ML20059J456

Proposed Tech Specs Re Reactor Vessel pressure-temp Limits

ML20059J456
Person / Time
Site:	Perry
Issue date:	09/14/1990
From:	CENTERIOR ENERGY
To:
Shared Package
ML20059J454	List: ML20059J456 PY-CEI-NRR-1188, Application for Amend to License NPF-58,revising Tech Specs Re Recalculation of Reactor Vessel pressure-temp Limits,Per Generic Ltr 88-11
References
NUDOCS 9009200002
Download: ML20059J456 (35)

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Proposed FNPP Technicali. Specification Changes forRegulatoryGuide1.99Revisidn.2

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i c.- J REACTOR COOLANT SYSTEM SASES l 1

- SPECIFIC ACTIVITY (Continuad)

Closing the main staan line isolation valves prevents the release of activity-l- to the environs should a staan line rupture occur outside containment. The sur-veillance requirements provide adequate assurance that excessive specific activity-levels in the reactor coolant will be detected in sufficient time to take corrective action.- .t 3/4.4.6 PRESSURE /TEMERATURE LIMITS i

All components in the reactor coolant system are designed to withstand

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the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by neraal load transients, reactor trips, 3 and startup and shutdown operations. f used for design purposes.are provided in Section 3.9 of the f94R. DuringThe various s startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with - -

the design assumptions and sausfy the stress limits for cyclic operation [tTSt'fM

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toughness- requirements of 10 CFR 50 Appendix G and ASME Code Section III,l  ;

.Ap x G. The curves are based on the RT ET and stress intensity factor-L informa n for the rear. tor vessel components. Fracture toughness limits'and '

- the basis ' r complianc.e are more fully discussed in Chapter 5 of the FSAR.

l The tea vessel materials have been tested to' determine their initial

! 'RT The res s of these tests are shown in Table B 3/4.4.6-1.' Reactor ET.

operation and resu t fast neutron, E greater than 1 MeV, irradiation will . ~i cause an increase in RT Therefore, an adjusted-reference, temperature ET.

based upon the fluence, p horus content and copper content.of the material' in question, can be predic sing Bases Figure B 3/4.4.6-1 and the recommenda--

tions of Regulatory Guide 1.99, vision'1, " Effects of Residual Elements on Predicted Radiation Damage to Re e Vesse1' Materials." The pressure / tempera-ture limit curve, Figure 3.4.6.1-1, es ' A', B' and C', includes assumed

. shift in RT ET for the end of life fl . 1 The actual shift in RT ET of the vesse terial will be established period-

ically during operation by removing and evalua ,,in accordance with'10 CFR 50, t Appendix H, irradiated flux wires installed near inside wall of the. reactor 1 vessel in the core area. The irradiated flux wires be used with confidence .,

in predicting reactor vessel material transition temper ce shift. The operat-L ing limit curves of Figure 3.4.6.1-1 shall be adjusted, a ired, on the_ basis of the flux wires data and recommendations of Regulatory Gui 1.99, Revision 1.

3 l The pressure-temperature limit lines shown in Figures 3.4. -1, curves  !

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C, and C', and A and A', for reactor criticality and for inservice. ak and

- hydrostatic testing have been provided to assure compliance with the. nieue  ;

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temperature requirements of Appendix G to 10 CFR Part 50 for reactor

! criticality and for inservice leak and hydrostatic testing.

l (A The number of reactor vessel irradiation surveillance capsules and the V frequencies for removing and testing the specimens in these capsules are pro-vided in Table 4.4.5.1.3-1 to assure c 11ance with the requirements of i PERRY - UNIT 1 B 3/4 4-5 j>

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PY-CEI/NRR-1188 L  :

INSERT "A" l' The purpose of this specification'is to establish operating limits that -

provide a vide margin. to brittle failure for major piping and pressure vessel components of the Reactor Coolant Pressure Boundary -(RCPB).1RCPB materials are subject to brittle failure _below their nil-ductility temperature (NDT), at relatively.lov stresses.. Below Ntrr, stresses are carefully limited by:

specifying both allovable pressure and heatup/cooldown rates. Of -the major l components within the RCPB,, the reactor vessel is the component most subject to brittle failure and therefore the component for which~these technical specification limits is most pertinent.

