

By electronic email dated March 8, 2007, the NRC sent a request for additional information (RAI). On April 26, 2007, a teleconference was held with the NRC to discuss the RAI. Enclosure 1 provides the response to the RAI for PNP.

Summary of Commitments This letter contains no new commitments and no revision to existing commitments.

hristoph ýe 7 ?.

Site Vice President Palisades Nuclear Plant CC Administrator, Region Ill, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC A4q 7

ENCLOSURE 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION PALISADES NUCLEAR PLANT NRC Request As discussed in a letter to Westinghouse dated January27, 2005, the staff expects that one-time requests to extend the inspection interval of tie reactor pressure vessel (RPV) welds by one cycle should include a discussionindicating that the likelihood of a significantpressurized thermal shock (PTS) e6 vent over the next operatingcycle is very low.

Your submittals dated March 31 and October 11, 2005, describedPalisades' response to three of the most significant PTS sequences identified in the ongoing PTS rulemaking work. To support the conclusion that the requestfor relief for this second one-cycle extension satisfies the risk-informed principalthat any proposed increase in risk is small, please provide an estimate of the annual frequency of these more severe PTS sequences and describe the process used to evaluate the frequency of these events which could challenge the integrity of the RPV, if a flaw was present.

ENO Response Palisades Nuclear Plant (PNP) was one of three pilot plants evaluated in the recent NRC effort to re-evaluate the risk of pressurized thermal shock. These efforts are summarized in NUREG-1806, "Technical Basis for Revision of the Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61): Summary Report." As part of the NRC effort, probabilistic risk assessment (PRA) models were developed for each of the pilot plants using plant specific information. The PNP PRA model is discussed in an NRC letter report, "Palisades Pressurized Thermal Shock (PTS) Probabilistic Risk Assessment (PRA)," dated October 6, 2004 (ADAMS Accession number ML042880473).

The analyses documented in this report were developed by the PNP PRA staff, and therefore, accurately models the PNP. The PRA model included detailed event tree and fault tree analyses that defined both the sequences of events that are likely to produce a PTS challenge to RPV structural integrity, and the frequency with which such events can be expected to occur. Due to the large number of sequences identified, it was necessary to group/bin sequences with like characteristics into representatives that could later be analyzed using thermal-hydraulic codes. This resulted in 65 binned sequences for PNP.

Thermal-hydraulic analyses were performed for each of these bins (i.e.,

representative transients) by Information Systems Laboratories, Inc. (ISL) to develop time histories of temperature, pressure, and reactor vessel wall heat transfer boundary conditions. The PNP staff assisted ISL in developing the appropriate RELAP boundary conditions, as well as providing a detailed design review of the developed model used in creating the transient histories.

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These histories were then input into the probabilistic fracture mechanics (PFM) analysis to determine conditional probability of reactor vessel failure for each transient.

From this analysis, it was determined that only a portion of the transients contribute to the total risk of RPV failure, while the remainder have an insignificant or zero contribution. The transients that were identified, to be contributors to PTS risk were then used for the PFM analysis in the PTS study and for the pilot plant studies in this report. Therefore, thirty transients were analyzed for PNP. After detailed PFM analyses, only eleven transients were identified to have a contribution to the frequency of reactor vessel failure greater than one percent of the total risk. The results of the PFM analyses are discussed in ORNL/NRC/LTR-04/18, "Electronic Archival of the Results of Pressurized Thermal Shock Analyses for Beaver Valley, Oconee, and Palisades Reactor Pressure Vessels Generated with the 04.1 version of FAVOR."

Information from this report for these eleven sequences/transients (identified by "TH Case #" from ORNL/NRC/LTR-04/18) is provided in Table 1. The column at far right identifies to which sequence category, from the October 11, 2005, submittal, the transient applies. The following sequences were previously documented in a request for additional information (RAI) dated August 23, 2005. The August 23, 2005, RAI was sent in regards to the March 31, 2005, submittal mentioned above.

These sequence categories are defined as follows:

Sequence 1 Any transient with reactor trip followed by one stuck-open pressurizer safety relief valve that re-closes after about one hour. Severe PTS events also require the failure to properly control high-head injection.

Sequence 2 Large loss of secondary steam from steam line break or stuck-open atmospheric dump valves. Severe PTS events also require the failure to properly control auxiliary feedwater flow rate and destination (e.g., away from affected steam generators), and failure to properly control high-pressure injection.

Sequence 3 Four- to nine-inch loss-of-coolant accidents. Severity of PTS event depends on break location (worst location appears to be in the pressurizer line) and primary injection systems flowrate and water temperature.

Table 1 provides the eleven transients that were identified to have a contribution to the frequency of reactor vessel failure greater than one percent of the total risk.

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Table 1: PTS Sequence/Translent Frequencles Sequence! System Failure Operator Action Sequence! Sequence Transient Transient Category TH Case # Frequency (Events/yr) 19 Reactor trip with 1 stuck-open None. Operator 2.29E-03 2 ADV on SG-A. does not throttle HPI.

40 40.64 cm (16 in) hot leg break. None. Operator 3.22E-05 3 Containment sump recirculation does not throttle included In the analysis. HPI.

48 Two stuck-open pressurizer SRVs None. Operator 7.67E-07 1 that reclose at 6000 sec after does not throttle initiation. Containment spray is HPI.

assumed not to actuate.

54 Main steam line break with failure Operator does not 4.26E-06 2 of both MSIVs to close. Break isolate assumed to be inside containment AFW on affected causing containment spray SG. Operator actuation. does not throttle HPI.

55 Turbine/reactor trip with 2 stuck- Operator starts 2.74E-03 2 open ADVs on SG-A combined second with controller failure resulting in AFW pump.

the flow from two AFW pumps into affected steam generator.

58 10.16 cm (4 In) cold leg break. None. Operator 2.66E-04 3 Winter conditions assumed (HPI does not throttle and LPI injection temp = 40 F, HPI.

Accumulator temp = 60 F) 60 5.08 cm (2 in) surge line break. None. Operator 2.09E-04 3 Winter conditions assumed (HPI does not throttle and LPI injection temp = 40 F, HPI.

Accumulator temp = 60 F) 62 20.32 cm (8 in) cold leg break. None. Operator 7.07E-06 3 Winter conditions assumed (HPI does not throttle and LPI injection temp = 40 F, HPI.

Accumulator temp = 60 F) 63 14.37 cm (5.656 In) cold leg None. Operator 6.06E-06 3 break. Winter conditions does not throttle assumed (HPI and LPI Injection HPI.

temp = 40 F, Accumulator temp =

60 F) 64 10.16 cm (4 in) surge line break. None. Operator 7.07E-06 3 Summer conditions assumed (HPI does not throttle and LPI injection temp = 100 F, HPI.

Accumulator temp = 90 F) 65 One stuck-open pressurizer SRV None. Operator 1.24E-04 1 that recloses at 6000 sec after does not throttle initiation. Containment spray is HPI.

assumed not to actuate.

Notes:

TH ### - Thermal hydraulics run number ### from NRC PTS Risk Re-evaluation IE - Initiating event ADV - Atmospheric dump valve SRV - Safety and relief valve AFW - Auxdliary feedwater HPI - High-pressure Injection LPI - Low-pressure Injection RCP - Reactor coolant pump SG - Steam generator The sequence/transient frequencies presented in Table 1 show that even if a flaw were present in the PNP reactor vessel beltline, the likelihood of having a PTS sequence/transient that could challenge the integrity of the reactor vessel is acceptably small.

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