ML20196F688

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Proposed Tech Specs Updating Heatup,Cooldown & Overpressure Protection Curves & Bases to Apply for Up to 15 EFPYs
ML20196F688
Person / Time
Site: Beaver Valley
Issue date: 06/17/1999
From:
DUQUESNE LIGHT CO.
To:
Shared Package
ML20137V023 List:
References
NUDOCS 9906290280
Download: ML20196F688 (40)
v • d • e


Text

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ATTACHMENT A Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 127 The following is a list of the affected pages:

Affected Pages:

3/4 4-31 3/4 4-32 3/4 4-37 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 B 3/4 4-10 B 3/4 4-11 B 3/4 4-12 B 3/4 4-13 B 3/4 4-14 3

B 3/4 4-14a l

B 3/4 4-15 B 3/4 4-15a B 3/4 4-15c B 3/4 4-15d 9906290280 990617 PDR ADOCK 05000412 P

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NPF-73 MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: INTERMEDIATE SHELL Pl. ATE B9004-1 INITIAL RTNDT:

60 F AFTEhFPY:

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RTNDT URVES APPLICABLE FOR HEATUP RATES UP TO S0*F/HR FOR THE SERVICE PERIOD UP TO EFPY;CO;;~AJ40 :/.AGOli; Or 10"I A!40 00 F0;O TOR POOO;0L' =~T ""'NT 0nn030.

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3/4.4.9 PRESSURE /TEMPERATUR LIMITS (Continued)

The heatup limit curv, Figure 3.4-2, is a composite curv which we.s pre-pared by determining the st conservative case, with either tr intice or out-side wall controlling, fo any heatup rate up to 60*F per hour.

The cooldown limit curvesj Figure 3.4-3 hare composite curves which were prepa ed based upon the same type analysis with the exception that the controlling loca. ion is always the l

inside wall where the cooldown thermal gradients tend to produc<

tensile stresses while producing compressive stresses at the outside wall.

The teatup and cool-down curves were prepared based upon the aos siting value of the predicted adjusted referaara +=aerature at the end of FPY.

(ni httAJ**)E1M)

The reacwr vess ma eriais n a been tested to determine their initial RTNOT; the results of sese tests a shown in Table B 3/4.4-1.

Reactor opera-tion and resultant fast neutron (E l'Mov) irradiation will ca se an increase in

.the RT Therefore,y }n adjust reference temperature, base up he fluence, ET.

copper conten hos ntent of the material in ques on,$

be pre-dicted using aures and Regulatory _ Guide 1.99, Rev sioni T 7 7.J..

ef "..id 1 : L;.;.... e. 7..C J e4 P.adiation e - - tm React Vessel Materials."

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The most limiting RTNOT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-indu ed ART RT is designated as the higher of either the drop weight ET.

ET ni uctility transition temperature (TE T) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expan-sion (normal' to the major working direction) minus 60*F G;sQ RT increases as the material is exposed to fast utron radiation.

Thus, NOT to find the most limiting RT at any time period in e reactor's life, ART ET NOT due to the radiation exposure associated with that ti period must be added to the original unitradiated RT The extent of the ift in RT I8 '"h'"C'd ET.

NOT by certair, chemical elements (such as copper and g M ) present in reactor vessel steels.

The Regulatory Guide 1.99 Revision @ curves which show the effect 3

of fluence copper contentrana onosnnaeus content on w g for reactor vessel steels at hown in Figure 5 3/4.4 gg

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REACTOR COOLANT SYSTEM Ay I

BASES M[C 3/4.4.9 PRESSURE /T PERA URE LIMITS (Continued) s.

Given the copper n'd 6hosohoroliB contents of th m at limitino material, the radiation-induced TNOT can beChstimated from Figure B 3/4.4-2] Fast-neutron fluence (E > 1

) at the 1/4 T (wall thickness) and 3/4 T (wall thick-ness) vessel locations are e v e as a function of full power service life, 4 r y., c. ", 2/0. 0 1.

The data Por all other ferritic materials in the reactor coolant pressure boundary are examined to insure that no other component will be limiting with respect to RTET*

The preirradiation fracture-toughness properties of the Beaver Valley Unit 2 reactor vessel materials are presented in Table B 3/4.4-1.

The fracture tough-ness properties of the ferritic material in the reactor coolant preuure boundary are determined in accordanca with the 1972 Summer Addenda to Sectico III of ASME Boiler and e fMS$2T u

The ASIE approach for calculating the allowable curves fo' var tous heatup and cooldown rates specifies that the t ress intensity

' acto r, K,

g for the combined thermal and p res at any time during eat nd cooldown cannot be greate the refer a stress intensity fa tor, for the metal temperat at that time, is obtained from the r rence acture gg toughness cu

,definedanAppendixratotheASMECode.

The K curve is given by t equation:

l bit =26.78+1.223exp[0.0145(T-RTmT + 160) p.typ (4-1) e i

where is the reference stress inti nsity factor as a function of the metal j

temperature T and the metal reference nilductility temperature, RTNOT.pThus, the covernina -tion for the heatupj cooldown analysis is defined 6n Anoendix D Go tne Awt W Jas follows:

CKgy + kit <

2) wher K

is the stress intensity factor caused by membrane (pressure) stress gg K

is the stress intensity factor caused by the thermal gradients t

K is a function of temperature the RT of the material ET C = 2.0 for Level A and Level B s ice limits C = 1.5 for hydrostatic and leak conditions during which the reactor core is not critical gg b_M1! -'M'--- 1-3---53 3b5!N5.555'1MUdllii3

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Attichr.cnt A

.CO3vsr VallGy Powar Station, Unit No. 2 License Amendment Reauest No. 127 INSERT E The pressure-temperature limit curves are developed using Code Case N-640.

One of the safety margins incorporated into the curves is the lower bound fracture toughness curve.

The lower bound fracture

. toughness curves available in Appendix G to Section XI use the reference stress intensity factor Kra.

The pressure-temperature limit' curves based on Code Case N-640 use the reference stress intensity factor Krc.

Era is a fracture toughness curve which is a lower bound on all static,. dynamic and arrest fracture toughness, and Kze. is a fracture toughness curve which is a lower bound on static fracture toughness only.

The only change.that is made when generating the revised pressure-temperature limits curve with K c is I

the lower bound fracture toughness curve selected.

All other margins involved in the generation process remain unchanged.

Since the heatup and'cooldown process is a very slow one, with the fastest rate allowed being 100'F per hour, the rate of change of pressure and temperature is considered constant so that the stress is essentially constant.

Both heatup and cooldown correspond to static loading, with regard to fracture toughness.

The only time when dynamic loading can occur and where the dynamic / arrest toughness Kza should be used for the reactor pressure vessel is when a crack is running.

This might happen during a pressurized thermal shock event, but not during heatup and cooldown.

Therefore, the static toughness Kc I

lcwer bound toughness is used to generate the pressure-ter.perature

. limit. curves.

