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Proposed Tech Specs Bases 3/4.4.9 Re Pressure/Temp Limits
	ML20077Q308
Person / Time
Site:	Beaver Valley
Issue date:	08/12/1991
From:	; DUQUESNE LIGHT CO.
To:	;
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ML20077Q301	List: ML20077Q298; ML20077Q308; ML20077Q324;
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	NUDOCS 9108210187
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_ ILV-l Technical Snocification Bases 3/4.4.9 ItoviS19D i Bemove Pacita Insert Pagga l I

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. iEARTolLCQ0LANT SYSTE

D\SES vessel inside radius are essentially identical, the measured transition shift for a cample can be applied with confidence to the adjacent section of the reactor vescel. The heatup and couldown curves must be recalculated when the RT NDT determined from the surveillance capsule is different from the calculated RT NDT f r the equivalent capsule radiation exposure.

The pressare-temperature limit lines shown on Figure 3.4-2 tor 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 for reactor criticality and for inservice leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in UFSAR Table 4.5-3 to assure compliance with the lequirements of Appendix 11 to 10 CFR 50.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressu-izer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

Pressure-temperature limit curves shown in Figure B 3/4 4-3 were developed for the limiting forritic steel component within an isolated reactor coolant loop. The limiting component is the steam generator channel head to tubesheet region. This Iigure provides the ASME III, Appendix G limiting curve which is used to define operational bounds, such that when operating with an isolated loop the analyzed pressure-temperature limits are known. The temperature range provided bounds the expected operating range for an isolated loop.

The OPERABILITY of two PORV's or an RCS vent opening of greater than 3.14 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CPR Part 50 when one or more of the RCS cold legs are s 275"F.

Either PORV has adequate relieving capability to protect the RCS from over-pressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator s 25*F above the RCS cold leg temperature or (2) the start of a charging pump and its injection into a water solid RCS.

3/4. 4 el0 STRUCIUBAL IRIEGRITl The innervice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50. 55a (g) (6) (i) . i BEAVER VALLEY - Unit 1 B 3/4 4-10

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@ 1000 ---

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

FIGURE B 3/4 4-3 ISOLATED LOOP PRESSURE-TEMPERATURE LIMIT CURVE BEAVER VA!_ LEY UNIT 1 B 3/4 4-10a PROPOSED

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. HEACJOR COOLAN'I fiYSTEM BASES n- ._ _

-- - -...-.-- = = =

3/4.4.11 RELIEF VALVES The relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

2/4.4.12 REACTOR COOLATE SYSTIM VillHis Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam f rom the primary systen that cc,uld inhibit natural circulation core cooling. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space, ensures the capability exists to perform this function.

The valve redundancy of the reactor coolant system vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring dat a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path.

The function, capabilities, and testing requirements of the reactor coolant system vent systens are consistent with the requit3ments of Item II.B.1 of NUREG-0737, " Clarification of TMI Action Pli.n Requirements", November 1980.

BEAVER VALLEY - UNIT 1 B 3/4 4-11 j

.' BV-2 Technical Spec i f icat io_D_IlD se s 3 / 4 . 4 . 9 rey 15.l_QD Reinove Paces Insert Ea.geq m B 3/4 4-14 B 3/4 4-14 B 3/4 4-14a B 3/4 4-15 B 3/4 4-15 i

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REACTOR COOLANT SYSTEM

' BASES.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS (Continued)

The pressure-temperature limit lines shown on Figure 3.4-2 for roactor criticality and for inservice leak and hydrostatic testing have .been provided to assure compliance with the minimum temperature requirements of- Appendix G to 10 CFH 50 for reactor criticality and for-inservice _ leak and hydrostatic testing.

The number of reactor vessel irradiation surveillance specimens and the- frequencies for removing and testing these specimens are provided in UFSAR -Table 5.3-6 to assure compliance with the requirements of

' Appendix H to 10 CFR Part 50.

.The limitations -imposed-on the pressurizer heatup and cooldown rates and auxiliary spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

Precsure-temperature limit curves shown in figure B 3/4 4-3 were developed ~for the limiting ferritic steel component within an isolated reactor coolant loop. The limiting component is the steam generator channel head to tubesheet region. This figure provides the L ASME_ _ III, Appendix G limiting curve which is used to define :

L operational bounds, :such that when operating with an isolated loop the- analyzed pressure-temperature limits are known. The temperature range provided bounds the expected operating range for an isolated loop.

The--OPERABILITY of tuo PORVs or an RCS vent opening of greater than-3.14 . 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 s 350"F.

Either PORV-has adequate relieving capability to protect the RCS from L overpressurization when 'the transient is limited to either (1) the start- ~ of an idle RCP with the secondary water temperature of the steam generator s 50*F above the RCS cold leg temperature or (2) the start of a charling pump and its injection into a water solid RCS, OVERPRESSURE-PROTECTION SYSTEMS The Maximum Allowed PORV Setpoint for the Overpressure Protection LSysthms (OPPS) is derived by analysis which models the performance of ther OPPS- assuming various mass input and heat input transients.

Operation with a. PORV -setpoint less .than or equal to the maximum

-setpoint ensures .that nominal 10 EPPY Appendix G limits will not be violated with consideration for: (1) a maximum pressure overshoot boycnd_ the PORV satpoint which can occur as a result of time delays in signal processing- and valve opening; (2) a 50*F heat transport Jeffect imade possible by the geometrical relationship of_the RHR cuction line and the RCS wide range temperature indicator used for OPPS;. (3) instrument uncertainties; and (4) single failure.

BEAVER VALLEY - UNIT 2 B 3/4 4-14 I

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2500 2000 6 1500 k

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[ 1000

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500 0

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

FIGURE B 3/4 4-3 ISOLATED LOOP PRESSURE-TEMPERATURE LIMIT CURVE BEAVER VALLEY UNIT 2 B 3/4 4-14a PROPOSED

. BEACTOR C_QO_LANT SYSTEM BASES OVEPPRSSURE PROTECTION SYSTEMS (Continued)

To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of 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 coolant temperature is more than 50*F above reactor coolant temperature. Exceptions to these requirements are acceptable as described below.

Operation above 350*F but less than 375'F with only one centrifugal charging pump OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . As shown by analysis LOCAs occurring at low temperature, low pressure conditions can be successfully mitigated by the operation of a single centrifugal charging pump and a single LHSI pump with no credit for accumulator injection. Given the short time duration 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 charging pump is not assumed.

Operation below 350*F but greater than 325'F with all centrifugal charging pumps OPERABLE is allowed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . During low pressure, low temperature operation all automatic Safety Injection actuation signals are blocked. In normal conditions a single failure of the ESP actuation circuitry will result in the starting of at nost one train of Safety Injection (one centrifugal charging pump, and one LHSI pump). For temperatures above 325'F, an overpressure event occurring as a result of starting these two pumps can be successfully mitigated by operation of both PORVs 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 overpressure 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 are not considered to be credible accidents.

The maximum allowed PORV setpoint for the Overpressure Protection 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 and in accordance with the schedule in UFSAR Table 5.3-6.

3Z4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (i) .

BEAVER VALLEY - UNIT 2 B 3/4 4-15

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ML20077Q308

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