ML20196D415
| ML20196D415 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 11/29/1988 |
| From: | Fay C WISCONSIN ELECTRIC POWER CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| CON-NRC-88-118, RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, VPNPD-88-580, NUDOCS 8812090090 | |
| Download: ML20196D415 (33) | |
Text
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l Msconsm Electnc w com, 231 W ViCKCA'lP O BOX 2046.VtwAUkEE.Wt $r01 14 t 0 2212345 VPNPD-88-580 NRC-88-118 November 29, 1988 U.S. NUCLEAR REGULATORY COMMISSION Document Control Desk Mail Station P1-137 Washington, D.C.
20555 Gentlemen:
DOCKETS 50-266 AND 50-301 RESPONSE TO NRC GENERIC LETTER 88-11 ANALYSIS OF EMBRPlTLEMENT OF REACTOR VESSEL BELTLINE MATERIALS POINT BEACH NUCLEAR PLANT, tNITS 1 AND 2 NRC Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor ('essel Materials and Its Impact on Plant Operations"
, dated July 12, 1988 requires licensees to examine the effects of neutron radiation on reactor vessel materials using the methods described in Regulatory Guide 1.99 Revision 2.
Additionally, the results of this technical analysis and proposed actions to continue to sr.tisfy the requirements of 10 CFR 50 Appendix G are to be presented.
i Using Regulatory Guide 1.99 Revision 2, we have estimated the embrittlement of Point Beach reactor vessel beltline materials, and have examined the effects that this methodology has on pressure-temperature (P-T) limits, low temperature overpressure protection (LTOP) petpoints, and pressurized thermal shock (PTS' screening calculations.
We conclude that our present P-T curves will necessitate modification when the more conservative methods of Revision 2 are utilized to project the RT shift.
We thereforeplantosubmitrevisedP-TcurvesgpTJuly 1989.
The l
Point Beach LTOP pressure setpoints are unaffected by the use of Revision 2 at low temperatures.
At higher temperatures (>370'F),
a modification may be performed to increase the pressure window available to operators because of the encroachment of the subcooling and reactor coolant pump net positive suction head curves.
We will also have to update Point Beach operating procedures to include the new anable temperature determined from the new P-T curves.
Table 1 in the Attachment presents a listing of our overall reactor vessel integrity program, plans, and schedules.
go0h 89120+0090 SS1129 gg FDR ADOCK 05000266 p
PDC i
s Document Control Desk November 29, 1988 Page 2 In the Attachment, we have also exauined the effect of replacing the equat. ions for RT in the PTS rule with the methods of Section C.1 of Regul$Ibry Guide 1.99 Revision 2.
We conclude that with planned flux : eductions the Point Beach reactor vessels will remain below the PTS screening criteria through license expiration.
We nevertheless object to NRC's planned amendment to the PTS rule.
The substitution of Revision 2 and future revisions to Regulatory Guide 1.99 into the PTS rule are inappropriate and unnecessarily changes the order of plants approaching the screening criteria without any real corresponding change in risk.
The repeated addition of new margins can result in physical plant and operational changes which may negatively affect overall plant safety while providing an unquantified, possibly negligible, margin change for reactor vessel integrity.
As part of the reactor vessel life extension study for Point Beach Nuclear Plant, we performed a comprehensive scoping risk assessment which considered the effects of Regulatory Guide 1.99 Revision 2, low upper shelf fracture toughness, and other concerns on vessel failure risk.
We have also recently joined the Materials committee of the Babcock and Wilcox owner's Group and plan to utilize its integrated surveillance program to satisfy our future materials data needs.
We request a meeting with the NRC staff to discuss thece and other vessel integrity initiatives, which are identified in Table 1.
We believe our evaluation of the impact of Regulatory Guide 1.99 Revision 2 demonstrates that our existing plans and programs justify operation of the Point Beach reactor vessels through 40 operating years.
Please contact us if additional information is r e qui re d.
Very truly yours,
/
':/
i at.
l C.
W.
Fay l
Vice President I
Nuclear Power Attachment 1
Copy to NRC Residenc Inspector
Pcgo 1 of 31 PCgCO Attachment November 29, 1988 IMPACT OF REGULATORY GUIDE 1.99 REVISION 2 ON PRESSURE-TEMPERATURE (P-T) LIMITS.
10W TEMPERATURE OVERPRESSURE PROTECTIVE (LTOP) SETPOINTS AND PRESSURIZED THERMAL SHOCA (PTS) SCREENING CALCULATIONS POINT BEACH NUCLEAR PLANT l
METHODS The adjusted reference temperatures (ARTS) for all of the materials in the beltline region of the Point Beach reactor vessels have been calculated using the methods of Regulatory Guide 1.99 Revision 2 (hereafter RG 1.99 Rev. 2) Section 1.1 - Adjusted Reference Temperature.
However, material test reports, surveillance capsule date, and statistical studies recently obtained from the Babcock and Wilcox Owner's Group Materials Comn Ltee were used to determine the initial RT for the Point Beach beltline materials.
Thesedataalsodeterminethestbarddeviation(o,)utilizedinthe calculation of the margin term.
Table 4 discusses the origin of initial RT data and derivation of the margin terms in greater detail.