NPF-73 04 l

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~3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued)

I At any time during the heatup or cooldown transient K

is determined by the i

metaltemperatureat.thetipofthepostulatedflad,'teappropriatevaluefor RTNDT, and the reference fracture toughness curve.

The thermal stresses result ing from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factors, kit, f r the reference flaw are computed.

From equation 4-2, the pressure stress intensity factors are and, from these, the allowable pressures are calculated.

For the calculation of the allowable pressure-versus-coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall.

During cooldown, the controlling location of the flaw is al-ways at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increases with increasing cooldown rates.

Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations.

From these relations, composite limit cuMs an.

enevres 4 he constructed for each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

During cooldown, the V4 T vessel location is at a higher temperature than the fluid adjacent to the vessel ID.

This condition, of course, is not true for the steady-state situation.

It fol-lows that, at any given reactor coola emperature, the AT developed during g eocidown results in a higher value at the V4 T location for finite cool down rates than for to operat on.

Furthermore, if conditions exist such that the increase exceeds kit, the calculated allowable pressure during cooldown will be gre r than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the V4 T location and, therefore, allowable pressures may unknow-ingly be violated if the rate of cooling is decreased at various intervals along I

a cooldown ramp. The use of the composite curve eliminates this problem and conservative' operation of the system for the entire cooldown period.

Three separate calculations are required te 4termine the limit curves for

~

finite heatup rates. As is done in the cooldew mlysis, allowable pressure-i temperature relationships are developed for sf eady~. ate conditions as well as finite heatup rate conditions assuming the pruonee sf a V4 T defect at the inside of the vessel wall. The thermal gradient. uuring heatup produce compres-sive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure.

The I temperature at the crack tip lags the coolant temperature; therefore, the for the V4 T crack during heatup is BEAVER VALLEY - UNIT 2 B 3/4 4-12 Ottpomi Wed&

NPF-73 REACTOR COOLANT SYSTEM BASES 3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued) lower than the the T crack during steady-state conditions at the Ic M

J same coolant temperature.

E During heatup, especially at the end of the transient, cond s may exist such that the effects of compressive thermal stresses and lower

's do n t IR offset each other, and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered.

Therefore, both cases have to 1

be analyzed in order to insure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

g The second hortion of the heatup analysis concerns the calculation of pressurc h limitations for the case in which a 1/4 T deep outside sur-face flaw is assumed.

Unlike the situation at 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 thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves re produced as follows:

A composite curve is constructed based on a point-by-oint comparison of the steady-state and finite heatup rate data.

At any given te erature, the a

allowable pressure is taken to be the lesser of the three values ken from the curves under consideration.

The use of the composite curve is nec sary to set conservative heatup limitations Decause it is possible for conditio s to exist wherein, over the course of the-heatup ramp, the controlling condit n switches from the inside to the outside and the pressure limit must at all ti s be based on analysis of the most critical criterion.

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The actual shift in RTNOT of the vessel material will be establis hed period-ically during operation by removing and evaluating, in accordance witt 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 th e neutron spectra at the irradiation samples and vessel inside radius are essent Ially iden-i tical, the measured transition shift for a sample can be applied with ':onfidence to the adjacent section of the reactor vessel.

The h'estup and cooldow i curves must be recalculated when the ART determined from the surveillance capsule ET is different from the calculated ART for the equivalent capsule rad,ation ET

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NPF '3 RncTOR coo *JL9T SYSTEM BASES 3/4.d.9 PRFREURF/ TEMPERATURE LTMTTE * (ccmtig The pressure-temperature limit lines anchwn on rigure 3.4-2 for reactor criticality and for inservice leak And hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50 fot 4 eactor criticality and for inservice leak and hydrostatic testing.

W;rpff4y g 1

The number of reactor vessel irradiation surveillance specisens and the frequencies for removing and testiag these specimens are provided in UFSAR Table 5.3-4 to assuga compliance with the requirements of Appendix N to 10 CFR Part to -

The limitations imposed on the pressurimer heatup and cocidown rates and auxiliary spray water temperature di4ferential are provided i

to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis perforset in accordance vi the ASME Code requirements.

Prsesure-temperature limit curves shown in figure B 3/4 44 were developed for the limiting ferritic stesA component with n an i

isolated reactor coolant loop. The limitin generator channel head to tubesheet region. g component is the steam' Ytnis figure provides the ASME III, Appendix G limiting curve whicth is used to define operational bounds, such that when operating with an isolated loop the analysed pressure-temperature limits ats known. The temperature range provided bounds the expected operatinf range for an isolated looQand Code Catt Al-HO]

OVERPRESEURE PROTECTIoM SYSTEMS nAcEcRotmn The overpressure protection system (oFPW controls RCs pressure at low temperatures so the integrity of the raiector coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature (F/T) limits of 10 CFR 50, app widix G.

The reactor vessel is the limiting RCPB component icht demonstrating such protection.

The maximum setpoint for the power operated relief valves (PCWs) and the maximus RCs pressure tot the existing RCS cold leg temperature during cooldown, shutdown, and heatup meet the 10 CFR 50, Appendix G requirements during.the OFF8 18 088.

0 The reactor vessel material is less toc 9h at low temperatures than at normal operating temperature.

As the v ssel neutron exposure e

accumulates, the material toughness decreases and becomes less resistant to pressure stress at low temperatures.

RCs pressure, therefore, is maintained low at low temperatures and is increased only as temperature is increased.

BEAVER VALLEY - UNIT 2 5 3/4 4-14 Amendment No.M-UMfY inbds

l Attechmant A B32 var Vallcy Pow 0r Stttien, Unit No. 2 License Amendment Recuest No. 127 l

INSERT F These pressure-temperature limit lines on Figures 3.4-2 and 3.4-3 (a, b, c,

d and e) for boltup temperature are provided to ensure compliance with the minimum temperature requirements of Appendix G to ASME Section XI for vessel closure head flange boltup.

It recommends that when the flange and adjacent shell region are stressed by the full intended bolt preload the minimum metal temperature in the stressed region is at least the initial RTNDT temperature for the material in the stressed regions.

l l

NPF-73 l

2500 l

l 2000 1

@ 1500 8

/

8

/

[

1000

, 71 i

I i

I 500 l

0 50 60 70 80 90 100 110 120 TEMPERATURE (*F)

FIGURE B 3/4 4 ISOLATED LOOP PRESSURE-TEMPERATURE LIMIT CURVE BEAVER VALLEY UNIT 2 B3 Oupas/4 4 d14,p)=:: n,: :m:-

4 tw 3/2/92

NPF ?3 prACTOR COOULNT SYSTEM BASES (Continued) 3/4.4.9 PR FE StJRE /TEMPERATUD E LIMITS (Continued)

BACKGROUND (Continued)

The potential for vessel overpressurization is most acute when the RCS is water solid, occurring only during shutdown; a prassure fluctuation can occur more quickly than an operator can react to relieve the condition.

