Text

Proposed Tech Specs Re Overpressure Protection Sys
	ML20043G688
Person / Time
Site:	Beaver Valley
Issue date:	06/12/1990
From:	; DUQUESNE LIGHT CO.
To:	;
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ML20043G686	List: ML20043G685; ML20043G687; ML20043G688;
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	NUDOCS 9006210008
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{{#Wiki_filter:._ r f ATTACfDENT A Revise the Beaver Valley Unit No. 1 Technical Specifications as follows. Remove Pages _I_nsert Paces 3/4 4-27a 3/4 4-27a 3/4 4-27c 3/4 4-27c B 3/4 4-10 B 3/4 4-10 B 3/4 4-11 B 3/4 4-11 t-t. ): 1 l' i k; i.;- ?! 4 1 9006210008 900612ADOCK 05000334 PDR PDC '!i - S: P h,

, ' REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At-least one of the following overpressure protection systems shall be OPERABIE: a. Two power operated relief valves (PORVs) with nominal maximum allowable lift settings which vary with the RCS r temperature and which do not exceed the limits established in Figure 7.4-4, or I b. A reactor coolant system vent of 2 3.14 square inches. APPLICABILITY: When the temperature of one or more of the non-isolated RCS cold legs is s 275'F. ACTION: a. With one PORV inoperable, either restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent l the.RCS' through a 3.14 square inch vent (s) within the next L 12 hours; maintain the RCS in a vented condition until both l PORVs have been restored to operable status. Refer to L Technical Specification 3.4.1.6 for further limitations. l b. With both PORV's inoperable, depressurize and vent the RCS l through a 3.14 square inch vent (s) within 12 hours; maintain the RCS in a vented condition until both PORVs have been 1 os;?ed to OPERABLE status. ] c. !Fa provisions of specification 3.0.4 are not applicable.. l SURVE',U v2CE UJ'FEMENT ,- e -e m- - 4.4.9.i.i tach <0RV shall be demonstrated OPERABLE BY: l l BEAVER VALLEY - UNIT 1 3/4 4-27a PROPOSED

' REACTOR COOLANT SYSTEM BASES vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the RT determined from the NDT surveillance capsule is different from the calculated RT f0F NDT the equivalent capsule radiation exposure. The pressure-temperaturn limit lines shown on Figure 3.4-2 for reactor criticality and for Inservice leak and hydrostatic testing have broM provided to assure compliance with the minimum temperature requi 8aah$5 of Appendix G to 10 CFR 50 for reactor criticality and T for 1&aervice leak and hydrostatic testing. The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provide in UFSAR Table 4.5-3 to assure compliance with the requirements of Appendix H 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 pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. 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 i CFR Part 50 when one or more of the RCS. cold legs are $ 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 5 25'F above the RCS cold leg temperature or (2) the start of a charging pump and its injection into a water solid RCS. The Low Temperature Overpressure Protection System (LTOPS) setpoint curve is provided for single PORV operation applicable to steady-state pressure temperature limits based on NRC Reg. Guide 1.99 Rev. 2 without 1nstrument uncertainty. LTOPS is considered to be a mitigation

system, as opposed to a protection system, and as such, pressure instrument error is not included in curve development, Omission of instrument-error in this application has been previously y

I reviewed by the NRC. The steady state pressure temperature limit l provides the greatest operational flexibility, and has been accepted by the NRC with the justification that most transients occur during isothermal metal conditions. i l 3/4.4.10 STRUCTURAL INTEGRITY l 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 l 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) (1). BEAVER VALLEY - UNIT 1 B 3/4 4-10 PROPOSED

' REACTOR COOLANT SYSTEM BASES 3/4.4.11 RELIEF VALVES l 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. 3/4.4.12 REACTOR COOLANT SYSTEM VENTS Reactor Coolant system Vents are provided to exhaust noncondensible gases and/or steam from the primary system that could 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 that 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 systems are consistent with the requirements of Item II.B.1 of NUREG-0737,

" Clarification of TMI Action Plan Requirements", November 1980. BEAVER VALLEY - UNIT 1 B 3/4 4-11 PROPOSED

.i O' s. t 800 PORV Piping Protection N Z g-APPENDIX G Protection g WO / y Acceptable a Setpoint 3 Range 7 / 1 / j / E 400 RCP Seal Protection 300 I 85 105 125 145 165 185 3)5 225 245 265 RCS TEMPERATURE Wg F) o Figure 3.4-4 Maximum Allowable Nominal PORV Setpoint For the Overpressure Protection System BEAVER VALLEY - UNIT 1 3/4 4-27C PROPOSED

4 X* ATTACHMENT B 4 4' Beaver Valley Power Station, Unit No. 1 O! Proposed Technical Specification Change No. 176 Revision of Technical Specification 3.4.9.3 LTOPS Setooint Based on Temnerature A. DESCRIPTION OF AMENDMENT REQUEST: o The proposed amendment would replace the Low Temperature i Overpressure Protection System (LTOPS) setpoint with a curve 4 based on temperature.- The pressure differential between the current LTOPS setpoint and the minimum pressure required to start the reactor coolant pumps, has created an oper:tional constraint Jh which is unduly limiting the rate at which the plant is able to heat up from cold shutdown and due to inadequate margin, between the LTOPS setting and the pressure required to operate reactor coolant

pumps, can challenge the pressurizer cold overpressure setpoint.

i B. BACKGROUND Overpressure protection for the RCS at Beaver Valley Unit 1 is provided by self-actuated, pressurizer safety valves. 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. The allowable system pressure is significantly less than the design pressure of 2485 psig necessitating additional means to prevent brittle fractures l. of the reactor vessel at temperatures below 275'F. Therefore, overpressure mitigation provisions for the reactor vessel, as: provided' by LTOPS, must be available when the RCS and the reactor vessel are at temperatures below 275'F. e During a normal plant

heatup, the RCS is open to the Residual Heat Removal System (RHRS) and may be operated for a short period o

of time in a water solid mode until a steam bubble is formed in-the pressurizer. The RHRS is provided with self-actuated o power operated relief valves to prevent overpressure caused -either within the system itself or from transients transmitted 4 from the RCS. However, the relief valve setpoint during normal operation is above the 10CFR50 Appendix G limits and cannot be relied upon to provide low temperature overpressure protection in accordance with the criteria which exists in Regulatory Guide 1.99, Revision 2. During these low-temperature, low-pressure operating conditions, the LTOPS is armed and in a ready status to mitigato pressure transients. In determining the

LTOPS, setpoints, no credit is taken for RHR relief valve operation.