NOT The calculations of ART in this Attachment (Figures b-12) aso a
.rrent low leakage loadirig pattern (L3P) cores are continued through th. exp:
tion of our current licenses (EOL).
The fluence accumulation associa h these loading patterns is given in WCAP-10638, "Adjoint Flux Program i int Beach Units 1 and 2." (1) Pertinent portions of WCAP-10638, which. ovide background documentation concerning Point Beach fluence calculations, were provided in our January 20, 1986 and March 14, 1986 PTS submittals.
We note, however, that we have recently conducted a reactor vessel life extension study (2) and that we are in process of implementing core flux reductions at the critical reactor vessel welds.
This will occur in 1989.
Therefore, the ART calculations in this Attachment become more and more conservative with time. We will update our fluence projections after power history and excore dosimetry data become available following the implementation of super low leakage loading pattern (L4P) cores and hafnium inserts in the guide l
tubes of peripheral asssemblies.
Table 1 provides additional details con-cerning the Point Beach reactor vessel integrity program.
The following sections discuss the bases for our ART calculations and the conclusions to be drawn from these calculations.
IDENTIFICATION AND LOCATION OF BELTLINE REGION MATERIALS l
Figures 1 and 2 identify and indicate the location of all beltline region materials for the Point Beach Unit 1 and 2 reactor vessels, respectively.
The beltline region is defined to be "the region of the reactor vessel l
(shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent l
regions of the reactor vessel that are predicted to experience sufficient neutron irradiation damage to be considered in the selection of the most limiting material with regard to radiation damage".
l
Pego 2 M 31 Pcgos l
l REACTOR VESSEL NEUTRON FLUENCE l
Througn September 30, 1988, both Point Bee. reactors had been operated l
for a total of 13.1 effective full power p:.rs (EFPY).
Table 2 documents the periods of reactor operation (EFPY) to present, for both current and plar.ned pressure-temperature limit curves, and until license expiration.
It l
i is also seen in Table 2 that the current Point Beach P-T curves expire in the beginning of 1990.
Figures 3 and 4 are reproduced from WCAP-10638, "Adjoint Flux Program for Point Beach Units 1 and 2," and show the aaximum fast neutron (E>l MeV) fluene.e at the beltline weld locations as a function of full power operating time for Units 1 and 2 respectively.
Table 3 results from converting the reactor operating periods in Table 2 to fluence utilizing Figures 3 and 4, as appro-priate.
The next section explains the evolution of core loading patterns at Point Geach Nuclear Plant.
It also explains the makeup and bases of Figures 3 and 4.
CORE LOADING PATTERNS For Point Beach Unit 1, beginning with Cycle 8 in 1980, core loading patterns employed a Low Leakage Loading Pattern (L3P) design.
Assemblies with several previous cycles of burnup were positioned at certain locations on the core periphery.
The L3P design was fully implemented for PBNP Unit 2 during Cycle 7, which also occurred in 1980.
Prior to this time, new fuel had bean placed on the core periphery.
In 1984 the Westinghouse Adjoint Flux Program was utilized to assess the effects that past and present core management strategies have had on neutron fluence levels in the reactor pressure vessel.
Figures 3 and 4 have been reproduced from WCAP-10638 and represen: the application of the "Adjoint Flux Program For Point Beach Units 1 and 2".
In regard to Figures 3 and 4, the solid portions of the curves are based directly on calculations that utilized the PBNP plant specific core power distributions through fuel cycle 11 for Unit 1 and fuel cycle 10 for Unit 2.
The dashed portions of these curves involve a projection into the future, based on the average neutron flux at the key locations over the low leakage fuel cycles.
For Unit 1 the neutron flux average over cycles 8 through 11 was used to projer,t future fluer.ce levels, while the neutron flux average over cycles 6 through l
10 was employed for Unit 2.
i 1
Optimized fuel assemblies (OFA) were utilized at PBNP in core reloads beginning in the Fall of 1984 for Unit 2 and the Spring of 1985 for Unit 1.
Complete l
transition to 0FA occurred at the fourth core reload of 0FA for each unit, specifica11v the Fall of 1987 for Unit 2 and the Spring of 1988 for Unit 1.
I The fluence proj,tettons in Figures 3 and 4 are applicable to both 0FA and normal fuel, because the core power distributions of 0FA and normal fuel are basically the same.
Hence, the plant specific fluence distributions, which are calculated by adjoint importance functions that directly relate the spatial distribution of fission density within the reactor core to the re-sponse of interest on the vessel, are unchanged.
The L3P cores described earlier and the transition to 0FA fuel have the benefit of both fuel cycle economy and lower neutron flux at the vessel wall.
Page 3 of 31 Pages I
In 1989, Point Beach will adopt a new approach to fuel management.
This new fuel management scheme was selected from fourteen possible techniques and consists of a super low leakage loading pattern (L4P) and hafnium (Hf) neutron absorber assemblies in the guide tubes of peripheral fuel assemblies.
Twelve hafnium assemblies will be installed in each unit at Point Beach.
Each assembly contains 16 part-length hafnium flux depression rods.
These essemblies are zoned axially and placed in a position near the vessels' cardinal axes to shield the critical welds in each unit.