Exceeding the RCS P/T limits by a significant amount could cause brittle cracking of the reactor vossel.

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O.i;-LCorprovices m.= averpressure protection ny naving a minimum coolant input capability and having adequate pressure relief cr9acity.

Limiting coolant input capability requires deactivating all but one charging pump and isolating the accumulators.

The pressure relief capacity requires either two redundant RCS relief valves or a depressurized RCS and an RCS vent of sufficient size.

One RCS relief valve or the open RCS vent is the overpressure protection device that acts to terminate an increasing pressure event.

With miniaun coolant input capability, the ability to provide core coolant addition is restricted.

The LCO does not require the makeup control system deactivated or the safety injection (SI) actuation circuits blocked.

Due to the lower pressures in the OPPS MODES and the expected core decay heat levels, the makeup system can provide adequate flow via the makeup control valve and, if needed, until the charging pump is actuated by SI.

The OPPS for pressure relief consists of two PORVs with reduced lift settings or a depressurized RCS and an RCS vent of sufficient size.

Two RCS relief valves are required for redundancy.

One RCS relief valve has adequate relieving capability to keep fron overpressurization for the required ec>olant input capability.

PORV REqEIIREMENTS As designed for the OPPS System, sach PORY is signaled to open if the RCS pressure approaches a limit determined by the OPPS actuation logic.

The OPPS actuation logic monitors both RCS temperature and RCS pressure and determines when a condition not acceptable in the limits is approached.

The wide range RCS temperature indications are i

auctioneered to select the lowest temperature signal.

The lowest-temperature signal is processed through a function generator that calculates a pressure limit for that temperature.

The calculated pressure limit is then compared with the indicated RCS pressure from BEAVER VALLEY - UNIT 2 B 3/4 4-15 Amendment No.O (lmeadWoedQ,

NPF-73 FMCTOR SCOLANT SYSTEM BASES (Continued) 3/4.4.9 PRFMSURE/ TEMPERATURE LTMTTS (Centinuedl PORV REOUIREMENTS (Continued) a wide range pressure channel.

If the indicated pressure meets or exceeds the calculated value, a PORV is signaled to open.

Having the setpoints of both valves within the limits ensures that the Appendix G limits will not be exceeded in any analyzed event.

When.a PORV is opened in an increcsing pressure transient, the release of coolant will cause the pressure increase to slow and reverse.

As the PORV releases coolant, the RCS pressure decreases until a reset pressure is reached and the valve is signaled to close.

The pressure continues to decrease below the reset pressure as the valve closes.

Res VrwT nrOUrnrMrwTs N

Once the RCS is depressurized, a vent exposed to the conts.inment atmosphere will maintain the RCS at containment ambient pressure in an RCS overpressure transient, if the relieving requirements of the transient do not exceed the capabilities of the vent.

Thus, the vent path must be capable of relieving the flow resulting from the limiting OPPS mass or heat input transient, and maintaining pressure below the P/T limits.

The required vent capacity may be provided by one or more vent paths.

For an RCS vent to meet the flow capacity requirement, it may be satisfied by removing a pressurizar safety valve or establishing an opening between the RCS and the containment atmosphere of the required size through any positive means available which cannot be inadvertently defeated.

The vent path (s) must be above the level of reactor coolant, so as not to drain the RCS when open.

APPLICABLE SAFETY ANALYSES Safety analyses demonstrate that the reactor vessel is adequately protected against exceeding the P/T limits when low RCS temperature conditions exist.

At the enable temperature and below, overpressure i

prevention is provid&d by two OPERABLE RCS relief valves or a depressurised RCS and a a :. icient sized RCS vent.

The actual temperature at which the pressure in the P/T limit curve falls below the OPPS setpoint increases as the reactor vessel material toughness decreases due to neutron embrittlement.

Each time the heatup and cooldown curves are revised, the OPPS must be re-evaluated to ensure its functional requirements can still be met.

i B3 Amendment No. m"l BEAVER VALLEY - UNIT 2 Chopo.r/4 4-15ael Wed,9)

l Att chm:nt A Beaver Valley Power Station, Unit No. 2

)

License Amendment Recuest No. 127 l

INSERT G The low. limit on pressure during the transient is typically established based solely on an operational consideration for the Reactor Coolant Pump (RCP)

No.

1 seal to maintain a

nominal differential pressure across the seal faces for proper film-riding performance.

The upper limit (based on the minimum of the steady-state 10 CFR 50 Appendix G

requirement and the PORV piping limitations) and the RCP No. 1 seal performance criteria create a pressure range from which the setpoints for both PORVs are selected.

When there is insufficient range between the upper and lower pressure limits to select the PORV setpoints to provide protection against violating both limits, setpoint selection to provide protection against the upper limit violation takes precedence.

i 4

NPF-73 REACTOR CCO ULNT SYSTEM BASES (Continued) 3/4.4.9 PREMSURE/ TEMPERATURE LIMITS (Continued)

MEAT TNPUT TYPE TRANSTENTS (Continued)

The analyses demonstrate that either one RCS relief valve or the

.depressurized RCS and RCS vent can maintain the RCS pressure below the limits when only one charging pump is actuated by SI.

Thus, the LCO allows only one charging pump OPERABLE during the OPPS MODES.

Since neither one RCS relief valve nor the RCS vent can handle a full SI actuation, the LCO also requires the accumulators isolated.

The isolated accumulators must have their discharge valves closed with power removed.

Fracture mechanics analyses established the temperature of OPPS Applicability at the enable temperature.

PORV PERFORMANCE The fracture mechanics analyses show that the vessel is protected when the PORVs are set to open at or below the limit.

The setpoint is derived by analyses that model the performance of the OPPS assuming the limiting transient of SI actuation of one charging pump.

These analyses consider pressure overshoot and undershoot beyond the-PORV opening and closing, resulting from signal processing and valve stroke times.

The PORV setpoints at or below the derived limit ensures the P/T limits will be met.

(Nems&D

,jjgj, g

3 Ine naximum Allowed PORV Setpoi for the OPPS is derived by analysis which models the performa e of the OPPS assuming various mass input and haat input transient.

Operation with a PORV setpoint less than or equal to the maximum etpoint ensures that Appendix G limits will not be violated with consideration for:

(1) a maximum pressure overshoot 'neyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening; (2) a 50*F heat transport affect

=ada aaasible by the geometrical relationship of the (RER suction linD and the RCS wide range temperature indicator used for OPPS; (3) instrument uncertainties;

-ene(4) single failur

,4 gg The PORV setpoint will be updated when the revised P/T limits conflict with the PS analysis limits.

The P/T limits are periodically modifi as the reactor vessel material toughness decreases due to neu on embrittlement caused by neutron irradiation.

