ML18113A701

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Forwards Proposed Changes in Form of an Amend,To Tech Specs of Oper Lics DPR-32 & DPR-37.Request Does Not Involve Any Unreviewed Safety Question.Check for Payment of Amend Fees Encl
ML18113A701
Person / Time
Site: Surry  
Issue date: 12/15/1978
From: Stallings C
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To: Harold Denton, Schwencer A
Office of Nuclear Reactor Regulation
References
NUDOCS 7812190090
Download: ML18113A701 (18)
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Text

{{#Wiki_filter:/ VIRGINIA ELECTRIC AND POWER COMPA.1"" RICHMOND, VIRGINIA 23209 December 15, 1978 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn: Mr. A. Schwencer, Chief Operating Reactors Branch No. Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Denton:

AMENDMENT TO OPERATING LICENSES Seri a 1 No. 702 LQA/ESG:esh Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 SURRY POWER STATION UNIT NOS. 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGE NO. 74 Pursuant to 10 CFR 50.90, the Virginia Electric and Power Company hereby requests an amendment, in the form of changes to the Technical Speci-fications, to Operating Licenses DPR-32 and DPR-37 for the Surry Power Sta-tion, Unit Nos. 1 and 2. The proposed changes are enclosed and have been designated as Change No. 74. By our letter Serial No. 081, dated February 15, 1978, we informed the Staff that our NSSS Supplier, Westinghouse Electric Corporation, had con-firmed that the material in the reactor vessel surveillance capsules was not spawned from the material contained in the vessels themselves. Consequently, the use of material fracture toughness data from these capsules is not con-sidered valid in evaluating radiation damage to the vessels, and the only appropriate means of determining this damage is by measurement of flux or fluence. It is also evident that present Technical Specification Figure 3.1-1, spe~ifying heatup and cooldown 1 imitations, incorporated a shift of 2B0°F in 1975, based on invalid surveillance specimen data, and is consequently extremely conservative. As discussed in Attachment 1, it has been demonstrated that the ap-plicability of present TS Figure 3.1-1 can be extended to 32 EFPY. The revised figure is included in Attachment 2. Also included are revisions to Specifica-tion 3. 18 which provide new procedures for updating Figure 3.1-1, if necessary, based on future fluence calculations on surveillance samples. Table 4.2-1 (Item 7.1) has been revised to specify dosimetry as the examination method for reactor vessel irradiation surveillance. The enclosed changes have been reviewed and approved by the Station Nuclear Safety and Operating Committee and the System Nuclear Safety and Opera-ting Committee. It has been determined that this request does not involve an unreviewed safety question. 7S1219 0-010

e e VIRGINIA ELECTRIC AND POWER COMPANY TO. Mr. Harold R. Denton, Director We have evaluated this request in accordance with the criteria specified in 10 CFR 170.22. Since it has been demonstrated that the 2 present limitations are based on extremely conservative data, and the speci-fication provides for updating the limitations should future data not be bounded by the present curves, the staff should be able to determine that this request does not involve a signi~icant hazards consideration. Accordingly, this request has been determined to be Class I II for Untt 1. The duplicate revision for Unit 2 has been designated Class I. A check in the amount of $4,400.00 is attached in payment of the amendment fees. Attachments: Very truly yours, ~m-~~ C.M. Stallings Vice President-Power Supply and Production Operations

1. Safety Evaluation and Description of Change
2.

Change No. 74

3.

Check in amount of $4,400.00 cc: Mr. James P. O'Reilly, Director Office of Inspection and Enforcement Reg ion 11

[_ .e COMMONWEALTH OF VIRGINIA CITY OF RICHMOND ) ) s. s. ) Before me, a Notary Publi~, in and for the City and Common-wealth aforesaid, today personally appeared C. M. Stallings, who being duly sworn, made oath and said (1) that he is Vice President-Power Supply and Production Operations, of the Virginia Electric and Power Company, (2) that he is duly authorized to execute and file the fore-going Amendment in behalf of that Company, and (3) that the statements in the Amendment are true to the best of his knowledge and belief°. Given under my hand and notarial seal this )._5r), day of Pr:cen:;hec , ~* My Commission expires tknv<1ry ~ 12$1 (SEAL)

