ML20081F713

Forwards Pressure Temp Limit Rept Methodology,As Originally Requested by TR Tjader

ML20081F713
Person / Time
Site:	San Onofre
Issue date:	03/20/1995
From:	Marsh W SOUTHERN CALIFORNIA EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9503220240
Download: ML20081F713 (9)

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f[5 Southem Califomia Edison Company 23 PAFtKER STREET IRVINE, CALIFORNIA 92718 March 20, 1995 wALTen e. uAnss m -.

U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C.

20555 Gentlemen:

Subject:

Docket Nos. 50-361 and 50-362 Pressure-Temperature Limit Report Methodology (PTLR)

San Onofre Nuclear Generating Station Units 2 and 3 i

Reference:

Amendment Applications No.137 and No.121 (PCN 299) for San Onofre Unit 2 and Unit 3, respectively As originally requested by Mr. T. R. Tjader, Review Coordinator for the San Onofre Technical Specification Improvement Program (TSIP) submittal, enclosed is "The Pressure-Temperature Limit Report (PTLR) Methodology."

This methodology will be used at San Onofre Nuclear Generating Station when updating its Pressure-Temperature curves.

The PTLR methodology has been formatted to follow the logic of the Table of Requirements taken from the Generic Letter in preparation as provided to the TSIP lead plants. To i

facilitate the timely review of the San Onofre Units 2 & 3 TSIP submittal, Proposed Change Notice (PCN) 299, reference to the PTLR was deleted-and the Pressure-Temperaturc curves re-instated.

Following NRC review and approval of j

the methodology it is Edison's intent to submit a License Amendment Request implementing the PTLR arovisions of NUREG 1432 (Standard Technical Specifications for Comaustion Engineering Plants).

Should you have any comments or questions concerning this submittal, please let us know.

Sincerely, Enclosures 9503220240 950320 PDR ADOCK 05000361 k

PDR l

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. Document Control Desk i cc:

.L. J. Callan, Regional Administrator, NRC Region IV A. B. Beach,-Director, Division of Reactor Projects, Region IV K. E. Perkins, Jr., Director, Walnut Creek Field Office, NRC Region IV J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 1

M. B. Fields, NRC Project Manager, San Onofre Units 2 and.

C.;I. Grimes, Chief, NRC Technical Specifications Branch M. W. Weston, NRC Technical Specifications Branch T. R. Tjader, Review Coordinator - CEOG STS, NRC Technical Specifications Branch H. Kocol, California Department of Health Services I

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Pressure Temperature Limit Report Methodology V,

1. Transport Calculation Methodolozy Used to Calculate Neutron Fluence The transport calculation used at SONGS accounts for differences in neutron energy spectra between surveillance capsule locations and positions within the vessel wall.

SONGS has adopted an energy dependant displacements per iron atom (dpa) function for the radiation analysis as specified in ASTM Standard Practice E-693 (Ref.1). The calculated neutron fluence is validated by comparing the calculated (transport analysis) vs. the measured (surveillance capsule dosimetry) exposure levels at the reactor vessel wall, and normalizing the calculated values to account for the bias between the two.

Specifically, the calculated neutron fluence values at SONGS are determined using two transport calculations which represent two fuelloading geometries. The fuelloading geometries in Cycles 1 through 3 consisted of fresh fuelloaded on the periphery of the core. In Cycle 4, SONGS began using a low-leakage core design (burned fuel on the periphery) which will be used for all future cycles. These two calculations are combined to produce average relative neutron distributions throughout the reactor geometry as well as establish relative radial distributions of exposure parameters (fluence and dpa) through the vessel wall integrated over time. The transport calculations use the DOT two-dimensional discrete ordinate code (Ref. 2) and the SAILOR cross-section library (Ref. 3). The neutron fluence values calculated at the p

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surveillance capsule center are compared with the dosimetry results obtained from the surveillance capsule monitors.