When the reactor coolant temperature has increased above about j 275'F, the LTOPS is manually disarmed. L D r i

i ATTACHMENT B ,7 Page 2 'C. JUSTIFICATION r The presently installed Beaver Valley Unit 1 LTOPS features a constant value setpoint program that is independent of temperature. In order to provide increased operating margin, Westinghouse has performed an analysis (Attachment D) to determine acceptable setpoints that can be used as a function of-Reactor Coolant System (RCS) temperature based on the 9.5 Effective Full Power Year (EFPY) pressure-temperature limits of the 10CFRSO Appendix G curves. The temperature-dependent setpoints have been determined for both the current technical specifications and the limits implemented by NRC Regulatory Guide 1.99, Revision 2. During a normal plant cooldown, the LTOPS is manually armed as -the reactor coolant temperature is decreased below 275'F. Note that. at this time there is a steam bubble in the pressurizer and the water level is at the normal level for no-load operation. The RHRS is normally placed in service by opening the suction isolation valves prior to the LTOPS being placed in service. When the coolant temperature has been decreased to 160*F, the steam bubble may be-quenched and the reactor coolant pumps stopped. From this point on in the cooldown, the LTOPS will be-in an active status ready to mitigate pressure transients which i might occur. D. SAFETY ANALYSIS Potential overpressurization transients to the RCS can be caused by either of two types of events to the RCS, mass input or heat input. Both types result in more rapid pressure changes when the RCS is water solid. Specifically, the LTOPS design bases transients are defined as:

1) the mass input transient caused by a

normal charging / letdown flow mismatch after the termination of letdown flow and 2) the heat input transient caused by the L restart of a RCP pump when a temperature asymmetry exists within the RCS due to the continued injection of cold seal injection water. i 1 For a particular mass input' transient to the RCS, the relief valve will be signalled to open at a specific pressure setpoint.

However, there will be a pressure overshoot during the delay time before the valve starts to move and during the time the valve is 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. Similarly, there will be a pressure undershoot, while the valve is relieving, both due to the reset pressure being below 'the setpoint and to the delay in stroking the valve closed. The maximum and minimum pressures reached in the

ATTACHMENT B Page 3 l transient are a function of the selected setpoint and must fall within the acceptable pressure range. Note that the pressure overshoot and undershoot for the mass input case is greatest at low temperatures. Thus, the overshoot calculation is limited to the most restrictive low temperature condition only,

Whereas, the heat input evaluation calculates the pressure overshoot for a range of reactor coolant temperatures.

Thus, the LTOPS at Beaver Valley Unit 1 is designed to provide the-capability, during relatively low temperature RCS operation, to prevent the RCS pressure from exceeding allowable limits.

The LTOPS. is designed with redundant components to assure it will perform its function assuming any single active component failure. It is provided in addition to the administrative controls to reduce the likelihood that pressure transients will exceed the technical specification. pressure temperature limits. Based on the

above, the proposed change incorporates the LTOPS setpoint curve (Figure 3.4-4) based on temperature in place of a fixed setpoint into specification 3.4.9.3 and modifies the associated bases to clarify the setpoint description.

UFSAR Section 4.2.2.7 will also be revised to further describe the operation of LTOPS. Therefore, the proposed changes reflect the analysis and have been determined to be safe and will not reduce the safety of the plant. E. NO SIGNIFICANT HAZARDS EVALUATION The no significant hazard considerations involved with the proposed amendment have been evaluated, focusing on the three standards set forth in 10 CFR 50.92(c) 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:

ATTACRMENT B Page 4 1. Does the change involve a significant increase in the-- probability or consequences of an accident previously evaluated? Reactor vessel-rupture due to low temperature overpressure events is not part of the design basis for Beaver Valley Unit 1. Thus, reactor vessel rupture is excluded from the Beaver Valley Unit 1 accident analysis. However,- the setpoint analysis performed confirms that the LTOPS continues to ensure that the probability of a reactor vessel rupture is low enough that its inclusion in the accident analysis is not required. In

addition, administrative procedures have been provided for minimizing the potential for any transient that could actuate the overpressure relief system.

The revised LTOPS. setpoints, as a function of RCS temperature, continue to provide the capability to mitigate pressure transients at low-temperature, low-pressure operating conditions and reduce the likelihood of reactor vessel ruptures in the event of an overpressure transient during low temperature operation. Therefore, the proposed change will not significantly increase the probability or consequences of an accident previously evaluated. 2. Does the change create the possibility of a new or different kind of accident from any accident previously evaluated? The LTOPS pressure-temperature limits are set to prevent brittle fractures in the reactor vessel at temperatures below 275'F.. Reactor vessel rupture due to LTOP events is not part of the Beaver Valley design basis and reactor vessel-rupture is not included in the accident analysis.

However, the setpoint analysis performed shows that the revised temperature-dependent setpoints continue to ensure that the probability of a

reactor vessel rupture is low enough that it is able to be excluded from the accident analysis. The LTOPS is designed with redundant components to assure it will perform its function assuming any single active component failure. Two PORVs are provided to mitigate any potential pressure transients. Analyses have shown that one PORV is sufficient to prevent violation of these limits due to anticipated mass and heat input transients. Revision of the LTOPS setpoints does not reduce the reliability of the LTOPS.

Thus, the proposed change will not create the possibility of a new or different kind of accident.

1 i _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ -. _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ - _ - _ _ _ =

t ' ATTACHMENT B Page 5 s 3. Does the change involve a significant reduction in a margin of. safety? The revised 'LTOPS setpoints continue to ensure 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 275'F. Either PORV 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 25'F above the RCS cold leg temperature, or 2) the start of a charging pump and its injection into a water solid RCS, as described in the bases to the technical specifications. The LTOPS is designed with redundant components to assure it will perform its function assuming any single active component failure. Two PORVs are provided to mitigate any potential pressure transient. Analyses have shown that one PORV is-sufficient to prevent violation of these limits due to anticipated mass and heat input transients. Revision of the LTOPS setpoints does not reduce the reliability of the

LTOPS, nor does it affect the likelihood of vessel damage / failure -in the event of a

overpressure transient during low temperature operation. Therefore, the proposed change does not involve a significant reduction in a margin of safety and will act to reduce. challenges on the LTOPS setpoint while running reactor coolant pumps by providing additional margin. F. NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the above safety evaluation, it is concluded that the activities associated with this license amendment request satisfy the no significant hazards consideration standards of 10 CFR 50.92(c) and, accordingly, a no significant hazards consideration finding is justified. G. ENVIRONMENTAL EVALUATION The proposed changes have been evaluated and it has been determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be ~ released

offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22 (c) (9). Therefore, pursuant to 10 CFR '51.22(b), an environmental assessment of the proposed changes is not required.