Revised limits are d termined using neutron fluence projections and the results of e mainations of the reactor vessel material irradiation surveil ance specimens.

The Bases for LCO 3.4.9.1,

" Pressure /Temperatu a Limits," discuss these examinations.

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  1. Y t>y klHb* terh BEAVER VALLEY - UNIT 2 5 3/4 4-15c Amendment No. 49l

@tefeSY Y0kl

NPi~73 REACTOR COOLUlT SYSTEM BASES (Continued)

/4.4.9 PREESURE / TEMPERATURE LTMITS (Continued [

3 PORV PER/ORMANCE (Continued)

The PORVs are considered active components.

Thus, the failure of one PORV is assumed to represent the worst case, single active failure.

RCE-VENT PFWFORMANCE With the RCS depressurized, analyses show that a PORV or equivalent opening with a vent size of 3.14 square inches is capable of mitigating the allowed OPPS overpressure transient.

The capacity of a vent this size is greater than the flow of the limiting transient for the OPPS configuration, SI actuation with one charging pump OPERABLE, maintaining RCS pressure less than the maximum pressure on the P/T limit curve.

The RCS vent size is based on the PORV sizg, therefore, the vent is bounded by the PORV analyses.

The RCS vent is passive and is not subject to active failure.

LCQ This LCO requires that the OPPS is OPERABLE.

The OPPS is OPERABLE when the minimum coolant input and pressure relief capabilities are OPERABLE.

Violation of this LCO could lead to the less of low temperature overpressure mitigation and violation of the limits as a result of an operational transient.

(ltN.5EA T & o T

limit the coolant input capability, the LCo requires one charging pump capable of injecting into the RCS and all accumulator discharge iso'.ution valves closed and immobilized.

The LCO is qualified by n note that permits two pumps capable of RCS injection for less than or equal to 15 minutes to allow for pump swaps.

This note also allows all charging pumps capable of injecting into the KCS during a change from MODE 3 to MODE 4 to be OPERABLE for a limited period of time.

The' Lc0 is also qualified by a note stating that accumulator isolation with power removed from the discharge isolation valves is only required when the accumulator pressure is greater than or at the maximum RCS pressure for the existing temperature, as allowed by the P/T limit curves.

This note permits the accumulator discharge isolation valve surveillance to be performed only under these pressure and temperature conditions.

BEAVER VALLEY - UNIT 2 B3/4f-15d Amendment No.W-l Utofoseg L00k l

Attcchment A Ceaver Valley Power Station, Unit No. 2 License Amendment Recuest No. 127.

INSERT H The Maximum Allowable Nominal Setpoint Curve defines the maximum nominal setpoint at which the PORVs can be set which will ensure that Appendix G limits are not exceeded.

To maximize operating margin, the setpoint for the higher PORV is set at the Maximum Allowable Nominal Operating Curve within the respective instrumentation loop calibration tolerance band.

The PORV.setpoint uncertainty is calculated with reference to the methodology in ISA 67.04-1994 for performing instrumentation uncertainty calculations.

The l

instrumentation calibration tolerances are provided in p3 ant procedures.

The overall setpoint calculation accounts for the instrumentation calibration tolerances in the uncertainty calculation.

Since actuation of both PORVs can result in excessive undershoot below the PORV setpoint, the lower PORV setpoints are staggered by an amount greater than or equal to the limiting overshoot (from either the mass injection or heat addition events).

The staggered setpoints are provided in plant procedures.

i i

)

1 I

ATTACHMENT B Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 127 UPDATE HEATUP AND COOLDOWN CURVES AND OVERPRESSURE PROTECTION SYSTEM (OPPS) SETPOINT CURVE A.

DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would extend the applicability of the heatup and cooldown curves of Technical Specification (TS),

Section 3.4.9.1, Pressure / Temperature (P/T) Limits.

A new Figure 3.4-2 provides the Beaver Valley Unit 2 Reactor Coolant System Heatup Limitation Applicable for the First 15 Effective Full Power Years (EFPY).

A new Figure 3.4-3 (a,

b, c,

d and e) provides the Beaver Valley Unit 2 Reactor Coolant System Cooldown Limitation Applicable for the First 15 EFPY.

The proposed changes include new heatup and cooldown curves developed in accordance with the methodology provided in Regulatory Guide (RG) 1.99 Revision 2 (Reference 1) and Code Case N-640 using the Surveillance ( nsule V results.

The applicability of Section 3.4.9.3, Overpt asure Protection Systems, is also updated to 15 EFPY.

A new Figure 3.4-4 provides the Maximum Allowable Nominal PORV (power operated relief valves) Setpoint for the Overpressure Protection System.

The methodology for the P/T limits has been changed from RG 1.99 Revision 1 to RG 1.99 Revision 2.

Bases Section B 3/4.4.9, Reactor Coolant System, has also been revised to incorporate new references and associtted methodology changes.

Figure B 3/4.4-1, titled " Fast Neutron Fluence as a Function of Full Power Service Life" has been deleted since it is not used by RG 1.99 Revision 2 which replaces the previous reference to Regulatory Guide 1.99 Revision 1.

Figure B 3/4.4-2, Predicted Adjustment of Reference Temperature, "A",

as a Function of Fluence and Copper Content, is replaced by a new figure, Predicted Decrease in Shelf Energy as a Function of Copper Content and Fluence, since RG 1.99 Revision 2 refers to the new figure to show the effect of fluence and copper content on upper shelf life for reactor vessel steels.

This figure has also been renumbered to B 3/4 4-1 and, as a result, Figure B 3/4 4-3 has been renumbered to B 3/4 4-2, and reference to these figures in the text has been modified as well.

An I

equation has been added to page B 3/4 4-11 which is used to predict the radiation-induced shift in the reactor vessel

{

reference temperature (ARTNDT).

Previously Figure B 3/4.4-2 was used.

j A new column was added to Table B 3/4.4-1, Reactor Vessel Toughness Data, for nickel since it is one of the predominant metals used to generate the Adjusted Reference Temperature (ART).

A footnote was added to this table to indicate

that, for evaluation of inservice reactor vessel irradiation damage i

assessments, the best estimate chemistry values reported in the latest response to Generic Letter (GL) 92-01 or equivalent document are applicable.

Reference to Kra has been replaced with l

1

ATTACHMENT B,-continusd LicsnBS Amsndmsnt Rcqusst No. 127 Page 2 K c in accordance with the change to the methodology used in Code I

Case N-640.

A description of the term "Boltup Temperature" has been added to the Bases in Insert F.

The purpose of the boltup remperature limit line is to ensure compliance with the minimum temperature requirements of Appendix G to ASME Section XI for vessel flange and closure head flange boltup.

Selection criteria has been added to the Bases to describe the basis used for selecting the PORV limits in Insert G.

It notes that the Appendix G requirement takes precedence over the reactor coolant pump (RCP)

No.

1 seal limit.

