ML20034B936

Proposed Tech Specs,Revising Pages 3/4 4-24,3/4 4-25, B 3/4 4-6 & B 3/4 4-9 Re Heatup & Cooldown Curves

ML20034B936
Person / Time
Site:	Beaver Valley, Trojan
Issue date:	04/19/1990
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20034B933	List: ML20034B932 ML20034B934 ML20034B936 ML20034B937
References
NUDOCS 9005010187
Download: ML20034B936 (13)

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{{#Wiki_filter:.... ATTACHMENT A Revise the Beaver Valley Unit No. 1 Technical Specifications as follows Remove Paaes Insert Paaes 3/4 4-24 3/4_4-24 3/4 4 3/4 4-25 B 3/4 4-6 .B 3/4 4-6 B 3/4 4-9 B 3/4 4-9 'I i L I i l l ) i I I . 4 9005010187 900419 PDR ADOCK 05000334 P PDC

4 MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERNEDIATE SHELL PLATE B6607-2 RT AFTER 9.5 EFPY: 1/47, 202'F NDT 3/4T,176'F 'l 1 CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE 9.5 EFPY. 2500 u

nx 1

TI = a I Leak Test r 7 - r 2250 Limit i [. [ 7-1 J J 1. I 2000 f t n a a W 1750 kM [' { l Heatup Rates ~ Ug To f. [ ~ 3 1500 60,F/Hr _u r t / / i i a e I w S S i-r 1250 I1000 4_. Unacceptsele / j p Operation +_ f F =. r

1 1

8 ~4 / Criticality I - Limit Sased - /- - on Inservie; 3* 750 "~ Hydrostatic _ ~ ~ e f Test Temp. - .R t (329'F)for- '1 r f the service : $00 M - Period Up TE 9.5 EFPY = Acceptable I 250 E Operation i ~ { < 1 1 1 e, o ,0 50 100 1%0 200 250 300 350 400 4he 500 4HDIC&iED igNPERafMRE (988.F) Beaver Valley Unit 1 Reactor Coolant Systen Heatup FIGURE 3.4-2 Limitations Applicable for the First 9.5 EFPY 3/4 4-24 0k0 POS&b ~

4 MATERIAL PROPERTY BASIS CONTROLLING MATERIAL:- INTERMEDIATE SHELL PLATE B 6607-2 RT AFTER 9.5 EFPY: 1/4T, 202*F NDT 3/4T 176'F CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100

• F/HR FOR THE VP TO 9.5 EFPY, 2500_u yt)r -

-l r 3 1 3 i yy 1 2250 t i a 2000 [ J + 1750 +- r f f i I .r 9 1500 ,A unaccepu u. a 1 l i g, Ope ra t ion [ 1250 l1000 - / Operation f L Acceptable I j I I l l ~ [ y / l l l g ygg , l a 2 Cooloown ry;y 9 z Rates

1 we 4,

'F/Mr --%w 500 $3 P[/ E s ..wp" 1 250 -J I I I 1 ' 'l i 04 50 100 150 <t00 200 See ' ano. 400 400 300 teeDICAft0 TEmptaatuRE (8ES.r) FIGURE 3.4-3 Beaver Valley Unit 1 Reactor Coolant System Cooldown Limitations Applicable for the First 9.5 EFPY 3/4 4-25 (next page is 3/4 4-27) fAOPOSE'b Taf

REACTOR COOLANT SYSTEM BASES The heatup analysis also covers the determination of l pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the veasel are tensile t I and are dependent on both the rate of heatup and tho time along the heatup ramp; therefore, a lower-bound curve similar to that described for the heatup of the inner wall cannot be defined.. Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling

location, each heatup rate of interest must be analyzed on an individual basis.