The surveillance capsules contain flux monitors, or sensors which are evaluated to yield reaction rates based on the nuclear parameters (for each monitor) and the irradiation history of the vessel. Neutron exposure parameters are then derived from these reaction rates using the FERRET least squares adjustment code (Ref. 4). These exposure parameters represent the neutron fluence at the surveillance capsule center, and establish the means for absolute comparison with the calculated values. The bias resulting from this comparison is then used to normalize the calculated neutron fluence at the vessel inner radius. Fluence values are derived for the current exposure period (ie. the time at which the capsule was removed), as well as key exposure periods relevant to the SONGS P-T curve development, including the end of design life (32 EFPY).

2. Description of Surveillance Prorram A description of the wrveillance program is contained in the SONGS 2 and 3 Updated Final Safety Analysis Report (UFSAR), Section S.3.1 (Ref. S). As described in Section S.3.1, the SONGS 2 and 3 surveillance program adheres to all the requirements of ASTM E-185 73 (Ref. 6), and 10CFR50 Appendix G and H (Ref. 7). The six AU 1

surveillance capsules in each Unit contain beltline plate and weld material. In Unit 2, O

'this consists of plate 6404-2 and a surveillance weld, (weld 9-203, Ref. 8). In Unit 3,

,C plate 6802-1 and a surveillance weld, (weld 9-203, Ref. 8) are used. Material property information for the surveillance program is contained in the SONGS 2 and 3 UFSAR as well as the SONGS 2 and 3 response to Generic Letter 92-01 (Ref. 8).

Evaluation of the surveillance capsules have been performed by Batelle Labs for Unit 2 (Ref. 9), and Westinghoilse for Unit 3 (Ref.10). The fluence values determined by the Westinghouse report were used in the development of the P-T curves in our current Technical Specifications-(TS). As stated above, the fluence values calculated via the transport model were validated by the dosimetry results from the surveillance capsules.

The resulting fluence values are then used to calculate the Adjusted Reference Temperature (ART) for the limiting beltline material. The withdrawal schedule for the surveillance capsules are contained e the UFSAR. Based on updated lead factors obtained from the surveillance capsule evaluations, the UFSAR will be revised as required to reflect a new schedule. Any revision to the withdrawal schedule will be in accordance with ASTM E 185-82 (Ref. 6).

3. Methodolozy for Calculatinz LTOP Setpoints The shutdown cooling (SDC) suction relief valve was determined to be a sufficient relieving device with the necessary capacity to provide adequate overpressure protection at low RCS temperatures. An analysis was performed by Combustion Engineering O

(Ref.11) to address the' initiating overpressure events and the effectiveness of the SDC

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suction line relief valve (LTOP) to terminate the limiting events. The initiating events that were tvaluated are the following: Safety injection actuation; High pressure safety injection pump actuation; Charging / letdown imbalance (Letdown isolated); One RCP start with positive steam generator to reactor vessel temperature differential; and actuation of pressurizer heaters. For conservatism, the analysis considered the RCS to be water solid, RCS pressure boundaries were considered rigid (no expansion),

negligible heat absorption by the RCS metal mass, no ambient heat loss, letdown flow isolated, and all pumps attain full speed instantaneously. The analysis found that the limiting events were the inadvertent safety injection actuation, and the RCP start transient. The analysis determined that the SDC relief valve setpoint specified in the Technical Specifications sufficiently mitigates the limiting overpressure events, and precludes fracture of the reactor vessel material.

