ET 08-0007, Followup Response to NRC Requests for Additional Information Related to License Renewal Application Time-Limited Aging Analysis
| ML080350012 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 01/25/2008 |
| From: | Garrett T Wolf Creek |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| ET 08-0007 | |
| Download: ML080350012 (14) | |
Text
- CREEK OPERATING CORPORATION Terry J. Garrett Vice President, Engineering January 25, 2008 ET 08-0007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555
Reference:
- 4) Telephone Conference Summary dated September 4, 2007, from V. Rodriguez, USNRC (ML072320487)
Subject:
Docket No. 50-482: Followup Response to NRC Requests for Additional Information Related to Wolf Creek Generating Station License Renewal Application Time-Limited Aging Analysis Gentlemen:
Reference 1 provided Wolf Creek Nuclear Operating Corporation's (WCNOC) License Renewal Application for the Wolf Creek Generating Station (WCGS).
References 2 and 3 provided WCNOC responses to NRC requests for additional information '(RAI) regarding the License Renewal Application Time-Limited Aging Analysis.
Reference 4 documents-telephone.
conference calls held on August 17, 2007 and-August 31, 2007 to discuss and clarify WCNOC responses and requested additional information. ý In response to the RAI..of Reference 4, WCNOC submitted Reference 5 and committed to complete actions by-January 31, 2008. This, commitment was numbered thirty-eight.
The Enclosure provides a summary of the actions taken to close commitment number thirty-,
eight.
P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET jj/Z*
ET 08-0007 Page 2 of 3 This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4084, or Mr. Richard Flannigan at (620) 364-4117.
Sincerely Terry J. Garrett TJG/rlt Enclosure WCNOC Followup Response to NRC Requests for Additional Information cc:
E. E. Collins (NRC), w/e V. G. Gaddy (NRC), w/e B. K. Singal (NRC),w/e T. Tran (NRC), wle Senior Resident Inspector (NRC), w/e
ET 08-0007 Page 3 of 3 STATE OF KANSAS COUNTY OF COFFEY
)
)
Terry J. Garrett, of lawful age, being first duly sworn upon oath says that he is Vice President' Engineering of Wolf Creek Nuclear Operating Corporation; that he. has, read the foregoing document and knows the contents thereof; that he has executed-the same for, and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.
By________________
Terry6JK'Garrett Vice President Engineering SUBSCRIBED and sworn to before me this,S*day of f,--1
)I L(j
,2008.
C GAYLE SHEPHEARD Notary Public - State of Kansas, My Appt. Expires 'r1/
/. 0/ l i Notary Ptlblic Expiration Date
'/c / /c9,O (1
Enclosure to ET 08-0007 Page 1 of 11 Followup Response to NRC Requests for Additional Information Related to Wolf Creek Generating Station License Renewal Application Time-Limited Aging Analysis
Enclosure ET 08-0007 Page 2 of 11 License Renewal Commitment #38 Backward. projection of CUF was used for NUREG/CR-6260 locations (Surge Line Hot Leg Nozzle, Charging Nozzles), and for several locations not, covered by NUREG /CR -6260 locations (Pressurizer Lower Head, Pressurizer Spray Nozzle, Pressurizer Surge Nozzle, Pressurizer Surge Line, S/G Feedwater Nozzles). While the ratios used for back-projection do incorporate accumulated fatigue effects from-all transients that occurred during PERIOD2, it does not account for transients that occurred more frequently in PERIOD1 than during PERIOD2. Therefore, Wolf Creek will prepare an updated baseline that adequately bounds transients experienced prior to the start of CUF monitoring. The existing baseline CUF for all monitored locations will be increased to bound the potential CUF contribution from the transients that were under-represented in the existing baseline.