4 ATTACHMENT C UFSAR Changes Beaver Valley Power Station, Unit No. 1 i ProDosed Technical Soecification Chance No. 176 i i f

.m ...,_m BVPS-1-UPDATED FSAR Rev. 2 (1/84) a temperature indicator-in the safety valve discharge manifolds - alert the operator to the passage of steam due to leakage or valves lifting. These pressure switches monitor the pilot valve chamber for leakage which could lift the safecy valve below setpoint. The pressurizer safety valve support is designed to withstand seismic, thermal, and dead weight forces in addition to the valve discharge reactions. The supports consist of: 1. Two circumferential anchor straps placed around the pressurizer vessel, and held in place by the forces of pressure and friction 2. Two built-up box sections which are welded to each other and are bolted to the straps 3. A flange welded to the box sections and bolted to the safety valve flange. Power Relief Valves The pressuriz'er is equipped with power-operated relief valves which limit system pressure for a large power mismatch and thus prevent actuation of the fixed high-pressure reactor trip. The relief valves are operated automatically or by remote manual control. The operation of these valves also limits the undesirable opening of the spring-loaded safety valves. Remotely operated stop valves are provided to isolate the power-operated relief valves if excessive leakage occurs. The relief valves are designed to limit the pressurizer pressure to a value below the high-pressure trip setpoint for all design transients up to and including the design percent step load decrease with steam dump but without reactor trip. In addition, the relief valves have a low pressure setpoint which is used when the reactor coolant system is cold and solid. The design basis for the low pressure setpoint is preventing over pressurization during the inadvertent starting of a high head safety injection pump and assumes no operator action for ten minutes. Design parameters for the pressurizer spray control, safety, and power relief valves are given in Table 4.1-8. Valve Operability Tests Full size proof tests to show that the pressurizer safety and relief valves and block valves would perform their intended function were performed. See Reference 6 for details. Loop Stop Valves The reactor coolant loop stop valves shown in Figure 4.2-9 are remotely controlled motor operated gate valves which permit any 1/J8N T-/ 4.2-12 l 1 n,

' NSERT 1 Low Temoerature Overoressure Protection System (LTOPS) The allowable system pressure is significantly less than the design pressure of 2485

psig, necessitating ; additional means to prevent brittle fractures of the reactor vessel at temperatures below 275'F.

Therefore, overpressure mitigation provisions for the reactor vessel, as provided by LTOPS, must be available when the RCS and the reactor vessel are at temperatures below 275'F. During a normal plant heatup, the RCS is open to the RHRS and may be operated for a short period of time in a water solid mode until the steam bubble is formed in the pressurizer. The RHRS is provided with self-actuated water relief valves to prevent overpressure caused either within the system itself or from transients transmitted from the RCS. During these low-temperature, low-pressure operating conditions, the LTOPS is armed and in a ready status to mitigate pressure transients. In determining the LTOPS setpoints, no credit is taken for -RHR relief valve operation. When the reactor coolant temperature has increased above about 275'F, the LTOPS is manually disarmed. During a normal plant cooldown, the LTOPS is manually armed as the reactor coolant temperature is decreased below 275'F. Note that at this time there is a steam bubble in the pressurizer and the water level is at the normal level for no-load operation. The RHRS is normally placed in service by opening the suction isolation valves prior to the LTOPs being placed in service. When the coolant temperature has been decreased to 160'F, the steam bubble may be quenched and the reactor coolant pumps stopped. From this point on in the

cooldown, the LTOPS will be in an active status ready to mitigate pressure transients which might occur.

Potential overpressurization transients to the RCS can be caused by either of two types of events to the RCS, mass input or heat input. Both types result in more rapid pressure changes when the RCS is water solid. Specifically, the LTOPS design bases transients are defined as: 1) the mass input transient caused by a normal charging / letdown flow mismatch after the termination of letdown flow and 2) the heat input transient caused by the restart of a RCP pump when a temperature asymmetry exists within the RCS due to the continued injection of cold seal injection water. For a particular mass japut transient to the RCS, the relief valve will be signalled to open at a specific pressure setpoint.

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

This overshoot is dependent on the dynamics of ---.x_-_-- x a-a

'IISERT 1 Page 2 the system and the input parameters and results in a maximum system pressure somewhat higher than the set pressure. Similarly, there will be a pressure undershoot, while the valve is relieving, both due to the reset pressure being below the setpoint and to the delay in stroking the valve closed. The maximum and minimum pressures reached in the transient are a-function of the selected setpoint and must fall within the acceptable pressure range. Note that the pressure overshoot and undershoot for the mass input case is greatest at low temperatures. Thus, the overshoot calculation is limited to the most restrictive low temperature condition only. Whereas, the heat input evaluation calculates the pressure overshoot for a range of reactor coolant temperatures.

Thus, the LTOPS at Beaver Valley Unit 1 is designed to provide the capability, during relatively low temperature RCS operation, to prevent the RCS pressure from exceeding allowable limits.

The LTOPS is designed with redundant components to assure it will perform its function assuming any single active component failure. It is provided in addition to the administrative controls to reduce the likelihood that pressure transients will exceed the technical specification pressure / temperature limits, t

5 ATTACHMENT D BEAVER VALLEY UNIT 1 I4W TEMPERATURE OVERPRESSURE PROTECTION SYSTEM (LTOPS) SETPOINT ANALYSIS AT 9.5 EPPY s

l

^ INIRODUCTION We IIM W ture Overpressure Protection System (IHOPS) limits pressure transients during cold shutdown heatup, aM cooldown operations in order to minimize the potential for inpairing reactor vessel integrity when operating at or near the vessel ductility limits. We imposition of this constraint, together with the requirements for reactor coolant punp operation, create a set of both high aM low, temperature depeMent, pressure bounds.

It is a regulatory requirement that the upper bound not be exceeded. Violatig the-lower bouM is not a regulatory concern, but does result in damage to the reactor coolant pump No. 1 seal. We goal of setpoint selection should be to prevent either bound from being exceeded below the IHOPS enable tenporature (275'F for Beaver Valley Unit 1). Westimhouse Electric Corporation was advised by Duquesne Light Ccxnpany that .the 9.5 Effective Full Power Year (EFPY) pressure-temperature limit (Appendix G curve) applicable to the Beaver Valley Unit 1 reactor vessel, and the minimum pressure required to start the reactor coolant pumps, has created an operational constraint which is unduly limiting the rate at which the plant is able.to heat up frcn cold shutdown. A meeting (May 11, 1988) was subsequently held at the Beaver Valley Site with DIro personnel in order to provide background and to explore possible solutions to the problem. As a result of that meeting, and a second meeting (July 14, 1988) held at the Westinghouse Energy Center, instruction was given to Westinghouse by DIro which defined the kind of analysis that would best meet the requirements of Beaver Valley Unit 1. Se second meeting authorized Westinghouse to proceed -with the analysis documented here. l The Westinghouse analysis is based on 3-loop water solid operation, the failure of one pressurizer power operated relief valve (PORV), and over-pressures resulting from mass injection events. The mass injection events L are assumed to result from the sudden loss of letdown concurrent with the operation of a single centrifugal charging pump. Se analysis is limited to mass' injection scenarios since a previous analysis for Beaver Valley Unit 1 l-has shown that, with a maximum temperature difference of 25'F between the RCS l-l

water in the steam generator tubes and the reactor vessel, heat injection events will not begin to dominate until temperatures well above the LiOPS enable temperature are reached. Setpoint evaluation is provided for steady-state Appendix G limits based on two different criteria:

1) the limits provided by the currently in-force technical specification, and 2) the limits calculated from the recently implemented revision 2 of NRC reg guide 1.99. Revision 2 changes the general procedure for calculating the effects of reactor vessel embrittlement.