The basis of the pressure and temperature-(P-T) limits _is found in Appendix G to 10 CFR 50. . Appendix G requires that the limits be based on specific' d fracture toughness-requirements for RCPB~ materials such that'an adequate l margin to brittle failure vill be provided during operational occurrences. 10 CFR 50 Appendix G mandatessthe use of ASME Section III , Appendix G. The  !

concern addressed by Appendix.G is that undetected flaws could exist in the RCPB components, which under. certain reactor coolant: system P-T combinations could'eause stress concentrations at' flav locations resulting in flav growth

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to failure before the ultimate strength of the material is attained.

Flav growth is resisted by the material 1 toughness, a property that-increases with temperature. Furthermore, the material. toughness. is affected by neutron fluence which causes the steel-ductility to decrease. - The effect of. fluence is cumulative, and ductility steadily decreases with exposure time. Toughness is also dependent on the chemistry of the-base metal and its impurities. '

Table B.3/4 4.6-1 provides initial and' predicted.end-of-life reactor vessel toughness data. Fracture-toughness' limits and the basis for compliance are more fully discussed in Chapter 5-of the,USAR. ,

Although any region within the pressure boundary is subject to non-ductile I failurs, the regions that provide the most restrictive limits' are the vessel, closure head flange, the feedvater nozzles,;the control rod drive nozzles, and the vessel beltline. At any specific pressure,' temperature,.and-temperature i rate of change, one location within the geometry of,the reactor vessel vill dictate the most restrictive limit. Across the entire pressure / temperature span of the limit curves,'the curves are composites of the most restrictive regions.

The reactor vessel beltline is the only RCPB material which experiences-

significant neutron fluence. Since fluence causes an increase in NDT, the vessel beltline becomes most limiting (requires the highest _ temperature when pressurized) after 4 years of operation. The value of NDT used to set operating limits is called t!e nil-ductility reference temperature-(RT which incorporates safety margin for variation in material properties NaN) l measurement. ,

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The actual shift in RT of the vessel material ~ vill be established periodicallyduringop!PItionbyr.movingandevaluatingirradiat.dreactor L vessel ~ material specimens installed near the inside vall of the reactor vessel =

L in the core area, _ in accordance with- ASTM E185 and 10 CFR50,: Appendix H. - The- J operating limit curves (A, B, C) are adjusted. as. required, on the basis of

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the specimen data and-recommendations of Regulatory Guide 1.99-Revision 2,

" Radiation Embrittlement of Reactor Vessel Materials" to assure a ductile pressure boundary.

Operating limits for pressure and: temperature are provided.for three I categories of operations- (a) hydrostatic pressure tests and' leak tests, )

referred to as Figure.3.4.6.1-1 Curve A;:(b) non-nuclear heatup/cooldown and '

low-power Physics Tests, referred to as: Curve B;-and (c) core criticals operation other than during Physics Tests, referred ~to as Curve C.' The-beltline region minimum temperature limits are adjusted to account for vessel ,

-irradiation in curves A',B',C'. 'I l

.These curves have been developed forc a. large number of operating cycles. and: l provide a conservative margin to non-ductile failure. Although they have been created to provide-limits for normal' operations, they also can be: used as a basis for determining if evaluations are necessaryL for abnormal: transients- J vhich;can:begin from power operation. ASME Section KI Appendix E provides-a recommended methodology.for evaluating-operating events.which:cause an  ;

excursion outside of the normal limits.  !

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Analysis of PNPP Unit 1 Reactor Vessel lt ' Compliance with Regulatory Guide 1.99 Revision 2 h

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FY-CEI/NRR-1188L I

TABLE OF ColffElff5 . *

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1.0 RACRGROUND 1-1 f 2.0 FLUX VIRE TEST 2-1 2.1 Introduction 2-1 i 2.2 Analysis

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2.3 Results 2-3 2.4 conclusions 2-4  !

t 3.0 REV 2 IMPACT EVALUATION 3-1 j 3.1 Chemistry and Initial RT 3'l NDT  !