'. ---~ --- --. I 'I ATTACHMENT 1 SAFETY EVALUATION Proposed Changes to Technical Specification 3.lB The proposed change to Technical Specification 3.lB is based on the desire to avoid further revisions to the document's text and figures. Analysis of the irradiation samples removed from the Unit One vessel during the first refueling revealed. dharP,y values inconsistent with predictions. As per our letter to Mr. Robert W. Reid, NRC Chief Operating Reactors Branch 4, of February 15, 1978, Serial No. 081,,it was confirmed by our NSSS supplier, Westinghouse, that the material contained in the surveillance capsules was not spawned from the material contained *in our reactor-vessel. The basic consequence of this new information is that the data regarding - material fracture toughness obtained from surveillance capsules are ~ot applica-ble to the reactor vessel from which it was removed.* Therefore, the use of charp:r ~'.* '* data obtained from surveillance capsule samples is not considered iialid in e~l-uating radiation damage to the reactor vessel in regard to meeting the require-.. ments of Appendix G of 10CFR50 for continued unit operation. It is now evident that the present heatup and cooldown curve, T.S. Figure 3.1-1, incorporated a -- shift of 280°F in 1975 based on invalid surveillance specimen data and consequently is overly c:onservativeo

  • Since the_ station has used,-. and is familiar withs, the current heatup and cooldown curve without operational difficulties, the validity of T.S. Figure 3.1-1 will be demonstrated to 32 EFPY rather than liberalizing the heatup and cooldown curve based on a revised T.S. Figure 3.1-1.

Presently, due to our inability to conform to 10CFRSO Appendix G, as stated \\ above, the only applicable means of determining radiation d~age to the reactor vessel, based on the surveillance capsule data, is by flux or fluence. The fluence data obtained from the initial specimens from.each unit coincides.with the Westinghouse prediction of fluence versus EFPY. The second and most recent sample was from :Uttlt 1 and ~as within the accuracy of the analysis*., The erro~ in the

    • --: ~

~. ~-- "\\".4* :* *.... . ~-....... e 2 - accuracy was associated with the uncertainty of the lead factor assigned by Westinghouse. The flux will be determined by dosimetry as the capsules con-tinue to be extracted in accordance with the Technical Specifications. However, charpy V-notch testing will not be performed, on the surveillance capsule samples. T.S. Figure 3.1-2 will be modified to omit Curve 1 since it is based on nonrepresentative ma~erial. The original llRTNDT versus fluence T.S. Curves. will be used. The individual curves will show base metal and weld metal copper content

  • Due to the_ previous 280.°F shift, the pres~nt T.S. Figures 3.1-1 is calcu-lated to.be valid beyond 32 EFPY.

T.S. Figure 3.1-3 will be added to show the relationship between operating time.and fluence at the 1/4 thickness of the vessel as described in the FSAR, Section 4. As the note on T.S. 3.1-3 infers, the.slope of this curve was established from a 280°F shift for 32 EFPY and is more conservative than any existing data. To prove the conservatism, we will look at three cases: (1) the slope at the 1/4 T for.32 EFPY, using the predicted data f:r9mWestinghouse, is 0.9375 x 1P18 n

(2) the slope at the 1/ 4 T for 32 EFPY based on the recent cm2.-EFFY...

surveillance data* and most conservative, is 1.201 x 1018 n

and

_cm2-EFPY (3) the slope at 1/4 _T for a 280°F shift (current shift in Surry's heatup and cooldown curve) for 32 EFPY is 1.203 x 1018 n Thus, Figure 3.1-3 is more conservative than all the applicable data

  • When, according to Technical Specification, another surveillance sample

. or samples are withdrawn and the fluence calculated, the data point s~ould be compared to the 1/4 T line of T.S. Figure 3.1-3. If this data point is above the line, a new line shall be constructed through the origin such that it is above all the applicable data.points. Once Figure 3.1-3 is revised, T.S.

3 - Figure 3 *. 1-1 must be updated, either by a temperature shift, or by revising the applicable period (EFPY) to match the new transition temperature from T.S. Figure 3.1-2. l

ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGE 74

i I .. --- ~** .~, e TS 3.1-6 B. HEATUP AND COOLDOWN Specification

1.

Unit 1 and Unit 2 reactor coolant temperature and pressure and the system heatup and cooldown (with the exception of the pressurizer) shall be limited in accordance with TS Figure 3.1-1. Heatup: Figure 3.1-1 may be used for heatup rates of up to 50°F/hr. below an indicated temperature of 440°F and l00°F/hr. above 440°F. 'i Cooldown: Allowable combinations of pressure and temperature for.specific cooldown rates are below and to the right of the limit line as shown in TS Figure 3.1-1. This rate shall not exceed 50°F/hr. for temperatures at or below an indicated temperature of 440°F. For temperatures.above an indicated temperature 440°F, the rate shall riot exceed 100°F/hr. Core Operation: During operation where the reactor core is in a critical condition, except for low level physics tests, vessel me~al and fluid tempera~ ture shall be maintained above*the reactor core criticality limits specified in Figure 3.1-1.

2.

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 70°F. I . /

  • 7; e

e TS 3.1-7

3.

The pressurizer heatup and cooldown rates shall not exceed l00°F/hr. and 200°F/hr., respectively. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320°F.

4.

