ULNRC-03664, Proposed Tech Specs 3/4.4.9 Re Pressure/Temp Limits
| ML20217K243 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 10/17/1997 |
| From: | UNION ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20217K233 | List: |
| References | |
| ULNRC-03664, ULNRC-3664, NUDOCS 9710240082 | |
| Download: ML20217K243 (26) | |
Text
.
ULNRC-03664 ATTACHMENT 1 I
TECHNICAL SPECIFICATION CHANGES Current Figure 3.4-2 (Page 3/4 4-30)
Replacement Figure 3.4-2 (New Page 3/4 4-30)
Current Figure 3.4-3 (Page 3/4 4-31)
Replacement Figure 3.4-3 (New Page 3/4 4-31)
Current Figure 3.4-4 (Page 3/4 4-36)
Replacement Figure 3.4-2 (New Page 3/4 4-36)
Current Bases Page B 3/4 4-7 Current Bases Page B 3/4 4-8 Current Bases Page B 3/4 4-15 Current Bases Page B 3/4 4-16 Current Bases Page B 3/4 5-2 ITS Bases Page B 3.4-60 ITS Bases Page B 3.4-71
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Applicable for the First 20 EFPY (With Margins for instrurnentation Errors)
Induces Vessel lange requremeres of 170*F and 561 psig per 10CFR50. Appendu G Canaway Unit i Heatup and Cooldown Lirnit Curves for Normat Ooeration 7/97
Material trenerty sar s 1/4" Li;iting Material: Plate, R2708-3 3/4T Liciting Material: Plate, R2706-1 Copper Centent: 0.07 wt. s Copper Centent: 0,07 wt 5 '
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0 50 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.F) l FIGURE 2 Callaway Unit i Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,20,40, 60 and 100*F/hr) Applicable for the First 20 EFPY (With Margins for Instrumentation Errors) traunes veessi sange reguremenis of 170*F and 561 psig per 10CFR50. Appenda G.
Cataway Unit 1 Heatup and Cooldown Umit Curves for Normal Operstion 7/97
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REACTOR COOLANT SYSTEM 4 BASES PRESSURE / TEMPERATURE LIMITS (Continued)
- 2. These limit lines shall be calculated periodically using methods provided below.
- 3. System preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler
, and Pressure Vessel Code,Section XI.
The '-ecture +^t;h e:: prei,e, tie; :f the f:rriti: ::terial: it, th: r;;.;ter vessel e-a dater = h 9 ia screrd =:: .;ith th 107: ',li,,te, ;,dde-de te 0;; tier, 1!!! ef the ASME Bethe =d nretter: Y ::el Cbd:.
- 2. 0 Heatup and cooldown limit curves are calculated using 11 g value of the nil-ductility reference temperature, RT the end of effective full power years (EFPY of service life. Ne, EFPY service life period is chosen such that thei l)miting RTat the 1/4T location in the core region is greater than the RT of the limning unirradiated material. The selection of such a limiting k assures that all components in the Reactor Coolant System will be operated c,onservatively in accordance with applicable Code requirements. n; cme The reactor vessel materials have been tested to d ermine their initial
', the results of these tests are shown in Table 3/4.4-1. Reactor RT.,;
opera tion and resultant fast neutron (E greater th 1 HeV) irradiation can cause an increase in the RT 7 Therefore, an mj ted reference temperature, based upon the fluence and copper content and phe_;h:re; content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART computed by either Regulatory Guide 1.99, Revision 2, ' Effects of Residual Nements on Predicted Radiation Damage to Reactor Vessel Materials," or the Westinghouse Copper Trend Curves shown in Figure B 3/4.4-2.
The heatup and cooldown limit curves of Figures 3.4-2 and 3 include-predicted adjustments for this shift in RT., at the end of EFPY as well as adjustments for possible errors in the pressure and tempe sensing instruments.
V:lue: ef' 19J d:t:-in:d in thi: ==n:7 =y b e;ed ;.atil the result;
' #,trve!11==
- pr:;r=, =:.1;eted ;;cerding t: AST", CIO;, ;re
=:i hth . Capsules v!" b; removed in accordance with the requirements of ASTM E185-73 and 10 CFR art 50, Appendix H. The lead factor-represents the rc CALLAWAY - UNIT 1 B 3/4 4-7 Amendment No. 2', 75,103
BASES PRESSURE / TEMPERATURE LIMITS (Continued) relationship between the fast neutron flux density at the location of the capsule and the inner wall of the reactor vessel. Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule. exceeds the. calculated ART NDT for the HDT equivalent capsule radiation exposure.
Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in W *^*2^ A AMAA- ltl The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology. In the calculation procedures a semi elliptical surface defect with a depth of one-quarter of the wall thickness, T, and a length of 3/2T is -
assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall. The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques. Therefore, the reactor l operation limit curves developed for this reference crack are conservative and provice sufficient safety margins for protection against nonductile failure.
To assure that the radiation embrittlement effects are accounted for in the j '
calculation of the limit curves, the most limiting value of the nil-ductility l
reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresp nding to the end of the period for which heatup and cooldown curves are generated.
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K
g, for the combined thermal and pressure stresses at any time during heatup nr cooldown cannot be greater than the reference stress intensity factor, K gg, for the metal temperature at that time. K gg is obtained from the reference f racture toughness curve, defined in Appendix G to the ASME Code. The K curve is given by the equation: IR KIR = 26.78 + 1.223 exp [0.0145(T-RTNOT + 160)) (1)
Where: K yp is the reference stress intensity factor as a function of the metal temperature 1 and the metal nil-ductility reference temperature RT Thus, NDT.
CAI1AWAY - UNIT 1 B 3/4 4-8 Amendment No. 76 S
- 4 REACTOR COOLANT SYSTEM BASES EME (Continued)
The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.
Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective :urves. l The OPERABILITY of two PORVs two RHR suction relief valves, one RHR suction relief valve and one PORV, or an RCS vent opening of at least 2 rquare inches ensures that the RCS will be protected from pressure transients wnich could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 368',F. Either PORV or either RHR suction relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either: (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50'F above the RCS cold leg temperatures, or (2) the start of a centrifugal charging pump and its injection into a water-solid RCS.
a6 As nwoul cA4Wsw
't In addition to opening 7c5 vents to meet tne requirement of Specification 3.4.9.3c., it is acceptable to remove a pressurizer Code safety valve, open a PORY block valve and remove power from the valve operator in conjunction with disassembly of a PORY and removal of its internals, or otherwise open the RCS.
COLD OVERPRESSURE The Maximum Allowed PORY Sotpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance of the CONS assuming various mass input and heat input transients. Operation with a PORY ;
l setpoint less than or equal to the maximum setpoint ensures that Appendix G criteria will not be violated with consideration for 1) a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing and valve opening; 2) a 50'F heat transport effect made CALLAWAY - UNIT 1 B 3/4 4-15 Amendment No. 4 , 83, 103
RLACTOR COOLANT SY51[M BASES COLO OVERPRES$URt
- - (Continued) /A/
possible by the geometrical relationship of the RHR suction line and t
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f"Y wide range temperature indicator used for CONS; 3) instrument uncertai ties; and 4) single failure. To ensure mass and heat input transients more evere than those assumed cannot occur, technical specifications require loc ut of bnth safety injection pumps and all but one centrifugal charging pump while in MODE 5 4, 5 and 6 with the' reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 50'F above primary temperature.
Exceptions to these mode requirements are acceptable as deu H had beinw.
b eer W / o 4 A T b / W a d Operation above 350*F but less than 375'F withAoniy one 6ensiivgai charging-pump OPERABLE and no safety injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
As shown by analysis LOCA's occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single RHR pump with no credit for accumulator injection. Given the short time curation that the condition of having only one centrifugal charging pump OPERABLE is allowed and the probability of a LOCA occurring during this time, the failure of the single centrifugal chargingj ump _is not ass - 1 UkssawdC % N W W=* &
Operation below 350*F but greater than 325'F witnpa si centet Nga~i charging
.ind safety injection pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. During low Prv.3 Cure, low temperature operation all automatic safety injection actuation signals except Containment Pressure - High are blocked. In normal conditions a
- i. ingle f ailure of the ESF actuation circuitry will result in the starting of at mobt one train of safety injection (one centrifugal charging pump, and one safety e injection pump). For temp'eratures above 325'F, an overpressure event occurring as a result of starting-two pumps can be successfully mitigated by operation of both :
PORV't, without exceeding Appendix G limit. Given the short time duration that this condition is allowed and the low probability of a single failure causing an nverpressure event during this time, the single failure of a PORV is not assumed.