3.2 Fluence 3-2 i 3.1 Surveillance Test Correction Factor 3-2 3.4 Shift and Adjusted Reference Temperature (ART) .3-3 $

3.5 Results of Impact Evaluation 3-4 i 4.0 PRESSURE-TEMPERATURE CURVES 4-1 4.1 Background 4-1 4.2 Non-Beltline Regions '4-1 4.3 Core Beltline F.gion 4-2 4.4 Closure Flange Region 4-3 '

4.5 Core Critical Operation Requirements of 4 1 10CFR50, Appendix G 5.0 SIGNIFICANT HAZARDS CONSIDERATION 5-1 ENVIRONMENTAL IMPACT CONSIDERATION 5-2 i l

6.0 REFERENCES

6-1 3 i

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Attcchment 2 PY-CgI/NRR-1188L 5- .6 1.0 bACKGR0' AD The pressure-temperature (P-T) curves in the Technical Specifications are established. to the requirements of 10CFR50, Appendix G [ Reference 1] to assure that brit'le fracture of the reactor vessel is prevented. Part of the analysis involved in developing the P-T curves is to account for irradiation embrittlement effects in the core region, or beltline. In the past, Regulatory Guide 1.99, Revision 1 [ Reference 2] has been used to predict the shift in nil-ductility reference temperature (RTNDT) as a function of fluence in the beltline assuming that copper (Cu) and phosphorus (P) were the key i

chemical elements influencing enhrittlement.

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Regulatory Guide 1.99, Revision 2 [ Reference 3) was issued in May 1988.

Revision 2 represents the results of statistical evaluation of commercial reactor surveillance test data accumulated through about,1984. The basic elements of the regulatory guide, a chemistry factor and a fluence factor, both remainad in Revision 2. However, each factor is significantly different.

The chemistry factor (CP) has been changed from an equation based on Cu and P in Rev 1 to tables of CP values based on Cu and Nickel (Ni), with separate tables for plates and for velds. The fluence factor has been modified in Rev 2 to a somewhat more complex form. i l

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l The surveillance program for the Perry Unit 1 vessel included a dosimeter intended for removal after the first fuel cycle. That dosimeter was removed I and tested by GE. This report provides the results of that dosimetry test, documents the impact of implementing Rev 2 with the fluence determined from the dealmetry test, updates the P-T curves based on Rev 2 shif ts, and concludes there are no significant hazards associated with the use of these '

P-T curves in the Technical Specifications. i 1

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PY-CEI/NRR-ll88L i

i 2.0 FLUX VIRE TEST ,

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2.1 INTRODUCTION

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i In February 1989, the Perry Nuclear Power Plant Unit 1 (Perry) completed its  ;

first fuel cycle. During the outage that followed, the flux vire dosimeter l attached to the surveillance capsule at the vessel 3' azimuth was removed.

The dosimeter was shipped to the General Electric Vallecitos Nuclear f

Center (VNC) in Pleasanton, CA for testing. The test results and the

, associated determination of peak flux and fluence are presented in this

! section.

The surveillance program for Perry consists of three surveillance capsules and one separate flux vire dosimeter. Each surveillan:e capsule contains Charpy specimens of the beltline base, veld and heat affected sone (BAZ) materials, s

and a set of flux vires used to determine the fluence experienced by the capsule. The rurveillance capsules are scheduled to be withdrawn periodically during plant life (the current schedule required by ASTM E185-82 is a capsule i at 6 and 15 ef fective full power yes.'s). In addition to the flux vires in the ,

surveillanca capsules, a flux vire dosimeter was attached to the capsule at 3', as shown in Figure 2-1, for removal af ter the first fuel cycle. Since  ;

the vessel fluence is proportional to thermal power produced, the results of l the flux vire dosimeter test are used to provide a calibration point of vessel l fluence versus accumulated thermal power. A linear extrapolation provides an l estimate of the fluence at 32 effective full power years (EFPY). It should be noted that the flux vires that vill be removed later with the surveillance capsules vill.have an irradiation history more typical of normal operation, and vill be useful for re-calibrating the 32 EFPY fluence estimate.