As part of the P-T calculation update, the temperature at which the LTOP valve is enabled during heatup and cooldown is determined. The temperature below which the LTOP valve shall be operable is the greater of a coolant temperature of 200 F, or a coolant temperature corresponding to a metal temperature of RTmyr+90 F at the beltline location (per Ref.12). This temperature is then adjusted for instrument uncertainties and the difference in temperature between the coolant and the crack location. The LTOP enable temperature is calculated for both beatup and cooldown 1

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conditions at the 1/4 and 3/4 thickness locations. The required temperature below Which the LTOP system must be operable, corresponds to the most limiting temperature h(N calculated for these four conditions. It should be noted that Code Case N-514 (Ref.13) permits the alternative calculation, RTum+50 F, when determining the LTOP enable temperature. Should Code Case N 514 be approved and adopted by the NRC, SCE may apply the procedures therein for calculating the LTOP enable temperature.

4. Methodolozy for Calculating the Adiusted Reference Temperature (ART)

The current methodology for calculating the ART follow the guidelines of Regulatory Position 1.1 of RG 1.99 Rev. 2 (Ref.14) explicitly. The methodology described by Regulatory Position 2.1 will be used when the SONGS' reactors have two credible surveillance data sets available for evaluation. The inputs to the ART calculation for each vessel are determined from the following: (1) Initial RTum vahi.es come from I

Reference 8; (2) Neutron fluence comes from surveillance capsule testing (currently Ref.

10; will use the most current testing results in the future); (3) Copper and nickel l

contents come from Reference 8; and, (4) The " Margin" is calculated in accordance with Reference 14. Current results of the ART calculation are contained in Reference 15.

5. Methodolozy for Constructing the P-T Curves Usine Fracture Mechanics

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The methodology used to develop the P-T curves (Ref.16) consists of the following steps:

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For any given case (heatup or cooldown at a certain rate), the reactor fluid temperature and the temperature distribution in the vessel wall are known functions of time. For a given fluid temperature, the thermal stresses and the corresponding stress intensity factors (K ) are calculated g

for a distance of 1/4 wall thickness from the inside surface (Ref.17 and 18).

The reference stress intensity factor (K g) is calculated in accordance with i

Reference 19 using the given metal temperature and the ART.

is divided by a safety factor of 2 to obtain K p, The difference K cKg i

i which is used to calculate the internal pressure corresponding to the fluid temperature.

For cooldown, the above pressure is compared with the allowable, quasi steady state pressure for the same fluid temperature, and the lower of the two is used.

For heatup, tlie abo'ce pressures are compared with the allowable pressure for a crack at the 3/4 thickness location, and the lowest of three is used.

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The above procedure is repeated for various temperatures to construct the P-T curve for the particular case.

_. s P-T curves are developed for various heatup and cooldown cases.

6. Apolication of the Minimum Temperature Reauirements of Aeoendix G.10CFR50 The allowable pressures add temperatures determined from the fracture mechanics evaluation (Section 5 above) represent actual values the vessel experiences in the beltline regionc In order to provide operating curves that are useful to the operators in the control room, the pressures and temperatures must be corrected to account for the static head, dynamic head,' instrument uncertainties, and cladding (Ref. 20).

Reference 12 is used to determine the LTOP enable temperature. Reactor vessel flange limits are provided by Reference 21. In the current P-T calculation, flange limits are provided for the heatup curves only, since during cooldown, the beltline region is controlling. For normal operation and inservice testing, flange limics are determined in accordance with Reference 7, using a limiting RT m. value specified by Reference 21.

N Hydrotest temperature and minimum boltup temperatures are determined.in accordance with Reference 19.

7. Use of Surveillance Capsule Data in the ART Calculation Q,m As stated above, the ART is current.y calculated in accordance with Regulatory Position 1

1.1. When two or more data sets become available for the reactor in question, the methodology for determining the ART will follow Regulatory Position 2.1 of RG 1.99 Rev. 2 (Ref.14). If the surveillance data yields a higher ART value than that given by the calculational method (Regulatory Position 1.1), then the surveillance data will be used. If the surveillance data yields a lower value, either may be used.

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. REFERENCES g-x

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ASTM E-693, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section

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12, American Society for Testing and Materials, Philadelphia, PA,' 1984.

i 2.