Initiation of both trains of safety injection during this 4-hour time frame due to operator error or a single failure occurring during testing of a redundant channel dre not Considered to be Credible accidents.
k Althnugh COMS is &ke requireo to befiord 6 MAN OPERABLL P%"
when RC5"Ntemperature is less than ViR*I, operation withaall centrifugal charging pumps and both safety injection pumps OPERABII is acceptable when RCS temperature is greater than 350*F. Should
.in inadvertent safety injection occur above 350*F, a single PORV has sufficient i.iparity tn relieve the combined flow rate of all pumps. Above 350'F, two RCP
.uut all pret.suriter safety valves are required to be OPERABLE. Operation of an RCP climinates llui possibility of a 50*f dif ference existing between indicated
.ind .ntu.il RCS temperature as a result of heat transport effects. Considering instrimient uncertainties only, an indicated RCS temperature of 350*F is suffi-eiently high to allow full RCS pressurization in accordance with Appendix G limitations. Should an overpressure event occur in these conditions, the pres-surizer safety valves provide acceptable and redundant overpressure protection, lhe Maximum Allowed PORV setpoint for the Cold Overpressure Mitigation System u ' bir updated based on the results of examinations of reactor vessel material i radiation surveillance specimens performed as required by 10 CFR Part bu, ppendix H.
l CAllAWAY - UNIT I B 3/4 4-16 Amendment No. 76 IS
EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS (Continued) g The limitation for a : 9 Of one centrifugal charging pumcdto be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE charging pumo to be inoperable in MODES 4 and 5 and in MODE 6 with thE reactor Yessel head on, provides assurance that a mass addi-tion pressure transient can be relieved by the operation of a single PORV or RHR suction relief valve. In addition, the requirement to verify all Safety Injection pumps to be inoperable in MODE 4, in MODE 5 with the water level above the top of the reactor vessel flange, and in MODE 6 with the r eactor vessel head on and with tne water level above the toc of the reactor vessel flange, provides assurance that the mass addition can be relieved by a single PORV or RHR suction relief valve.
With the water level not above the tcp of the reactor vessel flange and with the vessel head on. Safety Injection pumps may be auilacle to mitigate the effects of a lots of decay heat removal during partia'ily drained conditions.
The Surveillance Requirements, which are provided to ensure the OPERABILITY of each component, ensure that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained, the safety analyses make a pumptions with respect to: (1) both the maximum and minimum total system resistance (Z) t,oth the maximum and minimum branch injection line resistance, and (3) the maximum and minimum ranges of poten-tial pump perfonnance. These resistances and ranges of pump performance are used to calculate the maximum and minimum ECCS flows assumed in the safety analyses.
The centrifugal charging pump minimum flow Surveillance Requirement !
provides the absolute minimum injected flow assumed in the safety analyses.
The maximum total system resistance defines the range of minimum flows (including the minimum flow Surveillance Requirement), with respect to pump head, that is assumed in the safety analyses. Therefore, the centrifugal charging pump total system resistance ((P -PRCS)Nd d
) must not be greater than 1.004E-02 ft/gpm2 , where P d is pump discharge pressure in feet, P RCS is RCS pressure in feet, and Q d is the total pump flow rate in gpm.
The safety injection pump minimum flow Suaveillance Requirement provides the absolute minimum injected flow assumed in the safety analyses. The maximum total system resistance defines the range of minimum flows (in'lu h a -
the minimum flow Surveillance Requirement), with respect to pump he* s .s y' is assumed in the safety analyses. Therefore, the safety 1 @ cF .W4 y, RCS I/Od ) must not be greater tM ' A2% T total system resistance ((Pd-P y; ft/gpmt, where Pd is pump discharge pressure in feet P RCS 1'W## '
k in feet, and Qdis the total pumo flow rate in gpm. jj r:
CALLAWAY - UNIT 1 B 3/4 5-2 Amenanent Nu > %A ff
COMS B 3.4.12
~.
h M BASES 4
BACKGROUND With minimum coolant input capability, the ability to provide (continued) core coolant addition is restricted. The LC0 does not require the makeup control system deactivated or the safety injection f (SI) actuation circuits blocked. Due to the lower pressures in the COMS H0 DES and the expected core decay heat levels, the
{ makeup system can provide adequate flow via the makeup control valve. If conditions require the use of more than one centrifugal charging pump for makeup in the event of loss of inventory, then pumps can be made available through manual actions.