2.2 ANALYSIS l

The determination of the peak 32 EPPY fluence is basically a two-step process.

First, Go flux vires

• are analyzed to determine the flux and fluence at the

, dosimeter location. Then, lead factors are calculated which relate the flux l

magnitude at the dosimeter location to that at the location of peak flux.

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Attcchment 2 PT-CEI/NRR-ll88L The flux vire dosimeter was disassembled at VNC and the iron flux vires were cleaned and weighed. Camma spectrometr/ vas used to determine the rate of '

disintegrations. The daily power history of the first fuel cycle was used, along with cross-section data developed for BVRs to transform the disintegration data into rates of irradiation, or flux (n/cm 2 ,,),

4 The determination of lead factors was done for a generic 238 inch diameter vessel with 748 fuel bundles. This matches the Perry configuration. The lead j factors are essentially geometry dependent. Plant-specific characteristics of the flux are accounted for in the results of the flux vire test. Furthermore, the generic lead factors were calculated assuming an equilibrium fuel cycle, which is representative of a typical normal operation core power distribution.

Therefore, the 2eneric lead factors provide the best available means of predicting peak 32 EFPY fluence from the flux vire data.

Determination of the lead factors for the RPV peak location at the inside vall and 1/4 7 depth was done using a combination of two-dimensional finite I

difference computer analyses. One two-dimensional analysis established the I

relative fluence in the asinuthal direction at the vessel surface and 1/4 T depth. The other two-dimensional analysis was done to determine the core height of the axial flux peak and its relationship to the surveillance j capsule height. The combination of azimuthal and axial distribution results provides the lead factor between the dosimeter location and the peak flux location.

The two-dimensional DOT computer program was used to solve the Boltzman transport equation using the discrete ordinate method on an (R,9) geometry, assuming a fixed source. One quarter core symmetry was used with periodic

( boundary conditions at O' and 90'. Neutron cross sections were determined for l

26 energy groups, with angular scattering approximated by a third-order Legendre expansion. A total of 99 radial intervals and 90 azimuthal intervals i 2-2 I

i AttCchment 2 PY-CEI/NRR-1188L {

were used. The model consists of an inner and outer core region, the shroud,

water regions inside and outside the shroud, the vessel vall, and an air ,

region representing the dryvell. Flux as a function of asinuth was l calculated, establishing the asinuth of the peak flux and its magnitude j relative to the flux at the dosimeter location of 3'.

The other two-dimensional computer code (SN20) was used to calculate flux j distribution for the (R,Z) geometry at the peak asinuth angle. The elevation of the peak flux was determined, as well as its magnitude. relative to the flux t at the dosimeter elevation. This factor is the axial component of the lead l factor. The lead factor between the peak and dosimeter locations was  ;

calculated as the azimuthal component times the axial component.

2.3 RESULTS A summary of the >1 MeV flux and fluence values for the dosimeter are presented in Table 2-1. There is an uncertainty of 3,25% on the >l MeV flux and fluence. Table 2-1 shows the upper bound values with the nominal values.

t The lead factors for the peak location inside surface and 1/4 T depth are presented in Table 2-1 with the dosimeter test results. The lead factors are  !

used to predict the peak fluence according to the following equation:

Peak Fluence - (Dosimeter Flux)*(Full Power Seconds)/ Lead Factor l

The first fuel cycle for Perry consisted of 795 days of operation with an average capacity factor of 0.503. This is equivalent to 399.9 days at full power, or 1.09 EFPY. These values are used to calculate the fluence values at the end of cycle one (EOC1) and at 32 EFPY, as shown in Table 2-1.

, The fluences at the peak location I.D. and 1/4 T are plotted as a function of i EFPY in Figure 2-2.

2-3

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PY-CEI/NRR-1188L 6 j

2.4 CONCLUSION

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The flux vire test results summarized in Table 2-1 show a nominal peak fluence I8 2 on the vessel ID at 32 EPPY of 4.3x10 n/cm . The fluence determined by ,

dosimetry is somewhat lover than the calculated design fluence value of l 6.5x10 18 n/cm . This lower trend is consistent with the results of dosimetry 2

tests at other plants. l The 32 EPPY fluence value determined fron.the flux vire testing results is  ;

used to determine the Rev 2 impact in Section 3 and the pressure-temperature ,

curves in Section 4.