R.G. Soltesz, R.K. Disney, J. Jedruch, and S.L Ziegler, " Nuclear Rocket Shielding Methods, Modification, Updating and Input Data Preparation. Vol. 5--Two-Dimensional Discrete Ordinates Transport Technique", WANL-PR(LL)-034, Vol. 5, August 1970.

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3.

"ORNL RSCI Data Library Collection DLC-76 SAILOR Coupled Self-Shielded,47-Neutron, 20 Gamma-ray, P3, Cross Section Library for Light Water Reactors".

4.

F.A. Schmittroth, " FERRET Data Analysis Core", HEDL-TME 79-40, Hanford Engineering Development Laboratory, Richland, WA, September,1979.

)

5.

San Onofre Units 2 and 3, UFSAR Section 5.3.1, " Reactor Vessel Materials",

Volume 12.

6.

ASTM E-185, " Standard Practice for Light-Water. Cooled Nuclear Power Reactor

. Vessels, E706 (IF)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA,1984.

.j 7.

Code of Federal Regulations,10CFR50, Appendix G, " Fracture Toughness Requirements", and Appendix H, " Reactor Vessel Material Surveillance Program Requirements", U.S. Nuclear Regulatory Commission, Washington, D.C.

8.

Letter from W.C. Marsh, Manager Nuclear Regulatory Affairs, Southern California

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Edison to U.S. Nuclear Regulatory' Commission, " Revision to Supplemental Response to Generic Letter 92-01, Revision 1, San Onofre Nuclear Gericating -

Station Units 2 and 3, dated June 22,1994, 9.

Examination, Testing, and Evaluation of Irradiated Pressure Vessel Surveillance j

Specimens from the San Onofre Nuclear Generating Station Unit 2, December 1988.

1 10.

Analysis of the Southern California Edison Company San Onofre Unit 3 React 6r

- Vessel Surveillance Capsule From the 97 Location, WCAP 12920, Revision 2, May 1994.

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11.

" Specific Plant Report on Low Temperature Reactor Coolant System Overpressure

~ Protection for San Onofre Units 2 and 3", Prepared by Combustion Engineering, December 15,1977.

e NUREG 0800 Revisi. n 2, November 1988, Addendum: Branch Technical Position o

12. _-

RSB 5 2, " Overpressure Protection of Pressurized Water Reactors While Operating at Low Temperatures".'

13.

Code Case N 514, " Low Pressure Overpressure Protection,Section XI, Division 1",

Approved by ASME' Code Committee on February 12,1992.

Regulatory Guide 5.'99, Revision 2, " Radiation Embrittlement of Reactor Vessel 14.

Materials", U.S. Nuclear Regulatory Commission, May 1988.

15. -

SCE Calculation N 0220-020, " SONGS 2/3 Reactor Vessel Adjusted Reference Temperature".

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16.

SCE Calculations M 0011-062 and M-0011-064," SONGS 2 and 3 P-T Limits".

17.

" Development and Usage of P-T Calculator - A PC Based Computer Program for Constructing P-T Limit Curves", A.Y. Kuo and P.C. Riccardella, Proceedings of the ASME 1990 PVP Conference.

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" Stress Intensity Factor Influence Coefficients for Internal and External Surface V

18.

Cracks in Cylindrical Vessels", I.S. Raju and J.C. Newman, Jr., Aspects of Fracture Mechanics in Pressure Vessels and Piping, ed. S.S. Palusamy, and S.G. Sampath, PVP Vol. 85, ASME.

19.

ASME Boiler and Pressure Vessel Code,Section XI, Appendix G,1989 Edition.

20.

SCE Calculation M-0011-063, " Revised P-T Curves for SONGS 2 and 3".

21.

Letter from C.D. Stewart (ABB-Combustion Engineering) to M. Ogawa (SCE),

" Design Basis Reactor Vessel Flange P T Limits", dated January 19,1993.