The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60'F per hour. The cooldown limit

curves, Figure 3.4-3, are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce t

l tensile stresses while producing compressive stresses at the outside-l wall. The heatup and cooldown curves were prepared based upon the l most limiting value of the predicted adjusted reference temperature at the time in life indicated on the respective curves. l t The reactor vessel materials have been tested to determine their initial RTNTD; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E > 1 Mev). irradiation will cause an increase in the RTNTn. Therefore, an adjusted reference temperature, based upon the fluence and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2. The heatup and cooldown limit curves, Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RTNTD' The heatup and cooldown curves have been developed in accordance I with the methodology provided in Regulatory Guide 1.99 Revision 2 and i no longer contain the additional margin of 10'F and 60 psig for instrument error previously incorporated in these curves. BEAVER VALLEY - UNIT 1 B 3/4 4-6 PROPOSED

t REACTOR C001 ANT SYSTEM BASES The second portion of the heatup analysis concerns the-calculation of pressure-temperature limitations of the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at 1 the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the j thermal stresses at the outside are tensile and increase with p increasing heatup

rates, each heatup rate must be analyzed on an l

individual basis. Following the generation of pressure-temperature curves for both i the steady-state and finite heatup rate situations, the final limit curves are produced as follows: A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup' rate data. At any given. temperature, the allowable pressure is taken-to be the lesser of the three values taken from the curves-under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it la possible for - conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches'from the inside to the outside and the pressure limit must at all times be based on analysis-of the most critical criterion. l The actual shift in HDTT. of the vessel material will be established periodically during operation by removing and evaluating, i in accordance with 10 CFR 50 Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and . i B 3/4 4-9 PROPOSED 4 )

ATTACHMENT B Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change-No. 177 l REVISION OF TECHNICAL SPECIFICATION 3.4.9.1- .i HEATUP AND COOLDOWN CURVES AND BASES A. DESCRIPTION OF AMENDMENT REQUEST i The proposed amendment would modify _the heatup and cooldownLlimit

curves, Figures 3.4-2 and 3.4-3, to reflect the results of a-Westinghouse analysis (Attachment.D). performed to incorporate the requirements of Regulatory Guide 1.99, Revision 2.

'The analysis results indicate the current heatup and cooldown curves are . conservative with respect to implementation of the Regulatory Guide. This conservatism,. along with the accepted industry practice of deleting the margin; for. instrument error is incorporated into the new heatup and cooldown curves to provide margin for additional plant operating' flexibility. B. BACKGROUND Heatup and cooldown limit curves;are calculated'using the most limiting value of RTNDT (reference nil-ductility temperature) for -the reactor vessel. The most limiting RTHDT.of the material _in the core region of the reactor vessel is determined by.using the preservice reactor vessel material fracture toughness properties and estimating the radiation - induced shift.in.RTNDT. The RTNDT is designated as the higher: of either the drop-weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of' impact. energy and 35-mil lateral expansion (normal to the ~ major working-direction) minus 60*F. RTNDT increases as the material 'is exposed to fast-neutron radiation. Therefore, to find the most limiting RTNDT at any time period in the reactor's life, the shift in RTNDT due to the radiation exposure associated with that: time period must be added to the original unirradiated RTNDT. The extent of:the shift in RTNDT is enhanced by certain chemical elements (such as copper 3 and nickel) present in reactor vessel steels._ The:NRC has provided a method for predicting radiation embrittlement in Regulatory Guide 1.99, Revision 2. C. JUSTIFICATION The reactor vessel material surveillance program complies with 10 CFR 50 Appendix G and Appendix H to ensure the reactor vessel has an adequate margin of safety. with' respect to ' material toughness throughout the service life of-the plant. Specifically, the program develops operating limits (RCS heatup and cooldown limit curves) to prevent non-ductile failure. The heatup and cooldown operating curves have been adjusted in accordance with the NRC approved methodology of Regulatory Guide 1.59,- Revision 2 to account for the cumulative effects of radiation on the reactor vessel material properties and to maintain an adequate margin of safety.

q Propo cd Tcchnical Spacification Chango No. 177' Page 2 j D. SAFETY ANALYSIS Adjusted reference temperatures have been calculated using the-material property and neutron fluence data to determine the most limiting reactor vessel materials. Plates B6607-2 and B6903-1 were found to be the most limiting materials:in the-reactor ~; t vessel based on these calculations relative to the generation, of heatup and cooldown curves. The new cooldown limitations, provide significantly more operating flexibility. Deletion' of-the instrument. error produces the most benefit at low temperatures while the majority of the benefit at higher, temperatures results from the use of Regulatory Guide 1.99, Revision 2.:The reference flaw of Appendix G to the ASME code is assumed to~ exist at the inside of the vessel wall for calculating the allowable pressure versus coolant temperature during cooldown. <The controlling. location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the-inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations-are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest. As was done in the cooldown