A discussion of the Maximum Allowable Nominal PORV setpoint was also added to the Bases in Insert H.to indicate that instrumentation calibration tolerances are included within the setpoint.

B.

DESIGN BASES The proposed changes replace the P/T limits and the OPPS setpoint curves with those applicable to 15 EFPY based on the analysis of Surveillance Capsule V.

The revised heatup and cooldown curves do not account for instrument uncertainties since the OPPS i

setpoint curve is more conservative than the heatup and cooldown curves and is used for plant operation.

At long as the OPPS curve limits are satisfied the heatup and cooldown limits are not exceeded.

Instrument uncertainties have been included in the determination of the OPPS setpoints.

Additionally, the OPPS setpoints conservatively account for the pressure difference between the wide-range pressure transmitter and the reactor vessel limiting beltline region.

The OPPS design basis takes credit for the fact that overpressure events most likely occur during isothermal conditions in tha reactor coolant system (RCS).

According to WCAP 14040-NP-A, Rev.

2 (Reference 2),

it is appropriate to utilize the steady-state Appendix G

limit.

WCAP-14040-NP Revision 1 was approved by the NRC by letter dated October 16, 1995, which was incorporated in Revision 2 of the WCAP issued in January'1996.

In addition, OPPS also provides for an operational consideration to maintain the integrity of the PORV piping. Since the heatup and cooldown curves are generated Kc fracture toughness methodology, code case N-514 using the I

(which permits a 10% relaxation of the Appendix G pressure temperature limits up to the OPPS enable temperature) is not applicable.

The Code case N-640 methodology is used.

The

" enable" temperature, the RCS temperature below which OPPS is required to operate, is calculated in accordance with Code Case N-640.

The ability of the reactor vessel to resist fracture constitutes an important factor in ensuring safety in the nuclear industry.

The beltline region of the reactor pressure vessel is the most critical region of the vessel because it is subjected to significant fast neutron bombardment.

Generally, the overall effects of fast neutron irradiation on the mechanical properties of low alloy, ferritic pressure vessel steels such as SA533 Grade B Class 1, which is the base metal of the reactor pressure B-2

)

ATTACHMENT B, continu2d Licenna Amsndmsnt Rsquast No. 127 Page 3 vessel, show an increase in hardness and tensile properties and a decrease in ductility and toughness during high energy irradiation.

The Reactor. Vessel Radiation Surveillance Program, designed by Westinghouse, is described in WCAP-9615, Revision 1, "Duquesne Light Company. Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program" (Reference 3).

The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Vessels."

A method for performing analyses to guard against fast fracture in reactor pressure vessels has been presented in Appendix G to Section XI of the ASME Boiler and Pressure Vessel Code (Reference 4).

The method utilizes fracture mechanics concepts and is based on the reference nil-ductility temperature (RTNDT)-

RTNDT is defined as the greater of either the drop weight nil-ductility transition temperature (NDTT) per ASTM E-208 or the

-temperature 6 0*F less than the 50 ft-lb (and 35-mil lateral expansion) temperature as determined from Charpy specimens oriented normal (transverse) to the major working direction of the plate.

The RTNDT of a given material is used to index that material to a reference stress intensity factor curve (K c curve)

I

.provided in Code Case N-640.

The K c curve is a lower bound I

static fracture toughness curve based on results obtained from several heats of pressure vessel steel.

When a given material Ke curve, allowable stress intensity curve is indexed to the I

factors can be obtained for this material as a function of temperature.

Allowable operating limits can then be determined utilizing these allowable stress intensity factors.

The use of Kc adds substantial pressure margin to the heatup and cooldown I

curves than that provided by the stress intensity factor K a of I

Ka was used for the current heatup and cooldown Reference 5.

I curves.

RTNDT and, in turn, the operating limits of nuclear power plants, can be adjusted to account for the effects of radiation on the reactor vessel material properties.

The radiation embrittlement changes in mechanical properties of a given reactor pressure vessel steel can be monitored by a reactor surveillance program, such as the Beaver Valley Unit 2

Reactor Vessel Radiation j

Surveillance

Program, in which a

surveillance capsule is I

periodically removed from the reactor vessel capsula location and the encapsulated specimens tasted.

The increase in the average Charpy V-notch 30 f t-lb temperature, ARTNDT, due to irradiation j

is added to the initial RTNDT, along with margin to cover uncertainties, to adjust the RTNDT for radiation embrittlement.

l initial RTNDT + margin for uncertainties +

l This ART (ART

=

ART or) is used to index the material to the K c curve and, in N

I turn, to set operating limits which take into account the effects of irradiation on the reactor vessel materials.

B-3 I

ATTACHMENT B, continu1d

~

LicEnna Amsndannt Rtqusst No. 127 Page 4 Overpressure protection for the RCS is achieved by means of self-actuated safety valves located high in the system on the steam space of _ the pressurizer.

These safety valves have a set pressure based on the RCS design pressure of 2485 psig and are intended to protect the system against transients initiated in the plant when the RCS.is operating near its normal temperature.

To avoid, brittle fractures at reactor vessel metal temperatures below the OPPS enable temperature, the allowable system pressure is substantially less than the normal system design pressure of 2485 psig.

Therefore, overpressure mitigation provisions for the reactor vessel must be available when the RCS and, hence, the reactor

vessel, are at temperatures below the OPPS enable temperature.

Normally when the RCS is at a temperature below the OPPS enable temperature, the RCS is open to the Residual Heat Removal System (RHRS) for the purposes of removing residual heat from the core, providing a path for letdown to the purification subsystem and to control the RCS pressure when the plant is operating in a water solid mode.

The RHRS is provided with self-actuated water relief valves to prevent overpressure in this relatively low design pressure system (600 psig) caused either within the system itself or from transients transmitted from the RCS.

The RHRS relief valves will mitigate pressure transients originated in the RCS to maximum pressure values determined by the relief valves set pressure of 450 psig plus a pressure accumulation above the set pressure dependent on the liquid volume magnitude of the transient.

The low design pressure RHRS is normally isolated from the high design pressure

RCS, during reactor power operation at temperatures above 3 5 0*F,

by two isolation valves in series.

Therefore, the RHRS can be inadvertently isolated from the RCS by these same isolation valves.

The OPPS is intended to provide overpressure mitigation for the RCS by addressing those transients which may occur when the RHRS isolation valves inadvertently close thus isolating the RHRS water relief valves from the RCS.

The OPPS controls the RCS pressure at low temperature so the integrity of the Reactor Coolant Pressure i

Boundkr9 (RCpB) is not compromised by violating the P/T Limits.

This evaluation will demonstrate that plant operation with the revised heatup and cooldown and OPPS curves will not adversely affect the pressure boundary integrity or operability of the RCS, and does not represent a significant hazards consideration as defined in 10 CFR 50.92.

C.

JUSTIFICATION The current heatup and cooldown curves are being revised using the results of the evaluation of Surveillance Capsule V by i

changing the applicability from 10 EFPY to 15 EFPY.

