Text

Amend 76 to License NPF-30,revising TS Section 4.4.9.1.2, Figures 3.4-2,3.4-3 & 3.4-4,Tables 4.4-5 & B 3/4.4-1
	ML20126H941
Person / Time
Site:	Callaway
Issue date:	12/24/1992
From:	Hannon J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20126H945	List: ML20126H941; ML20126H954;
References
	NUDOCS 9301050369
	Download: ML20126H941 (15)
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n UNITED STATES y

,j NUCLEAR REQULATORY COMMISSION o

WA&HINoTON, D.C. 30006 j

a UNION ELECTRIC COMPANY CALLAWAY PLANT. UNIT I DOCKET NO. 50-483

[

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 76 License No. NPF-30 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by Union Electric Company--(UE, the licensee) dated December 18, 1991, complies with the standards-and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations-set forth in-10 4

CFR Chapter I; B.

The facility will operate in conformity with. the application, the i

provisions of the Act, and the rules and regulations of the Commission; i

C.

There is reasonable assurance-(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations:

D.

The issuance.of this amendment will: not be. inimical to the common

~

defense.and security or to' the-health ~ and safety of-the public;'

and, E.

The issuance of this amendment is in accordance with:10 CFR Part-~

~

51 of the Commission's regulations and all applicable-requirements have been satisfied..

2.

Accordingly, the_ license.is amended by changes to the= Technical

' Specifications as indicated in the attachment to this, license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-30 is:hereby!

amended to read as follows:.

t 9301050369 921224 DR ADOCK;0500 3-q u-

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(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.

76, and the Environmenial Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. UE shall operate the facility in accordanen with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance. The Technical Specifications are to be implemented within 30 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/t o

John N. Iannon, arrector Project Directorate III-3 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

December 24, 1992

ATTACHMENT TO LICENSE AMENDMENT NO. 76 OPERATING LICENSE NO. NPT-30 DOCKET NO. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change.

The corresponding overleaf pages are provided to maintain document completeness.

REMOVE INSERT 3/4 4-29 3/4 4-29 3/4 4-30 3/4-4-30 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-36 3/4 4-36 B 3/4 4-7 8 3/4 4-7 0 3/4 4-8 B 3/4 4-8 8 3/4 4-11 B 3/4 4-11 B 3/4 4-16 B 3/4 4-16

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RfACTOR C00LAN1 SYSTEM 3/4.4.9 PR[55URE/1(MPIRATURf LIMIT 5 REACTOR COOLANT SYSTEM IMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressuri2er) temperature and pressure shall be limited in accordance with the limit lises shown on figures 3.4-2 and 3.4 3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 100'f in any 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, b.

A maximum cooldown of 100'f in any 1-hour perios, and c.

A maximum temperature change of less than or to;41 to 10'f in any 1-hour period during intervice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABlllTY:

At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the 11mit within 30 minutes; perform an engineering evaluation to determine the ef fects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANDBY within the nest 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,yg and pressure to less than 200*F and 500 psig. respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

$URVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup. cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as reovired by 10 CFR Part 50, Appendix H.

The results of these examina-l l

tions shall be used to update Figures 3.4-2, 3.4-3, and 3.4-4.

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l CALLAWAY - UNll 1 3/4 4-29 Amendment No. 76 i

Material ProDerty Basis 1/4T Limiting Material: Plate, R2708-3 3/4T Limiting Material: Plate, R2708-1 Cop >er Content: 0.07 wt. 5 Cop >er Content: 0.07 wt 5 Niccel Content: 0.59 wt. %

Niccel Content: 0.59 wt. %

Initial RTNOT:

20'F Initial RTNDT 50'F t

Limiting ART after 17 EFPY:

1/47,95'F 3/4T,84'F

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Callaway Unit 1 Reactor Coolant System Heatup Limitations (Heat up rates up to 60'F/hr and 100'F/hr) Applicable for the First 17 EFPY (With Margins 10'F and 60 psig For Instrumentation Errors)

FIGU:'I 3.4-2 CALI.Ak'AY - l' NIT I 3/4 t.-30 Amendment No. 36. 76

m MLierial Pronerty Basis 1/4T Liciting Material: Plate, R2708 3 3/4T Liciting Material: Plate. R2708-1 l