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Minimum Requirements to be Included in PTLR

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N Provisions for Methodology From Administrative Controls Minimum Requirements to be Information to be in the STS Included in Methodology Information to be Provided in a Provided in the Main

" Developmental 1

Body of the PTLR Resources" Appendix to (Operator User Section) the PILR 3

1.

De methodology shall describe Describe the E.rspeit calculation Primde values of neutron how the neutron fluence is methods including computer codes fluences that are used in the calculated (reference new and formulas used to calculate ART calculation i

Regulatory Guide when issued).

neutron fluence. Provide references.

1. Provide the surveillance 2.

He Reactor Vessel Material Briefly describe the surveillance Surveillance progiein shall program. Identify the licensee report capsule withdrr*ral schedule, i

comply with App. H to 10CFR50.

by title and number thst contains the or reference by title and number the documents where

• Ihe reactor vessel material Reactor Vessel Surveillance program the schedule is located.

i irradiation surveillance specimen and surveillance capsule reports.

2. Reference the surveillance removal schedule shall be Reference App. H to 10CFR50.

capsule reports by title and provided, along with how the number if ARTS are calculated i

specimen examinatior.s shall be using surveillance data.

used to update the PTLR curves-

3. ' Normal and low temperature Describe how the LTOP setpoints are Provide setpoint curves, or overpressure protection (LTOP) calculated by applying setpoint values.

l system lift settmg limits for the system / thermal hydraulics and power operated relief valves fracture mechanics. Reference SRP l

(PORVs) developed using NRC 5.2.2.

l approved methodologies may be included in the PTLR.

4.

The adjusted reference Describe the method for calculating Identify both the limiting ART values and limiting matenals i

temperature (ART) for each the ART using RG 1.99, Rev. 2.

reactor vessel beltline material at the %'

11 %T locations i

(T=vess Itline thickness).

shall be calculated, accounting for radiation embrittlement, in accordance with Regulatory Guide (RG) 1.99, Rev. 2.

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5. 'the linutmg ART shall be Describe the application of fracture Provide P-T curves for

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incorporated into the calculation mechanics in constructing the P-T heatup, cooldown, criticality, of the pressure and temperature curves based on App. G to ASME and hydrostatic and leak limit curves in accordance with Code,Section XI, and SRP 53.2.

tests.

NUREG.0000 SRP 53.2, l

Pressure-Temperature fimi s.

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6. 'the mirumum temperature Describe how the minimum Identify minimum requirements of App. G to temperature requirements in App. G temperatures on P.T curves 10CFR50 shall be incorporated to 10CFR50 are applied to the P.T such as minimum boltup temperature and hydrotest into the pressure and curves.

temperature limit curves.

temperature.

7. - Licensees who have removed Describe how the data from multiple Provide' supplemental data and two or more capsules should surveillance capsules are used in the calculations of the chemisay factor in the FILR if the compare for each surveillance ART calculation.

material the measured increase survedlance data are used in l

in reference temperature Describe the licensee procedure if the ART calculation. -

(RTNor) to the predicted measured value exceeds predicted Evaluate the surveillance data increase in RTwar; h the value.

to determine if it meets the predicted increase in RTNm.is credibility criteria in RG 1.99,.

based on the mean shift in When Other Plant Data is Used Rev. 2. Provule results.

Nor P us the 2 standard

1. Identify the source (s) of data l

RT deviation value (2a) specified in when other plant data is used.

RG 1.99, Rev. 2. If the measured value exceeds the predicted

2. Identify by title and number the value (increase in RT ar + 2a), _

Safety Evaluation Report that w

the licensee should provide a approved the use of data for the supplement to the FILR to plant.

demonstrate how the results OR affect the approved

3. Compare licensee data with other methodology.

plant data for both the radiation environments (eg. neutron spectrum, irradiation temperature) and the '

surveillance test results.

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