The curves f

have been developed in accordance with the methodology provided B-4

ATTACHMENT B, continusd Licanso Amnndmsnt Rsqusst No. 127 Page 5 in RG 1.99 Revision 2 and Code Case N-640.

The use of the Code Case N-640 K c methodology, will result in an increase in the I

safety of the operating plant, as the likelihood of RCP seal failure and/or fuel problems will decrease. By changing from Kra to K c _ methodology, the operating window relative to RCP seal I

requirements is wider and the probability of damaging the RCP seals and initiating a small break LOCA (a potential pressurized thermal shock initiator) are reduced.

The OPPS setpoints are being revised for the same applicability. The current setpoints were based on a 20'F cooldown curve and a 60*F heatup curve.

The proposed setpoints are based on isothermal conditions.

The OPPS design basis takes credit for the fact that overpressure events most likely occur during isothermal conditions in the RCS.

According to WCAP 14040-NP-A, Rev.

2, it is appropriate to utilize the steady-state Appendix G limit.

In addition, the OPPS also takes into consideration those limitations required to maintain the integrity of the PORV. piping.

D.

SAFETY ANALYSIS Westinghouse performed an analysis to determine curves acceptable for heatup and cooldown based on the analysis results of Surveillance Capsule V.

The OPPS setpoints were then generated applicable to the 15 EFPY heatup and cooldown curves.

Heatup and cooldown curves are calculated using the most limiting value of RTNDT corresponding to the limiting beltline region material of the reactor vessel.

The adjusted RTwor of the limiting material in the core region of the reactor vessel is determined by using the unirradiated reactor vessel material fracture toughness properties, estimating the radiation-induced ARTwor and adding margin.

The RTwor is designated as the higher of either the drop weight 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.

RTwor increases as the material is exposed to fast-neutron radiation.

Therefore, to find the most limiting RTNDT at any time period in reactor life, ARTNDT due to the radiation exposure associated with that time period must be added to the unirradiated RTNDT.

The extent of the shif t in RTwor is enhanced

.by~certain chemical elements (such as copper and nickel) present in the reactor vessel steel.

The Nuclear Regulatory Commission (NRC)'

has published a

method-for predicting radiation embrittlement in RG 1.99 Revision 2.

RG 1.99, Revision 2,

is used for the calculation of ART values at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad / base metal interface.

The most limiting ART values are used in the generation of the heatup and cooldown curves.

B-5

ATTACHMENT-B, continusd License Amendment Request No. 127 Page 6 Heatup and cooldown curves. for normal heatup and cooldown have been calculated using the methods outlined in Reference 6.

These limit curves are developed using Code. Case N-640 as noted in Reference 7.-

One of the safety margins incorporated into the curves is the lower bound fracture toughness curve.

The lower bound. fracture toughness curves available in Appendix G

to Section XI uses the reference stress intensity factor Kra and is the basis for the current heatup and cooldown curves.

The proposed heatup and cooldown curves are based on Code Case N-640 and use the reference stress intensity factor Kn is a K c.

I I

fracture toughness curve which is a lower bound on all static, dynamic and arrest fracture toughness, and Kc is a fracture I

toughness curve which is a

lower bound on static fracture toughness only.

The only change in generating the revised heatup Kc is the lower bound fracture and cooldown curves with I

toughness curve selected.

All other margins involved in the generation process remain unchanged.

Since the heatup and cooldown process is very slow, with the fastest rate allowed being 100 F per hour, the rate of change of pressure and temperature is often constant so that the stress is essentially constant.

Both heatup and cooldown correspond to static loading, with regard to fracture toughness.

The only time when dynamic loading can occur and where the dynamic / arrest toughness Kra should be used for the reactor pressure vessel is when a crack is running.

This may occur during a pressurized thermal shock event, but not during heatup and cooldown.

Therefore, the static Xc lower bound toughness is used to generate the toughness I

heatup and cooldown curves.

The primary reason for incorporating this change in methodology is to reduce the excess conservatism in the current Appendix G approach that could, in fact, reduce overall plant safety.

By changing from Kra to K c methodology, I

the operating window relative to the RCP seal requirements is wider and the chances of damaging the RCP seals and initiating a small break

LOCA, a

potential pressurized thermal shock initiator, are reduced.

The pressure difference between the wide-range pressure transmitter and the limiting beltline region has not been accounted for in the heatup and cooldown curves generated for normal operation.

The pressure difference, however, has been incorporated into the generation of the OPPS curve and, since the OPPS setpoint curve is more conservative than the heatup and cooldown curves, and is used for plant operation, as long as the OPPS. curve limits are satisfied, the heatup and cooldown limits are not exceeded.

Inserts (with a letter designation; i.e.,

A, B, etc.) to be added to the marked-up pages are provided in Attachment A.

Insert (A) presents the heatup curve with no margin for instrumentation error using heatup rates up to 60'F/hr applicable for the first 15 EFUY.

Insert (B) presents the cooldown curve with no margin for instrumentation error up to 100*F/hr applicable for the first 15 EFPY.

Allowable combinations of temperature and pressure for B-6

ATTACHMENT B, continusd License Amendment Request No. 127 Page 7 specific temperature change rates are below and to the right of the limit lines shown in the new curves.

This is in addition to other criteria which must be satisfied prior to reactor criticality, as discussed in the following paragraphs.

Acceptable P/T limitations for reactor criticality are provided to the right of the criticality limit line shown on the new Figure 3.4-2.

The straight-line portion of the criticality limit is at the minimum permissible temperature of the 2485 psig inservice hydrostatic test as required by Appendix G to 10 CFR Part 50.

The governing equation for the hydrostatic test is defined in Appendix G to Section XI of the ASME Code.

The criticality limit curve specifies P/T limits for core operation to provide additional margin during actual power production as specified in Reference 8.

The P/T limits for core i

operation (except for low power physics tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown calculated as described in Reference 6.

For the heatup curve identified in Insert (A), the minimum temperature for the inservice hydrostatic leak tests for the reactor vessel at 15 EFPY is 19 6*F.

The vertical line drawn from this point on the P/T curve, intersecting a curve 4 0*F higher than the P/T limit curve, constitutes the limit for core operation for the reactor vessel.

The P/T limit curves define the above limits for ensuring prevention of heatup and cooldown nonductile failure for the reactor vessel.

l Technical Specification 3.4.9.3 requires that OPPS be operable with a maximum of one charging pump capable of injecting into the RCS and the accumulator isolated and either:

a)

Two power-operated relief valves (PORVs) with nominal maximum lift settings which vary with the RCS temperature l

and which do not exceed the limits established in Insert (C), or b)

The RCS depressurized and an RCS vent of greater than or equal to 3.14 square inches.

The requirements are applicable to Mode 4 when any RCS cold leg temperature is less than or equal to an enable temperature of 350 F, Mode 5,

and Mode 6 when the reactor vessel head is on.