Cop >er Content 0.07 wt. 5 Cup >er Content 0.07 wt. 5 Nic(el Content: 0.59 wt. 5 Nic(el Content 0.59 wt. %

Initial RTNOT:

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1 ScP C00LDOWN CURVCS ACC. CUl0C 1.99. REY.2 flTH WARCIN Callaway Unit 1 Reactor Coolant System Cooldown (Cooldown rates up to 100*F/hr) Limitations Applicable for the First 17 EFPY (WithMargin,s10'Fand60psigForInstrumentationErrors)

FIGL'RE 3.4-3 CALLAk'AY - L' NIT 1 3/4 4-31 Amendment No. 36,76 O

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CALLAWAY - UNIT 1 3/4 4-32 Amendment No. 36,76

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall oe demonstrated OPERABLE by:

a.

Perfonnance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding. valve operation, within 31 days prior to entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE; b.

Perfonnance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and c.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the PORV is being used for overpressure protection.

4.4.9.3.2 Each RHR suction relief valve shall be demonstrated OPERABLE when the RHR suction relief valves are being used for cold overpressure protection as follows:

a.

For RHR suction relief valve 8708B:

By verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that RHR RCS suction isolation valves (RRSIV) EJ-HV-8701B and BB-PV-8702B are open.

b.

For RHR suction relief valve 8708A:

By verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that RRSIV EJ-HV-8701A and BB-PV-8702A are open.

c.

Testing pursuant to Specification 4.0.5.

4.4.9.3.3 TheRCSvent(s)shallbeverifiedtobeopenat.leastonceper-12 hours

when the vent (s) is being used for overpressure protection.

Except when the~ vent pathway is provided with a valve which is locked..

sealed, or otherwise secured in the_ open position, then verify these valves open at least once per 31 days.

CALLAWAY - UNIT 1 3/4 4-35 Amendment No. 42

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FIGURE 3.4 4 -

MAXIMUM ALLOWED PORY SETPOINT FOR THE COLD OVERPRESSURE MmGATION SYSTEM :

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CALLAk'AY - UNIT 1

-3/4 4-36 Amendment No. 36,76 l

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REACIOR C00LAN1 SYSTEM BASE S P &$5URE/lEMPERATURE LIMITS (Continued) 2.

These limit lines shall be calculated periodically using methods provided below.

3.

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

4.

The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200'F/h, respectively.

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 583*F.

5.

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 fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the 1972 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code.

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNOT, at the end of 17 effec-ti"e full power years (EFPY) of service life.

The 17 EFPY service life, oerinri is chosen such that the limiting RT at the 1/4T location in the core region NDT is greater than the RT f the limiting unirradiated material.

The selection NDT of such a limiting RT assures nat all components in W Reactor Coola d NOT System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTNDl; the results of these tests are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature, NDT.

based upon the fluence and copper content and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ARI c mputed by either Regulatory Guide 1.99, Revision 2, " Effects of l

NDT Hesidual Elements 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 3.4-3 include predicted adjust-ments for this shift in RT at the end of 17 EFPY as well as adjustments for l

NDT possible errors in the pressure and temperature sensing instruments.

Values of ART determined in this manner may be used until the results HDT from the material surveillance program, evaluated according to ASTM E185, are available.

Capsules will be removed in accordance with the requirements of ASIM E 185-73 anti 10 CFR Part 50 Appendix H.

The lead factor represents the CALLAWAY - UNii 3 B 3/4 4-7 Amendment No. A6,76

REACTOR COOLANV SYSTEM 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 NDT determined from the surveillance capsule. exceeds the. calculated ART f r the NDT equivalent capsule radiation exposure, Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section !!! 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 WCAP-7924-A, 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 111 as the reference flaw, amply exceed the current capabilities of inservice inspection techniques.

Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide suf ficient safety margins for protection against nonductile failure.

To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, c rresponding to the end of the period for which heatup and cooldown curves are generated.

The ASME approach for Llculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal and prer.sure stresses at any time during heatup g

nr cooldown cannot be greater than the reference stress intensity factor, Kyg, for the metal temperature at that time.

K is obtained from the reference gg fracture toughness curve, defined in Appendix G to the ASME Code.