1.0 Background
Fatigue Management Program (FMP):
The Wolf Creek Generating Station (WCGS) FMP is an aging management program to monitor and track the metal fatigue effects of critical temperature and pressure transients experienced by the RCS pressure boundary. For a selected number of critical pressure boundary locations, Reference 1 Table 4.3-2, the FMP includes calculating estimates of the cumulative fatigue usage (CUF) caused by the transients that have occurred. These estimates are calculated in one of two ways, 1) cycle based fatigue monitoring (CBF) or, 2) stress based fatigue monitoring (SBF). The locations for which CUF is calculated by the FMP include six of seven (6 of 7) locations where the environmental effect of the reactor coolant on fatigue is evaluated in accordance with NUREG/CR-6260 [(NUREG/CR-6260 locations) (RPV inlet nozzles, RPV outlet nozzles, safety injection (BIT) nozzles, accumulator safety injection-RHR nozzles, hot leg surgeline nozzle, and charging nozzles)].
For the NUREG/CR-6260 locations, the FMP calculates CUF using appropriate environmental factors (Fen).
Stress Based Fatigue Monitoring:
The stress based fatigue (SBF) module computes CUF for several predetermined locations based on the actual plant operating history. Typically, SBF usage calculations produce a lower CUF than cycle based fatigue monitoring (CBF) usage calculations, because actual transients are less severe (i.e., have smaller peak to peak pressure and temperature changes and occur more slowly) than assumed by the design specification transient definitions.
Currently, SBF is not part of the plant licensing basis. SBF usage data are acquired for a small number of critical locations for potential future use should a corrective action limit on accrued cycles or CBF calculated CUF are reached.ý.: SBF data for. two locations will be used if necessary to support the NUREG/CR-6260 evaluation at WCGS: (1) the RCS hot leg surge line nozzle, and (2) the charging and alternate charging nozzles of the reactor coolant system.
Implementation of the SBF module requires a conservative baseline estimate of accrued CUF prior to beginning of monitoring and a conservative methodology for calculating usage from the time histories of temperature, pressure, and fluid flow rates.
Enclosure ET 08-0007 Page 3 of 11 Baseline CUF Estimates:
The baseline CUF for SBF monitored locations were calculated. using data accrued during almost ten years of operation of the data acquisition system and assumptions regarding the severity of transients during the period before monitoring. WCGS divided past operation into two periods, the, time before monitoring was begun is considered PERIOD1, and the time after.
PERIOD2 (i.e.,1/13/96-12/31/05).
CUF for PERIOD2 was calculated using the FatiguePro software, but CUF for PERIOD1 was determined by back projection. The Nuclear Regulatory Commission (NRC or Staff) in support of Wolf Creek Nuclear Operating Corporation (WCNOC)
License Renewal Application (LRA) (Reference 1) review requested additional information (RAI) on the use of monitoring data collected for. PERIOD2-and the values to derive backward projected initial CUFs for PERIOD1 (Reference 5). However, because a number of transients occurred more frequently during the period before monitoring than during the monitored period, the assumptions used in the baseline calculations could not be proved to be conservative.
Therefore, the baseline calculations have been revised to assure the CUF starting points for monitoring are above the actual accrued CUF for the components at the time monitoring was implemented.
2.0 Objective The purpose was to determine an additional CUF increment for each monitored FatiguePro location that demonstrably bounds the maximum usage accumulation from the "unaccounted" transients in PERIOD1. This was performed in two steps. First, it specifically identified the set of PERIOD1 plant transients that may not have been adequately addressed by the extrapolation approach. Second, it determined a CUF increment for each stress-based fatigue (SBF) location that bounds the collective fatigue impact of the transients identified in step one.
3.0 Methodology Identifying Unaccounted Transients:
In order to determine the unaccounted CUF, we first.established which transients occurred in the past that was not well represented in the template period (i.e., PERIOD2).
This was determined by examining the relative number of occurrences during PERIOD1 and PERIOD2, and by considering how the back-projection was completed. : Additionally, the transient counts were modified based on past events that were counted improperly.
In the baseline calculation (Reference 3), the back-projected CUF was determined by applying a multiplier X to the CUF calculated for the template period. X is a location-specific scaling factor based on ratios of the counted events that contribute the most CUF accumulation. At WCGS, ratios of 1.625 (the relative number of LoC/LoL events, for the Charging Nozzle locations) and 2.25 (the relative number of Heatup and Cooldown events, for all other locations) were.computed.
This treats past operation as if it was equivalent to X repetitions of the template period. This is reasonable as long as (the number of past transients) is less than or equal to (X times the number of transients during the template period).