With LAP reload cores and hafnium inserts installed, the following flux reduction factors are expected to be achieved:
Flux Reduction Factor (FRF) Achieved 0
15 t
Unit 1 2.0 1.8 (Lower Shell) 1.4 (Intermediate Shell)
Unit 2 2.0 Excore neutron dosimetry is being Installed to verify that the expected flux reductions are attained from this new fuel management scheme.
(See Table 1 for the schedule of implementation.)
DEFINITION OF PLANf-SPECIFIC MATERIAL PROPERTIES The pertinent chemical and mechanical properties of the beltline region l
plate, forging, and weld materials of the Point Beach Unit 1 and 2 reactor vessels are given in Table 4.
The copper and nickel values for each material are the same as provided in our PTS submittal.[3,4,5] The chemistry values for the shell plates and forgings were derived from vessel fabrication test cert'ficates and results from chemical analyses of surveillance capsule material performed by Westinghouse.
The Cu and Ni estimates for the welds were obtained from the Westinghouse Owners Group (WOG) Material Data Base, Revision 0.
These chemistry estimates represent the averagu of the records found in the data base for a given weld wire heat, wire type, flux type, and flux lot (s).
For the calculation of adjusted reference temperature using RG 1.99 Rev. 2, the initial RT values and uncertainties (o ) were taken from material g
testreportsaNOTfrom statistical evaluations performed by Babcock and Wilcox [6,7] The guidance in RG 1.99 Rev. 2 was employed to determine the 10 CFR chemistry factors and overall margin terms used to calculate RT 50.61wasthesourceformarginterms,aswellasthegenerici$k[,alRT 1
value for Linde 80 welds, used in the RT calculations.
Also,aswasUSde inour10CFR50.61 submittal,initialRkTS values for plates and forgings wereestimatedaccordingtoBranchTechnikkIPositionMTEB5-2fortheRT PTS calculations in Table 7.
ADJUSTED REFERENCE TEMPERATURE The information in Tables 3 and 4 was used to calculate the adjusted reference i
temperatures (ART) for all of the beltline materials.
The exponential attenua-v-- - - -
.-----w--
Pogo 4 of 31 Pagcc tion factor (dpa) in Equation 3 of RG 1.99 Rev. 2 was utilized in the calculations of ART at the one quarter thickness (1QT) and three quarters thickness (3QT)locNIonsinthevesselwall.
Figures 5-12 present ART as a function of fluence for the various Units 1 and 2 reactor vessel beltline materials.
PRESSURE-TEMPERATURE (P-T) LIMIT CURVES The current P-T curves were calculated to be applicable through 14 EFPY or early 1990.
Table 5 presents the assumptions inherent in the current Point Beach P-T curves.
For operator convenience, the current P-T curves are based on the most limiting combination of material and fast neutron fluence, which is the Unit 2 circumferential weld SA-1484.
Thus, one set of curves is appitcable to both Units 1 and 2.
Examination of Figures 5 through 12 shows that the Unit 2 circumferential shift at Point weld (Figure 10) is still limiting in terms of RT Beach.
TheARTsforUnit2atthelocationsofcNerninthevesselwall are as follows:
UNIT 2 WELD:
SA-1484 Jp(*F) 1_qT(*F) 30T(*F)
Present (09/30/88) 263.8 244.4 205.3 Expiration Current P-T Curves 266.9 247.6 208.4 14EFPY (01/01/90)
Expiration Planned P-T Curves 277.2 258.4 219.2 (01/01/95)
As can be seen, the RT shift predicted by RG 1.99 Rev. 2 presently exceedstheprojection$DInTable5, Item 5obtainedbyusingtheWestinghouse trend formula at fluence values corresponding to 14 EFPY.
Thus, the RG 1.99 Rev. 2 methodology and the neutron attenuation function therein combine to necessitete a revision to the current P-T curves in the Point Beach Technical Specifications to adopt the more conservative prediction techniques of RG 1.99 Rev. 2.
New P-T curves will be calculated using RG 1.99 Rev. 2 and submitted in July 1989.
Althcugh not included in the calculations of this submittal, we are examining the calculation procedures of Paragraph 2.1 - Adjusted Reference Temperature in RG 1.99 Rev. 2.
As we recently joined the Babcock and Wilcox Owner's Group (BWOG) Materials Committee, we expect to improve our computations of ARTNDT and ca based on the credible surveillance data in the BWOG integrated surveillance program in the future.
LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP)
The LTOP system at Point Beach is a fixed setpoint system which utilizes the pressurizer power operated relief valves (PORVs) as the diverse means of pressure relief during solid water operation.
Page 5 of 31 Pages setpoints of 415 psig and 425 psig, respectively.
A keylock switch is used under administrative procedure to enable the LTOP system whenever the reactor coolant system temperature is less than the minimum temperature for inser-vice pressure test (MPT).
In addition to the LTOP system, Point Beach has several mechanical relief valves set at 500 psig or 600 psig that protect low pressure piping and the reactor coolant system during solid operations on the residual heat removal (RHR) system.
Use of the methods in RG 1.99 Rev. 2 to predict the embrittlement for cal-culating P-T curves affects the calculation of MPT and the LTOP enable set-point, but does not affect the relief setpoint of the PORVs at Point Beach.
The latter occurs because the Point Beach reactor vessels have already experienced significant shifts in RT to date.