The j steady-state pressure-temperature limit provides the greatest operational flexibility, and has been accepted by the NRC with the justification that "most transients occur during isothermal metal conditions." A key assumption is the neglect of the pressure instrument error. This is consistent with standard Westinghouse practice for LTOPS analysis. The LTOPS is considered to be a mitigation system, as opposed to a protection system, and the neglect of the pressure instrument error is understood and approved by I the NRC. Summary of Results The result of the analysis is summarized by Figures S-1 and S-2, illustrating i the range, as a function of RCS temperature, of acceptable LTOPS setpoints for steady-state pressure-temperature limits based on the two different methods of reactor vessel evaluation. Figure 1 is based on the limits contained in the i ctrrently in-force technical specifications, and Figure 2 shows the range as a result of implementing the methodology of NRC Reg. Guide 1.99, Rev 2. Both figures assume a reactor vessel exposure of 9.5 EFPY. The current technical specification-limits (Figure S-1) preclude protection of the reactor coolant pump No. I seal, in order to prevent the opening of both PORV's, the PORV setpoint difference must be in excess of 82 psi. The plant condition which allows a setpoint spread of that magnitude does not occur until relatively high temperatures (i.e., in excess of 200*F). Opening both PORV's results in an underpressure which is less than the pressure required to maintain separation of the pump seal faces. 8494e:1d/121488 3

i Tigure S-1 ... :_.; _.j s._ -~...',';_. _. _ '. ~ _.. ' - _ _ ~ _ '.. _ =.=.--l.:.::..9 =_=_L=.__ ~~ ~~~ -~ 4 - LTOPS Setpoint Range for Ssnele PORV Operation :-j-. 25ZZ -- Applicable to Steady State Pressure Temperature E d ~6~d Limits Without Instrumentation Uncertainty at E51== ~~

==s _.... 9.5 EFPY Reactor Vessel Esposure With Currently

==9:===-

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Figuro S-3 .. ; _. _; =s=1 ---. -t _.,_;: ___,._.;, = a.__g ;.q:.=r.3.._ _ LTOPS Seipoint Range f or 5inele PORV Operation Applicab!o " l~... =~~~- ~~Z -? - - ~ - --E ~;is to Steady-State Pressure-Temperature Limits Based on NRC E_? _ _ r -_ - - ~ ~_-IZ y_ Reg. Guide 1.99 Rev. 2 Without Instrument Uncertainty at ___ .= _. = _. 8.5 EFPY Reactor Vessel Exposure

{. _.=

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e m - ~=-3 o J ~=w&___ ?-~ _.: i = _.. e w_.. =_. . RCP Seal Protection -- ~ - - - - '-- C_r. J.-'... 1 4 =E-'._.__:---_ -+ l

==. ._ _ = _... _. -,... =4 = -... .7

==-- =C .. -:= a... A =sy-~._ RCS Temperature ('F) 4 -+ _2 g_ l ....i '...- c ---_._.4...] ~ &.+_.- 4_-._. 4--._.. _.__4___ 4 - --- C.T.' ] l 5

[ The vessel limits based on NRC Reg. Guide 1.99, Rev. 2, proved to be much less restrictive than those of the current technical specifications. From Figure S-2, it can be seen that pump seal protection can be provided above an RCS temperature of about I?O'F. Below that, setpoints cannot be selected that will prevent both PORV's from opening, s e 9 l I l l. 8494e:1d/121488 6 ~

. ' 0 DESCRIPPION OF 'IEE 1 HOPS SEHTOINT Alf0RI'Im l. 'Ihe determination of the lw tenperature werpressure protection setpoint is based on a local version of the LOFIRAN code. 'Ihe IDETRAN code models the reactor coolant

systen, includirg the steam generators, pressurizer (includirg ICRVs), and reactor coolant punps, as well as the control and protection systems, selected valving, and same balance of plant systems. 'No versions of the LOFIRAN code are utilized: the first version, used for the mass input calculations, collapses the several RCS loops into a single loop model; the second version, used for the heat input calculation, nodels each loop explicitly.

For Deaver Valley Unit 1, only the mass input calculations were performed. 'Ihe IHOPS setpoint analysis requires the considerations of a number of s/ stem parameters. Among these aru the foll w ing: l. Volume of the reactor coolant involved in the transient. 2. RCS pressure signal transmission delay. 3. Volumetric capacity of the relief valves vs. openirg position. 4. Stroke time of the relief valves (open and closirg). If the pressure undershoot is important, the closure tire is required. 5. Mass input rate into the RCS. 6. Heat. transfer characteristics of the steam generators. 7. Initial temperature asymmetry between the RCS and steam generator secondary water. 8. Mass of steam generator secondary water. 9. RCP startup dynamics.

10. RCP No.

1 seal delta P requirements. Important if a lwer setpoint limit is to be specified for RCP seal protection. l

11. Apperdix G pressure / temperature limits for the reactor vessel.

.1.1 PRESSURE LIMITS SEIECTION 'Ibe function of the IHOPS is to prevent the RCS pressure from increasing above the FORV pipirq analysis limits and the limits described by the allwable 7

t pressure / temperature characteristics for the specific reactor vessel material -in accordance with the rules given in Appendix G to 10CRF50. For Beaver Valley Unit 1, a constant pressure setpoint, independent of temperature, is employed so that the limits considered for this particular case must meet the mest restrictive segment of the Appendix G curves when compared to the overpressure transient as a function of temperature. The PORV piping limits are well above those pressures of concern at low temperature, and are considerations only if setpoint changes are anticipated with increasing temperature. A characteristic pressure / temperature relationship is shown by Figure 1.1, illustrating the allowable system pressure increases with increasing temperature. This type of curve sets the nominal upper limit on the pressure which should not be exceeded during RCS increasing pressure transients. When a relief valve is actue ed to mitigate an increasing pressure transient, the release of a volume of coolant through the valve will cause the pressure increase to be slowed and reversed as illustrated by figure 1.2. The system pressure then decreases, as the relief valve discharges coolant, until a reset pressure is reached where the valve is signaled to close. Note that the pressurs continues to decrease below the reset pressure as the valve closes. The nominal lower limit on the pressure during the transient is selected based on a requirement of the reactor coolant pump No.1 seal to maintain a nominal 200 psit differential pressure across the seal faces. The nominal upper limit (based on the minimum of Appendix G requirements or the PORV piping limitations) and the nominal RCP No.1 seal lower limit pres-sure values create an acceptable pressure range into which the PORV setpoints must.be fit. An illustration of thesc limits along with the setpoint selection range is shown in Figures 1.3 and 1.4. In the event that the setpoint selection range is insufficient to accommodate both the Appendix G and the RCP No. 1 seal limit, the Appendix-G limit will take precedence. 9 8494e;1d/121488 8

[ Tigure 1.1 APPENDIX 6 PRESSURE-TEMPERATURE LIMITS (TYPICAL) D 2500 < - 1 0000, g' C 1500, l I u 'F/HR g' 1000, uW E o 300 E i 1 300 200 300 400 500-l: l INDICATED COOLANT TEMPlRATURI 'T L' l 9 i

***-"^^ss44 y, ,4 S Figure ~ 12 TYPICAL PRESSURE TRANS!ENT (ONE CYCLE) Pyg--......... 9 Over L e sneoint - -- _ _ __.a _ l c ...Rtsti r PMIN*~~~*t l i UPnder l TIME 10 L

Figure 1.3 SETPOINT DETERMINATION (MASS IPPUT) 'WX APPIND116 ExlMUM LIMIT *7 MX.I-.N).% p 8 -...Pgy, r W RCP #1 $[AL E l l M]NIMUM LIM]T f '] $tTPolNT UNCE a PDRV StTPOINT, Ps!G.