The number of transient cycles both before and after the start of the template period (i.e.,
1/13/96) is compared in Table 1. Transients where the past occurrences are proportional, the template projection are indicated in italic.
Transients with more than the expected past occurrences are indicated in bold.
I I Enclosure ET 08-0007 Page 4 of 11 Table 1: Comparison of Events PERIODI to PERIOD2 Event Type Initial Increment Period 1 Period 2.(template)
Accumulator SI Actuation 1
0 Aux Spray dT>320'F 0
0 COMS Operation 0
0 Complete Loss of Flow 0
0 Excessive FW Flow 0
0 HPSI Actuation 6
0 Inadvertent RCS Depressurization 0
0 Inadvertent Startup Inactive Loop 0
0 LPSI Actuation 1
0 Large Loss of Coolant Accident 0
0 Large Steam Line Break 0
0 Loop Out-Of-Service 0
0 Loss RC Flow I Loop @ Power 2
0 Loss of Charging (Loop 1) 19 15 Loss of Charging (Loop 4) 0 8
Loss of Letdown Flow 11 7
Loss of Load w/o Rx Trip 2
0 Operating-Basis Earthquake 0
0 PZR Cooldown 17 8
PZR Heatup 19 8
Enclosure ET 08-0007 Page 5 of 11 Table 1 (cont.)
Post-LOCA Operation 0
0 RCS Cooldown 17 8
RCS Heatup 19 8
RCS Hydrostatic Test 1
0 RCS Leak Test (Normal) 6, 0
Reactor Trip (CD and SI) 0, 0
Reactor Trip (CD no SI) 0 0
Reactor Trip (No Cooldown) 46 9
Reduced-Temp Return to Power 0
0 Refueling
- 7.
7 S/G-A Secondary Hydro 4
0 SIG-B Secondary Hydro 4
0 SIG-C Secondary Hydro 4
0 S/G-D Secondary Hydro 4
0 Small Loss of Coolant Accident 0
0 Small Steam Line Break 0
0 Steam Line/FW Line Break 0
0 Turbine Roll Test 9
0 Most of the counted transients were proportional to the back-projection scaling factors (1.625 or 2.25). However, 11 events exceed the expected number of past occurrences; see Table 2. Ten of those transients occurred only during the past (i.e., they did not occur at all during the-template period).
Only the Reactor Trip event appears during both periods (but at a much higher frequency before monitoring began).
Enclosure ET 08-0007 Page 6 of 11 Table 2: Comparison of Counted Events PERIODI to PERIOD2 Event Type PERIODI PERIOD2 (template)
Accumulator SI Actuation 1
0 HPSI Actuation 6
0 Table 2 (cont.)
LPSI Actuation 1
0 Loss RC Flow 1 Loop @ Power 2
0 Loss of Load w/o Rx Trip 2
0 RCS Hydrostatic Test 1
0 RCS Leak Test (Normal) 6 0
Reactor Trip (No Cooldown) 46 9
Secondary Hydrotest (each S/G) 4 0
Turbine Roll Test 9
0 Improperly Counted Events:
Due to unexpectedly large numbers of several unusual upset events (6 x HPSI Actuation, 7 x Loss of Off-site Power), a review of the transient records was performed to establish if all of the events occurred as counted. The evidence supported most of the events.
However two events - Loss of Load and Loss of Off-site Power - did not occur as indicated in the Westinghouse Fatigue History report (Reference 6). Plant operator logs were examined for each of the days where these events were determined to have occurred. In -all cases the counting was erroneous; something did occur, but it was not consistent with the plant transients that were counted. The evidence is presented below. Any event occurrences counted in error was excluded from the CUF correction.
Loss of Load (w/o Reactor Trip) Events The Westinghouse Systems Standard defines this event as follows (Reference 7):
This transient involves a step decrease in turbine load from full power (turbine trip) without immediate automatic reactor trip. These conditions produce the most severe pressure transient on the Reactor Coolant system under upset conditions. The reactor eventually trips as a consequence of a high pressurizer level trip initiated by the Reactor Protection System. Since redundant means for tripping the reactor are provided by the Reactor Protection System, a transient of this nature is not expected, but is included to ensure conservative design.