. The CONS for pressure relief consists of two PORVs with reduced lift settings, or two residual heat removal (RHR) suction relief valves, or one PORV and one RHR suction relief valve, 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 prevent overpressurization for the required coolant input capability.
The normal charging pump (NCP)inl.ve dra atte/ a.r e
.* : rend:r^d :p bi ef vel in yrsting c 4 g 4' 9te-the RCS. u+ W4at:tr:tive cent cl . een any RCS celd-e
' ^; t+ nw e 15 > 00C^F. This ensur:: that the current CO O craly:10 r cin: M'rd45. ,-//c46g #4,c,f, 4 gegmr a u q ar a.rizrm; m a n,8 9 eg,, au s P ou$reme t7 r '
Dd b' M N /
% A* @ ev/<4epy 9 As designed for the COMS, each PORV is signaled to open if the RCS pressure approaches a limit determined by the COMS actuation logic. The COMS actuation logic monitors both RCS temperature and RCS pressure and determines when a condition not acceptable with respect to the PTLR limits is approached. The wide range RCS temperature indications are 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, i The calculated pressure limit is then compared with the indicated
, RCS pressure from a wide range pressure channel, If the M-T indicated pressure meets or exceeds the calculated value, a PORV fgy is signaled to opn.
k g4y (continued)
?Wt . CALLA'AY PLANT ITS BASES B 3.4 60 5/15/97
v
. COMS
- B 3.4.12 BASES ACTIONS Cul (continued) .
mass input transient reasonable during the applicable H00ES.
This action is needed to protect the RCPB from a low temperature overpressure event and a possible brittle failure of the reactor vessel.
The Completion Time considers the time required to place the plant in this Condition and the relatively low probability of an overpressure event during this time period due to increased operator awareness of administrative control requirements.
SURVEILLANCE SR 3.4.12.1. SR 3.4.12.2. and SR 3.4.12.3 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, a maximum of zero safety injection pumps and a maximum of one centrifugal charging pump are verified to be capable of injecting into the RCS and the accumulator discharge isolation valves are verified closed with power removed from the valve operators (Refs.10 and 11).
Verification that each accumulator is isolated is only required when accumulator isolation is required as stated in Note 4 to the LCO.
! Thesafetylnjectionpumps.none centrifugal charging p MlIbaraina~oWo~THCEDare rendered incapable of injecting into the RCS through removing the power from the pumps by racking the breakers out under administrative control. An alternate method of cold overpressure protection control may be ,
employed using at least two independent means to render a pump i
incapable of injecting into the RCS such that a single failure or single action will not result in an injection into the RCS. This may be accomplished by placing the pump control switch in pull to lock and closing at least one valve in the discharge flow path, or by closing at least one valve in the discharge flow path aM removing power from the valve operator, or by closing at least one manual valve in the discharge flow path under administrative controls.
The Femuency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is sufficient, considering adminiscative controls and other indications and alarms available to the operator in the control room, to verify the required status of the equipment.
> (continued)
CALLAWAY PLANT ITS BASES B 3.4 71 5/15/97 x _J
a 9
ULNRC-03664 ATTACHMENT 2 SIGNIFICANT HAZARDS EVALUATION
ULNRC-03664 Attachment 2:
Page:1 of 6-
=
=SIGNIFICANT HAZARDS EVALUATION l
ThisETechnical Specification (TS) amendment. application 1 _
~
J requests that the plant heatup and cooldown curves and the maximum allowable _PORV__setpoint curve for cold overpressure protection, as found in Technical Specification Figures 3.4-2,_3.4 3, and 3,4-4 be' modified.
These changes are requested to-incorporate information gained from_ Surveillance _ Capsule V,-which was removed during Callaway Refuel 8 in the fall'of i
1996 after 9.85 effective full power years (EFPY) of
< exposure. Capsule V is'the third capsule to be removed from t the reactor vessel in_the continuing surveillance program that monitors the offects of neutron trradiation on the Callaway reactor vessel materials uncer actual plant-operating conditions.