Therefore, the low temperatureendofthePointBeachPkTcurves is already governed by the flat curve, which asympto-portion of the reference stress intensity factor - K tically approaches a value of 26.78 ksi,[lT, at low thperatures relative to Be b.
In other wnrds, the low temperature end of the P-T curves at Point RT does not experience any appreciable drop in allowable pressure due to indexing of the K curve via RT calculated using RG 1.99 Rev. 2.
- Thus, the LTOP pressure $etpoint does b need to be changed.
However, to further open the operating window at temperatures above 370 F, a modification may be performed to provide a higher, two stage or ramping setpoint for LTOP pressure.
i This modification would provide relief from the subcooling and reactor coolant pump net positive suction head curves which encroach on the LTOP setpoint at higher (>370'f) temperatures.
In terms of LTOP enable temperature, the methodology of RG 1.99 Rev. 2 has a significant impact, which is shown in the table below:
Current P-T Curves Planned P-T Curves (Applicable to January 1990)
(Applicable to January 1995)
MPT 354 F 392'F*
LTOP Enable Temper *.ture 370*F 405'F*
(per Administrat'.ve Procedre)
- Estimated with instrumented error included.
Overall, the methodology of RG 1.99 Rev. 2 narrows the plant operating window at the higher temperature end of the P-T curves only.
The low temperature pressure window remains as is.
The LTOP pressure relief setpoint and its bases are unaffected.
In conclusion, because the minimum differential pres-sure required by the seals in the reactor coolant pumps is 200 psid, Point Beach retains operating flexibility at low temperatures, although the methodology of RG 1.99 Rev. 2 is employed.
At higher temperatures, a modi-fication to LTOP circuitry may be performed to retain operating flexibility.
PRESSURIZED THERMAL SHOCK (PTS)
J The methods of Section C.1 of RG 1.99 Rev. 2 have been employed to calculate RT f r comparison to the PTS screening criteria.
Tables 4 and 6 provide NDT
Pago 6 of 31 Pagan the inputs for the calculations in Table 7.
Three cases were calculated for each of the welds ** in the Point Beach reactor vessels:
- Present (09/30/88)
- License expiration assuming L3P core configuration continues.
- License expiration assuming flux reductions are implemented in 1989 with L4P cores with Hf inserts and continued thereafter.
Table 7 shows that the flux reductions that we plan to implement in 1989 adequately reduce fluence ensuring the Point Beach reactor vessels do not reach the screening criteria.
Additionally, Point Beach has conducted a comprehensive scoping risk assessment [8] of the reactor vessels.
The its is two orders of overall risk of vessel failure for both Point Beach ug/R-YR defined in magnitude below the acceptance criterion of 5.0 x 10 Regulatory Guide 1.154 for the assessment vessel risk from PTS events.
The Point Beach scoping risk assessment considered the effects of RG 1.99 Rev.
2, low upper shelf fracture toughness, and plant specific concerns in the evaluation of vessel failure risk from PTS events.
CONCLUSION We have made a detailed calculation of the embrittlement of Point Beach reactor vessel materials using the methods of RG 1.99 Rev. 2.
The embrittle-ment projections herein do not take into account the expected neutron fluence reductions that will accrue from our L4P with hafnium insert reload cores, which will be implemented in 1989.
To incorporate RG 1.99 Rev. 2 methods, new P-T curves will be submitted in July 1989.
The pressure setpoints for LTOP remain adequate at low temperatures.
However, at higher temperatures a modification to the LTOP system may be performed to increase the operating pressure window because of the encroach-ment of the subcooling and reactor coolant pump net positive suction head curves.
Our administrative procedures will also need to be changed to reflect the new LTOP enable temperature appropriate for our new P-T curves.
With respect to PTS, the Point Beach reactor vessels approach the screening criteria near license expiration with current core designs.
We expect, however, to significantly reduce the shift in RTTa$idpriortothattimeby implementing planned flux reduction schemes.
1 presents a listing of our overall reactor vessel integrity program, plans and schedules.
- It can be seen in Figures 8, 9, 11, and 12 that the plates and forgings in the Point Beach reactor vessels are not limiting for PTS purposes,
- 1 Pago 7-of.31 Pagno REFERENCES 1.
- Anderson, S. L.
and Balkey, K.
R.,
"Adjoint Flux Program'for Point Beach Units 1 and 2," WCAP-10638, December 1984.
2.
Perone, V.
A.,
et al, "Flux Reduction Evaluation for the Point Beach Units 1 and 2 Reactor Vessel-Life Extension Study," WCAP-ll535, September 1987.
3.
Letter C. W. Fay to H. R. Denton, "Docket Nos. 50-266 and 50-301 Response to 10CFR50.61 Protection Against Pressurized Thermal. Shock (PTS) Point Beach Nuclear Plant, Units 1 and 2,"
January 1986.
4.
Letter C. W.
Fay to H. R. Denton, "Docket Nos. 50-266 and 50-301 Correction to Pressurized Thermal Shock (PTS)
Submittal Dated January 20, 1986 Point Beach Nuclear Plant, Units 1 and 2,"
March 1986.
5.
Letter USNRC, T. G. Colburn to C. W.
Fay "Projected Values of Material Properties for Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Point Beach Nuclear Plant Unit No. 1 and 2," July 24, 1986.
6.
Heller, A. S.
and Lowe, Jr.,
A.