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i FINENCE DETERMINATION PCR TNE PEAK 1hCATION .

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IN TNE PEasy vessel l

I Time at Power: ,

f 80C1 1.09 EFFY = 3.46x10 seconds f 32 EFFY 32 EFFY = 1.01x10' seconds t

Lead Factors:  !

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I.D. 0.40 t 1/4 T 0.58  !

I Dosimeter Flux (m/am .u) 1.7x10' (nominal) 2.1x10' (upper bound)

FIRENCE (n/cm ):

NOMINAL UPPER BOUND 17 EOC1 Peak I.D. 1.5x10 1.tx10 17 32 EFFY Peak I.D. 4.3x10 18 5.4xM I8 I 18 18 32 EFFY Peak 1/4 T 3.0x10 3.7x10 l

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' i Attachment 2  ;

PY-CEI/NRR-llB8L 3.0 REV 2 IMPACT EVALUATION ,

The beltline region in the Perry vessel consists of three shell ring plates and their associated welds. The process followed for each beltline material I is described below.

3.1 CHEMISTRY AND INITIAL RT NDT The chemistry data for the beltline shell plates and veld. filler materials were taken from Table 5.3-1 of the Perry USAR (Reference 4]. The values of i initial RT NDT vere taken from Table 5.3-2 of the Perry USAR [ Reference 4).

These values were based on 50 ft-lb impact energy verification testing, with transverse Charpy specimens used for plate, as required by ASME Code,  ;

paragraph NB-2300.  ;

For beltline materials, the methods of calculating adjusted RT yg in Rev 2 include a Margin term to be added to the calculated value RT NM. The Margin term includes a component for uncertainty in initial RTNDT' 'I. Rev 2 ,

discusses determination of og for two categories of initial RT yg t measured values and generic mean values. To generic mean values, ye is simply the standard deviation calculated for the data set used to compute the mean. For measured values, requirements for determination of e are y somewhat vague.

Rev 2 states, "If a measured value of initial RTyg for the material in question is available, e yis to be estimated from the precision of the test I,

method."" GE's position for RTgg values derived from measured data, as is the case for the Perry beltline materials, is that ey is zero.

t

• In the Rev 2 draft which was circulated after editing to incorporate public comments, the text stated, "e g , the standard deviation for the initial RT NM' may be taken as zero if a measured value of initial RT for the material in NDT question is available."

r I

3-1

,...-,__.-_-.m.. _ . - - ~ - . - ~ _ ~ .

Attachment 2 PY-CEI/NRR-llB8L The Charpy curves fit to surveillance data, which ultimately provided the .

ST yg data for development of Rev 2, vere best-estimate fits. An idealised example is provided as curve #1 in Figure 3-1. However, the ASME Code approach to determining RT NDT is based on the lowest value of three specimen's exceeding the required limits of impact energy and lateral expansion. A visualisation of a Charpy curve drawn on the Code RT 8PProach is shown as NDT curve #2 in Figure 3-1. In comparing curves el and #2, it is clear that curve #2, which is based on the lowest value rather than the mean value, i provides a conservative estimate of initial RT Therefore, the ASME Code NDT.

method of determining RT O' F NDT from measured data is conservative, and cy is appropriate.

3.2 FLUENCE The best-estimate 32 EFPY fluence calculated from the dosimetry test results in Section 2 was used in the Rev 2 shift calculations described below.

The Rev 2 method of calculating shi!t requires that the fluence at the vessel inside surface, fsurf' be calculated and then attenuated to the depth x according to the relationships fx *Isurf('

• This method results in the same fluence at the 1/4 T location as that calculated in Section 2 vith the lead factors.

3.3 SURVEILLANCE TEST CORRECTION FACTOR Rev 1 allows for consideration of credible surveillance esta when it becomes ai ' lab 1r Rev 2 requires that two sets of credible data be developed before 1

4 cc < eir \r use. However, no surveillance testing has been performed "

ye -

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..nce test ec *ection factors do not apply.