P.'
Enclosure ET 08-0007 Page 7 of 11 The cycle history indicates that two of these events occurred (page 82 of Reference 6):
Loss of Load w/o Rx Trip:
1: 8/4/86 Opened Generator Output Breaker (Gen, Turb, RxTrip)
(page 86, also Rx Trip counted 8/9/86 "Opened Gen. Output Breaker")
2: 9/10/87 Line Conductor Failure (Gen, Turb, Rx Trip)
(page 86, also Rx Trip counted 9/10/87 "Transmission Line Conductor Failure")
While the operator logs confirm the turbine trips on both days, they also indicate an immediate reactor trip. Without any delay between trips, there was a minimal pressure spike, and so no additional usage beyond that associated with a normal trip would be accrued (Reference 9).
These events should have been (and were) correctly counted as Reactor Trips, but neither of them should be counted as a Loss of Load event.
Note: the Reactor Trip appears to be counted on 8/9/86 rather than 8/4/86 due to a "typographical" error in the plant records. Review of operator logs show that a Rx Trip did occur on 8/4, and that the plant was steady-state, Mode 1 at 100% power on 8/9.
Loss of Offsite Power Events The Westinghouse Systems Standard defines this event as follows (Reference 7):
This transient applies to a blackout situation involving the loss of outside electrical power to the station, assumed to be operating initially at 100 percent power, followed by reactor and turbine trips.
The reactor coolant pumps are deenergized, as are all electrical loads connected to the turbine -generator bus, including the main feedwater and condensate pumps. As the reactor coolant pumps coast -down, RCS flow.reaches an equilibrium value under natural circulation.
This condition permits removal of core residual heat through the steam generators, which by this time are receiving feedwater, assumed to be at 320F, from the Auxiliary Feedwater System. For equipment design purposes it is conservatively assumed that all auxiliary feedwater pumps operate within one minute following the blackout. Later in the transient the auxiliary feedwater pumps are operated under manual control to obtain stable plant conditions. Steam is removed for the reactor cooldown through power operated relief valves provided for this purpose.
The cycle history records seven of these events (page 83 of Reference 6):.
Loss of Offsite Power:
1: 12/2/85 Potential Transformer Failure 2:2/26/86 Tripped Breaker 3: 6/26/86 Tripped Breaker 4: 9/2/86 Tripped Breaker 5: 7/20/87 Potential Transformer Failure (also counted as Rx Trip 100%)
6: 6/13/90 Potential Transformer Failure 7:10/23/90 Breaker Failure However, the operator logs show that none of these events involved a complete loss of power.
In all cases, the plant continued to operate normally using the remaining external power sources and/or the emergency diesel generators. In one case (7/20/87), a Rx Trip followed the "loss of power", but in all other cases the reactor remained at 100% power following the event.
In no case did any of the serious consequences of the design event (RCP's deenergized, auxiliary feedwater initiated, electrical equipment offline) occur. The plant records correctly count the Reactor Trip, which occurred on 7/20/87-but none of these events should have been counted as a Loss of Offsite Power.
Enclosure ET 08-0007 Page 8 of 11 Computing Bounding CUF Values:
After establishing which transients occurred in the past that were not well represented in the template period and modifying transient counts, a bounding CUF increment was computed for each SBF location in two parts. The sum of the two parts effectively bounds the fatigue impact of all of the unaccounted transients. One of the transients identified in step one is the Reactor Trip transient. This transient occurred 9 times during the template period. While this is too few cycles to effectively represent the cycle frequency during PERIOD1, it is enough to statistically determine a CUF increment that bounds.the average usage accumulation per Reactor Trip.
Part one of the CUF increment is the usage per Reactor Trip event times the number of unaccounted Trips.
To account for the remaining "unaccounted" transients, the calculation performed a fatigue usage calculation at each monitored location, consistent with ASME Section III NB-3200
.methods, but limited in scope to the transients identified in the first step. The resulting CUF values are part two of the CUF increment.