With respect to the charging pumps, a footnote is provided.

The footnote states that two pumps are allowed to be capable of RCS injection for less than or equal to 15 minutes for pump swaps.

This note also allows all charging pumps capable of injecting into the RCS during a change from Mode 3 to Mode 4 to be operable for a limited period of time.

This footnote existed prior to the I

proposed change to OPPS and is not affected by this change.

With l

respect to the accumulator, a footnote is provided to state that accumulator isolation with power removed from the discharge B-7

ATTACHMENT B, continusd License Amendment Request No. 127 Page 8 isolation valves is only required when the accumulator pressure is greater than or at the maximum RCS pressure for the existing temperature, as allowed by the P/T limit curve.

This footnote is provided to ensure that the accumulator discharge isolation valve surveillance is only performed under the P/T conditions allowed by the P/T limit curve.

The OPPS curve was derived from the steady state heatup and cooldown curves at 15 EFPY.

The setpoints are applicable to 30%

steam generator tube plugging.

The setpoints conservatively account for instrument uncertainties associated with the wide range pressure (99 psi) and wide range temperature (17*F)

)

instrumentation, per Reference 9.

The setpoints conservatively i

account for the pressure difference, between the wide-range pressure transmitter and the reactor vessel limiting beltline

region, identified in Nuclear Safety Advisory letter NSAL 005A.

The pressure differences according to Reference 9 are 0,

17, 30, and 47 psi for 0, 1,

2, and 3 RCPs running, respectively.

The minimum value of either the adjusted Appendix G limit, after subtracting the delta-P, or the 800 psig PORV piping limit, forms the limit used as the basis for setpoint selection.

Heat transport effects, which are applied to the heat injection i

transient results and account for a 50 F difference between the wide range temperature sensor and the reactor vessel, have also j

been incorporated.

Pressure overshoots during the design basis events are based on a pressurizer PORV stroke open/close time of 1.65/1.0 seconds.

The heatup and cooldown curves are generated using the Kc fracture toughness methodology; therefore, ASME I

Code Case N-514 is not applicable and Code Case N-640 is used.

Setpoints are selected so that RCS pressures will not exceed the 15 EFPY Appendix G pressure limits down to a reactor vessel temperature of 60*F.

The design basis transient for determination of the OPPS setpoints are both the mass injection event caused by the failure of the controls for a single charging pump to go to the full flow condition and the heat addition scenario in which an RCP in a single loop is started when the RCS temperature is as much as 50*F lower than the steam generator secondary side temperature.

The heat addition transient results in a sudden secondary to primary heat transfer and rapid increase in primary system pressure.

The heat addition event has been analyzed for RCS temperatures between 70 F and 3 0 0*F.

The influx of fluid into the relatively inelastic RCS during the mass injection event also causes a rapid increase in system pressure. The RCS temperature assumed for the mass injection event is 70 F.

For any particular design basis transient, the PORV will be signaled to open at a specific pressure setpoint for a given temperature.

However, there will be a pressure overshoot during the delay time before the valve starts to move and during the time the valve la moving to the full open position.

This overshoot is dependent on the dynamics of the system and the input parameters and results in a maximum system pressure somewhat higher than the set pressure.

B-8

l

)

ATTACHMENT B, continued l

Licanza Am2ndmsnt Raquest No. 127 Page 9 Similarly there will be a pressure undershoot, while the valve is relieving, both due to the reset pressure being below the setpoint and due to the delay in stroking the valve closed.

In l

order.to preserve the single failure criteria, the overshoots are calculated assuming the availability of one PORV during the design basis mass injection and heat addition events, when the RCS is water solid, concurrent with a loss of letdown and isolation of the RHRS.

The second PORV is assumed to have failed.

The maximum and minimum pressures reached in the transient are a function of the selected setpoint and fall within an acceptable pressure range.

l The upper pressure limit 'for OPPS is defined by Appendix G requirements or on PORV piping limitations, after consideration l

of uncertainties and the delta-P between the wide range pressure transmitter and the reactor vessel limiting region.

The lower limit on pressure during the design basis OPPS mass injection and l

heat addition transients is established based on operational consideration for the RCP number one seal limit which requires a nominal differential pressure across the seal faces for proper film-riding performance.

As part of the OPPS setpoint evaluation, margin to the RCP number one seal limit is evaluated.

Based on the analyses of record, this limit corresponds to a differential pressure across the seal of 200

psid, which corresponds to the gage pressures provided in Reference 9.

As demonstrated in Reference 9,

pressure undershoot below the PORV l

setpoint during a design basis mass injection or heat addition event can exceed 100 psi.

Therefore, with the PORV setpoints developed for the 15 EFPY heatup and cooldown curves, there is the potential for RCS pressure to violate the RCP number one seal j

limit at the lowest RCS temperatures.

l While analysis has not been performed that models the simultaneous relief from two PORVs, undershoot below the PORV setpoint can be significantly higher if both PORVs actuate during an OPPS event, and it is anticipated that the pump seal limit would be exceeded.

However, staggering the setpoints minimizes the likelihood that both PORVs will actuate simultaneously during credible OPPS events.

Similarly, WCAP 14040-NP-A (Reference 2) indicates that when there is insufficient range between the upper and lower pressure limits to select PORV setpoints that provide l

protection against violating both

limits, then the setpoint J

selection that provides protection against the upper limit violation takes precedence.

The final Maximum Allowable Nominal Setpoints are defined by a curve which passes through the most limiting points determined l

for the mass injection and heat addition

events, after consideration of applicable instrumentation uncertainties and i

heat transport effects.

Breakpoints are selected as input to the function generator so that setpoints calculated by the function generator protect the steady state Appendix G limits.

These B-9

r ATTACHMENT B, continusd Lican:o Am:ndmsnt R: quat No. 127 Page 10 l

breakpoints represent the higher PORV setpoints.

Since actuation l

of both PORVs can result in excessive undershoot below the PORV i

setpoint, the lower PORV setpoints are staggered by an amount l

greater than or equal to the limiting overshoot (from either the mass injection or heat addition events) at setpoint temperature.

l In order to prevent PORV actuation should the OPPS be l

inadvertently armed beyond a temperature of 367*F, the setpoints ramp linearly to a maximum pressure of 2335 psig at 425 F.

E.

NO SIGNIFICANT HAZARDS EVALUATION l

The no significant hazard considerations involved with the proposed amendment have been evaluated.

The evaluation focusing on the three standards set forth in 10 CFR 50.92(c) are 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 proposed heatup and cooldown curves have been revised by changing the applicability from 10 effective full power years (EFPY) to 15 EFPY.

The curves have been developed in accordance with the methodology provided in Regulatory Guide 1.99 Revision 2 and Code Case N-640.

The proposed heatup and cooldown curves define limits that still ensure the prevention of nonductile failure for the reactor vessel.