3-2

s .

Attcchment 2 FY-CEI/NRR-1188L 3.4 SRIFT AND ADJUSTED REFERENCE TEMFERATURE (ART) .

The RT NDT shift calculations are based on the procedures in Rev 1 and Rev 2.

For Rev 1, the equation for SRIFT is SHIFT = (STF)*(40 + 1000(%Cu .08) + 5000(%F .006)]*(f)0.5 where STF - surveillance test correction factor f = fluence for the given EFFY/10 19 For Rev 2, the SRIFT equation consists of two terms:

SHIFT = STg + Margin where RTNDT = (CFl*f(0.28 - 0.10 log f)

Margin = 2(ey2 ,2)J Chemistry factors (CF) are tabulated for velds and plates in Tables 1 and 2 respectively of Rev 2. The margin term e has set values in Rev 2 of 17'F for 3

plate and 28'F for veld. .Bovever,-e 3need not be greater than 0.5* aT NDT' >

The values of adjusted reference temperature (ART) are computed by adding the SRIFT terms to the values of initial RT NDT. ART versus EFFY is plotted for the most limiting beltline conditions in Figure 3-2. As seen, the beltline plate with the highest initial RTg is limiting through about 12 EFFY. For operation beyond that time, the beltline veld with the highest MTg is limiting.

h l

3-3 i

Attachment 2 PY-CE1/NRR-1188L 3.5 RESULTS OF IMPACT EVALUATION The impact of implementing Rev 2 can best be determined by comparing the ART values based on Rev 1 and Rev 2. Table 3-1 shows the ART values at 32 EFPY for each beltline material. The following conclusions are drawn from the results in the table:

1. The Rev 2 ART values at 32 EPPY are belov 200'F, vbich is the allowable limit in 10CFA50, Appendix G. Therefore, implementation of Rev 2 vill not result in any additional requirements for analysis, testing or provisions for thermal annealing.

2. Rev 2 increased the maximum ART value by 36'F (43.5'F to 79.9'F). As a result, the P-T curves A', B', C' currently in thw Tech Spec are valid for only about 13 EFPY, not 32 EFPY as they were for Rev 1.

3. P-T curves valid for 8 EFPY have been developed consistent with the first full surveillance capsule being withdrawn at 6 EFPY per ASTM E185. These curves will be updated to 32 EFPY vith dosimetry results from the test at 6 EFPY.

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l CCMPARISON OF REV 1 AND REV 2 ART VAIRES I

f. T2 PERRY .

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Rev 1 Rev 2 l ART (*F) ART f'F) l Saltlina Cannonant i 4

Plates: ,

r C2557-1 43.5 59.6  ;

B6270 1 10.2 19.6 A1155-1 23.5 39.6 Velds: t c

627260 37.1 79.9 626677 30.3 6.8 SP62145 3.6 -3.8 624063 -19.8 5.0 627069 16.4 ' 33.2 h

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Attachment 2 PY-CEI/NRR-1188L 4.0 PRESSURE-TEMPERATURE CURVES 4

4.1 BACKCROUND Operating limits for pressure and temperature are required for three categories of operations (a) hydrostatic pressure tests and leak tests, referred to as Curve Al (b) non-nuclear heatup/cooldown and low-power physics tests, referred to as curve B and (c) core critical operation other than Physics Tests, referred to as Curve C. There are three vessel regions that affect the operating limits the closure flange region, the core beltline region, and the remainder of the vessel, or non-beltline tsgions. The closure flange region limits are controlling at lower pressures primarily because of 10CFR50 Appendix G [ Reference 1] requirements. The non-beltline and beltline region operating lialts are evaluated according to procedures in 10CFR50 Appendix G, Appendix G of the ASME Code [ Reference 5), and Velding Research Council (VRC) Bulletin 175 [ Reference 6], with the beltline region minimum temperature limits adjusted to account for vessel irradiation as discussed in Section 3.