L.,
BAW-1803, "Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds", January 1984.
7.
Lowe, Jr., A.
L.,
et al, BAW-1895, "Pressurized Thermal Shock Evaluation In Accordance With 10CFR50.61 for Babcock and Wilcox Owners Group Reactor Pressure Vessels", January 1986.
8.
- Balkey, K.
R.,
et al "Scoping Risk Assessment for the Point Beach Units 1 and 2 Reactor Vessel Life Extension Study,"
WCAP-11676 (Westinghouse Proprietary) December 1987.
2 I
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'1 l-ll
TABLE 1 POINT BEACH NUCLEAR PLANT REACTOR VESSEL INTEGRITY PROGRAM 1984 - PRESENT y
d2 0
Date Planned M
Project Date Complete For Implementation 0
Neutron exposure evaluation of Point Beach reactor vessels.(1)
Hi December, 1984 1.
ta E'
Tested Unit 1 Surveillance capsule T.I December, 1984 2.
'o 10 CFR 50.61 - Pressured Thermal Shock (PTS) Submittal.(3)
January, 1986 d$
3.
e Correction to PTS submittal.{4)
March, 1986 Safety evaluation report received from NRC.
July, 1986 4.
Reactor Vessel Life Extension Study.
Initiated study in May, 1986.
Evaluationoffuelmanagementtechniquesygjinternalsmodifications (shielding) to meet flux reduction goals.
September, 1987 Identification of critical components in NSSS, including the reactor vessel, and 99mpilationoftransientdataassociatedwiththese October, 1987 components.
ComprehensivescopingriskassessmenttoexaminePoint{gjchspecif:
concerns and the propriety of the flux reduction goals.
December, 1987 Developed bases an i
i for a plantwide on-line fatigue monitoringsystem.g9ypecifcatons December, 1987
TABLE 1 POINT BEACH NUCLEAR PLANT j
REACTOR VESSEL INTEGRITY PROGRAM 1984 - PRESENT 4
i e -
so Date Planned Project Date Complete For Implementation O
5.
Inservice Inspection ta H
a.
Second Unit 1 Reactor Vessel Ten-year Exam:
y*
Performed ASME Code exam utilizing SWRI standard data May, 1987 j
a acquisg g system, including 50/70 tandem near surface search units.
Performed exam using NES/Dynacon Ultrasonic Data Recording and (37)
May, M87 Processing System (UDRPS) concurrent with ASME Code exam above.
b.
Second Unit 2 Reactor Vessel Ten-year Exam:
i SWRIS EDAS or NES/Dynacon's UDRPS system will be utilized.
October, 1989 6.
Joined Babcock and Wilcox Owner's Group (BWOG) Materials Comunittee.
August, 1988 Full participant in BWOG Reactor Vessel Integrity Program (RVIP).
August, 1988 Participant in BWOG Reactor vessel Life Extension Surveillance Prograan (RVSP).
1989 Developing master integrated reactor vessel surveillance program to include Westinghouse utilities with Linde 80 welds in their reactor vessels. This document will describe the new Point Beach Nuclear Plant surveillance capsule schedule and will be referenced in the Point Beach Technical Specifications.
March, 1989
I
~
TABLE 1 POINT BEACH NUCLEAR PLANT REACTOR VESSEL INTEGRITY PROGRAM y
D 1984 - PRESENT soo M
Date Planned w
Date Complete For Implementation Project on Installation of excore neutron dosimetry (radiometric monitors 7.
and solid state track recorders) over one octant of both unit's y
reactor vessels.
D so November, 1988 o
Install mounting hardware and first set of dosimetry in Unit 2.
os May, 1989 Install mounting hardware and first set of dosimetry in Unit 1.
Analysis of sensor sets and correlation of cavity measurements with transport calculations will be performed after each fuel cycle until sufficient confidence exists to extrapolate neutrcn exposure data through expected plant life.
November, 1988 8.
Pilot project: On-line fatigue monitoring of Unit 2 pressure surge nozzle (related to reactor vessel life extension study fatigue evaluation).
Implement super Low Leakage Loading Pattern (L4P) cores and 9.
axially-zoned hafnium inserts in the guide tubes of peripheral assemblies.
May, 1989 Unit 1 November, 1989 Unit 2 1989 Perform image enhancement of selected radiographs of important 10.
reactor coolant,ystem components (reactor vessels, piping, steam generators, etc.) and retain radiograph image on more permanent media.
Submit revised heatup and cwldown curves using the guidance of July, 1989 11.
Regulatory Guide 1.99, Revision 2.
i
Pego 23 of 31 Pagao SUPPORTING DOCUMENTATION FOR TABLE 1 1.
- Anderson, S.
L.
and Balkey, K.
R.,
"Adjoint Flux Program for Point Beach Units 1 and 2,"
WCAP-10638, December 1984.
2.
- Perone, V.
A.,
et al, "Flux Reduction Evaluation for the Point Beach Units 1 and 2 Reactor Vessel Life Extension Study," WCAP-11535, September 1987.
2.
- Yanichko, S.E.,
et al, "Analysis of Capsule T From the Wisconsin Electric Power Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Radiation Surveillance Program,"
WCAP-10736, December 1984.
3.
Letter C.
W, Fay to H.