The maximum pressure limit is the minimum of the Appendix $ Limit, the peak RC$ pressure based upon piping / structural analysis limits, or the peak RC$ pressure based upon SG tube sheet &P limits.

11

F6gure 1,4 w' SETPOINT DETERMINATION (HEAT INPUT) Appendix 6 Maximum Limit' f ... ' MAX. p MIN ..PM1N -... e y 1 e RCP di Sea) Minimum Limit $ETPOINT RANGE L / _i PS PORY SETPOINT, P$14

The maximum pressure limit is the minimum of the Appendix 6 Limit. the peak RCS pressure based upon piping / structural analysis limits, or the peak RCS pressure based upon SG tube sheet &P limits.

12

-1.2 MASS INPUT CONSIDERATIONS For a particular mass input transient to the RCS, the relief valve will be signaled to open at a specific pressure setpoint. However, as shown in Fievia ~ 1.2, there will be a pressure overshoot during the delay time before ths valve starts to move and during the time the valve is moving to the full open position. This overshoot is dependent on the dynamics of the system and the input parameters (e.g. mass flow rate), and results in a maximum system pressure somewhat higher than the set pressure. Similarly, there will be ar, undershoot while the valve is relieving, both due to the reset pressure being below the setpoint and to the delay in stroking the valve closed. The maximum and minimum pressures reached in the transient are a function of the selected setpoint and must fall within the acceptable pressure range shown in Figure 1.3. A number of mass input cases are run at various input flow rates selected to bound the expected setpoint range. From these runs, a locus of the maximum and ninimum pressure values is generated over the expected setpoint range, as shown in Figure 1.3. The shaded area represents the acceptable range from which to select the setpoint. 2.3 HEAT INPUT CONSIDERATION The heat input case is analyzed in about the same way as the mass input case except that the locus of transient pressure values vs. selected setpoints are determined for several values of the initial RCS temperature. This heat input evaluation provides a range of acceptable setpoints dependent on the reactor coolant temperature, whereas the mass input case is limited to the most restrictive low temperature conditions only. The shaded area on Figure 1.4 highlights the acceptable range for a heat input transient for a particular initial reactor coolant temperature. 2.4 FINAL SETPOINT SELECTION By superimposing the results of the several mass input and heat input cases evaluated (from a series of figures such as 1.3 and 1.4), the range of L allowable setpoints can be determined that will satisfy both mass input and heat input considerations. As previously stated, the selection of the a494e:1d/121488 13 l

,a pressure setpoint for the PORVs is based on the use of nominal upper and lower -limits. The LTOPS is considered to be a mitigation system, as opposed to a

-j protection system, and the use of nominal limits is understood and approved by j

1 the NRC.- 1 -] l l 4 -l i i i a 6494e:1d/121488 14

1 i 1 2.0 LTOPS SETPOINT ANALYS1$ The LTOPS setpoint analysis presented in this section was developed for Beaver Valley Unit 1 using the algorithm discussed in the previous section. In addition, as part of the discussion, justification of selected plant parameters used for input to the LOFTRAN deck is provided. 1 The LTOPS currently installed in Beaver Valley Unit 1 features a constant value setpoint program (i.e., program independent of temperature) with an enable / disable RC3 temperature of 275'F. At various meetings with DLCo, j Westinghouse was informed that plant operations would like the option of resetting the LTOPS setpoints at selected RCS temperatures in order to minimize operational constraints, for this reason, specific setpoints have not been recommended. Rather, a range of acceptable setpoints as a function of RCS temperature is provided. The range is upper bounded by Appendix G, and lower bounded by RCP seal requirements. 2.1 OPERATIONAL LIMITS 2.1.1 Technical Specification Pressure-Temperature Limit The pressure-temperature limit curves specified by the currently in-force technical specifications are shown in Figures 2.1 and 2.2. These curves are applicable to 9.5 effective full power years (EFPY). The limit curves in digitized form, without pressure instrument error, are provided by Tables 2.1 and 2.2. The digitized values were obtained from the digitizer of the S.E.L. computer. 2.1.2 NRC Rec. Guide 1.99, R6v j, Pressure-Temperature Limit L The recently implemented Revision 2 of NRC Reg. Guide 1.99, revises the l general procedure for calculating the effects of reactor vessel embrittlement. The pressure temperature limits resulting from the new procedure is shown by figures 2.3 and 2.4, with the digitized values provided in Tables 2.3 and 2.4. The pressure-temperature limits are applicable to 9.d EIPY, a494e:1d/121468 15

I $ttRIALPROPttTYBA$tl Tisure 8.1 l COWfl0LLtW MTERIAlt WLD BETAL COPPER 10NTEWit

9. 1 Vf4 PN0$PMORV5 CONTtwis

9. Il W"I b.6(FPT:

1/47,!?4'F 3/47.187'F RND .k' TNT S NTY1 N 4k b10$ 0 N P$h0k 0 0 i 1 PossitLE!strThimtwfERRORs 3000 3 1 ;,,;. i I I i i g 8 l 1 I n"i i l l I ~ I i l i l! 1 j ie l i il l i li ! i i I i. t. I ~ ~l l 11 fi' 18 ? l l i llll si l i Ti c11til .i I*

,i 111

tiI (- i I 1 J li!*I l t! f t il I l.I J 111 I Is'I ii UECCEPTABLE f IiI: ji, OPERAtlh6 R1410u / l ;i: i l J

l'!

( l.< 1000 J. '41: l 'lj. i F l 2 AE/ i ACCEPTABLE C00LDOWN M V OPLRATpm 41610N ^ RAtts 'r/NR i, j:ggy [ yaillF7 i 'i '~ I 40 - IW i4 'I 40 - i .j_ ~ ~ 100 3 7 i 9 100 200 NC 400 500 le!CATED TDIPERATWtt (*F) TECHNICAL $PECIFICATION C00LDOWN RATE LIMITATIONS KAVER VALLEY UNIT 1 16

l TABLE 2.1 DIGIT 12E0 RCS PRESSURE / TEMPERATURE CODLOOWN LIMITS

PRESSURE LIMIT (PSIG) AT INDICATED COOLDOWN RATE ('F/HR)

RCS Temp ('F) 0.0 20.0 40.0 60.0 100.0 80.0 491.1 454.4 406.7 364.9 276.0 100.0 499.1 459.4 412.5 370.1 283.5 120.0 505.5 463.5 417.4 374.4 288.3 140.0 511.9 468.4 422.9 379.7 293.0 160.0 519.9 475.6 430.9

387.7 300.0 180.0 531.1 486.9 442.9 400.2 311.8 200.0 547.1 503.8 460.7 419.0 330.8 220.0 569.5 527.9 485.9 445.8 359.4 240.0 600.0 560.6 520.3 482.3 400.1 260.0 640.0 604.2 565.6 530.5 455.3 280.0 691.3 659.2 623.5 591.9 527.6 300.0 755.4 728.7 695.6 668.5 619.3
The pressure limits do not include the 60 psi pressure instrument uncertainty included on the curves shown on Figure 2.1.