The design basis events that were protected against have not changed; therefore, the probability of an accident is not increased.

f The overpressure protection system (OPPS) has been revised such that the applicability has changed from 10 EFPY to 15 EFPY.

This system protects the Reactor Coolant System i

B-10 i

ATTACHMENT B, continutd Lican2s Amsndannt Rsquast No. 127 Page 11 (RCS).at low temperatures so that the integrity of the Reactor' Coolant Pressure Boundary (RCPB) is not compromised by violating the pressure / temperature (P/T) limits.

These changes were determined in accordance with the methodologies 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 associated with postulated faulted events.

The lower limit on pressure during-the design basis OPPS mass injection and heat addition transients is established based on operational consideration for the. RCP number one seal limit which requires a nominal differential pressure across the seal faces for proper. film-riding performance.

As part of the OPPS setpoint evaluation, margin to the RCP numbe'r one seal

-limit is evalua*ad.

This limit corresponds to a differential pressure across the seal of 200 paid, which corresponds to the gage pressures.

The pressure undershoot below the PORV setpoint 'during a design basis mass injection or heat addition event can exceed 100 psi.

Therefore, with the PORV setpoints developed for the 15 EFPY heatup and cooldown curves, there is the potential for RCS pressure to violate the RCP number one seal limit at the lowest RCS temperatures.

Undershoot below the PORV setpoint can be significantly higher if-both PORVs actuate during an OPPS event, and it is anticipated that the pump seal limit would be exceeded.

However, staggering the setpoints minimizes the likelihood that both PORVs will actuate simultaneously during credible OPPS events.

Similarly, WCAP 14040-NP-A indicates that when there is insufficient range between the upper and lower pressure limits to select PORV setpoints that provide protection against violating both limits, then the setpoint selection that provides protection against the upper limit violation takes precedence.

WCAP-14040-NP Revision 1 was approved by the NRC by letter dated October 16, 1995, which was incorporated in Revision 2 of the approved WCAP issued in January 1996.

Modification of the heatup and cooldown curves and OPPS setpoints does not alter any assumptions previously made in the radiological consequence evaluations nor affect mitigation of the radiological consequences of an accident described in the Updated Final Safety Analysis Report (UFSAR).

Therefore, the proposed changes will not significantly increase the probability or consequences of an accident previously evaluated.

B-11

)

ATTACHMENT D, continu d Lic nso Am:ndssnt Rsquant No. 127 Page 12 2.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed heatup and cooldown curves applicable for the first 15 EFPY were generated using approved methodology and Code Case N-640.

Generating these curves with Code Case N-640 reduced the excess conservatism that exists in the current curves and results in an increase in the safety of the plant, as the likelihood of RCP seal failures and/or fuel problems will decrease.

The change does not cause the initiation of any accident nor create any new single failure.

The modification of the OPPS setpoints ensures that the RCPB integrity is protected at low temperatures.

The new setpoints were selected using conservative assumptions to ensure that. sufficient margin is available to prevent violation of the P/T limits due to anticipated mass and heat j

input transients.

The modification of the setpoints does not i

change, degrade, or prevent the safe response of the RCS to accident scenarios, as described in UFSAR Chapter 15.

The proposed change does not cause the initiation of any accident nor create any new credible single failure.

4 i

Therefore, the proposed licanse amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Does the change involve a significant reduction in a margin of safety?

4 The new P/T curves define the limits for ensuring prevention of_nonductile failure for the reactor vessel, and does not significantly reduce the margin of safety for the plant.

The methodology provided in Code Case N-640 removed some of the excess conservatism from the current Appendix G

analysis.

However, this improved overall plant safety by expanding the operating window relative to the RCP seal requirements.

The probability of damaging the RCP seals is reduced.

Therefore, the margin of safety is not significantly reduced.

The OPPS setpoints will continue to ensure the RCS pressure boundary will be protected from pressure transients.

They were generated using the proposed heatup and cooldown curves as input.

The OPPS setpoints include additional margin by including instrument uncertainties not included in the current setpoints.

Therefore, the margin of safety is not significantly reduced.

B-12

ATTACHMENT B, continund Lic;nO@ Amsndmsnt Rsqusrat No. 127-Page 13 F.

NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the considerations expressed above, it is concluded that the activities associated with this license amendment request satisfy the requirements of 10 CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified.

G.

ENVIRONMENTAL CONSIDERATION This. license amendment request changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.

It has been determined that this license amendment request involves no significant increase in the amounts, and no significant change in the types of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

This license amendment request may change requirements with respect to installation or use of a facility component located within the restricted area or change an inspection or surveillance requirement; however, the category of this licensing action does not individually or cumulatively have a significant effect on the human environment.

Accordingly, this license amendment request meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c) (9).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this license amendment request.

H.

UFSAR CHANGES No UFSAR changes to implement this License Amendment Request are required.

REFERENCES 1.

Regulatory Guide 1.99, Revision 2,

" Radiation Embrittlement of Reactor Vessel Materials,"

U.S.

Nuclear Regulatory Commission, May 1988.

2.

WCAP 14040-NP-A, Rev.

2,

" Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J.

D. Andracheck et al, January 1996.

3.

WCAP-9615, Revision 1,

"Duquesne Light Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program,"

P.

A.

Peter, June 1995.

4.

Section XI of the ASME Boiler and Pressure Vessel Code, Appendix G,

" Fracture Toughness Criteria for Protection Against Failure."

5.

WCAP-14485, " Beaver Valley Unit 2 Heatup and Cooldown Curves for Normal Operation," P. A. Grendys, March 1996.

B-13

ATTACHMENT B, continund I

License Amendment Request No. 127 Page 14 6.

WCAP-15139,

" Beaver Valley Unit 2 Heatup and Cooldown Limit Curves During Normal Operation at 15 EFPY Using Code Case N-626,"

T. J.

Laubham, January 1999.

7.

EDRE-SMT-98-135,

" Technical Basis for Revised P-T Limit Curve Methodology," W.

H.

Bamford et al, December 1998.

8.

Code of Federal Regulations, 10 CFR Part 50, Appendix G,

" Fracture Toughness Requirements,"

U.S.

Nuclear Regulatory Commission, Washington, D.

C.,

Federal Register, Volume 60, No. 243, dated December 19, 1995.

9.

NPD-OPES (99)-055, " Low Temperature Overpressure Protection System Setpoint Review for Beaver Valley Unit 2 15 EFPY Heatup and Cooldown Curves," March 1999.

B-14

ATTACHMENT C Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 127 UPDATE HEATUP AND COOLDOWN CURVES AND OVERPRESSURE PROTECTION SYSTEM (OPPS) SETPOINT CURVE I

WCAP-15139 i

Beaver Valley Unit 2 Heatup and Cooldown Limit Curves During Normal i

Operation at 15 EFPY Using Code Case N-626 (January 1999) l l

i

)

l

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