R. Denton, "Docket Nos. 50-266 and 50-301 Response to 10CFR50.61 Protection Against Pressurized Thermal Shock (PTS) Point Beach Nuclear Plant, Units 1 and 2,"
January 1986.
4.
Letter C.
W.
Fay to H.
R.
Denton, "Docket Nos. 50-266 and 50-301 Correction to Pressurized Thermal Shock (PTS)
Submittal Dated January 20, 1986 Point Beach Nuclear Plant, Units 1 and 2," March 1986.
5.
Letter USNRC, T.
G.
Colburn to C.
W.
Fay "Projected Values of Material Properties for Fracture Toughness Requirements for Protection Against Pressurized hermal Shock Events, Point Beach Nuclear Plant Unit No. J
.d 2,"
July 24, 1986.
7.
- Bond, C.
B.
and Pepki, J.
M.,
Transient Monitoring Program for Wisconsin Electric Power Company Point Beach Units 1 and 2,
Phase 1 Final Report," WCAP-13501 (Westinghouse Proprietary) October 1987 8.
- Balkey, K.
R.,
et al "Scoping Risk Assessment for the Point Beach Units 1 and 2 Reactor Vessel Life Extension Study,"
WCAP-ll676 (Westinghouse Proprietary) December 1987.
9.
- Bond, C.
B. and Pepka, J.
M.,
"Point Beach Units 1 and 2 Transient and Fatigue Cycle Monitoring System Functional Requirements," WCAP-11706, Decemoer 1987.
10.
- Enoch, H.
D.,
"1987 Inservice Examination of the Reactor Pressure Vessel at Point Beach Nuclear Plant Unit 1,"
Final Report (Volumes 1-5) SwRI Project 1597, August 1987.
11.
- Fong, R.
and Martens, G.,
"UDRPS Report on Automated Data Acquisition and Processing for the Point Beach Nuclear Power Plant Unit 1,
10-Year Reactor Pressure Vessel Examination, Spring 1987 Outage," July 9, 1987.
4 TABLE 2 POINT BEACH UNIIS 1 AND 2 ACCtMJLATED YEARS (EFPY) OF REACIOR OPERATION *
,,ee m
DJ b
UNIT 1 UNIT 2 o
n Predicted Predicted Period Date EFPY Date EFPY U
- o
- To present 09/30/88 13.1**
09/30/88 13.1**
j-em
- Until expiration of current pressure -
4 1/01/90 14.0 4 2/15/90***
14.0 temperature (P-T) limits in P8NP Technical Specifications
- Through period of applicability of planned pressure 01/01/95 18.0 01/01/95 18.0 tsaperature (P-T) limits generated using R.G. 1.99, Rev. 2.
- Until license expiration 10/05/2010 30.6 03/08/2013 32.5 EFPY - Effective Full Power Years
- Assumes a cumulative capacity factor of S0% for projected EFPY values.
- Denotes actual years of reactor operation through September 30, 1988.
- Point Beach Technical Specifications estimate that the Unit 2 P-T curves are applicable to September 1990. However, I
above average generation has been sustained in Unit 2 causing cumulative reactor operation to reach 14 EFPY j
sooner than previously anticipated.
. ~ - -.. - -,
Paga 25 of 31 PGgas f
TABLE 3 POINT BEACH UNITS 1 AND 2 ACCUMULATED NEUTRON FLUENCE (E>l MeV) i 2REDICTEDCUMULATIVEFLUENCE (10 n/cm )
j 19 2
VESSEL AXIMUTHAL LOCATIONS i
UNIT 1 OATE 0*
15' Present 09/30/1988 1.65 0.97 l
Expiration of Current P-T Limit curves 01/01/1990 1.65 1.05 Expiration of Planned P-T Lim:ts 01/01/1995 2.05 1.30 (Generated using RG 1.99 Rev. 2) l License Expiration 10/05/2010 3.30 2.10 r
i VESSEL AZIMUTHAL t
i UNIT 2 DATE LOCATION-0*
i present 09/30/1988 1.55 1
Expiration of Current P-T Limits
$02/15/1990 1.65 i
l Expiration of Planned P-T Limits 03/01/1995 2.05 License Expiration 03/08/2013 3.50 r
i i
i i
8 Figures 3 and 4 were utilized to convert reactor operating period (EFPY) to i
the inside surface fluence values for the reactor vessels at the azimuth t
angles shown.