Reactor vessel exposure = 9.5 EFPY e u p :1onataan 17

-~ i Figure 2.2 j $Tp!AL PMPBTY AAlli. i 49fft0L1!sII MfDIAL: ELD E TAL COPPtt CONTDft 0.31 tfrs PMb5PN0tW5 ColffDit 0.015 Ifft RTggIn!TIAL: t'F Rf APTR 4.5 EFFT 1/47. 274'F gy 3/4T.13?'F CUWt APPLICABtt FOR NLATUP Rafts UP TO 90*F/m POR TE SCWICE Ppt 00 W TO l 9.8 (FPY AnD CONTAllen MA41115 0F 10*F AND to Pill POR POMISLE Ilt871ptLiff LAA043 .I a.1 o I! j i l l 1 ,6'

1 i.. 1, i

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il j

/ /1 i tj! 0l [ [i .ill'i. 'l j j l g it':Ti i i i ni I J I ..:ti'. i / (8 i. g h / / litll l ...il't,' I i' I 5 g .- i.i l, s i..c Ac,et,f. C .- m 1 s . uAfili. uu0li I. 4 e i 3 jt / L I i .i i l' / i . ii.t

h r

d J gggyy., q q p p : RA .L d "p u n um... mum : 60*F/* _: ;;i,, 1-1 L,,, e etttitAlftY LfMf7 ~~ y f i1'I W LASO 014 tit $tRVICE +;I-el 11

j i

WYD40$7ATIC 7t!T TEMPILAfullt (414'F) I II I i'. ..j. 9. FOR twt $ tty! I.

~ Ptit!00 W TO

.5 EF g, 7 ip, p., g 7 e 4 100 800 300 400 000 tlGICATD TDFRATWtt (*F) TECHNICAL SPECIFICATION 60 'F/HR EATlr ftATE LIMIThTION SEAVER VALLEY UNIT I 18 ~

TABLE 2.2 DIGITIZED RCS PRESSURE / TEMPERATURE HEATUP LIK;T AT 60'F/HR* RCS Temp ('F) Press (psic) RCS Temp ('F) Press (psio) 80.0 487.0 200.0 545.2 100.0 497.5 220.0 567.0 120.0 505.1 240.0 597.1 140.0 511.8 260.0 637.6 160.0 519.5 280.0 690.2 180.0 530.0 300.0 756.9

The pressure limits do not include the 60 psi pressure instrument uncertainty included with the curve shown on Figure 2.2.

Reactor vessel exposure = 9.5 EFPY e494..wirisse 19

IthTERIAL PROPERTY 4A515 rigure 2.3 CONTROLLING ETERIAL: Intermediate Plate INITIAL RT 8

I3*I ET RT AFTER 9.5 EFPY: 1/47. 202'F ET 3/4T. 176'F i i CURVI5 APPO.ABLE FOR C00LD0tfN RATES UP TO 100'F/HR FOR THE SERVltt PERIOD UP TO 9.5 EFPY. 110 ERGIN5 ARE $1VEN FOR P0551tLE INSTRUMENT ERROR 5. l-rl00 m 2250 l r i 2000 l 2 I 1,50 i I r E 1500 -/ unacceptable Operation / 1250 / l Acceptable I1000 f Operation r /_ g s 750 Cooldown f) l Rates ure 'F/Hr ' L ""w/ 500 ; 0 ar g m-- 250 0 3 to le 100 150 ' t00 itle 300 350 400 el0 500 i 10001C418D TEMPERA 1WRC (DCo.F) NRC RE6. SUIDE I.09, REY. I C00LDOWN RATE LIMITATIONS BEAVER VALLEY UNIT I 20

e' ~ TABLE 2.3 O! GIT!IED RCS PRESSURE / TEMPERATURE C00LOOWN LIMITS

BASE 0 ON NRC REG. GU10E 1.99, REV. 2

' PRESSURE LIMIT (PSIG) AT INDICATED CODLOOWN RATE ('F/HR) RCS Temp ('F) 0.0 20.0 40.0 60.0 100.0 85.0 523.7 485.3 446.3 406.3 323.9 100.0 533.7 495.5 456.4 416.7 334.7 120.0 550.9 513.0 474.5 435.3 354.6 140.0 573.9 536.9 499.2 460.9 382.5 160.0. 604.6 568.9 532.7 496.0 421.3 180.0 645.8 611.9 577.9' 543.7 474.7 200.0 700.5 669.7 639.0 608.3 547.7 220.0 773.8 747.1 721.1 695.6 646.8 240.0 871.6 850.8 831.4 813.0 780.9 260.0 1001.7 989.4 -979.1 970.B 961.2 280.0 1175.2 1174.4 N/A N/A N/A ~ 300.0 1405.6 N/A N/A N/A N/A Notes:

1) Reactor vessel exposure = 9.5 EFPY 2)

Instrumentation uncertainties not included

3) N/A = Data not available u p.:1en214:e 21

i ngue 2.4 . IETERIAL PROPERTY Balls CONTROLLING MATERIAL: Intemediate Platt INITIAL RTNDT' II*I RT AFTER 9.5 EFPY: 1/47,202'F hDT 3/47,176'F CURVIS APPLICABLE FOR NEATUP RATES UP 70 60*F/HR FOR THE SERVICE PERIOD UP TO j 9.5 EFPY. NO MARGINS ARE GIVIN FOR P0551BLE INSTRU, MENT ERROR $. I 2$00 w i n, J ] E Leak Test l 2250 Limit 1 f 1 J E I 2000 f l 1 1 l . f I l 1750 Heatup Rates J l UgF/Hr - f f To 60 r G 1900 . l I' l ? 1250 Unacceptable f ) Operation . / 1000 Criticality l r Limit Based ~ 9 / 'on Inservic;- Hydrostatic l 3 73, f Test Temp. e l / (329'F)for e the Service - L 500 - Period Up Ti._- 9.5 EFPY Acceptable 250 Operation i i > > i i i., ,e le 100 140 200 250 300 350 400 "she 600 enescatto truptaatunt (ets.F) NRC RES. SUIDE 1.99, REV 2 69 'F/HR HEATUP RATE LIMITATION BEAVER VALLEY UNIT I 22

TABLE 2.4 DICIT12ED RCS PRES $URE/ TEMPERATURE HEATUP LIMIT AT 60'F/HR* BASED ON NRC REG. GUIDE 1.99, REV. 2 RCS Temp'('F) Press (psic) RCS Temp ('F) Press (psic) 85.0 469.2 200.0 602.3 100.0 469.2 220.0 672.3 120.0 471.2 240.0 766.0 140.0 486.1 260.0 890.8 160.0 512.4 280.0 1057.1 180.0 550.3 300.0 1277.2 Notes: 1) Reactor vessel exposure = 9.5 EFPY

2) Instrumentation uncertainties not included e404..ian 214ss 23

i 4 2.1.3 Limit for Setpoint i The functional limit used as a basis for setpoint selection that provides the greatest operational flexibility is the steady-state pressure-temperature limit. This limit has been accepted by the NRC with the justification that

most transients occur during isothermal metal conditions." This position has recently (1986) been reconfirmed by the NRC.