TS Figure 3.1-1 shall be updated periodically. in',accordance with the following procedures, before the calculatedmaximum exposure of the vessel exceeds the exposure for which TS Figure 3.1-1 applies. The curve based on 0.25% Cu weld in TS Figure 3.1-2 shall be used to predict the increase in transition tempera-ture based on.integrated power.

a.

If measurements on the most recently examined irradiation specimen show that its data point is above the 1/4T (thick~ ness) line of T.S. Figure 3.1-3 then a new line shall be con-structed through the origin such that it is above all the applicable data points. Once T.S. Figure 3.1-3 is revised, T.S. Figure 3.1-1 must be updated, either by a temperature shift, as by T.S. 3.l.B.4c below, or by revising the applica-ble period (EFPY) to match the new transition temperature from TS Figure 3.1-2.

b.

At or before the erid of the integrated power period for which TS Figure 3.1-1 applies, the limit lines on the figure shall be updated for a new integrated power period as follows. The total integrated reactor thermal power from startup to the end of the new period shall be converted to an eq~ivalent integrated neutron exposure. The predicted increase in transition temperature at the end of the new period shall then be obtained from TS Figure 3.1-2. /

  • Basis

-* e TS 3.1-8

c.

The limit lines in TS Figure 3.1-1 shall be moved parallel to the temperature axis (horizontally) in the direction of increasing temperature a distance equivalent to the transition temperature increase obtained from TS Figure 3.1-2 less the increment used for the end of the present period. All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to reactor system temperature and pressure changes.(1) These cyclic loads are introduced by normal unit load transients, . reactor trips, and startup and shutdown operation. The number of thermal and loading cycles used for design purposes are shown in Section 4.1 of the

  • FSAR.

During unit startup and shutdown, the rates of temperature and pressure are limited. The maximum plant heatup and cooldown rate of 100°F/hr. is consistent with the design number of c;ycles and satisf.ies stress limits for cyclic operation.(2) The allowable pressure vs. temperature is based on a temperature scale relative to the RTNDT" The RTNDT is basically the drop weight NDTT of the material, as determined by ASTM E208. However, to assure that this value is conservative, and to guard against the possibility that material with low upper shelf toughness, or with a -low rate of increase of toughness with temperature, is not properly evaluated, Charpy tests may be performed. If 35 mils lateral

---~--- TS 3.1~9 expansion or SO.ft-lbs is not obtained at NDTT + 60, the RTNDT is shifted upward until these criteria are met. This procedur~ of selecting RTNDT.assures that the KIR curve used to calculate allowable pressures will be conservatively applicable to the material *.. \\, The procedure for dktermining the limiting RTNDT for the Reactor System is as follows:

1.

Determine the highest RTNDT of the material in the core region of the reactor vessel, using original values and adding to this the predicted shift in RTNDT due to radiation during the service period for which this RTNDT applies. This takes into account the copper content of the material.

2.

Examine the data for all other ferritic materials in the reactor system

  • to assure that the RTNDT so selected is the highest in the system.

If drop weight data are not available for all materials, the RTNDT of these shall be estimated in a conservative manner using trend data for the materials concerned.

3.

For succeeding service periods, the same procedure as given in (1) above will be used unless test data from the surveillance.program indicates that this will not be appropriate. In this event, the results of these tests will be used to predict the limiting RTNDT" Test results on material from the Surry Unit 1 reactor vessel is presented in FSAR Table 4.A-1. Using the above procedure,.the highest original RTNDT of the core region plates is +20°F. No drop weight NDTT value is /

\\,I. -- TS 3.1-10 available for the core region weld material but on the basis of actual drop weight data on many similar weld materials, plus the actual Charpy values on this material, the drop weight NDTT is estimated to be o°F. The RTNDT for the first two years of* operation included a conservative estimate of the shift in RTNDT caused by radiation of 100°F. This added to the original RTNDT of QOF assumed for the welds, gave a reference RTNDT of lQOOF to be used for the first two*years of operation, or until tp.e radiation shift was estimated tobe over 100°F. In examining the data for the rest of the material in the vessel; as well as the properties for the other ferritic components of the reactor system, it is certain that all other materials initially had RTNDT values significantly lower than lOOOF. Since the neutron spectra at the samples and vessel inside radius are identical, the measured (RT)NDT shift for a sample can be supplied-with confidence to the adjacent section of reactor vessel for some later stage in plant life. The maximum exposure of the vessel is obtainable from the

  • measured sample data by appropriate application of the calculated azimuthal neutron flux variation.