l f
L I
h i
I i
l
,.j' k TA8LE 4 POINT BEACH UNITS 1 AND 2 REACTOR VESSEL BELILINE REGION MATERIAL PROPERIIES
- Chemical Comoosition *)
RG 1.99 Rev. 2[16]
10 CFR 50.61[13,14, 5 I
R{nitial(b) Chemistry Margin nitial Margin g{NOT(*F)
Cu Ni 2o(*F)
UNIT 1 (Wt.%) (Wt.%)
NDT(*F)
Factor 20 (*F) 34 ')
-2 48 I
Intermediate Shell Plate A-9811:
0.20 0.056
-25[8,9]
87.6 Lower Shell Plate C-1423:[2,3]
0.12 0.065
-20[9,10]
54.85 34(*)
-20 48 3
0.19(d) 0.63(d)
-6 162.1 68(I) 0(C) 59 II13 Axial Weld - Inter. Shell SA-775/812:
Weld Wire Heat Hos. IP0815/IP0661 Linde 80 Flux Lots 8350/8304 169.0 68(I) 0(c) 59 E43 EI13 Axial Weld - Lower Shell SA-847:
0.25 0.55
-6
'deld Wire Heat No. 61782 Linde 80 Flux Lot 8350 IC) 152.25 68(#)
O 59 Ell 3 EN 0.20 0.55
-6 Circumferential Weld - Inter.:
to Lowcr Shell SA-1101 Weld Wire Heat No. 71249 Lhae 80 Flux Lot 8350 UNIT 2 f9) 0.09 0.70
+3[12]
58.0 69 IntermediateShellForging153V500:[5,6]
40 48 I9)
Lower Shell Forging 122W195:[6,7]
0.05 0.72
+3[123 31.0 69 40 48 Circumferential Weld - Inter.:
0.2ti 0.60
-6 180.0 68( )
O(c) 59 E#3 EII3 to Lower Shell SA-1484 Weld Wire Heat No. 72442 Linde 80 Flux Lot 8579
- Figures below are the same as was t. M in the Point Beach Nuclear Plant January 1986 PTS submittal.
( ) indicate notes; [ ] indicate documentation for Table 4.
Pago 27 of 31 Prego::
a NOTES TO TABLE 4 (a) The chemistry values for the shell plates and forgings were derived from vessel material test reports and surveillance capsule chemistry measure-ments.
The chemistry valt.es for welds were derived from searches in the WOG Materials Data Base, Rev. O and represent rounded, average values.
RT (b) Initial NDT was determined according to the rules of the ASME Boiler and Pressure Vessel Code,Section III, Paragraph NB-2331.
Additionally, where noted, data were treated statistically to obtain the mean value of initial RT and the correspondino ttandard deviation (01).
NDT RT (c) The initici NOT values for weld are ep.1eric mean values defined by the PTS Rule at 10CFR50.61 (b)(2)ii.
(d) The chemistry data for SA-775 was utilized c'.:.e this will result in a conservative calculation for this weld.
RT (e) The plate initial NDT value was measured from material-specific drop weight and charpy data.
Hence, el equals zero (0) as the test methods precisely determined initial RT The margin added to obtain conserva-tive, upper bound values of adjbe.d reference is therefore 2ca or 34*F.
(f) The statistical evaluation in BAW-1803[11] Table 3-5 concludes that the mean initial reference temperature for Linde 80 welds is -6*F and the standard deviation (al) about this mean is 19 F.
FromReg. Guide 1.99 Rev. 2[16], the standard deviation for ARTNOT (ca) is 28 F for welds.
Thus, the margir. applied equ. sis M = 2/(19)2 + (287 = 67.67 s 68 F
=
(g) The statistical evaluation in BAW-1ts95[12] concludes that the average initial reference temperature for SA 508 CL.2 forgings is +3*F with a standard deviation of 30*F.
From Reg. Guide 1.99 Rev. 2 [16), the standard deviation for ARTNOT (ca) is 17'Ffor base metal.
The margin applied is therefore.
s 69'F Margin = 2/(30)2 + (17)2 68.8 =
=
Pcgo 28 of 31 Pagan j
. DOCUMENTATION FOR TABLE 4 i
1.
Lukens Steel Company Test Certificate No. RM12965-NS, January 3, 1966 i
for Babcock and Wilcox Company, 2.
WCAP-10736, "Analysis of Capsule T from the Wisconsin Electric Power l
Company Point Beach Nuclear Plant Unit No. 1 Reactor Vessel Radiation f
Surveillance Program", December 1984.
l 3.
Lukens Steel Company Test Certificate No. RM61766-BB, January 20, 1966 l
for Babcock & Wilcox Cc.apany.
j r
4.
Westinghouse Owner's Group (WOG), "Reactor Vessel Materials Data Bast",
t' Revision 0, March 1985.
5.
Bethlehem Steel Corporation Test Report No. 911, July 15. 1968 for Babcock
(
& Wilcox Company.
A.
WCAP-7712, "Wisconsin Michigan Power Co. and the Wisconsin Electric Power Co. Point Beach Unit No. 2 Reactor Vessel Radiation Surveillance Program", June 1971.
i 7.
Bethleham Steel Corporation Test Report No. 917, July 18, 1968 for Babcock
& Wilcox Company.
8.
Lukens Steel Company, A-9811 Materials Test Certificate - File No. 602, January 3, 1966.
[
9.
WCAP-7513. "Wisconsin Michigan Power Co. Point Beach Unit No. 1 Reactor i
Vessel Radiation Surveillance Program", June 1970.
10.
'ukens Steel Company, C-1423 Materials Test Certificate - File No. 602, June 20, 1966, 11.
BAW-1803, "Correlations for Predicting the Effects of Neutron Radiation i
on i.inde 80 Submerged-Arc Welds", January 1984.
[
12.
BAW-1895, "Pressurized Thermal Shock Evaluation In Accordance With L
10CFR50.61 for Babcock and Wilcox Owners Group Reactor Pressure Vessels",
I January 1986.
r 13.
U.S. NUCLEAR REGULATORY COMMISSION, 10CFR Part 50.61, "Fracture Toughness j
Requirements for Protection Against Pressurized Thermal Shock Events",
July 23, 1985.