The K3g curve is given by the equation:

Kgp = 26.78 + 1.223 exp [0.0145(T-RTNOT + 160))

(1)

Where:

K is the reference stress intensity factor as a function of the metal IR l.

temperature T and the metal nil-ductility reference temperature RT

Thus, NDT.

Call AWAY - UNIT 1 B 3/4 4-8 Amendment No. 76

TABLE B 3.4.4-1 (Continued)

S REACTOR VESSEL TOUGHNESS h

50 FT-LB AVG. SPPER SHELF 3E ASME 35 Mil COMP MATERIAL CU P

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FWD **

i COMPONENT CODE TYPE

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10 0 10 20 30 40 50 l EFFECTIVE FULL POWER (YEARS) i FIGURE B3/4.4-1 FAST NEUTRON FLUENCE (E > 1 MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE CALLAWAY - UNIT 1 B 3/4 4-12 i I i i l l I t-

V REAC10R COOLANT SYSTEM BASES COLD OVERPRES$tlRE (Continued) poswihle by the geometrical relationship of the RHR suction line and the RCS wide range temperature indicator used for COMS; 3) instrument uncertainties; and 4) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, technical specifications require lockout of both safety injection pumps and all but one centrifugal charging pump while in MODES 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 described below. Operation above 350'F but less than 375'F with only one centifugal charging pump OPERABLE and no safety injection pumps OPERABLE is allowed for up to 4 hours. 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 shurt time duration that the condition of having only one centrifugal cnarging pump OPERABLE is allowed and the probability of a LOCA occurring during this time, the failure of the single centrifugal charging pump is not assumed. Operation below 350 F but greater than 325'F with all centrifugal charging and utety injection pumps OPERABLE is allowed for up to 4 hours. During low pr w.ure, low temperature operation all automatic safety injection actuation signals enept Containment Pressure - High are blocked. In normal conditions a . ingle failure of the ESF actuation circuitry will result in the starting of at most one train of safety injection (one centrifugal charging pump, and one safety injection pump). For temperatures above 325*F, an overpressure event occurring as a result of starting two pumps car be successfully mitigated by operation of both p0RV's without exceeding Appendix G limit. Given the short time duration that t his condition is allowed and the low probability of a single failure causing an overpressure event during this time, the single failure of a PORY 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 M credible accidents. Although COMS is required to be OPERABLE when RCS temperature is less than M # f. operation with all centrifugal charging pumps and both safety injection pumps OPERABLE is acceptable when RCS temperature is greater than 350*F. Should inadvertent safety injection occur above 350*F. a single PORV has-sufficient an iapacity to relieve the combined flow rate of all pumps. Above 350 F, two RCP and all pressurizer safety valves are required to be OPERABLE. Operation of an RCP climinates the possibility of a 50 f difference existing between indicated and attual RCS temperature as a result of heat transport effects. Considering instrument uncertainties only, an indicated RCS temperature of 350'F is suffi-i ient ly high to allow f ull RCS pressurization in accordance with Appendix G l imi ta t. iuns. Should an overpressure event occur in these conditions, the pres-surizer safety valves provide acceptable and redundant overpressure protection. 1he Maximum Allowed PORV setpoint for the Cold Overnressure Mitigation System will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50. Appendix H. l CAltAWAY - UNIT 1 B 3/4 4-16 Amendment No. 76

REACTOR COOLANT SYSTEM BASES HEATUP(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 curves. Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the f atigue analysis performed in accordance with the ASME Code requirements. The OPERABILITY of two PORVs, or two RHR suction relief valves, or an RCS vent opening of-at least 2 square inches ensures that the RCS will be protected from pressure transients which 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 capabil-ity 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. l In addition to opening RCS vents to meet the requirement of Specifica-tion 3.4.9.3c., it is acceptable to remove a pressurizer Code _. safety valve, open a PORV block valve and remove power from the valve operator in conjunction with disassembly of a PORV and removal of its internals, or otherwise open the RCS. _ COLD OVERPRESSURE The Maximum Allowed PORY Setpoint for the Cold Overpressure Mitigation System (COMS) is derived by analysis which models the performance'of the COMS. assuming various mass input and heat input transients. Operation with a PORV 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. 42 -}}

ML20126H941

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