The steady-state pressure-temperature limit from the technical specifications, and the limit derived 3 from revision 2 of NRC Reg. Guide 1.99 are reproduced, without instrument uncertainty, in Figures 2.5 and 2.6 respectively. 4 2.2 PORY STROKE TIME The PORV opening and closing times assumed for the analysis were determined by DLCo. The times chosen by DLCo. were 3 seconds for both the opening and the closing times. Included within the 3 seconds is'the delay time from when system pressure reaches the setpoint value and the PORV begins to move. The closure time has no impact on the overpressure transient (an Appendix G consideration), but does impact the underpressure transient (important for the protection of the reactor coolant pump number 1 seal). The characteristic of the transient is such that as the closure time increases, the underpressure becomes more severe. 2.3 PORV OPERATION The design basis for an overpressure event requires that either PORV provide adequate relieving capability in the event of a single valve failure. Therefore, the setpoint analysis is based on the assumption of single valve operation. The setpoint analysis includes the effect of time delays associated with the transmission of the wide range reactor coolant system pressure signal. A i a494e id/121488 24

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C.'_ x: E Steady-State Cooldown Pressure-Temperature Limits. E ' T.d l rr.::r: Without Instrumentation Uncertainty. Based on the F M - a _, m-E-M M=5E"_=_

= M Currently In-Force Technical Specifications at W=9 "_ =a

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- -4

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o ~ Figure 206 .4

4

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= = ' =

= =- .ig A.:;; f.;.; _; g ;_=a=- 1.- Withowt Instrunentation Uncertainty i 35 " E5ELi-

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-.=- conservative value of 0.9 seconds was utilized for the analysis. The breakdown of the time delay is as follows: Capillary delay 0.15 sec. Pressure transmitter delay 0.20 sec. Electronics delay 0.10 sec. Solenoid valve actuation delay 0.10 sec. Pneumatic delay 0.35 sec. Total 0.90 sec. The opening characteristics (valve C(v) vs. valve stroke) for the pressurizer power operated relief valves are shown by figure 2.7. 2.4 MASS IN]ECTION LOFTRAN SPECIFICATIONS AND RESULTS In order to determine the maximum overpressure, the mass input mechanism assumed the operation of a single centrifugal charging pump supplying fluid to the reactor coolant system at the maximum rate allowed by system pressure. The maximum flow from a charging pump as a function of system pressure, is provided in Table 2.5 and Figure 2.8. l The specifications for this transient are summarized as follows: l Temperatures Reactor coolant system temperature = 90'F. Reactor Coolant System Volume l 3 RCS volume = 9,619.5 ft for Beaver Valley Unit 1 Initial Reactor Coolant System Pressure The initial RCS pressure was assumed to be 200 psi less than the setpoint pressure. This is conservative and assures that the puew1214ss 27

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,e l TABLE 2.5 MA';IMUM RCS CHARGING FLOW BEAVER VALLEY UNIT 1 Mass Injection Rate RCS Press (psic) lbm/see gpm* 2400.0 20.4 147.6 2200.0 30.0 217.1 2000.0 34.2 247.4 1800.0 38.8 280.7 1600.0 42.6 308.2 1400.0 46.1 333.5 1200.0 49.2 356.0 1000.0 52.5 379.8 800.0 55.0 397.9 600.0 57.6 416.7 400.0 59.7 431.9 200.0 62.3 450.8 0.0 64.8 468.8

1bm/sec=(0.138214)(gpm)

Note: 1) Flow due to the operation of a single centrifugal charging pump e4s4 1on214es 29

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i ) \\ transient is well def ned by the time the PORV setpoint is reached. Given this definition, the overpressure is essentially insensitive to the initial pressure selection. Also, a starting pressure of about 200 psi less than the PORV setpoint provides a conservative undershoot j for the RCS number 1 seal evaluation. ) i Reactor Coolant Relief Capacity The transient is analyzed assuming the failure of one PORV. PORV Characteristics LTOPS setpoints selected for the parameter study = 400, 500, 600, and 700 psig. 6 PORV characteristic curve is shown in Figure 2.7 Opening time = 3.0 sec., including signal delay Closing time = 3.0 sec., including signal delay C(v) = 46 Mass injection Flow Mass injection rates selected for the parameter study = 40, 100, 200, 300, 400, 500, and 600 gpm. Pressure Signal Transmission Characteristics Time delay to PORV stem motion = 0.9 sec. (Ref. Section 2.3) The mass injection cases were conservatively analysed at low temperature where the pressure transient resulting from a mass input shows the greatest over-g l shoots and undershoots. No other calculations were performed at different temperatures. s494e:1d/1214aa 31 l

Wass injection rates ranging from 40 to 600 gpm were selected for the mass input parameter study portion of this analysis. This range was based on analy-ses performed for units similar to Beaver Valley Unit 1 and provides a good definition of the relationship between mass input and the resulting pressure transient. The results of the LOFTRAN runs are sumarized in Table 2.6, and presented graphically in Figures 2.9 and 2.10 for both the overpressure and underpressure transients, respectively. 2.5 SETPOINT EVALUATION The pressure overshoot and undersheet values resulting from overpressure events at maximum mass injection rater for specific LTOPS setponts is provided by Table 2.7 and in Figure 2.11. The ocershoot values are obtained from Figure 2.9 for the maximum mass injection rate at the system pressure defined by the LTOPS setpoint (ref. Figure 2.8). The undershoot data was conserva-tively estimated as the minimum value indicated by Figure 2.10 for the setpoint in question. This defines the minimum setpoint to provide reactor coolant number 1 seal protection for all overpressure events, regardless of the mass injection rate. The data from these tables and figures is presented in Figures 2.12 and 2.13 showing the maxima and minima system pressure as a function of LTOPS setpoint. Figure 2.12 has superimposed the pressure limits at specific RCS temperatures for the steady-state pressure-temperature limit contained in the current technical specifications. Figure 2.13 shows the same limits based on NRC Reg. Guide 1.99, Rev. 2. Both figures show the minimum pressure required to provide reactor coolant pump seal protection. The minimum system pressure required to maintain number 1 seal separation is shown in Figure 2.14. The intersection of these limits (Appendix G and RCP seal) with the most limiting of the maxima and minima curves forms the basis for the construction of Figures 2.15 and 2.16. These figures show, respectively, the range of acceptable LTOPS setpoints, as a function of RCS temperature, that will meet the technical specification and the Reg. Guide 1.99, Rev. 2 steady-state pressure-temperature limits. sas4. tonn4se 32 1