Figure 4-1 has curves applicable per Rev 2 for 32 EFPY of operation, for use in the USAR. Figure 4-2 has curves applicable per Rev 2 for'8 EFPY of operation, for use in the Tech Spec. The requirements for each vessel region influencing the P-T curves are discussed below.

4.2 NON-BELTLINE REGIONS Non-beltline regions are those locations that receive too little fluence to cause any RT increase. Non-beltline components include the nozzles, the NDT closure flanges, some shell plates, top and bottom head plates and the control rod drive (CRD) penetrations. Detailed stress analysis, specifically for the purpose of fracture toughness analysis, of the non-beltline comporents were performed for the BVR/6. The analysis took into account all mechanical 4-1 l l

'Atttchment 2 PY-CEI/NRR-1188L l

t dings and. thermal transients anticipated. Detailed stresses were used i 3rding to Reference 6 to develop plots of allowable pressure (P) versus iperature relative to the reference temperature (T-- RTyg) . Plots were

'31 ped for the two most limiting regions: the feedveter nossle and the CRD etration regions. All other non-beltline regions are categorized under one those two regions.

t gan:ric BVR/6 non-beltline region results were applied to Perry by adding i high3st RT for the non-beltline discontinuities to the appropriate P NDT

sus (T - RTyg) curves for the BVR/6 CRD penetration or feedvater nostle.-

i liciting RT values are 10'F for the CRD penetration limits and -20'F NDT th3 feedvater nostle limits.

5 CORE BELTLINE REGION )

e pressure-temperature (P-T) limits for the beltline region are determined I c3= ding to the methods in ASME Code Appendix G [ Reference 5]. As the ylins fluence increases during operation, these curves shif t by an enount -

h::.ssed in Section 3. Typically, the beltline curves shift to become more L liiting than the non-beltline curves at some point during operating life.

Ing Rav 2 for Perry, this occurs after only 4 ETPY of operation. The curves j iulting from shif ting the beltline limits are shown as A', B', C' in Jure 4-1 for 32 EFPY of cperation, and in Figure 4-2 for 8 EFPY of arcticn. Curve A' at 8 EFPY is still less limiting than the non-beltline sits in Figure 4-2, so in the Tech Spec's for the first 8 EFPY of operation,  !

rve A' was overlaid onto Curve A for clarity to the operators.

1 o stross intensity factors (K ),y calculated for the beltline region eerding to ASME Appendix G procedures, vere based on a combination of

ssuro and thermal stresses for a 1/4 T flav in a flat plate. The pressure ecss:s were calculated using thin-valled cylinder equations. Thermal rossos were calculated assuming the through-vall temperature distribution of flot plate subjected to a 100'F/hr thermal gradient. A 32 EFPY ART of 80'F '

d n 8 EFPY ART of 37'F vere used to adjust the (T - RTNDT) values from gute G-2210-1 of Referrence 5.

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i soLTUP [ CURVES A',B',C' ARE VALIO I 70'r / FOR 32 EFPY OF OPERATION '

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100 200 300 400 500 600 MINIMUM REACTOR VESSEL WCTAL TEMPERATURE (*F)

Figure 4-1. Minimum Temperotures Required vs.

Reoctor Pressure, Unit 1. Volid to 32 EFPY l

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Attcchment 2

.. ., PY-CEI/NRR-l188 L 1600 A'A BB' CC' 1400 [ i; 7 E 5 li

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$ CURVES A',B'.C' ARE VALID FOR 8 EFPY OF OPERATION 200 ,

hU' CURVE A' - iNFORMATION ONLY CURVES A, 8 AND C VAllD FOR 12, 4 AND 4 EFPY, RESPECTIVELY 0

0 100 200 300 400 500 600 MINIMUM REACTOR VESSEL METAL TEMPERATURE (*F)

Figure 4-2, Minimum Temperatures Required vs.

Reoctor Pressure, Unit 1, Votid to 8 EFPY

i Attachment 3 PY-CEI/NRR-11881.