The following is a! discussion of the proposed changes: <
- 1) Figure 3.4-2 is the heatup limitation curve. This curve is revised to reflect the RTmn calculated from the surveillance capsule data. The revised curve is valid-for 20 EFPY. The curve is based on Regulatory Guide-1.99, Revision 2,.and 10 CFR 50.61.
. ;2 ) Figure 3.4-3 is the cooldown limitation curve. This-curve is also revised to reflect the RTmn. calculated for 20 EFPY in the surveillance. capsule report. The curve is also based on Regulatory Guide 1.99, Revision 2, and 10 CFR 50.61. _ __
3)- Figure 3.4-4 is'the maximum allowable PORV setpoint curve-for cold overpressure protection. This curve is
-a) revised to account-for the changes made in the_heatup
-and cooldown--limitation = curves, b) allows for the operation of the Normal Charging Pump, and c) ' accounts for instrument accuracy and other uncertainties.
-4): Technical Specification Bases k.4.9 and h.5.2 through k.5.4 is revised by correcting miscellaneous items and-knBadding discussion of the-Normal Charging Pump.
The modification of Figures 3.4-2, 3.4-3, and 3.4-4 incorporates the RTmn as . determined from Capsule V. _ The
= revisions to the Bases ~are editorial in nature-orLadd discussion and clarification on the use of the Normal
-Charging Pump.
ULNRC-03664 Attcchm:nt 2 Page 2 of 6
Background
The surveillance program for the Callaway Plant reactor pressure vessel material covers the 40-year plant design life and is based on ASTM E185-73, " Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels." Capsule V Charpy V-notch impact and tensile surveillanco specimens were subjected to postirradiation mechanical testing. The postirradiation data from surveillance Capsule V is summarized in WCAP-14895 (attached) .
The heatup and cooldown curves (Figures 3.4-2 and 3.4-3) were calculated using the NRC-approved methods described in NCAP-
, 14040-NP-A, Revision 2 (Reference 2) and meet the requirements of 10 CFR 50, Appendices G and H. The curves
! are applicable for the first 20 EFPY of operation and include margins of 10 'F and 60 psig to allow for possible instrumentation crrors. The changes to these curves are consistent with the design basis as described in the Bases section of the Technical Specifications. The heatup and cooldown curves are calculated in WCAP-14894 (attached).
The PORV Setpoint Curve (Figure 3.4-4) for Cold Overpressure Mitigation System (COMS) is based on the heatup and et91dewn curves. Therefore, anytime the heatup/cooldown curves are revised, the PORV Setpoint Curve and the COMS setpoints must be evaluated and revised if necessary. For the purpose of this submittal, the heatup/cooldown curves were revised such that a revision to the PORV Setpoint Curve was required. The evaluation of Pressurized Thermal Shock is contained in WCAP-14896 (attached).
In Reference 1, we requested an exemption from th?
requirements of 10 CFR 50.60, " Acceptance Criteria for Fracture Prevention for Lightwater Nuclear Power Reactors for Normal Operation". This request for exemption was made in order to apply the guidance of American Society of Mechanical Engineers (ASME) Code Case N-514, " Low Temperature Overpressure Protection," in lieu of those specified by 10 CFR 50, Appendix G. Code Case N-514 was used in developing Figure 3.4-4 per the methodology of Reference 2.
The ASME Working Group for Operating Plant Criteria developed Code Case N-514 as an alternative methodclogy to the safety margin requirements of Appendix G to 10 CFR 50. The Code case provides criteria to determine pressure limits during COMS events that avoid certain operational restrictions, provide adequate margins against failure of the reactor vessel, and reduce the potential for unnecessary activation of the relief valves used for COMS. Code Case N-51t. allows 1
l
ULNRC-03664 Attachmsnt 2 Page 3 of 6 determination of the COMS setpoints such that for COMS events the maximum pressure in the reactor vessel would not exceed 110% of the pressure / temperature (P/T) limits of the existing ASME Appendix G curves. Code Case N-514 has been approved by the ASME Coda Committee and its content has been incorporated in Appendix G of the ASME Section XI and published in the 1993 Addenda and 1995 edition. Code Case N-514 has not been approved for use in Regulatory guide 1.147, " Inservice Inspection Code case Acceptability, ASME Section XI;"
however, it has been included in the Draft Regulatory Guide 1.147 (DG-1050).