analysis, allowable pressure-l temperature relationships

_are developed for' steady-state conditions as well as finite.heatup rate conditions assuming the presence of a 1/4 T defect at-the inside of the wall-that-alleviates the tensile stresses produced by internal pressure. i The heatup analysis also concerns the calculation of pressure-temperature limitations assuming a-1/4-T deep outside surface flaw. Following the-generation of pressure-temperature curves for both-the steady-state and finite heatup rates, the final limit curves are produced by constructing a composite curve: based on a point-by-point comparison of.the steady-state and finite i heatup rate data. The composite curve is-necessary to provide i conservative heatup limitations because it is' possible for conditions to exist wherein, over the course of-the heatup ramp, the controlling condition switches from the inside to the

outside, and the pressure limit must at all times be based on the most limiting criteria.

Based on the above considerations, these changes reflect the application of methodologies recognized by the NRC and Industry as providing a sufficient margin of safety.- The-fracture toughness requirements of_10 CFR 50 Appendix G are satisfied and 4 conservative operating restrictions are applied in the proposed heatup and cooldown

curves, therefore, these changes are considered to be safe and will not reduce the safety of the l

plant.

j Propo? d.TCchnic21.Sp;cific tien Changa No. 177 [ _-Page 3 E. NO SIGNIFICANT HAZARDS EVALUATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) as quoted below: The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that a proposed amendment to an operating license for a. facility licensed under paragraph 50.21(b)- or paragraph 50.22 or for a testing facility involves no significant hazards consideration,- if operation of the facility in accordance with the proposed amendment would not: (1) Involve a significant increase in-the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new.or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety. The following evaluation is provided for the no~ significant hazards consideration standards. 1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? The heatup and cooldown curves have been revised to' reflect the results of a Westinghouse analysis performed to incorporate the requirements of Regulatory Guide _l.99, Revision 2.- The analysis results indicate the current heatup j and cooldown curves are conservative with respect to l implementation of the Regulatory Guide. This conservatism, l along with the accepted industry practice of deleting the l margin for instrument

error, is incorporated into the new

( heatup and cooldown curves and provides margin that is i available for additional plant operating flexibility and will act to reduce challenges on the pressurizer power relief valves while running reactor coolant pumps at the current cold overpressure setpoint. The reactor ~ vessel material surveillance program complies with 10 CFR.50, Appendix G and-H to ensure the reactor vessel has an adequate margin of safety with regard to material toughness throughout the service life of the plant. The program develops operating limits to prevent non-ductile failure and the operating limits are adjusted to account for the cumulative effects of radiation on the reactor vessel material properties. The operating limits provided by these new curves were determined in accordance with the methodology set forth in the. regulations to provide an adequate margin of safety to ensure the reactor vessel will withstand the effects of normal cyclic -loads due to temperature and pressure changes as well-as the loads l associated with postulated faulted conditions. Therefore, the l. proposed changes do not involve a significant increase in the I probability or consequences of an accident previously l evaluated.

l l iroposed' Technical sp;cificatien Chang 3 No. 177 i Page 4 4 2. Does the change create the possibility.of a new orfdifferent' kind of accident from any accident previously evaluated? -{ Reactor vessel rupture is not part of the Beaver Valley design basis and is not included in the accident analysis. The new heatup and cooldown curves have been determined in.accordance-with Regulatory Guide.l 99, Revision 2 and contain sufficient margin to. ensure that the probability of a reactor vessel rupture is low enough that it is able to be excluded-from the accident analysis. Changing the heatup and cooldown curves F does not reduce the. reliability of the' reactor vessel or the procedures involved in plant heatup and.cooldown.. Therefore, the proposed-changes do_not create the possibility. of a new or different kind of accident from any accident previously evaluated. i i 3. Does the change involve a significant reduction in a margin of ~! safety? + The revised heatup and cooldown curves were established in- 'I accordance with current regulations and the latest regulatory; . guidance on RTNDT determinations. Because operation will be within these limits, the reactor vessel materials will' behave-in a. nonbrittle manner, thus, maintaining-the original safety design basis. Therefore, the proposed changes do not involve a significant reduction in a~ margin of safety. F. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION-l Based on the above safety