14.
Letter from C. W. Fay to H. R. Denton (NRC), "Response to 10CFR50.61 Protection Against Pressurized Thermal Shock (PTS) Point Beach Nuclear Plant, Units 1 and 2", January 20, 1986.
l 1
15.
Letter from C. W. Fay to a Denton (NRC), "Correction to Pressurized i
Thermal Shock (PTS) Submitia: Dated January 20, 1986 Point Beach Nuclear
[
Plant, Units 1 and 2", March 14, 1986, t
i 16.
U.S. NLC Regulatory Guide 1.99, Revision 2, Radiation Embrittlement of
[
Reactor Vessel Materials", May 1988.
[
Ptge 29 of 31 Pegos i
TABLE 5 1981 PRESSURE-TEMPERATURE (P-T) LIMIT CURVE SUBMITTAL (1)
POINT BEACH NUCLEAR PLANT i
i I
l I
Assumptions of Current P-T Curves a
l n/cm2 1/4T=1.05X1h9 1.
14 EFPY:
Fluence 4
j (E>l MeV) 3/4T = 3.3 X 10 n/cm2 4
i i
I 2.
Controlling Welds:
Unit 1 SA-847 Lower Shell Axial Weld I
Unit 2 SA-1484 Inter.-to-Lower Shell Girth Weld i
I2) 3.
Trend Formula:
Westinghouse Trend Formula I
4.
t f
5.
RTNOT 1/4T = 225'F RTNOT 3/4T = 170*F
{
i j
6.
Unit 2 P-T curves utilized for both units for convenience of the PBNP 1
operators.
Additionally, Unit 1 Weld SA-847 is located 15' off the cardinal (pech fluence) axes, and thus, the Unit 2 P-T curves are more restrictive.
l 5
I Noten
- 1) In the Point Beach Technical Specifications, the current P-T curves were approved in the following:
Unit 1 License Amendment #53 dated August 28, 1981 Unit 2 License Amendment #57 dated July 10, 1981
- 2) See WCAP-7924-A "Basis for Heatup and Cooldown Limit Curves" by W.S. Hazelton et. al. dated April 1975.
.h TABLE 6 REACTOR VESSEL ID FLUENCE (E > 1 kev) PROJECTIGNS*
FOR :'RESSURIIED IHLRMAL SHOCK SCREEMING CALCULATIONS m=
4 POINT BEACH UNIIS 1 AND 2 uo Unit 1 Unit 2 Q
EFPY 0*
15' EFPY 0*
"r w
1.
Present: 09/30/88 13.1 1.55
.97 13.1 1.55-m-
j 2.
License Expiration:
Unit 1 - 10/05/2010 30.6 3.30 2.10 32.5 3.50 Unit 2 - 03/08/2013 (w/L3P cores continued) 3.
Implement Flux Reductions 13.5 1.59 1.00 13.8 1.62 Unit 1 - March l' S Unit 2 - October 1989 (w/L4P cores and hafnium inserts) 4.
License Expiration: (See dates in No. 2 above) 30.6
-2.45 1.61**
32.5 2.55 1.79'**
(w/L4P + Hf cores implemented when planned and continued t9 EOL) 1
- Assum% celative capacity factor of 80%
Fluence values are x10 n/cm2
- Lowe; shell axial weld: SA-847
- Intermediate shell axial weld: SA-775/812 l
v -w, - -
~. w,-
e e - m m
,e, m,
--w-e
,m vn+,--mm-,w--
---~nem-,--,-,-,n~
-v-om
,-,,,-,m v,,-- = -, -,
,n-~- - - ---
,-v-----
,-n
Pcgo 31 of 31 Pag 3e a
e TABLE 7 i
COMPARISON FOR PTS RULE PURPOSFS REFERENCE TEMPEP.ATURES FOR REACTOR VESSEL BELTLfNE WELD MATERIALS
- POINT BEACH NUCLEAR PLANT RG 1.99 REY. 2 vs. 10CFR50.61 h
Unit 1 i
SA-775/812**
SA-847**
SA-1101***
Axial Weld Axial Weld Circumferential Weld
~
PERIOD Inter. Shell Lower Shell Inter. to Lower Shell RG 1.99 RTPTS RG 1.99 RTPTS RG 1.99 RTPTS f
Present (09/30/88) 222.7 17.4.2 229.6 213.4 232.7 196.9 License Expiration (10/05/2010) 256.8 207.1 265.1 249.1 261.9 228.1 w/L3P Cores Continued License Expiration (10/05/2010) 250.0 200.8 253.2 236.0 251.0 215.0 w/ Expected Flur Reductions from Planned L47 and Hf Insert Core Designs l
Unit 2 l
SA-1484***
Circumferential Weld Inter. to Lower Shell RG 1.99 RTPTS
[
i Present (09/30/88) 263.8 246.8 i.
i.
License Expiration (03/08/2013) 300.8 292.9 l
w/L3P Cores C ntinued License Expiration (03/08/2013) 287.2 273.8 l
w/ Expected Flux Reductions from i
Planned L4P and Hf Insert Core i
Designs j
- Predicted PTS values assume a cumulative lifetime capacity factor of 80%.
- Applicable PTS screening criterion - 270*F.
- Applicable PTS screening criterion - 300'F.