The need for implementation of Code Case N-514 at Callaway involves the avoidance of certain operational restrictions associated with low temperature operation of the plant. Use of Appendix G P/T limits to determine the PORV setpoints would result in pressure setpoints within the operating window; consequently, no margin would be available for normal operating pressure surges. Therefore, operating with these limits could result in an unnecessary challenge to the PORVs and cavitation of the reactor coolant pumps during normal operation.
The selection of the pressure setpoints for the PORVs is based on the use of nominal upper and lower pressure limits specified by Appendix G to 10 CFR 50 and the reactor coolant pump shaft seals, respectively. From the standpoint of determining the maximum setpoint and proximity to Appendix G, the mass input mechanisms considered in the analysis involve the operation of a single charging pump at maximum flow and the operation of the normal charging pump at maximum flow with inadvertent isolation of letdown flow and RHR suction relief valves. Inadvertent actuation of a safety injection pump was not explicitly analyzed since its operation is prevented by the Technical Specifications. Plant specific p essure and temperature uncertainties and the pressure difference between the wide range pressure transmitter and the limiting beltline region of the reactor vessel were also factored Anto the analysis.
The heat input mechanism considered for analysis involved a RCS coolant pump startup in one loop with a temperature asymmetry in the RCS, whereby the steam generator is at a temperature 50 *F higher than the rest of the RCS. The magnitude of the temperature difference would normally depend on the previous operation of the plant that allowed the asymmetry to develop. However, 50 F is considered a maximum temperature difference that is controlled in the Technical Specifications, l .
ULNRC-03664 Attachmsnt 2 Page 4 of Evaluation The proposed change to Technical Specifications does not involve a significant hazards consideration because operation l
of the Callaway Plant with this change would not:
- 1. Involve a significant increase in the probability or consequences of an-accident previously evaluated.-
Pressure and temperature limits for the reactor pressure vessel (RPV) are established to the requirements of 10 CFR 50, Appendix G to ensure brittle fracture of the vessel does not occur. This amendment revises the P/T curves in the TS to reflect the material capsric-surveillance results from the sample removed during the 1 ' outage of 1996.
The RPV surveillance capsule contained flux wires for neutron flux monitoring and Charpy V notch impact and tensile test specimens. The irradiated material properties were compared to available unirradiated properties to determine the effect of irradiation on material toughness for the base and weld materials through Charpy testing. Irradiated tensile testing results are compared with unirradiated data to determine the l effect of irradiation on the stress-strain relationship of l the materials.
The P/T curves are modified to reflect the results of the above examination. These curves and their operating limits were generated using the NRC-approved methods described in WCA?-14040-NP-A, Revision 2 and meet the requirements of 10 CFR 50, Appendices G and H as modified by the provisions of ASME Code Case N-514. The new curves therefore represent the latest information available on the state of the reactor vessel materials. The P/T curves are generated for reactor vessel protection against brittle fracture, they do not
-affect the recirculation piping. Accordingly, the probability of occurrence of a design basis Loss of Coolant Accident (LOCA) is not increased. Likewise, no other previously evaluated accident and transients, as defined in Chapter 15 of the Final Safaty Analysis Report are affected by this proposed change to the Callaway P/T curves.
Additionally,.this proposed revision does not affect the design, operation, or maintenance of any safety-related system designed for the mitigation or prevention of previously analyzed events.
Since no previously evaluated accidents or transients are affected by this change, their probability of occurrence and consequences is not increased.
- Attachment 2 Page 5 of 6
- 2. Create the possibility of a new or different kind of accident from any previously evaluated.
Implementing the proposed P/T curves into the TS does not alter.the design or operation of any system or piece of equipment designed for the prevention or mitigation of accidents and transients. As a result, no new operating modes are introduced from which a new type accident becomes possible. Existing systems will continue to be operated per
-present design basis assumptions.
The proposed P/T limits were generated from the evaluation of the material capsule removed during the fall outage of 1996 l using the NRC-approved methods. described in WCAP-14040-NP-A, Revision 2. As a result, these limits include che latest available information on the reactor vessel materials.
Furthermore, they will continue to be monitored per the requirements of-the TS and 10 CFR 50, Appendices G and H.
For the above reasons,-the changes do not create the possibility of a new-type of accident.
- 3. Involve a significant reduction in a margin of safety.