analysis, it. is' concluded that the e

l activities associated with this license amendment request satisfies the no significant hazards consideration standards of.10 CF3 50.92(c)

and, accordingly, a

no significant hazards consideration finding is justified. G. ENVIRONMENTAL EVALUATION The proposed changes have been evaluated and it has been; c determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a-significant increase.in. individual or cumulative occupational radiation exposure. Accordingly, the l proposed changes meet the eligibility criterion'for categorical exclusion set forth in 10 CFR 51. 22 (c) (9). Therefore, pursuant to 10 CFR 51.22 (b), an environmental assessment of the proposed changes is not required.

4 ATTACHMENT C UFSAR~ Changes -Beaver Valley Power Station, Unit No.1 Proposed Technical' Specification Change No. 177 'l 'l t i, ) i l I - i i k

F BVPS-1-UPDATED FSAR Rav. 0 (1/82)- TNS&W I heatup and cooldown curves for the unit are based on tV act measured fracture toughness properties of the' v sel materi determined in accordance with the above-menti d new 1 fracture ughness requirements. Where sufficient ests to. comply with t new requirements for fracture tou ess testing' were not perform conservative estimates of f cture toughness properties are use. Allowable ' pressures a a function of ' the rate of temperature cha and the actual perature relative.to i the vessel RT(NDT) are est ished acco ing to the methods given ~ in the 1980 issue of ASME, S-ion , Appendix G 2000. Curves incorporating allowances for ins ment error in measurement of temperature and pressure re-ven in the Technical Specification. These curves are.basef on temperature scale re tive to the-RT.of the vessel, includ (g appropriate estimates of R DT) caused-by radiation. Pr cted RT (NDT) values are derived using ~ the curve in the echnical Specification (also Figure 4.2-8 nd the fluence 1/4T. corresponding.to the maximum for the s ice period pplicable. Initial RT (NDT) will include an acau RT ) corresponding to that predicted after two integrated' full-wer years of operation. l The results of the radiation surveillance program will be used to l Verify that the RT(NDT) predicted from the curve in the Technical Specifications is appropriate and,to make any changes necessary to correct this curve, r ? The use of an RT(NDT) that. include a ' RT (NDT) to account for radiation effects on the. core region

material, automatically; l

provides additional conservatism for the nonirradiated regions. Therefore, the flanges, nozzles, and ' other regions not affected by radiation will be favored .by additional conservatism approximately equal to the assumed RT(NDT). Details of the radiation effects surveillance program will be based on the evaluation of the test results on the actual vessel material. Changes in fracture toughness of the core ' region plates-or forgings, weld metal, and associated heat treated zone due to radiation damage will be monitored by a surveillance program which conforms with ASTM E-185-79. The evaluation' of the ~ radiation damage in this surveillance program is based on' pre-irradiation and post-irradiation testing by Charpy V-notch, dropweight

test, and tensile specimens and post-irradiation

-i testing of Charpy V-notch, tensile, and wedge opening loading specimens carried out during the lifetime of the reactor vessel. Specimens are irradiated in capsules located near the core midheight and removable from the vessel at specified intervals. 4.3-6 i

t INS &AT l Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel. The most limiting R.TNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture tough-ness ' properties and estimating the radiation-induced ART RT is NDT. WDT designated as the higher of either the drop weight nil-ductility transition j temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F. RT increases as the material is exposed to fast-neutron radiation. NDT Therefore, to find the most limiting RT at any time period in the NDT raaetor's life, ART due to the radiation exposure associated with that NDT time period must be added to the original unirradiated RT The extent of' NDT. the shift in RTNDT is enhanced by certain chemical elements (such'as copper and nickel) present in reactor vessel steels. The Nuclear Regulatory j Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2-(Radiation Embrittlement of Reactor Vessel ] Materials), i l l 1 '.l

k .l ATTACHMENT D. BEAVER VALLEY UNIT l' { Reactor vessel Heatup and Cooldown curves for. Normal Operation b f ,~ { b I l 1 l ~ -}}