During cooldown and steady state; the thermaL stress varies from tensile at* f the inner wall to compressive at the outer wall. The internal pressure super-imposes a tensile stress on this thermal stress pattern, increasing the stress at the inside wall and.relieving the stress at the outside wall. Therefore, the limiting stress always appears at th~ inside wall and the limit line has a

e TS 3.1-11 direct dependence on cooldown rate. For heatup, the thennal stress is reversed and the location of the limiting stress is a function of heatup rate. The 1/4T location is considered conservative since the enhanced metallur$ical properties of the surface are not used for the determination of NDTT. The 1/4T location is used for cooldown and steady state and 3/4T location is used for heatup but the 1/4T location is the most restrictive so it will be the controlling curve. In addition, the limiting NDTT for the reactor vessel after operation is based on the NDTT shift due to irradiation. Since the fast neutron dose is highest at the inner surface, usage of the 1/4T NDTT criterion is conservative (FSAR Section 4). The S0°F/hr. heatup and cooldown line on TS Figure 3.1-1 bounds all limit lines for heatup and cooldown rates up to ~0°F/hr. for indicated temperatures at or below 440°F, and 100°F/hr. above 440°F. TS Figure 3.1-1 is based on the Standard Review Plan as modified by measured irradiation sample temperature shifts and appropriate vessel attenuation factors and azimuthal neutron flux variations. TS Figure 3.1-1 defines stress limitations only. For nonnal operation other inherent plant characteristics, e.g., pump parameter and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure ranges. The heatup and cooldown rate of 100°F/hr. for the steam generator is consistent with the.remainder of the Reactor Coolant System, *as discussed in the first paragraph of the basis. The stresses are within acceptable limits for the anticipated usage. Temperature requirements for the steam generator correspond with the measured NDT for the shell. The spray should not be used if the temperature difference between the pressurizer and spray fluid is greater than 32QOF, This limit is imposed to maintain the thennal stresses at the pressurizer spray line nozzle below the design limit.

. ---~.. --*-* *---.. -- **--*---**---* - e TS 3.1-12

References:

(1) E'SAR, Section 4.1.5 (2) ASME Bo:l.ler & Pressure Vessel Code, Section III, N-415 (3) ASME Boiler & Pressure Vessel Code, Section III, proposed non-mandatory Appendix G2000 (4) 10 CFR 50, Appendix A, G, & H (5) Regulatory Guide 1.99, Revision 1, April 1977, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials" (6) USNRC Standard Review Plan, Section 5.3.2, 11/29/75, "Pressure - Temperature Limits" (7) Welding Research Council (WRC) Bulletin 175, "PVRC Recommendation on Toughness Requirements for Ferritic Materials" (8) WCAP - 7924-A, "Basis for Heatup and Cooldown Limit Curves" (9) Surry Reactor Vessel Radiation Surveillance* Program.WCAP 7723-Surry 1 (July, 1972), WCAP 8085-Surry 2 (June, 1973) (10),Battelle Columbus Laboratories Research RepQrts for Surry Pressu~e Vessel Irradiation Capsule Program. (a) Surry 1 examination and analysis of capsule T (June, 1975) (b) Surry 2 examination and analysis of capsule X (Sept., 1975) (11) ASTM: El85-73, E208, & E23 (12) Surry T.S. Change 27 (Proposed Change 35) (13) Vepco letter to Mr. Robert W. Reid, NRC Chief Operating Reactors Branch 4, of February 15, 1978, Serial No. 081

TABLE 4.2-1 SECTION F. VALVE PRESSURE BOUNDARY (Continued) 6.7 Category (Continued) Required Examination Areas Required Examination Methods Extent of Examination Planned During First 5-Year Interval SECTION G. MISCELLANEOUS INSPECTIONS 7.1 7.2 7.3 Mater::Lals Irradiation Capsule. 1 shall be removed and examined at the first region replacement. Capsule 2 shall be examined after 5 years. Low Head SIS Visual

i'::l.p::l.ng Located (See Relllarks) in Valve Pit Low Pressure Turbine Rotor Visual and

.100% of blades magnetic particle or dye penetrant Tentative Inspec-tion During

  • 10-Year Interval Capsules shall be removed and examined after 10 years Not Applicable Not Applicable Remarks The support settings of constant and variable spring-type e:

hangers, snubbers and shock -absorbers would be inspected to verify proper distribution of design loads among the associated support components. Capsule 4 shall.be removed and examined after 20 years. Capsules 5~8 are ext;a capsules for complementary or duplicate testing. This pipe shall be visually inspected at each refueling shutdown.

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I I I i I TT 3/4T=0.155(0T) innervesselwall:1 1 1 1 111 1 1111 1,,--;-- 1 ' ( lid t f 11 ) 'I I I I I I I I I I I I I I I I I ~.-*'/' 4 ~ va O u . vessel life l I I I '

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3. 85x10 19N/ cm2

~ = 32 EFPY = 1.203xl018N/(cm2-EFPY)}}

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