5.0 SIGNIFICANT HAZARDS AND ENVIRONMENA1. IMPACT CONSIDERATIONS l

The Nuclear Regulatory Commission (NRC) has promulgated star.dards in 10CFR50.92(c) for determining whether a proposed amendment to a facility operating license involves no significant hazards considerations. A proposed l amendment to an operating license involves no significant hazards considerations if operation of the facility in accordance with the proposed l 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 than previously evaluated; or (3) involve a significant reduction in a margin of safety.

CEI has reviewed the proposed amendment (Attachment 1 to this letter) with ]

respect to these standards and has determined that the proposed changes do not j involve a significant hazard because f l

(1) The proposed changes do not involve a significaat increase in the probability or consequences of a previously evaluated accident because the proposed changes are determined in accordance with i 10CFR50 Appendix G and H, using the methods described in Generic 1.etter 88-11 and Regulatory Guide 1.99 Revision 2. The changes to Figure 3.4.6.1-1 vill result in equivalent or more conservative j

limits on reactor vessel pressure as a function of temperature for all operational conditions (hydrostatic and leak testing, ]

non-nuclear heatup/cooldown, and core critical operations) during t

the first eight years of ef fective full power operation. The methodology used to derive these values produces limits which continue to ensure that suf ficient margin is unintainu to meet the criteria of GDC 31 " Fracture Prevention of Reactor Coolant Pressure Boundary."

i S-1

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• e' Attechnent 2 PY-CEI/NRR-ll88L Therefore, the probability of occurrence of a previously evaluated accident is not impacted by these proposed changes but remains unchanged.: Nor do these proposed changes increase the consequences of an accident since any temperature shifts are well within equipment operating ranges. The changes to Table 4.4.6.1.3-1 vould-

• incorporate newly derived lead factors for the surveillance programs and revise the withdrawal schedule for the surveillance capsules as requested by ASTM E 185-82. These are merely administrative program changes which are incorporated to permit timely update to P-T limits based on actual' plant experience, and therefore cannot affect either the probability or consequences of any previously evaluated accidents.

(2) The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because the proposed change do not involve any new modes of operation. The only change vill be operation of the plant within operating pressure limits which are determined in a more conservative manner. Therefore no new failure mode or accident sequence is introduced by this change.

(3) The proposed changes do not involve a significant reduction in the margin of safety because the changes described in Generic Letter 88-11 and its references result in greater assurance that margins to nil ductile failure are acceptable under plant operating and test conditions. The required margins are incorporated in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G, and in 10CFR50 Appendix G. The revised curves are based on the latest NRC  !

guidelines along with actual neutron fluence data for Perry. The new limits conservatively account for irradiation embrittlement effects, thereby maintaining the margin of safety. The program changes to'the surveillance capsule withdrawal schedule are administrative in nature and do not impact any margins of safety.

5-2 a

+<> ' e Atttchment 2 PY-CEI/NRR-1188L Environmental Impact Considerations The Cleveland Electric illuminating Company has reviewed .the proposed Technical Specification change against the criteria of 10 CFR 51.22 for environmental considerations. As shown above, the proposed change does not involve a significant hasards consideration, nor increase the ?ypes and amounts of effluents that may be released of fsite, nor sigt.ificantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, CEI concludes that the proposed Technical Specification changes meet the criteria given in 10 CFR 51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement. 1 1

i 4

i i

i i

i 5-3

'e o .- Attachment 2 PY-CEI/NRR-ll88L

6.0 REFERENCES

[1] " Fracture Toughness Requirements," Appendix G to Part 50 of Title 10 of the Code of Federal Regulations, July 1983.

I

[2] " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 1, April 1977.

{3) " Radiation Embrittlement of Reactor Vessel Materials," USNRC Regulatory Guide 1.99, Revision 2, May 1988. '

I i [4] Perry Nuclear Power Station Updated Safety Analysis Report, Revision 1,  ;

i March 1989. i i

[5] " Protection Against Non-Ductile Failure," Appendix G to Section III of the ASME Boiler & Pressure Vessel Code, 1986 Edition with leSS Addenda.  !

[6] "PVRC Reconsendations on Toughness Requirements for Ferritic Materials,"

Velding Research Council Bulletin 175, August 1972.

i NJC/ CODED /3576 I i

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