For simplicity (and to provide additional margin), the fatigue analyses determined a single stress intensity (SI) range that bounds all possible stress pairs, which can be constructed using at least one stress taken from the unaccounted transients.
The corresponding number of allowable cycles (Naiow) was computed using the appropriate ASME fatigue curve. With this construction, a CUF increment of N/ Naiow (where N = the total number of unaccounted transient cycles) demonstrably bounds the fatigue impact of those transients.
The bounding SI range for each monitored location was determined in one of two ways. For three. locations, Pressurizer Spray Nozzle [SPR_Noz}, Reactor Coolant System Surge Nozzle
[HLNOZZLE},
and Charging Nozzles [CHRGNOZ and ALTCHRGNOZ], load pair information is available from the corresponding design fatigue analyses. That information was used to select the most significant stress pair that involves one of the unaccounted transients.
For the remaining locations, a bounding SI range was determined using FatiguePro. Plant data simulations were prepared for each of the unaccounted transients. The simulations were run in FatiguePro to compute the stress peaks and valleys that would occur during those transients.
Two SI ranges were developed, by pairing the maximum peak stress from the simulated history with the global minimum stress from the PERIOD2 analysis, and the minimum stress valley from the simulation with the global maximum stress from PERIOD2. The incremental CUF (i.e.
l/Naiow) from both pairs were added, and the sum multiplied by the total number of unaccounted transient cycles, to produce a bounding CUF increment.
4.0 CALCULATIONS Transients with No CUF Impact:
The back-projection of CUF was only performed for the SBF locations monitored in the Wolf Creek FatiguePro system. Those locations consist of:
" Normal & Alternate Charging Nozzles,
" Pressurizer Spray Nozzle,
" Pressurizer Lower Head & Heater Penetration,
Enclosure ET 08-0007 Page 9 of 11
" Pressurizer Surge Nozzle,
" Surge Line Piping,
" RCS (Hot Leg) Surge Nozzle,
- All 4 S/G Feedwater Nozzles.
Two of the plant transients identified in Table 2 have insignificant effects on all of the monitored locations: Accumulator SI Actuation and LPSI Actuation. Therefore, those events need not be considered in the collection of "unaccounted" events.
Summary of Unaccounted Events:
Based on the above evidence, the complete set of "unaccounted" events is given in Table 3.
The first column ("PERIOD1") shows how many cycles were recorded prior to 1/13/1996.The second column ("template") shows how many cycles were recorded during the FatiguePro template period (i.e, PERIOD 2). The third column ("projected") shows how many cycles were accounted for by extrapolating the template period (i.e. 2.25 x the template cycles). The final column ("Unaccounted") is the difference, which tells us how many cycles are not directly accounted for by the back-projection.
Table 3: Unaccounted Events from the WCGS Baseline Event Type PERIODI Template Projected Unaccounted HPSI Actuation 6
0 0
6 LossRCFlowl 2
0 0
2 Loop @ Power RCS 1
0 0
1 Hydrostatic Test RCS Leak Test 6 0
0 6
(Normal)
Reactor Trip 46 9
20 26 (No Cooldown)
S/G-#
4 ea.
0 0
4 ea.
Secondary Hydro Turbine Roll 9
0 0
9 Test 5.0 RESULTS OF ANALYSIS CUF Increment:
For each SBF monitored location, the resulting fatigue increment is equal to the increment to bound the additional Reactor Trip events plus the increment to bound all other unaccounted transients - see Table 4.
.I Enclosure ET 08-0007 Page 10 of 11 Table 4: Fatigue Increments by Location SBF Location AU(Rx Trip)
AU(other)
AU(total)
CHRG NOZ 0.000152 0.0000333 0.000185 and ALT CHRG NOZ Table 4 (cont.)
HLNOZZLE 0.000152 0.02594 0.02609 LHEAD 0.000152 0.002137 0.00229 PZRHTRPEN 0.000152 0.03167 0.03183 SGA FW NOZ1 0.01906 0.11609 0.13515 through SGDFWNOZ4 SPRNOZ 0.000152 0.11931 0.11946 SRGNOZ.