TABLE 2.6 BEAVER VALLEY UNIT 1 LTOPS MASS INJECTION LOFTRAN

SUMMARY

Overpressure Underpressure Massinj. LTOPS' Setpt Value Delta P Value Delta P Run Rate (com) (psic) (psic) (psi) Ipsic) (psi) DLWM101 40.0 400.0 404.0 4.0 285.0 115.0 500.0 504.0 4.0 373.0 127.0 600.0 603.0 3.0 463.0 137.0 700.0 703.0 3.0 555.0 145.0 DLWMIO2 100.0 400.0 411.0 11.0 294.0 106.0 500.0 511.0 11.0 381.0 119.0 600.0 610.0 10.0 470.0 130.0 700.0 710.0 10.0 560.0 140.0 DLWM103 200.0 400.0 428.0 28.0 310.0 90.0 500.0 527.0 27.0 397.0 103.0 600.0 626.0 26.0 486.0 114.0 1 700.0 725.0 25.0 575.0 125.0 DLWM104 300.0 400.0 448.0 48.0 323.0 71.0 l 500.0 546.0 46.0 411.0 89.0 600.0 645.0 45.0 500.0 100.0 l 700.0 743.0 43.0 590.0 110.0 i DLWM105 400.0 400.0 473.0 73.0 333.0 67.0 500.0 571.0 71.0 421.0 79.0 600.0 668.0 68.0 510.0 90.0 700.0 766.0 66.0 599.0 101.0 DLWM106 500.0 400.0 500.0 100.0 343.0 57.0 500.0 596.0 96.0 430.0 70.0 600.0 694.0 94.0 519.0 81.0 700.0 791.0 91.0 609.0 91.0 t 1 DLWM107 600.0 400.0 526.0 126.0 354.0 46.0 500.0 622.0 122.0 441.0 59.0' 600.0 720.0 120.0 529.0 71.0 i 700.0 817.0 117.0 619.0 81.0 Notes: 1) Single PORV actuation.

2) PORV opening time = 2.1 sec.

3) PORV closing time = 2.1 sec.

4) PORV actuation delay time : 0.9 sec.

5) Initial pressure = LTOPS setpoint - 200 psi 64s4. w mass 33

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TABLE 2.7 BEAVER VALLEY UNIT 1 NAX OVERSHDDT/UNDERSHOOT VALUES

Overshoot Undershoot Setpt Press WassInject.

Delta P Value Delta P Value (psic) Rate (apm) (psi) (psis) (psi) (esic) 400.0 432.0 82.0 482.0 121.0 279.0 500.0 426.0 77.0 577.0 135.0 365.0 600.0 417.0 73.0 673.0 146.0 454.0 700.0 408.0 68.0 768,0 154.0 546,0 Notes:

1) Single PORV actuation

2) PORV actuation delay time = 0.9 see

3) PORV opening time = 2.1 sec

4) PORV closing time

2.1 sec

5) Mass injection rate obtained from Charging Flow vs RCS Pressure (Figure 2.8)

6) Overshoot obtained from tax delta P overshoot vs mass injection flow curve (Figure 2.9)

7) Undershoot obtained from max delta P undershoot vs mass injection flow (Figure 2.10) 6 so4. w m ss 36

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~ 2.6 REACTOR COOLANT PUMP SEAL PROTECTION The current technical specification limits (Figure 2.15) precludes protection of the reactor coolant pump No. I seal. In order to prevent the opening of both PORV's, the PORY setpoint difference must be in excess of 82 psi at a setpoirit pressure of 400 psig (ref. Table 2.7). The plant condition which allows a setpoint spread of that magnitude does not occur until relatively high temperatures (i.e., in excess of 200*F). Opening both PORV's results in an underpressure which is less than the pressure required to maintain separation of the pump seal faces. The vessel limits based on NRC Reg. Guide 1.99, Rev. 2, proved to be much less restrictive than those of the current technical specifications. From Figure 2.16, it can be seen that pump seal protection can be provided above an RCS temperature of about 120'F. Below that, setpoints cannot be selected that will prevent both PORV's from opening. l 2.7 PORV PIPING LIMITS Westinghouse practice, for all LTOPS analysis, has been to select setpoints such that the peak overpressure will.c exceed 800 psig or Appendix G, which ever is the most limiting. The 800 phs limit results from an analysis of water hammer effects on relief valve piping for certain classes of rapidly opening valves (e.g., Garrett solenoid valves) under water solid conditions. f Given the right combination of rapid stroke time and a sharply rising charac-teristic curve, the valve becomes effectively full open or full closed within a few tenths of a second, thus setting the conditions for a water hammer. Tne flow through an air operated valve, such as the Masoneilan valve featured at Beaver Valley Unit 1, is relatively less sensitive to stem position than te solenoid valves. When combined with the relatively slow opening and clos-ing times, the water hamer forces will be much reduced, if not effectively I eliminated. Evaluation of water hammer forces on the piping of air operated valves has not been performed by Westinghouse. The practice has been to assume the conservative psition of taking the worst case results and applying them { to all LTOPS setpoint t valuations, regardless of the relief valve employed. 8'.d4 e :1d/121488 43 ( 1

QLd-90 'JJI 40 s', 4 REFERENCES

1. SATO-CSDA-291/SATO-PDE-185, ' Water Solid Cold Overpressure Analysis Report for Beaver Valley Unit l', K.A. Gaydos and'S. Abedin, 5/16/83

2. Calc-note CSDA-84-96

Cold Overpressure Mitigation System for KORI 5 & 6 Setpoint Determination', Peter Santchev. 7/12/85

3. MDQ-PVE-698, 'Masonellan Valve Characteristics", L.I. Ezekoys and R.E. Kelley, 8/26/82

4. Calc-note CSDA-87-4,

Trojan Unit Cold Overpressure Mitigation System (CDMS)

Setpoint Evaluation With an 51 Pump Charging Configuration *, John P. Mutz, 1/27/86

5. Beaver Valley Unit 1 Technical Specification, LCO 3.4.9.1, Amendment 111.

( Attachment 4 to thae cale-note)

8. Calc-note FA-82-87 *.0verpressure Natigetton System letpoint Study for Byron /Braidwood Plants', L.E. Engelhart, 18/01/82

7. MT-SMART-195(88 ),

Data Pointe for Beaver Valley Unit t Heatup and Cooldown Curves for 9.5 EFPY', N.K. - Ray, 11/18/88

8. Calc-note CSDA-85-52 " Cold Overoressure Mitigation System Setpoint Determination for Beaver Valley Unit 2 (DNW)',

E.M. Mann, 7/19/85

9. Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 54 to Facility Operating License No. DPR-39 and Amendment No. El to Facility Deerating License No. DPR-48 Commonwealth Edison Company Zion Station Units I and 2, Docket Nos. 50-295 and 50-304.

19. CSAS-CSDA-569.,'Sizeve11 B COMS Wide Range Pressure Channel Signal Transmission Decay Time", L.E. Engelbert, S/19/86

11. MT/ SMART /1247, 'LTDPS Setpoints", R.L. Turner, 9/29/06

12. 86TD-4.8.4.1-4196, ' Steam Benerator Information Report *,.R.M. Wilson, Feb. 1995

13. Beaver Valley Unit I (DLW) Power Capab.111ty Parameters, Issue No. 84-3, 11/16/84 ( Attachment 5 to this Calc-Note)

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