The purpose of the P/T limits is to avoid a brittle fracture of the reactor vessel. As such, material capsules are removed periodically to determine the effects of neutron irradiation on reactor vessel materials. This change to the Callaway curses is proposed to incorporate-the evaluation results of the latest capsule removed during the fall outage of 1996. Accordingly, these curves represent the-latest information available on the reactor vessel materials. Also, the curves were generated using the approved methodologies of 10 CFR 50, Appcndix G.
The Cold Overpressure Mitigation System Curve (Figure 3.4-4) is also revised to reflect exposure dependencies. This curve was generated for 20 EFPY using approved methodologies and reflects the results of this latest material capsule report.
Utilizing the methodology set forth in ASME Section XI, Appendix G, which includes the provisions of Code Case N-514, and 10 CFR 50, Appendices G and H ensures.that proper limits and conservative safety factors are maintained.
The. proposed changes do not affect the evaluation of any FSAR Chapter 15 transient and accident. Furthermore, the proposed change does not affect the operation of systems or equipment important to safety.
_~
' Attechmsnt 2 Page 6 of 6 The Limiting condition for Operation of Specification 3.4.9 will not change. Also, no Technical Specification surveillance or surveillance frequencies are revised as a result of this Technical Specification submittal, besides the fact that the P/T su ve111ance t will now refer to the revised curves. Procedures regarding the monitoring of the P/T limits during reactor startup, cooldown, and leakage costing will not change as a result of this proposed Technical Specification change with respect to frequency of the surveillance ol the methods used to perform the surveillance.
Thub, the P/T limits will continue to be surveilled as before per the same procedures and at the same frequencies.
No other Technical Specificationc are affected by the ,
proposed revision. Tae margin of safety to any 'techn1 cal '
Specification safety limit therefore is not reduced. For the above reasons the new curves do not represent a significant reduction in the margin of safety.
Conclusion Given the above discussions, the proposed changes do not adversely affect or endanger the hcalth or safety of the general public or involve a significant safety hazard.
l l
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4 ULNRC-03664 ATTACliMENT 3 ENVIRONMENTAL CONSIDERATION 1
.= . . . _ . . _ _ _ _
1 ULNRC-03664 Attachm2nt 3 Page 1 of 1 l
I ENVIRONMENTAL CONSIDERATION l
This Technical Specification amendment requests that the plant heatup and cooldown curves and the maximum allowablo PORV ,
setpoint curve for cold overpressure protection, as found in l' Technical Specification Figures 3.4-2, 3.4-3, and 3.4-4 be modified. These changes are requested to incorporate '
information gained from surveillance capsule V, which was removed during Callaway Refuel 8 in the fall of 1996 after 9.85 effective full power- years (EFPY) of exposute. Capsule V is the third capsule to be removed from the reactor vessel in the :
continuing surveillance program that monitors the effects of
- neutron irradiation _or._the callaway reactor vessel materials under actual plant operating conditions.
The proposed amendment _ involves changes with respect to the use >
of facility components located within the restricted area-as defined in 10 CFR Part 20. Unien. Electric has determined that the proposed amendment does not involve
- 1. A significant hazard consideration, as discussed in Attachment 2 of this amendment application;
- 2. A significant change in the types or significant increase in the. amounts of any effluents that may be released offsite;
- 3. A significant increase in. individual or cumulative occupational radiation exposure, ,
Accordingly the proposed amendment meets the eligibility criteria for categorical' exclusion sot forth in.101 CFR 51.22 (c) (9) . Pursuant to 10 CFR Sl.22 (b) no: environmental ,
impact statement or environmental assessment need be prepared in connection'with'the issuance of this amendment.
v,wyp.e---W-M rart-9 woe y*r $w y9FT-**
_ - .= . ..- _ _ . - .. _ . . - - . . . . . . .
ULNRC 3664 ATTACllMENT 4 WCAP-14894 CALLAWAY UNIT 1 IIEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION i
L
ULNRC-3664 ATTACliMENT 5 WCAP-14895 ANALYSIS OF CAPSULE V FROM Tile UNION ELECTRIC COMPANY CALLAWAY UNIT 1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM
4 O
ULNRC-3664 ATTACllMENT 6 WCAP-14896 EVALUATION OF PRESSUR17.ED TilERMAL S110CK FOR CALLAWAY UNIT 1
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