0.000152 0.01686 0.01701 SRGLINE 0.000152 1.84e-5 0.000170 Revised CUF Baseline & Projection The CUF increments given in Table 4 can be added to the baseline and projections from Reference 3 (up through 12/31/05) and Reference 4 (up through 12/6/06).
The revised baseline and projections from the latest fatigue update Reference 4 are reported in Table 5.
Table 5: Revised Baseline CUF (up through 2/6/06)
Location UBL U60Y Uallow Baseline 60-year
% Used
% Used ACCSI NOZ 0.05121 0.06255 0.0651 79%
96%
ALT CHRG NOZ 0.02214 0.02789 0.1823 12%
15%
CHRG NOZ 0.09343 0.15579 0.1823 51%
85%
HL NOZZLE 0.05099 0.07771 0.1164 44%
67%
HPSI NOZ 0.11638 0.16057 0.1807 64%
89%
LHEAD 0.00306 0.00399 1.0 0%.
0%
PZR HTR PEN 0.03358 0.03584 0.0651 52%
55%
RPVIN NOZ 0.06345 0.13501 0.4082 16%
33%
RPVOUT NOZ 0.11475 0.21541 0.4082 28%
53%
SGA FW NOZI 0.30137 0.46851 1.0 30%
47%
SGA FW NOZ2 0.21960 0.29675 1.0 22%
30%
SGA FW NOZ3 0.17740 0.21192 1.0 18%
21%
SGA FW NOZ4 0.18192 0.22308 1.0 18%
22%
SGB FW NOZI 0.32240 0.52132 1.0 32%
52%
Enclosure ET 08-0007 Page 11 of 11 Table 5 (cont.)
SGB FW NOZ2 0.23011 0.31961 1.0 23%
32%
SGB FW NOZ3 0.17877 0.21710 1.0 18%
22%
SGB FW NOZ4 0.17951 0.21801 1.0 18%
22%
SGC FW NOZI 0.32397 0.51584 1.0 32%
52%
SGC FW NOZ2 0.23426 0.33162 1.0 23%
33%
SGC FW NOZ3 0.18101 0.22732 1.0 18%
23%
SGC FW NOZ4 0.17895 0.22461 1.0 18%
22%
SGD FW NOZ1 0.28638 0.39100 1.0 29%
39%
SGD FW NOZ2 0.24162 0.33569 1.0 24%
34%
SGD FW NOZ3 0.16289 0.18670 1.0 16%
19%
SGD FW NOZ4 0.24962 0.34490 1.0 25%
34%
SPR NOZ 0.12971 0.13945
.1.0 13%
14%
SRG NOZ 0.02198 0.02869 0.0651 34%
44%
SRGLINE 1.80E-04 2.OOE-04 1.0 0%
0%
6.0 REFERENCES
- 1.
Letter ET 06-0038, dated September 27, 2006, from T.J. Garrett, WCNOC, to USNRC.
- 2.
Letter ET 07-0046, dated October 3, 2007, from T.J. Garrett, WCNOC, to USNRC.
- 3.
SI Calculation, "Baseline Evaluation and 60-Year Projection for Wolf Creek", Revision 0, 5/25/06, SI File No. FP-WOLF-304.
- 4.
SI Calculation, "2006 Wolf Creek Baseline and 60-Year Projection Update", Revision 0, 2/11/07, SI File No. FP-WOLF-305.
- 5.
Telephone Conference Summary dated September 4, 2007, from V. Rodriguez, USNRC (ML072320487).
- 6.
Westinghouse Report ICE-ICAT(97)-012, Rev. 0, "Transient and Fatigue Cycle Monitoring-Transient and Fatigue History Evaluation Report of Wolf Creek Nuclear Operating Corporation, Wolf Creek Plant," April 1998, SI File No. UE-01Q-518.
- 7.
Westinghouse Document No.
1.3.X, "Systems Standard 1.3.X, Nuclear Steam SupplySystem, Auxiliary Equipment Design Transients for All Standard Plants", Sept.
1978, SI File No. PI-05Q-229P.
- 8.
Westinghouse Systems Standard 1.3, Revision 2, date, SI File No. WOLF-05Q-204P.
- 9.
Wolf Creek Control Room Logs, various days, SI File No. WOLF-05Q-201.